karen d. fili southern nuclear site vice president ...sep 18, 2015 · karen d. fili southern...
Post on 24-Jan-2021
3 Views
Preview:
TRANSCRIPT
Karen D. Fili Southern Nuclear Site Vice President Operating Company, Inc. Plant Vogtle Units 3&4 7825 River Road Waynesboro, GA 30830 Tel 706.848.7717 Fax 706.826.5796 kdfili@southernco.com
September 18, 2015 Docket Nos.: 52-025 ND-15-1333
52-026 10 CFR 55.46(b) U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Mr. Glenn M. Tracy Director, Office of New Reactors U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Southern Nuclear Operating Company
Vogtle Electric Generating Plant Units 3 and 4 Request for a Commission-Approved Simulation Facility
Ladies and Gentlemen:
Pursuant to 10 CFR 55.46(b), Southern Nuclear Operating Company (SNC) requests a
Commission-Approved Simulation Facility for Vogtle Electric Generating Plants (VEGP) Units 3
& 4.
The enclosures provide information required by 10 CFR 55.46(b) for facility licensees that
propose use of a simulation facility other than a plant-referenced simulator in the administration
of operating tests under 10 CFR 55.45(b)(1). Enclosure 1 provides summaries of SNC
evaluations of open simulation facility discrepancies with respect to; AP1000 simulation facility
Unresolved Items (UIs) issued by the Nuclear Regulatory Commission (NRC), their cumulative
effect on operator performance, simulator conformance with the AP1000 plant design, and
variances between AP1000 simulation facilities. Subsequent enclosures provide supporting
details.
Pursuant to 10 CFR 2.390, SNC requests that the specified information be withheld from public
disclosure. In support of this request for withholding, SNC has attached to this letter the
following documents:
Enclosure 5P contains an evaluation of AP1000 simulation facility Unresolved
Items (UIs) that were issued by the NRC.
U. S. Nuclear Regulatory Commission ND-15-1333 Page 2 of 5
Enclosure 6P contains an evaluation of open simulator deficiencies and their
aggregate impact on 10 CRF 55.45 criteria.
Enclosure 8P contains an evaluation of Priority One (1) Potential Human
Engineering Discrepancies (PHEDs) from Integrated Systems Validation (ISV)
Daily Assessments.
Enclosure 9P contains a list of simulator discrepancies that were open as of May
15, 2015.
Enclosure 12 contains an affidavit for withholding proprietary information from
public disclosure, executed by Westinghouse. The Affidavit sets forth the basis
on which the information may be withheld from public disclosure by the
Commission and addresses with specificity the considerations listed in paragraph
(b)(4) of Section 2.390 of the Commission’s regulations.
Accordingly, it is respectfully requested that the information which is proprietary to WEC be
withheld from public disclosure in accordance with 10 CFR 2.390. Correspondence with respect
to the copyright or proprietary aspects of the items listed above or the supporting WEC Affidavit
should reference CAW-15-4260 and should be addressed to James A. Gresham, Manager,
Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3
Suite 310, Cranberry Township, Pennsylvania 16066.
To support the operator licensing schedule, SNC respectfully requests NRC approval of this
request by December 18, 2015.
This letter contains no regulatory commitments. If you have any questions, please contact
Michael Yox at (706) 848-6459.
Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY
Karen D. Fili
Vice President,
VEGP 3&4 Operational Readiness
Nuclear Development
KDF/MC/sdc
U. S. Nuclear Regulatory Commission ND-15-1333 Page 3 of 5 Enclosure 1: Information Provided Pursuant to a 10 CFR 55.46(b) - Request for a
Commission-Approved Simulation Facility
Enclosure 2: Description of the Components of the Simulation Facility Intended to be Used for Each Part of the Operating Test - 10 CFR 55.46(b)(1)(i)
Enclosure 3: Description of the Performance Tests for the Simulation Facility and Results of the Tests - 10 CFR 55.46(b)(1)(ii)
Enclosure 4: Description of the Procedures for Maintaining Examination and Test Integrity Consistent with the Requirements of 10 CFR 55.49 - 10 CFR 55.46(b)(1)(iii)
Enclosure 5: Evaluation of AP1000 Simulation Facility Summary of Unresolved Items (UIs) Issued By the NRC
Enclosure 5P: Evaluation of AP1000 Simulation Facility Summary of Unresolved Items (UIs) Issued By the NRC (Withhold from Public Disclosure)
Enclosure 6: Commission Approved Simulator Aggregate Study - Simulator Training System Deficiency Impact on 10 CRF 55.45
Enclosure 6P: Commission Approved Simulator Aggregate Study - Simulator Training System Deficiency Impact on 10 CRF 55.45 (Withhold from Public Disclosure)
Enclosure 7: List of Westinghouse Simulator Corrective Actions
Enclosure 8: Evaluation of Priority One (1) Potential Human Engineering Discrepancies (PHEDs) from Integrated Systems Validation (ISV) Daily Assessments
Enclosure 8P: Evaluation of Priority One (1) Potential Human Engineering Discrepancies (PHEDs) from Integrated Systems Validation (ISV) Daily Assessments (Withhold from Public Disclosure)
Enclosure 9: List of Open Simulator Discrepancies
Enclosure 9P: List of Open Simulator Discrepancies (Withhold from Public Disclosure)
Enclosure 10: BEACON
Enclosure 11: Acronyms & Definitions
Enclosure 12: Westinghouse Authorization Letter CAW-15-4260, Application for Withholding Proprietary Information From Public Disclosure, Accompanying Affidavit, Proprietary Information Notice and Copyright Notice
U. S. Nuclear Regulatory Commission ND-15-1333 Page 4 of 5 cc:
Southern Nuclear Operating Company / Georgia Power Company
Mr. S. E. Kuczynski (w/o enclosures) Mr. J. A. Miller Mr. D. A. Bost (w/o enclosures) Mr. M. D. Meier Mr. M. D. Rauckhorst (w/o enclosures) Mr. J. T. Gasser (w/o enclosures) Mr. D. H. Jones (w/o enclosures) Ms. K. D. Fili Mr. D. R. Madison Mr. T. W. Yelverton Mr. B. H. Whitley Mr. C. R. Pierce Mr. D. L. Fulton Mr. M. J. Yox Mr. T. R. Takats Mr. W. A. Sparkman Mr. J. P. Redd Document Services RTYPE: VND.LI.L00 File AR.01.02.06
Nuclear Regulatory Commission
Mr. V. M. McCree (w/o enclosures) Mr. M. Delligatti (w/o enclosures) Mr. L. J. Burkhart (w/o enclosures) Mr. P. Kallan (w/o enclosures) Mr. C. P. Patel Ms. D. L. McGovern Mr. B. M. Bavol Ms. R. C. Reyes Ms. M. A. Sutton Mr. M. E. Ernstes Mr. G. J. Khouri Mr. L. M. Cain Mr. J. D. Fuller Mr. C. B. Abbott Ms. S. E. Temple Mr. I. A. Anchondo
Oglethorpe Power Corporation
Mr. M. W. Price Ms. K. T. Haynes Ms. A. Whaley
U. S. Nuclear Regulatory Commission ND-15-1333 Page 5 of 5 Municipal Electric Authority of Georgia
Mr. J. E. Fuller Mr. S. M. Jackson
Dalton Utilities
Mr. D. Cope Mr. T. Bundros
Westinghouse Electric Company, LLC
Mr. R. Easterling (w/o enclosures) Mr. J. W. Crenshaw (w/o enclosures) Mr. C. D. Churchman (w/o enclosures) Mr. L. Woodcock Mr. P. A. Russ Mr. T. G. Rubenstein Mr. G. F. Couture Mr. M. Y. Shaqqo
Other
Mr. J.E. Hesler, Bechtel Power Corporation Ms. L. Matis, Tetra Tech NUS, Inc. Dr. W. R. Jacobs, Jr., Ph.D., GDS Associates, Inc. Mr. S. Roetger, Georgia Public Service Commission Ms. S. W. Kernizan, Georgia Public Service Commission Mr. K. C. Greene, Troutman Sanders Mr. S. Blanton, Balch BinghamMr. R. Grumbir, APOG
Southern Nuclear Operating Company
Vogtle Electric Generating Plant (VEGP) Units 3 and 4
ND-15-1333
Enclosure 1
Information Provided Pursuant to a 10 CFR 55.46(b) Request for a Commission-Approved Simulation Facility
(This Enclosure consists of 6 pages, including this cover page)
ND-15-1333 Enclosure 1, Page 2 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility TABLE OF CONTENTS
1.0 Summary
2.0 Description of Simulator Discrepancies
2.1 Cumulative Effect of Simulator Discrepancies on Operator Performance
2.2 Evaluation of AP1000 Simulation Facility Unresolved Items (UIs) Issued by the NRC
2.3 Simulator Conformance with the AP1000 Plant Design
3.0 Variances between AP1000 Simulator Facilities
4.0 Conclusion
5.0 References
Special Note:
When referring to the “VEGP Units 3&4 Simulator Training System (STS)”, the word “simulator”
will be used throughout this and subsequent enclosures.
ND-15-1333 Enclosure 1, Page 3 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 1.0 Summary
Pursuant to 10 CFR 55.46(b), Southern Nuclear Operating Company (SNC) requests a
Commission-Approved Simulation (CAS) Facility for Vogtle Electric Generating Plants
(VEGP) Units 3 & 4 for the administration of operating tests under 10 CFR 55.45(b)(1).
This document and the related enclosures provide the information required by 10 CFR
55.46(b) for facility licensees that propose use of a simulation facility other than a plant-
referenced simulator in the administration of operating tests under 10 CFR 55.45(b)(1).
10 CFR 55.46(b)(1) states:
Facility licensees that propose to use a simulation facility, other than a plant-referenced
simulator, or the plant in the administration of the operating test under §§ 55.45(b)(1) or
55.45(b)(3), shall request approval from the Commission. This request must include:
(i) A description of the components of the simulation facility intended to be used, or the
way the plant would be used for each part of the operating test, unless previously
approved; and
(ii) A description of the performance tests for the simulation facility as part of the
request, and the results of these tests; and
(iii) A description of the procedures for maintaining examination and test integrity
consistent with the requirements of § 55.49.
Enclosure 2 contains a description of the components of the simulation facility per
paragraph (i) above.
Enclosure 3 contains a description of the performance tests for the simulation facility and
the results of those tests per paragraph (ii).
Enclosure 4 contains a description of the procedures for maintaining examination and test
integrity per paragraph (iii).
10 CFR 55.46(b)(2) states:
The Commission will approve a simulation facility or use of the plant for administration of
operating tests if it finds that the simulation facility and its proposed use, or the proposed
use of the plant, are suitable for the conduct of operating tests for the facility licensee's
reference plant under § 55.45(a).
Southern Nuclear Operating Company commissioned a team to evaluate the known
discrepancies in the simulator to determine if the 13 criteria established in 10 CFR
55.45(a), “Operating Tests,” would be challenged. The team was comprised of
representatives from SNC (Operations, Training and Engineering), SCANA (Training) and
Westinghouse (Human Factors Engineering).
ND-15-1333 Enclosure 1, Page 4 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
The team examined all Discrepancy Reports (DRs) that were open as of May 15, 2015,
and determined that 101 DRs were relevant to acceptability of one or more of the first nine
(9) criteria of 10 CFR 55.45(a). No DRs were found to be relevant to the last four criteria;
55.45(a)(10) through 55.45(a)(13). The team also determined that no singular DR posed a
challenge to the suitability of the simulation facility for the conduct of operating tests;
however, when considered in the aggregate, 42 of the DRs challenged criterion (3) and (5)
of 10 CFR 55.45(a) (See section 2.1 below for additional details).
In order to ensure the simulator is suitable for the conduct of operating tests, corrective
actions were initiated to resolve the subject 42 DRs. This assessment was communicated
to Westinghouse Electric Company (WEC) and WEC committed to implement
improvements aimed at resolving these issues in a patch deliverable to SNC by August
14, 2015. Based on this commitment, the CAS Aggregate Study Team reconvened on
July 7, 2015 and determined that the proposed changes would be adequate so that the
aggregate impact of the remaining discrepancies would not pose a challenge to any of the
10 CFR 55.45(a) criteria.
On August 14, 2015, WEC delivered a patch to SNC which contained corrections for the
42 items previously identified along with some additional corrections. After performing
Verification and Validation (V&V), 11 were determined to require further
investigation. After confirming the corrections that successfully passed the V&V process,
the CAS Aggregate Study Team reconvened on September 1, 2015, to review the impact
of the 11 outstanding items. The Aggregate Study Team determined that, in aggregate,
the impact of the 11 outstanding items, combined with the improvements in the area of
Alarm Response and the other remaining open items, would not impact the suitability of
the simulator for the conduct of operating tests.
Enclosure 6 contains the Aggregate Study mentioned above. Enclosure 7 contains a list
of the items WEC corrected. Enclosure 9 contains a list of the open DRs as of May 15,
2015.
2.0 Description of Simulator Discrepancies
Simulator discrepancies identified by SNC, other domestic AP1000 simulator owners and
those discrepancies that were issued as Unresolved Issues (UIs) by the NRC were
evaluated for applicability to SNC’s simulator. Discrepancies that were determined to be
applicable were entered into SNC’s Configuration Management System (CMS) Mantis
database as Simulator Change Requests (SCRs). If the SCR could be corrected by SNC,
it was corrected. Those SCRs that could not be corrected were evaluated by performing a
Training Needs Assessment to determine the impact on training. If the Training Needs
Assessment determined that there was an impact on training, a Training Needs Analysis
was performed to determine the extent of effect and to develop mitigation under the
Systematic Approach to Training (SAT) process. If the Training Needs Assessment
determined that there was no impact on training, then the discrepancy was entered into
the global tracking book as a historical record.
ND-15-1333 Enclosure 1, Page 5 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
2.1 Cumulative Effect of Simulator Discrepancies on Operator Performance
As stated in Section 1.0 above, SNC commissioned a team to determine if, in
aggregate, open discrepancies would present a challenge to the simulation facility’s
suitability for the conduct of operating tests. The results of this study are
documented in the report “Commission Approved Simulator Aggregate Study -
Simulator Training System Deficiency Impact on 10 CFR 55.45” (Enclosure 6).
Initially, the team determined that the aggregate of the open discrepancies would
challenge the ability of licensed operators to respond to simulator scenarios in
normal, off-normal, and emergency conditions based on criterion (3) and (5) of 10
CFR 55.45(a). The team analyzed the items that challenged these two criteria and
determined that the suitability of the simulator for the conduct of operating tests
would not be challenged if 42 of the discrepancies were corrected. SNC requested
Westinghouse Electric Company (WEC) to correct these discrepancies.
Based on WEC’s commitment to correct these items, the CAS Aggregate Study
Team reconvened on July 7, 2015 to determine if the remaining DRs would still
present a challenge to SNC’s ability to conduct an operating examination in
accordance with 10 CFR 55.45. The team concluded that the aggregate impact of
the remaining items would not pose a challenge to any of the 10 CFR 55.45(a)
criteria.
2.2 Evaluation of AP1000 Simulation Facility UIs Issued By the NRC
SNC performed a review of AP1000 simulation facility UIs issued by the NRC
(References 1 and 2). UIs were screened for applicability to the VEGP 3&4
simulation facility. Applicable UIs were entered into SNC’s CMS using NMP-TR-422,
“Simulator Configuration Control Procedure.”
Enclosure 5 contains the results of SNC’s evaluation of AP1000 Simulation Facility
UIs issued by the NRC.
2.3 Simulator Conformance with the AP1000 Plant Design
SNC accepted turnover of the VEGP Units 3&4 Simulator Training System (STS)
from Westinghouse, as described by letter (Reference 3) on December 30, 2014.
Subsequent to the STS turnover, SNC identified simulator discrepancies that were
determined to be plant design issues. Westinghouse retains design authority of the
plant configuration until Unit 3 turnover. Simulator discrepancies that are determined
to be plant design issues prior to Unit 3 turnover will be tracked until the AP1000
plant design changes have been approved. Approved design changes will be
incorporated into the simulation facility in accordance with SNC simulator fidelity and
configuration management programs per 10 CFR 55.46(d).
ND-15-1333 Enclosure 1, Page 6 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 3.0 Variances between AP1000 Simulator Facilities
It has been noted that variances exist between the various AP1000 simulators.
As described in this letter and Enclosures, SNC reviews simulator discrepancy reports
from other licensees and the vendor. The item is entered into the Configuration
Management System (CMS) Mantis database and tracked to resolution.
For example, if SNC identifies an issue or is able to duplicate an issue identified on one of
the other AP1000 simulators, SNC will generate an SCR. If it is within SNC’s capability to
do so, SNC will develop and implement a resolution to address the issue. SNC shares
these solutions with WEC. WEC, at its discretion, may immediately distribute the solution
with other AP1000 simulator owners or wait to incorporate the correction as part of a future
software update.
4.0 Conclusion
SNC evaluated all open simulator discrepancies that existed through May 15, 2015 and
corrected discrepancies that, in aggregate, could impact the suitability of the simulators for
the conduct of operating tests. SNC has determined that there is no open simulator
discrepancy, individually or in aggregate, that would challenge the ability of licensed
operators to respond to simulator scenarios in normal, off-normal, or emergency
conditions.
The material SNC is presenting is required for a Commission-Approved Simulation Facility
as defined in 10 CFR 55.46(b). SNC is requesting the Commission’s approval of the
VEGP Units 3&4 simulation facility for use in the administration and conduct of operating
tests in accordance with 10 CFR 55.46(b)(2).
5.0 References
1. NRC Email dated 2015-05-13, Meeting Materials for May 14, 2015- VCSNS 2 and 3
Commission-Approved Simulator - CAS-Summer-RAI 5-7-15_b Redacted,
ML#15133A497
2. NRC Letter dated 2015-07-02, Virgil C. Summer Nuclear Station Units 2 and 3 - Request
For A Commission Approved Simulation Facility - ML15182A097
3. SNC Letter dated 2014-12-30, Vogtle Electric Generating Plant, Units 3 & 4 - Response to
SVP_SVO_002964 and Simulator Training System (STS) Acceptance
Southern Nuclear Operating Company
Vogtle Electric Generating Plant (VEGP) Units 3 and 4
ND-15-1333
Enclosure 2
Description of the Components of the Simulation Facility Intended to be Used for Each Part of the Operating Test - 10 CFR 55.46(b)(1)(i)
(This Enclosure consists of 5 pages, including this cover page)
ND-15-1333 Enclosure 2, Page 2 of 5 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
1.0 Summary Description
The Vogtle simulation facility is comprised of two AP1000 full scope simulators,
designated “3A” and “3B.” Both simulators are referenced to Vogtle Unit 3 and are
intended to be maintained functionally identical. The simulators are licensed to conform to
the requirements of ANSI/ANS-3.5-1998, “Nuclear Power Plant Simulation Facilities for
Use in Operator Training and License Examination,” as endorsed by Revision 3 of NRC
Regulatory Guide 1.149, “Nuclear Power Plant Simulation Facilities for Use in Operator
Training and License Examinations.”
2.0 Functional Description
Instructor-controlled normal plant evolutions, system malfunctions, Component Level
Failures (CLFs) and Local Operator Actions (LOAs) are used to provide simulated plant
performance and failure or degradation of simulated plant systems or equipment. To
achieve this level of functionality, plant systems listed in Table E2-1 are simulated.
3.0 Detailed Description of the VEGP Simulators
The simulation facility design, models, and software are based upon the Westinghouse
“Baseline 7” milestone for Instrumentation and Controls (I&C) design. The Baseline 7
milestone document established a set of requirements to ensure the integrated I&C
system design is consistently implemented within various core I&C platforms and systems.
The Vogtle simulation facility has also been updated with various modifications, in
coordination with Westinghouse as new I&C issues or design changes have been
identified.
The Vogtle Unit 3A and 3B simulators are referenced to Unit 3. Unit 3 is approximately
one year ahead of construction for Unit 4. The only meaningful difference noted between
Unit 3 and Unit 4 design documentation at this time is the switchyard. Unit 3 is tied to the
230kV switchyards. This is shown in the ZBS Ovation screens in both simulators. Unit 4
will be tied to the 500kV switchyard. Presently, this is the only identified difference in Unit
3 and Unit 4.
The VEPG Unit 3 Simulator Training System is tested to the AP1000 design. Simulator
fidelity is maintained in accordance with 10 CFR 55.46(d) as documented by the NRC
(Reference 1). Any significant outstanding discrepancies will be resolved during the
finalization of the AP1000 design and/or initial test program. SNC continues to update the
simulator with corrections to minor programming discrepancies as they are identified and
information associated with these updates is forwarded to Westinghouse for inclusion in
the next baseline update or patch as appropriate.
ND-15-1333 Enclosure 2, Page 3 of 5 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 4.0 Detailed Description of Simulated Systems
The systems listed in Table E2-1 are simulated. A detailed description of each of these
systems can be found in the VEGP 3&4 Updated Final Safety Analysis (UFSAR), Rev. 4.0.
The systems listed in Table E2-2 are systems that are listed in the UFSAR, but are not
modeled for the reasons stated in the table.
Table E2-1 List of Plant Systems Simulated
System Code
System Title
ASS Auxiliary Steam Supply System
BDS Steam Generator Blowdown System
CAS Compressed and Instrument Air Systems
CCS Component Cooling Water System
CDS Condensate System
CES Condenser Tube Cleaning System
CFS Turbine Island Chemical Feed System
CMS Condenser Air Removal System
CNS Containment System
CPS Condensate Polishing System
CVS Chemical and Volume Control System
CWS Circulating Water System
DAS Diverse Actuation System
DDS Data Display and Processing System
DOS Standby Diesel Fuel Oil System
DTS Demineralized Water Treatment System
DWS Demineralized Water Transfer and Storage System
ECS Main AC Power System
EDS Non Class 1E DC and UPS System
EHS Special Process Heat Tracing System
ELS Plant Lighting System
FPS Fire Protection System
FWS Main and Startup Feedwater System
GSS Gland Seal System
HCS Generator Hydrogen and CO2 Systems
HDS Heater Drain System
HSS Hydrogen Seal Oil System
IDS Class 1E DC and UPS System
IIS Incore Instrumentation System
LOS Main Turbine and Generator Lube Oil System
MES Meteorological and Environmental Monitoring System
MHS Mechanical Handling System
MSS Main Steam System
MTS Main Turbine System
OCS Operation and Control Centers System
PCS Passive Containment Cooling System
PGS Plant Gas Systems
PLS Plant Control System
PMS Protection and Safety Monitoring System
PSS Primary Sampling System
ND-15-1333 Enclosure 2, Page 4 of 5 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E2-1 (continued)
System Code
System Title
PWS Potable Water System
PXS Passive Core Cooling System
RCS Reactor Coolant System
RMS Radiation Monitoring System
RNS Normal Residual Heat Removal System
RWS Raw Water System
RXS Reactor System
SDS Sanitary Drainage System
SFS Spent Fuel Pool Cooling System
SGS Steam Generator System
SJS Seismic Monitoring System
SMS Special Monitoring System
SSS Secondary Sampling System
SWS Service Water System
TCS Turbine Building Closed Cooling Water System
TDS Turbine Island Vents, Drains and Relief System
TOS Main Turbine Control and Diagnostics System
VAS Radiologically Controlled Area Ventilation System
VBS Nuclear Island Nonradioactive Ventilation System
VCS Containment Recirculation Cooling System
VES Main Control Room Emergency Habitability System
VFS Containment Air Filtration System
VHS Health Physics and Hot Machine Shop HVAC System
VLS Containment Hydrogen Control System
VRS Radwaste Building HVAC System
VTS Turbine Building Ventilation System
VUS Containment Leak Rate Test System
VWS Central Chilled Water System
VXS Annex/Aux Building Nonradioactive Ventilation System
VYS Hot Water Heating System
VZS Diesel Generator Building Heating and Ventilation System
WGS Gaseous Radwaste System
WLS Liquid Radwaste System
WRS Radioactive Waste Drain System
WSS Solid Radwaste System
WWS Waste Water System
ZAS Main Generation System
ZBS Transmission Switchyard and Offsite Power System
ZOS Onsite Standby Power System
ZRS Offsite Retail Power System
ZVS Excitation and Voltage Regulation System
ND-15-1333 Enclosure 2, Page 5 of 5 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E2-2 List of Plant Systems NOT Simulated
System Code
System Title Not Simulated Because . . .
DFS Diesel Fuel Offloading System No control room interface
DRS Storm Drain System No control room interface
EFS Communication Systems
The simulator does not model the EFS networking scheme but does provide similar functions for communication systems. The simulator mimics the plant communication systems with a Private Branch Exchange (PBX) phone system for the Training Center.
EGS Grounding and Lightning Protection System No control room interface
EQS Cathodic Protection System No control room interface
FHS Fuel Handling and Refueling System No control room interface
NCS Network Connection System No control room interface
OWS Offsite Water Treatment System No control room interface
RDS Gravity and Roof Drain Collection System No control room interface
RLS Radiochemistry Laboratory System No control room interface
SES Plant Security System No control room interface
TVS Closed Circuit TV System No control room interface
VDS Demineralized Water Treatment Building HVAC System
No control room interface
VGS Auxiliary Boiler Building Ventilation System No control room interface
VIS Transmission Switchyard Ventilation System No control room interface
VNS Switchyard Control Building HVAC System No control room interface
VPS Pump House Building Ventilation System No control room interface
VQS Chlorination Workshop HVAC System No control room interface
VVS Waste Water Treatment Plant Ventilation System
No control room interface
YFS Yard Fire Water System No control room interface
ZFS Offsite Communications System
The simulator does not model the ZFS but does provide similar functions for communication systems. The simulator mimics the plant and offsite communication systems with a PBX for the Training Center.
5.0 References
1. Vogtle Electric Generating Plant Units 3 and 4 - NRC Simulator Inspection Reports
05200025/2015301 and 05200026/2015301, dated April 21, 2015 - ML15113A028
Southern Nuclear Operating Company
Vogtle Electric Generating Plant (VEGP) Units 3 and 4
ND-15-1333
Enclosure 3
Description of the Performance Tests for the Simulation Facility and Results of the Tests - 10 CFR 55.46(b)(1)(ii)
(This Enclosure consists of 6 pages, including this cover page)
ND-15-1333 Enclosure 3, Page 2 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 1.0 Summary Description of the Performance Tests for the Simulation Facility and
Results of the Tests
Performance tests were conducted on site, in addition to the earlier factory acceptance
tests performed by the vendor at the vendor’s facility, in order to demonstrate simulator
fidelity. The performance testing concluded that simulator performance met the
requirements of ANS/ANSI-3.5 testing for the current simulator design. On April 8, 2015,
the NRC completed an inspection of the VEGP Units 3&4 simulation facilities to “ensure
that the Vogtle 3A and 3B simulation facilities were being tested in accordance with
ANSI/ANS-3.5-1998, ‘Nuclear Power Plant Simulators for Use in Operator Training
Examination,’” (Reference 1) with no findings of significance. The following is a summary
of the simulator performance test licensing basis, a description of the performance tests
conducted and the test results.
2.0 Detailed Description of the Performance Tests for the Simulation Facility and
Results of the Tests
2.1 Vendor and Other Testing
The VEGP Units 3&4 Simulator Training System was developed by WEC and turned
over to SNC on December 30, 2014. In parallel, WEC continued design finalization
activities which included Human Factors Engineering Validation tasks such as
Integrated System Validation (ISV). The ISV used and exercised the simulator
extensively. The ISV shakedown, pilot and final ISV testing identified a number of
integration issues. The NRC had requested licensees to report and assess these
issues for NRC consideration. Some of the issues reported by other licensees were
not able to be duplicated in the performance tests on the VEGP Units 3&4 STS.
Refer to Enclosure 1 Section 3.0 for a discussion on simulator variations.
A preliminary review of ISV testing identified twenty one potential Priority-1 Human
Engineering Discrepancies (HEDs). The NRC had requested licensees to report and
assess those potential HEDs. The assessment results, including resolutions, were
provided to the NRC. The NRC did not fully accept some of the resolutions and
indicated that they would need additional information (Reference 2). SNC reviewed
the information requested by the NRC and developed resolutions which are provided
in Enclosure 8.
2.2 Simulator Performance Testing Licensing Bases
Simulator licensing bases are described in Vogtle 3&4 UFSAR Chapter 1, Appendix
1A. Per the UFSAR, Vogtle conforms to section C.1 of Regulatory Guide 1.149,
Revision 3 “Nuclear Power Plant Simulation Facilities for Use in Operator License
Examinations.” Operator Licensing examinations are conducted on a simulator
meeting the applicable requirements of ANSI/ANS-3.5-1998.” This Regulatory Guide
endorses ANSI/ANS 3.5-1998 “Nuclear Power Plant Simulation Facilities for Use in
ND-15-1333 Enclosure 3, Page 3 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Operator License Examinations.” Section 4.4.3 of the ANSI/ANS 3.5 describes the
simulator performance testing.
2.3 VEGP Units 3&4 Simulator Performance Testing Description
Simulator performance testing is made up of operability testing and scenario-based
testing. Simulator performance testing is performed in a fully integrated mode of
operation. The test procedures are documented in Vogtle procedures NMP-TR-422-
006, “Plant Vogtle 3-4 Simulator Testing Instruction,” and NMP-TR-422-006-001,
“Simulator Configuration and Performance Criteria Instruction.” These tests are
based on the ANS-3.5-1998 standard. The test cases included:
1. Simulator Operability Testing - Simulator Operability Testing is conducted to confirm
overall simulator model completeness and integration. Operability testing:
Is intended to demonstrate overall simulator model completeness and
integration
Includes simulator transient performance for a benchmark set of transients as
shown below
Item Title Test Type
1 Manual Reactor Trip Transient
2 Simultaneous trip of Main Feedwater Pumps Transient
3 Simultaneous Trip of all Feedwater Pumps Transient
4 Simultaneous closure of All Main Steam Isolation Valves Transient
5 Simultaneous trip of All Reactor Coolant Pumps Transient
6 Single Reactor Coolant Pump Trip Transient
7 Main Turbine Trip Without a Reactor Trip Transient
8 Maximum Rate Power Ramp Transient
9 Maximum Size Reactor Coolant System Rupture with Loss of Offsite Power
Transient
10 Maximum Size Unisolable Main Steam Line Rupture Transient
11 Slow Primary System Depressurization to Saturated Condition (Pzr Safety)
Transient
12 Slow Primary System Depressurization to Saturated Condition (ADS) Transient
13 Maximum Design Load Rejection Transient
Includes real time and repeatability tests
Item Title Test Type
14 Computer Real Time Test Real Time
15 Simulated Limits Exceeded Test Real Time
16 Repeatability Test Repeatability
Includes simulator steady-state performance tests
Item Title Test Type
17 Steady State Performance at 50% Power Steady State
18 Steady State Performance at 75% Power Steady State
19 Steady State Performance at 100% Power Steady State
ND-15-1333 Enclosure 3, Page 4 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Includes the following normal evolutions
Item Title Test Type
20 Plant Startup from Cold to Hot Standby Normal Evolution
21 Nuclear Startup from Hot Standby to Rated Power Normal Evolution
22 Reactor Trip with Recovery to Rated Power Normal Evolution
23 Plant Shutdown from Rated Power to Cold Shutdown Normal Evolution
24 Surveillance Testing Normal Evolution
Includes the following core tests
Item Title Test Type
25 Shutdown Margin Determination Core
26 Core Reactivity coefficients Test Core
27 Isothermal Temperature Coefficient Core
28 Control Rod Worth Test Core
2. Malfunction testing is conducted on an as needed basis. Malfunction testing was
conducted to gather base line data.
Item Title Test Type
29 Steam Generator Tube Rupture Malfunction
30 Loss of Coolant Outside Containment Malfunction
31 Large Break Loss of Coolant Accident Malfunction
32 Small Break Loss of Coolant Accident Malfunction
33 Loss of Instrument Air Malfunction
34 Loss of IDS Division A Instrument Busses Malfunction
35 Loss of IDS Division B Instrument Busses Malfunction
36 Loss of IDS Division C Instrument Busses Malfunction
37 Loss of IDS Division D Instrument Busses Malfunction
38 Loss of Offsite Power with Loss of Diesel Generators Malfunction
39 Loss of Electrical Distribution Bus ES-1 Malfunction
40 Loss of Electrical Distribution Bus ES-2 Malfunction
41 Loss of Electrical Distribution Bus ES-3 Malfunction
42 Loss of Electrical Distribution Bus ES-4 Malfunction
43 Loss of Electrical Distribution Bus ES-5 Malfunction
44 Loss of Electrical Distribution Bus ES-6 Malfunction
45 Loss of Electrical Distribution Bus EK-11 Malfunction
46 Loss of Electrical Distribution Bus EK-12 Malfunction
47 Loss of Electrical Distribution Bus EK-13 Malfunction
48 Loss of Electrical Distribution Bus EK-14 Malfunction
49 Loss of Electrical Distribution Bus EK-21 Malfunction
50 Loss of Electrical Distribution Bus EK-22 Malfunction
51 Loss of Electrical Distribution Bus EK-23 Malfunction
52 Loss of Electrical Distribution Bus EK-24 Malfunction
53 Loss of Electrical Distribution Bus EK-31 Malfunction
54 Loss of Electrical Distribution Bus EK-41 Malfunction
55 Loss of EDS Instrument Buses Malfunction
56 Loss of IDS Division B and C 72 Hour Instrument Busses Malfunction
57 Loss of Electrical Distribution Bus ES-7 Malfunction
58 Loss of Condenser Vacuum Malfunction
59 Loss of Condenser Level Control Malfunction
60 Loss of Service Water Malfunction
61 Loss of Shutdown Cooling Malfunction
ND-15-1333 Enclosure 3, Page 5 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Item Title Test Type
62 Loss of Component Cooling Water Malfunction
63 Loss of Normal Feedwater Malfunction
64 Loss of All Heat Sinks Malfunction
65 Loss of Division A PMS Malfunction
66 Loss of Division B PMS Malfunction
67 Loss of Division C PMS Malfunction
68 Loss of Division D PMS Malfunction
69 Rod F06 Stuck Malfunction
70 Rod B06 Uncouples Malfunction
71 Rod G07 Drops Malfunction
72 Misaligned Rod Malfunction
73 Inability to Drive Rods Malfunction
74 Fuel Clad Failure Malfunction
75 Main Generator Trip Malfunction
76 Inadvertent Operation of Core Makeup Tanks at Power Malfunction
77 Inadvertent Actuation of Passive Residual Heat Exchanger at Power Malfunction
78 Increase in RCS Inventory Malfunction
79 Failure of Pressurizer Pressure Control Malfunction
80 Main Steam Line Break Inside Containment Malfunction
81 Main Steam Line Break Outside Containment Malfunction
82 Main Feed Line Break Inside Containment Malfunction
83 Main Feed Line Break Outside Containment Malfunction
84 Failure of Power Range Nuclear Instrument Malfunction
85 Failure of Intermediate Range Nuclear Instrument Malfunction
86 Failure of Source Range Nuclear Instrument Malfunction
87 Failure of the Alarm Presentation System Malfunction
88 Anticipated Transient without SCRAM without DAS Malfunction
89 Anticipated Transient without SCRAM with DAS Malfunction
3. Simulator Scenario-Based Testing (SBT) – The VEGP Units 3&4 simulator facility is
committed to the SBT methodology described in the 1998 ANS-3.5 standard as
endorsed by Reg. Guide 1.149 Rev. 3 and in NEI 09-09. SBT is the parallel testing
and evaluation of simulator performance while instructors validate NRC Initial license
examination scenarios, licensed operator requalification annual examination
scenarios, and scenarios used to satisfy the reactivity control manipulation
requirements for license candidates in 10 CFR 55.31 (a)(5). As instructors validate
satisfactory completion of training or evaluation objectives, procedure steps and
scenario content, they are also ensuring satisfactory simulator performance in
parallel, not series, making the process an “online” method of evaluating simulator
performance. SBT is conducted to ensure the simulator was capable of producing
the expected “reference unit” response to satisfy predetermined learning or
examination objectives by utilizing the existing training and examination scenario
validation process. The term, “reference unit,” as used above, refers to the AP1000
plant design since both of the VEGP units are still under construction.
ND-15-1333 Enclosure 3, Page 6 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
2.4 Simulator Operability Test Results
Operability tests were completed satisfactorily in 2014 with one test (Reactor Trip
Recovery) deviation. A Condition Report was initiated to document the deviation.
The performance tests resulted in the documentation of SCRs. These discrepancies
were captured in the final test report and the Simulator CMS SCR database.
On April 8, 2015, the NRC completed an inspection that included a selection of
simulator test procedures and test records. Based on the results of that inspection,
no findings of significance were identified (Reference 1).
3.0 Maintenance of Simulator Fidelity
As mentioned in Enclosure 2, the simulators are maintained in conformance with the
requirements of ANSI/ANS-3.5-1998, “Nuclear power Plant Simulation Facilities for Use in
Operator Training and License Examination,” as endorsed by Revision 3 of NRC
Regulatory Guide 1.149, “Nuclear Power Plant Simulation Facilities for Use in Operator
Training and License Examinations.”
The NRC performed an inspection of the VEGP Units 3&4 simulation facility on April 8,
2015. The inspection included a review of SNC’s programs and processes related to
continued assurance of simulator fidelity in accordance with 10 CFR 55.46(d). The
inspection yielded no findings of significance and determined that SNC’s programs to
assure continued simulator fidelity were adequate (Reference 1).
4.0 Summary Conclusion
Simulator operability tests and simulator scenario-based tests were conducted and
completed with no major differences identified between the AP1000 plant design and the
simulator. As a result, SNC believes NRC examiners should be able to make pass-fail
judgments with confidence as required by Reg. Guide 1.149.
5.0 References
1. Vogtle Electric Generating Plant Units 3 and 4 - NRC Simulator Inspection Reports
05200025/2015301 and 05200026/2015301, dated April 21, 2015 - ML15113A028
2. Virgil C. Summer Nuclear Station Units 2 and 3 - Request for a Commission-Approved
Simulation Facility, Dated July 2, 2015 - ML15182A097
Southern Nuclear Operating Company
Vogtle Electric Generating Plant (VEGP) Units 3 and 4
ND-15-1333
Enclosure 4
Description of the Procedures for Maintaining Examination and Test Integrity Consistent with the Requirements of 10 CFR 55.49 - 10 CFR 55.46(b)(1)(iii)
(This Enclosure consists of 2 pages, including this cover page)
ND-15-1333 Enclosure 4, Page 2 of 2 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
1.0 Summary Description of the Procedures for Maintaining Examination and Test
Integrity Consistent with the Requirements of 10 CFR 55.49
Security for examination development and implementation is accomplished with NMP-TR-423, “Regulatory Exam Development.” This procedure conforms to the requirements of NUREG-1021, “Operator Licensing Examination Standards for Power Reactors,” which is founded in the requirements of 10 CFR 55.49. The procedure includes:
Door Security Access Control;
Encryption of Initial Condition (IC) sets and Application (APP) and Trigger files;
Disabling of Video Recording Equipment; and,
Physical Security of Examination Material.
Southern Nuclear Operating Company
Vogtle Electric Generating Plant (VEGP) Units 3 and 4
ND-15-1333
Enclosure 5
Evaluation of AP1000 Simulation Facility Summary of Unresolved Items (UIs) Issued By the NRC
Redacted (Non-Proprietary)
(This Enclosure consists of 19 pages, including this cover page)
ND-15-1333 Enclosure 5, Page 2 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 1.0 Summary Evaluation of AP1000 Simulation Facility Unresolved Items (UIs) Issued
by the NRC
SNC performed a review of AP1000 simulation facility UIs issued by the NRC (References
1 and 2). UIs were screened for applicability to the VEGP 3&4 simulation facility.
Applicable UIs were entered into SNC’s CMS using NMP-TR-422, “Simulator
Configuration Control Procedure.”
2.0 Detailed Description of SNC’s Process
Simulator discrepancies are captured by the SNC Simulator Change Request
process. The main method SCRs are identified is through direct simulator response
during training, validation, or performance testing. Noticeable differences and
discrepancies in expected simulator response are entered into the SCR CMS database
(Mantis). The Mantis system is used for issue reporting, change management,
tracking/querying issues, software change documentation, hardware change
documentation, and other simulator related administrative issues. Alterations to the
simulator models, simulated I&C, and Design Change Package (DCP) implementation are
all documented via Mantis. SNC also receives simulator discrepancies from VC Summer
and Westinghouse. These issues are examined for applicability to the Vogtle 3 STS and
processed through the SNC simulator configuration management process where
applicable.
During analysis of reported simulator discrepancies, design documents are reviewed in
order to determine the response that should be expected from SNC’s simulator. Design
documentation is the main method of analyzing appropriate response due to the lack of an
operating reference unit.
Westinghouse is informed of any I&C or design issue via the SNC corrective action
process. Model issues are corrected by SNC where feasible and appropriate via the SNC
Fleet SCR process. Vendor issued fixes are also processed through SNC’s corrective
action process. All simulator changes are tested in accordance to the ANSI/ANS-3.5.1998
standard. Verification and Validation are performed to examine the effectiveness of the
simulator repair.
The Simulator Review Committee (SRC) reviews SCR disposition at least quarterly. The
SRC also determines if uncorrected simulator discrepancies introduce negative training to
the licensed operator curriculum. Issues which do not introduce negative training and
remain uncorrected for any given length of time are presented to the students at the
beginning of each segment/class. Issues having the potential to introduce negative
training undergo a Training Needs Analysis to exercise the SAT process to prevent
negative training.
ND-15-1333 Enclosure 5, Page 3 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 3.0 Nuances Related to the Resolutions in Table E5 and the Aggregate Study
There are numerous times within the Aggregate Study (Enclosure 6), where it states, “This
issue was dispositioned as unacceptable by the Simulator Review Committee (SRC),” yet
the issue was ultimately dispositioned as acceptable in Table E5-1. The review of these
items by the aggregate study team focused on the impact the lack of designed protective
functions (UI #26) and the invalid indications (UI #15) had on an operator when combined
with all other issues under review. The SRC was focused solely on the impact of the
individual items. However, an understanding of the SRC’s dispositioning process and its
use of the terms “acceptable deviation” and “unacceptable deviation” is beneficial.
Acceptable Deviation - After an issue is entered as an SCR in Mantis and proved valid via
investigation, the issue is evaluated by the SRC, supplemented by Subject Matter
Experts. If the issue screens as an acceptable deviation by the SRC, using the guidance
provided in section 4.2.1.4 of ANSI/ANS-3.5-1998, the item remains in Mantis as a
historical record for reference. A running log of these acceptable deviations is available to
the Operations Instructional staff for reference and is provided to licensed operator
candidates at the beginning of simulator training as a reference.
Unacceptable Deviation - If the issue screens as an unacceptable deviation by the SRC,
then a detailed evaluation of the deviation is conducted via a Training Needs Analysis
(TNA). The TNA will determine if any compensatory actions can be taken to mitigate the
deviation’s impact to students until a software or hardware solution is implemented. If
compensatory actions can be taken or the issue is corrected, the deviation status
becomes acceptable.
4.0 Simulator Review Committee (SRC)
The Simulator Review Committee (SRC) is composed of one member of Operations,
selected by the Operations Director; the Operations Training Manager or designee; and
the Simulator Coordinator or designee.
The SRC is supplemented by additional personnel, as necessary, to serve as subject
matter experts to conduct a Training Needs Assessment (an appraisal by a subject matter
expert of a simulator deviation, deficiency, or modification, and its relative importance to
the operator as required tasks are performed). Additional members of the SRC are
Operations individuals who have completed AP1000 certification training.
Representatives from Engineering attend the SRC when required to discuss plant design
changes and their impact on the simulator.
ND-15-1333 Enclosure 5, Page 4 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E5-1
SNC Evaluation of NRC UIs and Cross-Reference List
NRC
# {*} NRC UI Description
{*}
Ref
# {*}
SNC SCR
#
SNC Evaluation
1
Subcriticality Critical Safety Function (CSF) alarm block is turning magenta (bad input) intermittently
TO-40 5627
Closed. Fixed with patch WEC provided to SNC on August 14, 2015.
V&V testing was performed successfully.
For V&V, two different tests were conducted under simulator conditions replicating those under which the issue was first identified. The first test was a steady state 100% power run for one hour with no operator action. The second test was conducted for an RCS leak. During both tests, the Mode 1/2 Critical Safety Function status wall panel display was observed and [ ]a,c.
2 Control rods rejecting to manual {Rod Control Urgent Alarm}
TO-45 5808
Closed. Fixed with Patch Version 1.0.1.
For Validation and Verification (V&V), a loss of ES-1 for six different plant conditions was conducted. During each tested condition, rod control remained in automatic.
SNC has not seen this since the patch was installed.
3
Wall Panel Information System (WPIS) is cycling between different displays
{Mode 2 is procedurally called when all AO bank rods are off the bottom. Currently, PLS Auto Plant Mode selector changes from Mode 3 to Mode 2 when the RTBs are closed (P-3 is cleared).}
TO-52 6144
Closed. Fixed with patch WEC provided to SNC on August 14, 2015.
V&V testing was performed successfully.
For V&V, a General Operating Procedure startup was conducted. Simulator conditions under which the issue was first identified were replicated as nearly as possible. When all banks of AO rods indicated 1 step, it was verified that the auto plant mode selector NAP changed to Mode 2. This test was repeated 3 times with the same results.
4 Feedflow oscillations TO-54 and 58
6151
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
The Aggregate Study and the SRC dispositioned this issue as acceptable.
These feedwater oscillations are associated with Startup Feedwater during shutdown conditions with little steam and feed demand.
This is in accordance with the current AP1000 plant design. Engineering calculations of expected flow characteristics were used to establish the initial controller tuning values. The Simulator controller tuning values have been established at these calculated values.
ND-15-1333 Enclosure 5, Page 5 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E5-1 (continued)
NRC
# {*} NRC UI Description
{*}
Ref
# {*}
SNC SCR
#
SNC Evaluation
During shutdown conditions with low steam and feed demand, Start-up Feedwater Control Valves (SFCVs) are in automatic. Under these conditions, the SFCVs cycle over the entire operating range in short periods of time resulting in start-up feedwater oscillations even though Steam Generator Water Level (SGWL) remains steady in the program band with no noticeable perturbations. If an operator deems it necessary, current operational procedures do allow placing the SFCVs in MANUAL and maintaining SGWL in that mode of operation. At some point, operators would contact maintenance to troubleshoot and repair the cause of the oscillations.
Since the system is performing its design function of maintaining SGWL, SNC has determined that this issue does not impact the suitability of the simulator for the conduct of operating tests.
SNC does consider cycling of the SFCVs and the resulting feedflow oscillations as an undesirable condition for an operating plant. WEC is aware of this issue and plans are in place to obtain more precise tuning data during hot functional testing. Hot functional testing will provide as-built flow characteristics and more accurate controller tuning values. Once hot functional testing has been completed, the expectation is that SGWL will continue to be maintained in the required operating band with the SFCVs in AUTO and with no observable feedflow oscillations.
5 Unexpected hotwell low level during trip recovery
TO-89 5987
Closed. Fixed with patch, Version 1.0.9.
V&V testing was performed successfully.
For V&V, hotwell makeup valves were fully opened at 100% power. Simulator conditions under which the issue was first identified were replicated as nearly as possible. Makeup flow was verified to be greater than the required value per AP1000 design documents.
6 Modeling of baseline vs design certification configuration
TO-96 None
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This issue relates to how a secondary trip function is administered during training scenarios. During the evaluation of operators, any scenario affecting the operation of the turbine control valves, the instructor will either have all automatic trips fail, requiring operators to take manual action to trip the turbine, OR insert a spurious trip of the turbine, requiring operators to take action based on the sudden loss of turbine load. In either case, the issue is transparent to the operators.
The Cause and Effect document for TOS02 on the Instructor Station references future design information. This is an administrative issue on the Instructor Station. The functionality can be updated by WEC when the appropriate design resolution is available. The current modeling for TOS02 does prevent an automatic turbine trip. When evaluating operators, SNC will have all automatic turbine trips fail so as to require operators to take manual actions in accordance with plant procedures.
7 Control rods reject to manual
TO-101 and 104
5659
Closed. Fixed With patch, Version 1.0.1.
V&V testing was performed successfully.
ND-15-1333 Enclosure 5, Page 6 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E5-1 (continued)
NRC
# {*} NRC UI Description
{*}
Ref
# {*}
SNC SCR
#
SNC Evaluation
SNC has not seen this since the patch was installed. For Validation and Verification (V&V), the simulator was placed in run for 14 hours during a steady state test created for ANSI testing and in that 14 hour period, the rods did not reject to manual. A manual turbine trip was inserted [
]a,c.
8
Moisture Separator Reheater (MSR) valve response is incorrect and causes a reactor coolant system (RCS) temperature transient
TO-128 5618
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This is not in accordance with the current AP1000 plant design. The implementation of the AP1000 design requirements into the I&C control scheme was incorrect. The simulator correctly models the installed I&C control scheme. This issue will be corrected when the implementation of the I&C control scheme is updated by WEC in accordance with SNC’s configuration management program.
[
]a,c.
[ ]a,c. This does not affect the performance of simulator
operations as the alignment of the steam dump control system, [ ]a,c, is procedurally controlled during the load reduction and the lineup is established prior to reaching these lower turbine load conditions.
The time period that the incorrect control signal is in effect will vary dependent upon the down-power rate. Throughout this control sequence, operators are performing other actions that are directed by the controlling procedure. The controlling procedure will direct the use of the SOP to verify the proper position of the valves. However, when this direction is provided the valves have already cycled to the closed position and are in the expected position when checked. SNC has developed an APP file to override the controller and close the valves at the required turbine load. This results in the controller error being transparent to the operator.
SNC will continue to use this APP file until a permanent correction is provided by WEC.
9
Aux steam pressure not meeting design requirements
{GSS Header pressure will not maintain pressure as required}
TO-131 5609
Closed. Fixed with patch, Version 1.0.10.
V&V testing was performed successfully.
For V&V, testing was performed at 100% power. Initial simulator conditions under which the issue was first identified were replicated as nearly as possible during V&V testing. Gland Sealing Steam header pressure maintained [ ]a,c psig under full power conditions.
This issue has not been noted on the simulator since the correction was implemented.
10 Unexpected Main Turbine System alarm
1411-03 5722 Closed. Fixed with patch WEC provided to SNC on August 14, 2015.
V&V testing was performed successfully.
ND-15-1333 Enclosure 5, Page 7 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E5-1 (continued)
NRC
# {*} NRC UI Description
{*}
Ref
# {*}
SNC SCR
#
SNC Evaluation
at power For V&V, turbine load was lowered [ ]a,c. Simulator conditions under which the issue was first identified were replicated as nearly as possible. A second test was then conducted where turbine load was lowered [ ]a,c. During both turbine load reductions, APS was monitored. The unexpected MTS alarm was not received during either test.
This issue has not been noted on the simulator since the correction was implemented.
11
Rod control urgent failure on loss of EK-12 appears inconsistently without loss of power
1501-08 6726
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
The plant response identified by this DR is per current design.
SNC has confirmed that the indications and automatic actions are consistent with expected plant response per current rod control software design.
Results of investigation/observations:
[ ]a,c. This cabinet provides the processing and amplification of half the self-powered flux
detector signals. These signals are provided to [ ]a,c When power
is lost to this cabinet it [ ]a,c. The inconsistent results are due to different initial conditions when the event occurs.
[
]a,c. The small variations in plant parameters that are part of normal fluctuations during steady state operations will ultimately change the response to this power loss.
The cause and effect in this situation is per current plant software design and therefore replicates expected actual plant response. Operators will take action accordingly if this were to happen in the plant and therefore are taking the same actions if this event occurs in the simulator. The expected operator action for automatic rod motion that is not expected or is occurring due to a detector failure is to take rod control to manual. This expectation is consistent with fleet expectations.
If/When WEC develops a plant design change, it will be applied per current configuration control management procedures.
12 Axial Offset (AO) rods move inconsistently between tests
1502-10 5585
5659
Closed. Fixed in patch, Version 7F8.1.0.1.
V&V testing was performed successfully.
These issues were corrected by RITS 41468 which was delivered in a patch from WEC. The patch was tested on the SDS under conditions that attempted to duplicate the conditions that existed at the time the issue was
ND-15-1333 Enclosure 5, Page 8 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E5-1 (continued)
NRC
# {*} NRC UI Description
{*}
Ref
# {*}
SNC SCR
#
SNC Evaluation
first identified. The patch was implemented in November of 2014. The patch resolved both issues and the SDRs were closed.
AO rods randomly rejecting to manual has not been observed since the implementation of this patch.
13 RCS wide range pressure dropped from 1400 to 700 psig
1503-03 and 04
6741
This issue was closed as invalid.
An investigation determined that this is the expected plant response to proper operation of the plant passive cooling capabilities.
This observation was made during a simulator scenario which was evaluating operator response to a Loss of Coolant Accident (LOCA). The dynamic conditions of the plant at the time were that plant pressure was lowering and the Pressurizer (PZR) had completely emptied. This resulted in the reactor vessel coolant conditions [
]a,c during their re-creation of this event.
The continued cooling, [
]a,c.
14 Alarm avalanche 1503-16
HED #14
5612
5813
Closed. Fixed with patch WEC provided to SNC on August 14, 2015.
V&V testing was performed successfully.
SNC has determined that this issue no longer impacts the simulator’s suitability for the conduct of operating tests.
The large volume of alarms, often referred to as the “Alarm Avalanche”, was significantly lowered due to the combination of WEC’s alarm prioritization project and SNC’s use of the APS “Consequence” feature.
The alarm prioritization project by WEC led to a reevaluation of the alarm points and the priority assigned to each. [
]a,c and response but do not require immediate attention as they did previously.
[ ]a,c. To date; SNC has developed 8 specific consequence files based on
identified transients where the use of the consequence logic has been proven beneficial. An example of this is that [ ]a,c. Those alarms can be set to populate the Consequence tab on APS vice appearing on the current tab. The consequence alarms can be reviewed at any time by selecting the correct tab or by turning off individual consequence functions.
ND-15-1333 Enclosure 5, Page 9 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E5-1 (continued)
NRC
# {*} NRC UI Description
{*}
Ref
# {*}
SNC SCR
#
SNC Evaluation
Through the combination of these two efforts the overall number of audible alarms received during transients is reduced to those [ ]a,c. This removes a major distraction from operators and allows efforts and attention to be focused upon monitoring and controlling the plant.
Examples of alarm reduction: Reactor Trip:
[ ]a,c
[ ]a,c
Loss of offsite power concurrent with main generator trip:
[ ]a,c
[ ]a,c
This item was also identified during the ISV. See Enclosure 8, HED #14.
15
Inconsistent VRS and VHS radiation monitor indications on a loss of process flow
TO-75 and 76
5828
5914
Closed. Fixed with patch WEC provided to SNC on August 14, 2015.
Two separate V&V tests were performed successfully.
Two tests were conducted to verify response of VRS and VHS. The first test was conducted at 100% power. ECS-ES-1 was de-energized. It was verified that [ ]a,c and no alarm was received upon loss of power. This test was completed successfully.
The second test was conducted at 100% power. Test personnel ensured [ ]a,c. This test
was completed successfully.
16 BEACON operability TO-102 5583
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This issue has been determined to be a simulator I&C implementation issue. The issue is not an AP1000 I&C design issue, nor is it a modeling issue.
The ability of BEACON to perform its intended function is directly related to the functionality of [
]a,c. In the event that BEACON is not functional, operators are required to carry out actions in specific Technical Specifications.
For training scenarios where it is desired to fail BEACON (or one of the inputs to BEACON), OPDMS, as it is currently implemented on AP1000 simulators, fails to alert operators that it is no longer operable.
Currently, the status of BEACON is passed to students by instructors when OPDMS displays an incorrect operational status. Therefore, the lack of the ability of BEACON to determine its operability status when the BDP NAP provides a BAD quality signal to BEACON does not impact the suitability of the simulator for the
ND-15-1333 Enclosure 5, Page 10 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E5-1 (continued)
NRC
# {*} NRC UI Description
{*}
Ref
# {*}
SNC SCR
#
SNC Evaluation
conduct of operating tests.
17
Inconsistent navigation to Protection and Safety Monitoring System (PMS) mimics in Ovation
1504-02 6670
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
PMS mimics in Ovation have several graphics [ ]a,c, you will be
changed to PMSA.
The PMS mimic in Ovation is an operator aid and not needed for plant operation or PMS actuations. Therefore, this issue does not impact the conduct of operating tests.
Operators are trained to apply Human Performance (HU) tools when operating the plant, including changing from one Ovation screen to another.
CR 10070361 for WEC resolution.
18
Confusing PMS status display
{Stage 3 ADS box unused on PMS Divisions C and D Display}
TO-122 5619
This issue was closed as invalid.
An investigation determined that this is in accordance with the AP1000 design and that this is the expected plant indication.
The ADS Summary graphic provides the following possible indications in regards to ADS Actuation:
[
ND-15-1333 Enclosure 5, Page 11 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E5-1 (continued)
NRC
# {*} NRC UI Description
{*}
Ref
# {*}
SNC SCR
#
SNC Evaluation
]a,c, just before the sheet navigation flag for the respective valve control sheet.
For this reason SNC has determined that this SCR is not valid. These indications are the designed indications and therefore will be the same indications available to and require the same actions from an operator.
19
Unexplained Steam Generator (SG) level rise following trip of all Reactor Cooling Pumps (RCPs)
1502-03 6471
SNC evaluated this item and determined that the simulator is modeling the AP1000 plant design.
SNC has closed this issue.
The rise in level is due to the downcomer dropping the last bit of water it has as it goes to near dryout conditions.
During AP-OPS-T-004, “Trip of ALL RCPS,” a level rise of [ ]a,c (with a slight delta between SGs). At this time all inputs and outputs from the SGs have been isolated for over [
]a,c
A detailed evaluation of the conditions internal to the SGs determined the level rise is due to the last bit of water dropping out of the downcomer as it enters near dryout conditions. WEC agreed with this. Therefore, the simulator modeling is correct and this is a correct plant response for the transient.
20
Pressurizer (PZR) Level went down in 2 of 3 training scenarios with the leak through the PZR safety
1502-08 6484
Closed. Could Not Duplicate.
This issue was reported at a non-SNC AP1000 simulator.
The same initial conditions were established on SNC’s simulator. This was facilitated by using the same Simulator APP file that was used by the discovering simulator group. The APP file provides the ability to save a set of malfunctions such that the same scenario can be reset and the same exact malfunctions can be re-inserted. SNC used this file during three trial runs in an attempt to duplicate this issue. This ensured the same exact faults were used for the diagnosis. SNC monitored Wide Range and Narrow Range pressurizer level response and observed no significant difference in the indications. The maximum difference (delta %) between the three runs for wide range was [ . ]a,c Since these values represent no significant difference given the indication response and since this does not indicate a leak through the PZR safety, this issue was closed.
21
Over power control permissives did not respond to steam leak as designed
1502-09 6122
Closed. Fixed with patch WEC provided to SNC on August 14, 2015.
V&V testing was performed successfully.
V&V testing was conducted from 100% power under initial simulator conditions similar to those that existed at the time the issue was identified. Turbine control was placed in MWe IN with rods in automatic. Both Power Operated Relief Valves (PORVs) were manually opened to 100% for an excess steam demand. [
]a,c. V&V testing was completed satisfactory.
ND-15-1333 Enclosure 5, Page 12 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E5-1 (continued)
NRC
# {*} NRC UI Description
{*}
Ref
# {*}
SNC SCR
#
SNC Evaluation
22
PZR Water Level response during Safety valve malfunctions has variations in tests
1502-12 6484
Closed. Could Not Duplicate.
The APP file provides the ability to save a set of malfunctions such that the same scenario can be reset and the same exact malfunctions can be re-inserted. SNC used this file during three trial runs in an attempt to duplicate this issue. This ensured the same exact faults were used for the diagnosis. SNC monitored Wide Range and Narrow Range pressurizer level response and observed no significant difference in the indications. The maximum difference (delta %) between the three runs for wide range was [
]a,c Since these values represent no significant difference given the indication response and since this does not indicate a leak through the PZR safety, this issue was closed.
23
During Load Rejection events, Load Unbalance response is inconsistent causing noticeable deltas in several key parameters
1502-13 6483
Closed. Fixed with patch WEC provided to SNC on August 14, 2015.
V&V testing was performed successfully.
V&V testing was conducted with three identical tests under initial simulator conditions similar to those that existed at the time the issue was identified. Each test initiated a 100% load rejection and graphed the response of the turbine intercept valves. All valves responded identically throughout all three tests. The problems that were initially reported under this issue were not observed during these tests. V&V testing was completed satisfactory.
24
TCS heat transfer characteristics through the H2 coolers are unrealistic
1503-33 6181
SNC has determined that this issue is acceptable and that it does not impact the simulator’s suitability for the conduct of operating tests.
The simulator is correctly modeling the present plant design.
This issue is based on the inability of the TCS temperature control valve, controlling H2 cooler temperature, to establish a steady state position. Corrective Action Program And Learning (CAPAL) 100221278 was sent to CB&I for a design or I&C change. CB&I responded by informing SNC that the heat exchanger is too large. This corresponds to the Simulator response. Because the heat exchanger is too large, the temperature control valve is forced closed to prevent over-cooling. Because the valve is fully closed, when the temperature reaches a point where the valve needs to open, the response time is too slow and temperature doesn’t begin to lower before a high temperature alarm is received. As the valve continues to open, temperature turns and begins to lower, but the temperature drop occurs faster than the valve can respond and the valve is once again forced closed. However, even in steady-state conditions a small modulation of the temperature control valve will result in temperature lowering. CB&I will need a better control scheme or an actual heat exchanger design change.
This cycling of temperature from a low to a high value occurs over approximately 2 hours (from points where the over-cooling has occurred and high temperature alarm is received). A mitigating strategy has been put in place to establish the initial conditions and then save those initial conditions immediately after the temperature has been lowered. This provides the maximum amount of time before a high temperature alarm is received. Most training scenarios are either less than 2 hours OR result in placing the plant in a condition where H2 cooling is no longer required prior to this 2 hour window expiring. [
ND-15-1333 Enclosure 5, Page 13 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E5-1 (continued)
NRC
# {*} NRC UI Description
{*}
Ref
# {*}
SNC SCR
#
SNC Evaluation
]a,c.
25 "Instrument Air" alarm tile has no points assigned to it
1504-01 6669
SNC has determined that this issue is invalid. The issue is closed.
The Alarm Presentation System (APS) has multiple tiles where alarm points are associated with each tile. This issue was initially discovered at an alternate domestic AP1000 simulator. SNC confirmed that APS has associated alarm points with Instrument Air and therefore, this is not an issue at the SNC simulation facility.
26
Control logic functions associated with solid plant operations do not function as described in the design documentation
1504-09 5968
Closed. Fixed with patch WEC provided to SNC on August 14, 2015.
V&V testing was performed successfully.
For V&V testing, two testing scenarios were used to verify the proper function of the protective features under simulator conditions replicating those that existed when the issue was first identified. One test was performed to verify the high pressure related functions and another for the low pressure related functions. The observed responses were verified correct per AP1000 design documentation.
The first test was conducted at Mode 5 with the RCS in solid pressure mode. [
]a,c. The test was reset to initial conditions and the “B” CVS makeup pump trip setpoint tested. Both tests resulted in satisfactory V&V.
The second test conditions for V&V were to verify RCP trip setpoints on low pressure while in solid plant operations. [
]a,c. The test was reset to initial conditions and repeated with the same result. Both tests resulted in satisfactory V&V.
27
Control rods rejecting to manual during Anticipated Transient Without Scram (ATWS)
TO-47 None
SNC was unable to duplicate this issue.
A number of Rod Control issues have been observed at SNC and evaluated as a whole. Multiple ATWS scenarios have been reviewed and rods did not reject to manual. All rod control issues at SNC have been associated with one (1) SCR. Refer to SCR 5659 for SNC testing of rod control issues. Over 10 test runs were completed under various plant conditions to verify rods reject to manual. Following patch V3.R1.7F8.1.0.1, five different tests were run. A steady state test at [ ]a,c was allowed to run for [ ]a,c to verify rods did not reject to manual. All tests were completed satisfactory.
Since this particular issue was never observed or duplicated, SNC did not create an SCR.
28 Steam dump capacity appears to be larger than expected
1410-07 6830
SNC evaluated this issue and has developed and implemented a solution that results in turbine bypass valve flow being simulated per design.
V&V testing was conducted under initial simulator conditions similar to those that existed at the time the issue was identified. Main Steam Header pressure and Turbine Bypass valve response was monitored during the spurious trip of the main generator circuit breaker from 100% power. Steam flow through turbine bypass valve,
ND-15-1333 Enclosure 5, Page 14 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E5-1 (continued)
NRC
# {*} NRC UI Description
{*}
Ref
# {*}
SNC SCR
#
SNC Evaluation
MSS-V001, was verified correct when the valve was full open. [ ]a,c. V&V testing was
successful.
29
Determine if ventilation system response is correct (VAS, VRS, (VFS) systems)
1501-02 6410
SNC evaluated this issue and determined that it constitutes a plant design issue that does not impact the suitability of the simulator for the conduct of operating tests.
After discussions with WEC and other licensees, the simulator was verified to be functioning per plant design. WEC has determined that a plant design change will be necessary to alleviate the condition. The WEC update is the result of a specialized test that was developed by VC Summer. The test was designed to create a LOCA outside the reactor containment through the letdown flow path. [
]a,c.
This condition would only exist for [ ]a,c before the VFS fans trip on low flow. Operators will respond to alarming conditions per the ARPs associated with Containment (CTMT) pressure which require a flow path to be aligned to the VFS exhaust fans.
30
Following SG dryout, SG Wide {Narrow} Range level does not stay at zero. The Level will oscillate
1502-14 6434
SNC evaluated this item and determined that the instruments are responding as per the AP1000 plant design. This issue has no impact on the suitability of the simulator for the conduct of operating tests.
SNC closed this item as invalid.
SNC discovered that as the compartment pressure rises, the narrow range differential pressure slowly falls. This is what is causing the narrow range level to exhibit a slight rise.
This response is expected per the AP1000 design.
31
SG parameters have unexplained damped oscillation following “Main Steamline Break Outside Containment”
1502-15 6482
Closed. Could not duplicate.
The issue was originally noted on one of two simulators at another licensee’s site. The oscillations were noted around 1800 seconds in a malfunction test. While attempting to mimic the conditions that existed at the time the issue was first discovered, SNC identified that the issue was due to the data collection interval being shortened to 0.5 seconds. SNC shortened the data collection interval to 0.5 seconds, but was not able to recreate the issue using this interval with the same malfunction inserted in either of two test runs. This issue was closed.
32 Difficulty determining CMT actuation
1503-08 and 09
HED #2
5998
SNC evaluated this issue and determined that it does not impact the suitability of the simulator for the conduct of operating tests.
This issue is not an AP1000 design issue nor is it a simulator modeling issue.
HED #2 was driven by inconsistencies in determining whether CMTs were in service by crews during ISV. To
correct this, WEC is issuing procedural changes to the Emergency Operating Procedure network. SNC has received the updated procedures and is processing them in accordance with the normal procedural change
ND-15-1333 Enclosure 5, Page 15 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E5-1 (continued)
NRC
# {*} NRC UI Description
{*}
Ref
# {*}
SNC SCR
#
SNC Evaluation
process.
Related to Enclosure 8, HED #2.
33 Problems during transfer to remote shutdown room
1503-21 6075
SNC evaluated this issue and determined that it does not impact the suitability of the simulator for the conduct of operating tests.
This is a simulator I&C network related issue. It is not an AP1000 plant design issue nor is it a simulator modeling issue.
This issue occurs when simulator servers 216/217 are logged out and the displays are disconnected. The software will crash when an operator attempts to log in. If the drops are logged in and the displays are reconnected, the software will lock the interface.
As a safeguard to prevent this from happening during training, the simulation guide directs the instructor to test the functionality of the transfer switch as part of the scenario set-up.
The strategy to prevent future impact also involves the soft control functions at the instructor station. The booth operator has the ability to use these soft controls to ensure the transfer to the RSR occurs in a manner that is transparent to the operating crew.
SNC also determined that the issue only affects RSR operation. Neither the availability of the RSR or the ability of operators to use or shift operations to the RSR is a requirement for licensed operators.
34 (RNS) system over-pressurization
1503-13
HED #11 None
SNC evaluated this issue and determined it to be invalid at the SNC site simulator.
SNC evaluated this issue against its procedures and determined that its procedures were adequate, providing sufficient detail and guidance to prevent an operator or crew from performing this action.
Specifically:
In accordance with 3-RNS-SOP-001 version D 0.3, [ ]a,c (Attachment 4 section 3.0). [
]a,c which is inside containment.
By following regulatory requirements, management expectations to follow procedures and SNC’s “Conduct of Operations” procedure, an event where the need for an interlock on RNS-V061 should not occur. For this event to occur, an SNC operator or an operating crew would have to intentionally violate one or more procedures.
Therefore, SNC has determined that this issue does not affect operator training or the development of exams.
Related to Enclosure 8, HED #11.
35 CCS low surge tank 1503-15 None Closed. Fixed with patch WEC provided to SNC on August 14, 2015.
ND-15-1333 Enclosure 5, Page 16 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E5-1 (continued)
NRC
# {*} NRC UI Description
{*}
Ref
# {*}
SNC SCR
#
SNC Evaluation
level alarm priority is incorrect
HED #13 V&V testing was performed successfully.
The priority of the CCS surge tank low level alarm was raised to [ ]a,c which is commensurate with the effects of a low tank level upon the plant.
Related to Enclosure 8, HED #13.
36
RNS pump does not restart on Diesel Generator (DG) Sequencer
1410-09 6000
Closed. Fixed with patch, Version 1.0.6.
V&V testing was performed successfully.
For V&V, the simulator was initialized in a Mode 4 initial condition with both RNS pumps running under conditions that attempted to duplicate the conditions that existed at the time the issue was first identified. The power supplies for each RNS pump were de-energized individually. The RNS pumps loaded onto the diesel in the proper load sequence.
This issue has not been noted on the simulator since the correction was implemented.
37 EDS battery performance
TO-04 5679
Closed. Fixed with patch, Version 1.0.7.
V&V testing was performed successfully.
SNC replicated the initial conditions that existed at the time the issue was first identified and re-tested the EDS battery performance. The V&V test consisted of a manual turbine trip, loss of offsite power, and failure of the diesel generators to load their respective busses. [
]a,c.
38
Main Steam System (MS) radiation] monitors do not respond during Steam Generator Tube Rupture (SGTR)
TO 89
{TO-10} 5682
SNC evaluated this item and determined that the detectors are responding per the AP1000 plant design.
SNC closed this item as invalid based on WEC input.
The RMS is functioning correctly as designed. The simulator results are correct.
There two sets of radiation detectors on each Main Steam Line (MSL), SGS-RE026B/RE027B and SGS-RE026A/RE027A. SGS-RE026B/RE027B are for detecting low level radioactivity such as a primary-to-secondary leak. SGS-RE026A/RE027A are for detecting high levels of radioactivity during post-accident conditions. Low level primary-to-secondary leakage would only be detected by SGS-RE026B/RE027B not SGS-RE026A/RE027A. If a SG tube rupture does not emit large concentrations of activity it will only be detected by RE026B/RE027B.
The measuring range for SGS-RE026A/RE027A rad monitors is consistent with operating plants and the DCD as dictated by Reg. Guide 1.97. It is a Reg. Guide 1.97 Variable Type E with a dictated range of 1E-1 to 1E+3 µCi/cc for the purpose of measuring noble gas effluent releases. NUREG-0737 and Reg. Guide 1.97 mandated plants add high range detectors on the MSL for the purpose of measuring accident level (magnitude)
ND-15-1333 Enclosure 5, Page 17 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E5-1 (continued)
NRC
# {*} NRC UI Description
{*}
Ref
# {*}
SNC SCR
#
SNC Evaluation
effluent releases from the MSL relief or atmospheric dump valves for the purpose of evaluation off-site dose releases. SGS-RE026A/RE027A are intended for accident (high range) measurements (Range: [ ]a,c µCi/cc). SGS-RE026B/RE027B are intended for low range measurements (Approximate Range: [
]a,c µCi/cc). SGS-RE026B/RE027B were added to the design to support Tech Spec 3.4.7, RCS Operation Leakage, specifically to address the tech spec limit for detecting 150 gpm per day per SG resulting from primary-to-secondary tube leakage.
Note, not all operating plants have N-16 detectors on their MSLs and cannot detect primary-to-secondary leakage via on-line radiation measurements. They use alternate indications.
In summary, SGS-RE026B/RE027B are for activity or low range measurements. SGS-RE026A/RE027A are for post-accident measurements. The measurement range is dictated by Reg. Guide 1.97. The AP1000 RMS present design for the MSL is in compliance with the licensing basis.
39
When Containment Air Filtration System had no flow, VFS-RY102 alarmed for high iodine
1503-25 6192
5914
Closed. Fixed with patch WEC provided to SNC on August 14, 2015.
V&V testing was performed successfully and the correction was subsequently incorporated into simulator load V3.R1.7F8.1.1.0, which was deployed on August 29, 2015.
40
As the licensee notes in their RAI response, the computer support applications provided by (NAPs) would not be used for Job Performance Measures because they do not assess the applicant’s knowledge. Calculations would be performed manually. This is why many of the discrepancies were considered to be not significant. However, NAPs provides data to the operator during event diagnosis and response. Given the number of NAPs
SNC evaluated this issue and determined that it does not impact the suitability of the simulator for the conduct of operating tests.
WEC provided a patch to SNC on August 14, 2015, that included corrections to Nuclear Applications (NAPs). SNC conducted V&V testing for each of the corrections under simulator conditions replicating those under which each issue was first identified.
The following four NAPs corrections successfully passed V&V testing:
1. The Plant Mode Application automatically updates plant mode from [ ]a,c. V&V testing was performed with
similar conditions to those when the issue first identified. V&V testing was satisfactory.
2. The Redundant Sensor Algorithm for Power Range Nuclear Power was updated to [
]a,c. SNC determined that V&V testing was satisfactory.
3. The Inverse Count Ratio application was corrected for proper response of the Intermediate Range 1/M plot. [
ND-15-1333 Enclosure 5, Page 18 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E5-1 (continued)
NRC
# {*} NRC UI Description
{*}
Ref
# {*}
SNC SCR
#
SNC Evaluation
discrepancies the staff concludes that they could impact operator workload in an inconsistent manner. The staff concludes that there needs to be a reduction in the number of NAPs related discrepancies including those already identified as significant.
]a,c. V&V testing was satisfactory.
4. The Critical Shutdown Safety Function application was corrected to properly display whether a cooldown or heatup was uncontrolled on the Mode 5/6 Critical Safety Function WPIS display. V&V testing was conducted under similar initial conditions to when the issue was first identified. [
]a,c. After waiting for RCS temperature to
stabilize, the “Control HU or CD” was no longer illuminated. V&V testing was satisfactory.
The following two NAPS corrections failed V&V testing: SNC evaluated this issue and determined that it does not impact the suitability of the simulator for the conduct of operating tests.
These three issues were evaluated individually and in aggregate by members of the team that performed the initial Aggregate Study using the same evaluation criteria as before. The team determined that these items do not substantially impact the simulator’s suitability for the conduct of operating tests for the reasons given at the end of each issue.
1. The Leak Rate Monitoring Application was updated as part of this patch. After performing V&V testing, SNC determined that the update was not successful. The operators were still unable to perform a leak rate determination with the NAP. The V&V test consisted of operator performance of an RCS and Main Steam Leak Determination Surveillance in accordance with the surveillance procedure. [
]a,c. The surveillance was unable to be performed. This V&V test failed.
The Leak Rate Monitoring Application is informational only and does not drive any alarms based upon the calculated leakage. For this reason, any leak rate calculations would have to be performed manually per plant procedures vice using the NAP calculated values.
2. Updates to the Time to Boil indications were included as part of this patch. After performing V&V testing, SNC personnel determined that the update was not successful. Time to Boil indication on the Mode 5/6 Primary trend WPIS was observed to be displayed in exponential minutes for the RCS time to boil. The same was true for the Spent fuel pool time to boil. V&V testing was conducted under similar initial conditions as when the issue was first identified. The values indicated by each of these displays should be in hours and minutes. This V&V test failed.
The Time to Boil NAP is a tool that is used for information only. The NAP is active when in Mode 5/6 conditions for RCS time to boil and when fuel is present in the spent fuel pool for spent fuel pool time to boil.
ND-15-1333 Enclosure 5, Page 19 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E5-1 (continued)
NRC
# {*} NRC UI Description
{*}
Ref
# {*}
SNC SCR
#
SNC Evaluation
The value is displayed in exponential units versus hh:mm. This means that operators will need to convert the scientific notation values into hours and minutes. Although this does take a short amount of time, the net effect is that it does not remove the ability to monitor the time to boil and it has no impact on actions the operator may, or may not, take in response to plant conditions.
41
Provide documentation that the Westinghouse Electric Company’s resolution of HED-1 discrepancies is consistent with the VC Summer (VCS) conclusions provided in the Commission-approved simulator request and its supplements.
See Enclosure 8.
42
Include all open discrepancy reports when the docketed list of simulator discrepancies is submitted.
See Enclosure 9
Note: {*} Numbers and descriptions correspond to the table “Summary of Unresolved Items as of 06-30-2015” as it appeared in an NRC letter dated July 2, 2015
(Reference 2) with the following exceptions. If the information in the “NRC UI Description” or “Ref #” columns was found to be incorrect, that information
was retained, but indicated by using strikethrough. The correct information was added immediately following and was contained in brackets “{ }.”
2.0 References
1. NRC Email dated 2015-05-13, Meeting Materials for May 14, 2015- VCSNS 2 and 3 Commission-Approved Simulator - CAS-
Summer-RAI 5-7-15_b Redacted, ML#15133A497
2. Virgil C. Summer Nuclear Station Units 2 and 3 - Request For A Commission-Approved Simulation Facility dated July 2, 2015,
ML#15182A097
Southern Nuclear Operating Company
Vogtle Electric Generating Plant (VEGP) Units 3 and 4
ND-15-1333
Enclosure 6
Commission Approved Simulator Aggregate Study - Simulator Training System Deficiency Impact on 10 CRF 55.45
(Non-Proprietary)
(The Aggregate Study is a standalone document consisting of 105 pages.)
SOUTHERN NUCLEAR COMPANY
Commission Approved Simulator Aggregate Study
Simulator Training System Deficiency Impact On 10CFR55.45(a) Compliance
7/17/2015
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
2 | P a g e
Contents
Introduction ....................................................................................................................................................... 9
Executive Summary ........................................................................................................................................... 9
Method of Study ................................................................................................................................................ 9
Participants ...................................................................................................................................................... 10
Simulator Review Committee (SRC) ................................................................................................................ 10
Aggregate Study Evaluation Results ................................................................................................................ 11
Index of Proposed Corrections......................................................................................................................... 12
10 CFR 55.45(a)(1) ............................................................................................................................................... 14
Executive Summary ......................................................................................................................................... 14
Reactor Coolant Pump (RCP) Net Positive Suction Head (NPSH) Curve has Inadequate Range for Operation 14
Rod Withdrawal button deselects During Continuous Operation ................................................................... 15
Issue with Automatic Control of Deaerator Storage Tank (DST) level and Auto Start of Standby Condensate Pump ................................................................................................................................................................ 15
Model Instability during pressurizer (PZR) Fill to Solid (no vapor bubble remaining) ..................................... 15
OPDMS Rod Insertion Limit (RIL) Indication Does Not Align to Combined Operating Limits Report (COLR) Rev. 0 ....................................................................................................................................................................... 16
Decay Heat Calculation Summary - Assembly Move NAP Function Not Functional ........................................ 16
M Control Rod Banks B & C Reversed on DRPI Health Screen ......................................................................... 17
Liquid Radwaste System WLS-MP-08C improperly Pumps Monitor Tank C .................................................... 17
Excessive Startup Feedwater (SFW) Control Valve Cycling .............................................................................. 18
Redundant Sensors Algorithm Application NAP Does Not Process Failed Channels Correctly ........................ 18
Manual Reactor Trip Alarm Occurred without a Reactor Trip Request ........................................................... 19
10 CFR 55.45(a)(2) ............................................................................................................................................... 20
Executive Summary ......................................................................................................................................... 20
Rod Withdrawal button deselects During Continuous Operation ................................................................... 20
Issue with Automatic Control of DST level and Auto Start of Standby Condensate Pump .............................. 21
OPDMS RIL Indication Does Not Align to COLR Rev. 0 ..................................................................................... 21
NAP for 1/M Intermediate Range Not Functional ........................................................................................... 22
MA Bank Rods Sometimes Stop at 263 steps during a CRE ............................................................................. 22
Excessive SFW Control Valve Cycling ............................................................................................................... 23
Audible Rod Step Skips .................................................................................................................................... 23
Improper Bank Overlap Occurs when Data Point OCB07CE00C_OUTAV Increments Improperly During
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
3 | P a g e
Startup ............................................................................................................................................................. 24
Manual Reactor Trip Alarm Occurred without a Reactor Trip Request ........................................................... 24
10 CFR 55.45(a)(3) ............................................................................................................................................... 25
Executive Summary ......................................................................................................................................... 25
Containment Cooling System (VCS) fan response due to loss of power .......................................................... 25
EDS Power Supply Assignments to PLS/DDS Cabinets Incomplete .................................................................. 25
Modeled BEACON Data Cannot Determine Quality ......................................................................................... 26
Rod Withdrawal button deselects During Continuous Operation ................................................................... 26
Containment Radiation Alarm Reset Points Incorrect ..................................................................................... 27
Pressurizer Heater Current Indicates BAD Quality at Limits ............................................................................ 27
Unidentified and Identified Leak Rate Always Indicates BAD Data .................................................................. 28
Low Flow Alarm on TCS-FT007 Occurs Earlier than Expected .......................................................................... 28
MFP 'B' Alarm Response Differs For Identical Fault ......................................................................................... 28
Unexpected Response of Alarm Cutout of RWS Pressure Alarms ................................................................... 29
Pressurizer Pressure Out of Range Indication Not Properly Displayed ............................................................ 30
Subcriticality Indication on Critical Safety Function Screen Drops to Bad Quality ........................................... 30
Issue with Automatic Control of DST level and Auto Start of Standby Condensate Pump .............................. 30
Degasifier Level Alarm Limits ........................................................................................................................... 31
PMS Mimic Screens .......................................................................................................................................... 31
OPDMS RIL Indication Does Not Align to COLR Rev. 0 ..................................................................................... 32
Nuisance Valve Modulating Status Alarms ...................................................................................................... 32
Unexpected VRS High Rad Alarm ..................................................................................................................... 32
VFD Transformer Temperature........................................................................................................................ 33
Print Feature from NAP non-functional ........................................................................................................... 33
VHS Rad Monitor Response to Loss of Process Flow ....................................................................................... 34
Pressurizer Narrow Range Pressure Does Not Indicate Bottom of Scale ......................................................... 34
CDS-TE040A/B Range is Inadequate ................................................................................................................ 34
DRPI Health Screen Alarms for Data Cabinet A and B Crossed ........................................................................ 35
Digital Rod Position Indication (DRPI) Health Screen Incorrect Logic Cabinet Alarms ..................................... 35
Containment Recirculation Actuation Indication Issue .................................................................................... 35
Uncontrolled Heat-up (H/U) Indication Incorrect ............................................................................................ 36
DHC Summary - Assembly Move NAP Function Not Functional ...................................................................... 36
RCP Vibration Alarm Naming ........................................................................................................................... 37
HSS Display does not Include ESOP Discharge Pressure .................................................................................. 37
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
4 | P a g e
NAP for 1/M Intermediate Range Not Functional ........................................................................................... 38
ECS Penetration Temperature off Scale Low ................................................................................................... 38
Plant Mode Selector NAP Inconsistent with Procedure ................................................................................... 38
NAPS display issues .......................................................................................................................................... 39
WPIS Downscale Arrow Absent ....................................................................................................................... 39
RSA NAP Does Not Process Failed Channels Correctly .................................................................................... 39
ZVS and ZBS Alarm Scaling Incorrect ............................................................................................................... 40
Flux doubling difference between divisions .................................................................................................... 40
Time to Boil Calculation ................................................................................................................................... 41
Audible Rod Step Skips .................................................................................................................................... 41
WPIS Display VARs ........................................................................................................................................... 41
CMT WR Level Indications go Bad Quality ....................................................................................................... 42
Unexpected Bank Sequence Out of Sequence Alarm ...................................................................................... 42
Urgent Alarm Occurs During Case 2 CRE.......................................................................................................... 43
Improper Bank Overlap Occurs when Data Point OCB07CE00C_OUTAV Increments Improperly During Startup ............................................................................................................................................................. 43
Manual Reactor Trip Alarm Occurred without a Reactor Trip Request ........................................................... 44
Controller Fault Alarms Received on Turbine Trip ........................................................................................... 44
Diesel Fuel Oil Day Tank Level Transmitter Operation..................................................................................... 45
Inconsistent UAT Line Voltage Alarm Priorities ............................................................................................... 45
Any Rods at Bottom Alarm .............................................................................................................................. 46
WGS Sample Package Ovation Interface ......................................................................................................... 46
WGS Sample Package Digital Indication .......................................................................................................... 47
RSA NAP for Power Range Power does not Eliminate Erroneous Input .......................................................... 47
Inconsistent DPU Alarm Priority Levels ............................................................................................................ 47
Safety Mimic Display for SGS-V255A& B Indicates Bad Quality Following a SFW Isolation ............................. 48
10 CFR 55.45(a)(4) ............................................................................................................................................... 49
Executive Summary ......................................................................................................................................... 49
Stage 3 ADS Box Unused on Divisions C and D ................................................................................................ 49
Subcriticality Indication on Critical Safety Function Screen Drops to Bad Quality ........................................... 49
VWS-TE079 Point Named Incorrectly .............................................................................................................. 50
Calorimetric Data Precision ............................................................................................................................. 50
Inconsistent OPDMS QPT Indications .............................................................................................................. 50
VFD Transformer Temperature ........................................................................................................................ 51
CDS-TE040A/B Range is Inadequate ................................................................................................................ 51
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
5 | P a g e
CVS-PT040 does not Provide Proper Protective Functions .............................................................................. 52
M Banks B & C Reversed on DRPI Health Screen ............................................................................................. 52
Quality of RWS-V503 BAD at Limits ................................................................................................................. 53
Reactor Coolant Pump (RCP) Stator Temperature Indication off Scale Low at Lower Speeds ........................ 53
HSS Display does not Include Emergency Seal Oil Pressure (ESOP) Discharge Pressure ................................. 54
DWS-LT006 has Insufficient Range .................................................................................................................. 54
MA Bank Rods Sometimes Stop at 263 steps during a CRE ............................................................................. 54
ECS Penetration Temperature off Scale Low ................................................................................................... 55
Improper function of C-2 reactor power control interlock .............................................................................. 55
WPIS RCS Inventory Issues ............................................................................................................................... 55
WPIS Downscale Arrow Absent ....................................................................................................................... 56
Tuning of VBS Required for Stability ................................................................................................................ 56
Condensate Polisher Bypass Valve Control ...................................................................................................... 57
Time to Boil Calculation ................................................................................................................................... 57
CMT WR Level Indications go Bad Quality ....................................................................................................... 57
Manual Reactor Trip Alarm Occurred without a Reactor Trip Request ........................................................... 58
Main Generator Output breaker logic ............................................................................................................. 58
Excitation Transformer Graphic Issue .............................................................................................................. 59
IDS Charger Capacity and Design Float Voltage Requirement are Incompatible ............................................. 59
Graphic 1805 has reversed rods ...................................................................................................................... 60
Residual Bus Transfer Issues ............................................................................................................................ 60
Diesel Fuel Oil Day Tank Level Transmitter Operation..................................................................................... 61
WGS Sample Package Digital Indication .......................................................................................................... 61
RSA NAP for Power Range Power does not Eliminate Erroneous Input .......................................................... 62
Safety Mimic Display for SGS-V255A& B Indicates Bad Quality Following a SFW Isolation ............................. 62
Safety Mimic Display Navigation Issue............................................................................................................. 63
10 CFR 55.45(a)(5) ............................................................................................................................................... 64
Executive Summary ......................................................................................................................................... 64
EDS Power Supply Assignments to PLS/DDS Cabinets Incomplete .................................................................. 64
Modeled BEACON Data Cannot Determine Quality ......................................................................................... 64
Repeatability issues involving CL 1B ................................................................................................................ 65
Unidentified and Identified Leak Rate Always Indicates BAD Data .................................................................. 65
Primary Dedicated Safety Panel Screens Do Not Update during MCR/RSR Transfer ...................................... 66
PMS Mimic Screens .......................................................................................................................................... 66
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
6 | P a g e
Hot Leg Fluctuations at Mid-loop .................................................................................................................... 67
OPDMS RIL Indication Does Not Align to COLR Rev. 0 ..................................................................................... 67
Inconsistent OPDMS QPT Indications .............................................................................................................. 68
Print Feature from NAP non-functional ........................................................................................................... 68
CDS-TE040A/B Range is Inadequate ................................................................................................................ 69
CVS-PT040 does not Provide Proper Protective Functions .............................................................................. 69
Containment Recirculation Actuation Indication Issue .................................................................................... 70
CVS-V094 Power Failure Response .................................................................................................................. 70
DHC Summary - Assembly Move NAP Function Not Functional ...................................................................... 71
HSS Display does not Include Emergency Seal Oil Pump (ESOP) Discharge Pressure ...................................... 71
DWS-LT006 has Insufficient Range .................................................................................................................. 71
MA Bank Rods Sometimes Stop at 263 steps during a CRE ............................................................................. 72
Improper function of C-2 ................................................................................................................................. 72
Excessive SFW Control Valve Cycling ............................................................................................................... 73
SWS temperature control ................................................................................................................................ 73
FWS-V037 Control Issue ................................................................................................................................... 74
SGS MSL drain pot erratic indication ............................................................................................................... 74
Stuck Rod Recovery Malfunction ..................................................................................................................... 74
Tuning of VBS Required for Stability ................................................................................................................ 75
RSA NAP Does Not Process Failed Channels Correctly .................................................................................... 75
Flux doubling difference between divisions .................................................................................................... 76
Time to Boil Calculation ................................................................................................................................... 76
Audible Rod Step Skips .................................................................................................................................... 76
VFS Radiation Monitoring Issue ....................................................................................................................... 77
CMT WR Level Indications go Bad Quality ....................................................................................................... 77
Urgent Alarm Occurs During Case 2 CRE.......................................................................................................... 78
Improper Bank Overlap Occurs when Data Point OCB07CE00C_OUTAV Increments Improperly During Startup ............................................................................................................................................................. 78
Manual Reactor Trip Alarm Occurred without a Reactor Trip Request ........................................................... 79
Main Generator Output breaker logic ............................................................................................................. 79
Residual Bus Transfer Issues ............................................................................................................................ 80
Diesel Fuel Oil Day Tank Level Transmitter Operation..................................................................................... 80
VES Supply Header Pressure Response to Temperature Changes ................................................................... 81
ECS-EC-313 Loads not modeled ....................................................................................................................... 81
D/G Sequencer Operation ............................................................................................................................... 82
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
7 | P a g e
RSA NAP for Power Range Power does not Eliminate Erroneous Input .......................................................... 82
VZS Dampers do not Fail As-Is after Loss of Power .......................................................................................... 83
Turbine Bypass Control Valve Control Logic cannot Support Design Power Supplies ..................................... 83
Battery Temperature does not change ............................................................................................................ 84
Fire Protection System is not modeled in Containment .................................................................................. 84
IRWST Temperature Response ........................................................................................................................ 84
10 CFR 55.45(a)(6) ............................................................................................................................................... 86
Executive Summary ......................................................................................................................................... 86
Rod Withdrawal button deselects During Continuous Operation ................................................................... 86
Unstable VFS Containment Exhaust Flow ........................................................................................................ 87
GSS Header Pressure Response ....................................................................................................................... 87
Issue with Automatic Control of DST level and Auto Start of Standby Condensate Pump .............................. 88
Model Instability during PZR Fill to Solid ......................................................................................................... 88
Steam Generator Level Instability with Control Valves Shut ........................................................................... 89
CVS-V094 Power Failure Response .................................................................................................................. 89
DHC Summary - Assembly Move NAP Function Not Functional ...................................................................... 89
WLS-MP-08C improperly Pumps Monitor Tank C ............................................................................................ 90
MA Bank Rods Sometimes Stop at 263 steps during a CRE ............................................................................. 90
WRS Sump Pump B Discharge Pressure Inadequate ....................................................................................... 91
Excessive SFW Control Valve Cycling ............................................................................................................... 91
Stuck Rod Recovery Malfunction ..................................................................................................................... 92
Polisher Bypass Valve Control .......................................................................................................................... 92
Urgent Alarm Occurs During Case 2 CRE.......................................................................................................... 93
IDS Charger Capacity and Design Float Voltage Requirement are Incompatible ............................................. 93
Residual Bus Transfer Issues ............................................................................................................................ 94
ECS-EC-313 Loads not modeled ....................................................................................................................... 94
D/G Sequencer Operation ............................................................................................................................... 95
Turbine Bypass Control Valve Control Logic cannot Support Design Power Supplies ..................................... 95
Fire Protection System is not modeled in Containment .................................................................................. 96
10 CFR 55.45(a)(7) ............................................................................................................................................... 97
Executive Summary ......................................................................................................................................... 97
Repeatability issues involving CL 1B ................................................................................................................ 97
CVS-PT040 does not Provide Proper Protective Functions .............................................................................. 98
DHC Summary - Assembly Move NAP Function Not Functional ...................................................................... 98
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
8 | P a g e
Excessive SFW Control Valve Cycling ............................................................................................................... 99
Time to Boil Calculation ................................................................................................................................... 99
CMT WR Level Indications go Bad Quality ..................................................................................................... 100
Turbine Bypass Control Valve Control Logic cannot Support Design Power Supplies ................................... 100
IRWST Temperature Response ...................................................................................................................... 101
10 CFR 55.45(a)(8) ............................................................................................................................................. 102
Executive Summary ....................................................................................................................................... 102
EDS Power Supply Assignments to PLS/DDS Cabinets Incomplete ................................................................ 102
CVS-V094 Power Failure Response ................................................................................................................ 102
CMT WR Level Indications go Bad Quality ..................................................................................................... 103
Fire Protection System is not modeled in Containment ................................................................................ 103
IRWST Temperature Response ...................................................................................................................... 104
10 CFR 55.45(a)(9) ............................................................................................................................................. 105
Executive Summary ....................................................................................................................................... 105
Simulator MCR missing Rad Monitoring Panel .............................................................................................. 105
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
9 | P a g e
Introduction As part of the Commission Approved Simulator (CAS) request, Southern Nuclear Operating Company (SNC) commissioned a team to evaluate the known deficiencies in the simulator to determine if the 13 criteria established in 10 CFR 55.45(a), “Operating Tests,” would be challenged. The team was comprised of representatives from SNC (Operations, Training and Engineering), SCANA (Training) and Westinghouse (Human Factors Engineering).
Executive Summary The team examined all Discrepancy Reports (DRs) that were open as of May 15, 2015, and determined that 101 DRs were relevant to acceptability of one or more of the first nine criteria of 10 CFR 55.45(a). No DRs were found to be relevant to the last four criteria; 55.45(a)(10) through 55.45(a)(13). The team also determined that no singular DR posed a challenge to the suitability of the simulation facility for the conduct of operating tests; however, when considered in the aggregate, 42 of the DRs challenged criterion (3) and (5) of 10 CFR 55.45(a) (See section 2.1 below for additional details).
In order to declare the simulator suitable for the conduct of operating tests, corrective actions were initiated to resolve the subject 42 DRs. This assessment was communicated with Westinghouse Electric Company (WEC) and WEC committed to implement improvements aimed at resolving these issues in a patch deliverable to SNC by August 14, 2015. Based on this commitment, the CAS Aggregate Study Team reconvened on July 7, 2015 and determined that the proposed corrective actions would be adequate so that the aggregate impact of the remaining discrepancies would not pose a challenge to any of the 10 CFR 55.45(a) criteria.
On August 14, 2015, WEC delivered a patch to SNC which contained corrections for the 42 items previously identified along with some additional corrections. After performing Verification and Validation (V&V), 11 were determined to require further resolution. After confirming the corrections that successfully passed the V&V process, the CAS Aggregate Study Team reconvened on September 1, 2015, to review the impact of the remaining 11 items. Based on the combination of these successful corrections and additional improvement in the area of Alarm Response, through the use of the Consequence Alarm feature of the Alarm Presentation System (APS), the team concluded that the aggregate impact of the remaining items would not impact the suitability of the simulator for the performance of operating tests.
Method of Study A multi-disciplined team was formed (see “Participants” below) consisting of individuals from both internal and external of SNC. The participants were requested to answer the following questions for the individual and aggregate impact of the deficiencies.
General Questions for each deficiency of the study
• Does an individual item fit in the assigned category?
• Could an individual item affect another part of 55.45 that it is not currently assigned to?
• Could an individual item come off the list entirely? (I.e. does not affect 55.45)
• Does an individual item make sense? Do you understand the problem it introduces?
• Are we working around the issue with procedure changes or special training? If so, how is that documented?
Each of the identified deficiencies were evaluated using the above criteria and categorized into the appropriate criteria of 10 CFR 55.45(a). Individual deficiencies were not limited to one category, but
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
10 | P a g e
were added to all applicable category bins associated with the nature of the deficiency. Deficiencies associated with design of the plant or deficiencies that did not apply to the criteria in 10 CFR 55.45(a), were removed from the list. Upon completion of binning of the individual deficiencies, the team evaluated each of the criterion of 10 CFR 55.45(a)(1) – (13) to determine if the simulators are suitable for conducting operating tests with the existing deficiencies.
Participants The team was composed of participants with a diverse mix of backgrounds including Operations, Instrument & Controls (I&C) Engineering, Training, and Human Factors Engineering from both Vogtle 3&4, V.C. Summer Units 2&3 and Westinghouse. The team was composed of the following members:
Tom Arnette – Shift Manager, Vogtle Units 3&4 (12.5 years Nuclear Navy, Reactor Operator; SRO license holder and Shift Manager at Kewaunee)
Chris Parkes – Shift Supervisor, Vogtle Units 3&4 (BS- Computer Information Systems; 23 years Nuclear navy, Qualified Engineering Officer of the Watch and Engineering Watch Supervisor for 18 years)
Shawn Wolfgong – Shift Support Supervisor, Vogtle Units 3&4 (BS – Applied Nuclear Technology; 20 years Nuclear Navy, EWS & EOOW)
Chris Cannon – Nuclear Plant Operator, Vogtle Units 3&4 (BS – Nuclear Engineering Technology; 6 years Nuclear Navy, Reactor Operator; 3 years Lockheed Martin Electronics testing and repair)
Matt Schmader – Training Lead, Vogtle Units 3&4 (BS – Physics; 9 years Nuclear Navy Officer; 5 years Operations training at Watts Barr, SRO-certified.
Allahondra Manning – Engineer, Vogtle Units 3&4 (BS- Electrical Engineering; 6.5 years Electrical and I&C systems engineer, design authority SRS)
Kim Yennerell – Engineer, Vogtle Units 3&4 (BS – Electrical Engineering; 10.5 years Nuclear design and program engineering)
Korrie Hoffman – Simulator Engineer, Vogtle Units 3&4 (BS – Nuclear Engineering; 3.5 years core design engineer with WEC)
Kevin Balch – Simulator Engineer, V.C. Summer Units 2&3 (BS – Nuclear Engineering; 20 years simulator software engineer, 10 years nuclear fuel engineer)
H. Adrian Fletcher – Human Factors Engineering Operations Specialist, Westinghouse (AS – Nuclear Engineering Technology; NRC license holder for 18 years, 6 years Reactor Operator, 8 years Unit Supervisor, 4 years Shift Manager)
Simulator Review Committee (SRC) The SRC is composed of one member of Operations, selected by the Operations Director; the Operations Training Manager or designee; and the Simulator Coordinator or designee. The SRC is supplemented by incumbents, as necessary, to serve as subject matter experts to conduct a Training Needs Assessment (an appraisal by a subject matter expert of a simulator deviation, deficiency, or modification, and its relative importance to the operator as required tasks are performed). Incumbents are Operations individuals who have completed AP1000 certification training.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
11 | P a g e
Aggregate Study Evaluation Results The team evaluated each of the individual deficiencies and determined that none of the issues, by themselves, constituted a challenge to any of the 13 criteria of 10 CFR 55.45(a). The team did determine that, in the aggregate, some of the deficiencies could challenge 10 CFR 55.45(a) criterion (3) and (5).
10 CFR 55.45(a)(3):
“Identify annunciators and condition-indicating signals and perform appropriate remedial actions where appropriate.”
10 CFR 55.45(a)(5):
“Observe and safely control the operating behavior characteristics of the plant.”
The reasoning for the determination involved four (4) main areas: 1. Indication deficiencies 2. Alarms management deficiencies and challenges 3. Rod Control System deficiencies 4. Secondary control challenges
Indication deficiencies: The key drivers for this area are associated with challenges to the simulator in providing the operator with the necessary and correct information in order to monitor and control the plant. The effect of the identified deficiencies is that operators do not always have the correct information presented to them to make appropriate decisions required for safe operation. Additionally, with many identified deficiencies, operators will tend to question the validity of all indications, including the ones that are working correctly.
Specific DRs associated: 6169, 6621, 5689, 5599, 6175, 6315, 6159, 6089, and 5623.
Alarm Management deficiencies and challenges: The key drivers for this area are associated with the excessive number of alarms, the absence of some required alarms, and the distraction presented to the operators in managing the Alarm Presentation System (APS). Alarm management is a significant operator burden placed on the crew throughout all scenarios and plant conditions. Alarm management currently makes simple and routine evolutions difficult.
Specific DRs associated: 5813, 5613, and 6651.
Rod Control System deficiencies: The key drivers for this area are associated with the inconsistent and unpredictable nature of the Rod Control System during performance of reactivity related tasks. In effect, whenever the operator operates the controls to move rods, the operator could encounter no issues or he/she may encounter many issues:
- The control button can deselect during manual rod motion - The audible clicking sound can skip during rod motion or continue after rod motion has stopped - The Bank Out-of-Sequence alarm can come in when conditions do not warrant - The rod group indication may indicate values below the fully inserted position or above the fully
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
12 | P a g e
withdrawn position - The bank overlap could be incorrect - During a Rod Exchange, an Urgent Failure alarm may or may not occur when AO bank temperature
control is demanded The overall affect is that these issues will inhibit the operator’s ability to timely perform reactivity manipulations in a precise and controlled manner.
Specific DRs associated: 5584, 6259, 6302, 6186, and 6267.
Secondary control challenges: The key drivers for this area are associated with the inability of the automatic control of the secondary plant and subsequent operator required action to manually control systems in order to respond to simulator scenarios to prevent automatic actuation of standby components.
Specific DRs associated: 6151, 5655, and 6156.
Index of Proposed Corrections The following is a summary of the proposed corrections that the team determined would result in a valid combination of corrections that the aggregate impact of the remaining discrepancies would not pose a challenge to any of the 10 CFR 55.45(a) criteria. [See Enclosure 7 of the CAS Submittal Letter for an updated status of the following items.]
1. SCR-DR-5584 (Parts 1, 2, 3, 6)1 - Rod Withdrawal button deselects During Continuous Operation 2. SCR-DR-5597 (Part 3) - Containment Radiation Alarm Reset Points Incorrect
3. SCR-DR-5599 (Parts 3, 5)1 - Unidentified and Identified Leak Rate Always Indicates BAD Data 4. SCR-DR-5627 (Parts 3, 4) - Sub-criticality Indication on Critical Safety Function Screen Drops to Bad
Quality 5. SCR-DR-5643 (Part 4) - VWS-TE079 Point Named Incorrectly (TE079 is a temperature indication) 6. SCR-DR-5644 (Not part of aggregate study) – Display 17600 indicating wrong flowpath 7. SCR-DR-5688 (Not part of aggregate study) – RCS graphic 50308 incorrect 8. SCR-DR-5689 (Parts 3,5) – PMS mimic screens
9. SCR-DR-5702 (Not part of aggregate study) – IDS screens show inaccurate power supplies 10. SCR-DR-5712 (Part 4) - Calorimetric Data Precision
11. SCR-DR-5813 (Part 3)1 - Nuisance Valve Modulating Status Alarms 12. SCR-DR-5909 (Not part of aggregate study) – Graphic 11181 has orphaned “n” 13. SCR-DR-5920 (Part 3) - Pressurizer Narrow Range Pressure Does Not Indicate Bottom of Scale 14. SCR-DR-5924 (Part 3) – Digital Rod Position Indication (DRPI) Health Screen Alarms for Data Cabinet A
and B Crossed 15. SCR-DR-5925 (Part 3) - DRPI Health Screen Incorrect Logic Cabinet Alarms 16. SCR-DR-5968 (Parts 4, 5, 7) - CVS-PT040 does not Provide Proper Protective Functions (PT040 is a
pressure transmitter) 17. SCR-DR-6009 (Part 3) - Uncontrolled heatup or cooldown indication incorrect 18. SCR-DR-6030 (Parts 1, 4) - M Banks (control rods) B & C Reversed on DRPI Health Screen 19. SCR-DR-6078 (Parts 3, 4, 5) – Hydrogen Seal Oil System (HSS) Display does not Include Emergency Seal
Oil Pump (ESOP) Discharge Pressure
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
13 | P a g e
20. SCR-DR-6089 (Parts 2, 3)1 – NAP(Nuclear Application) for 1/M(inverse count rate ratio plot) Intermediate Range Not Functional
21. SCR-DR-6102 (Parts 2, 4, 5, 6) - MA Bank Rods Sometimes Stop at 263 steps during a CRE (Control Rod Exchange)
22. SCR-DR-6129 (Not part of aggregate study) – Display 40023 units issue
23. SCR-DR-6144 (Part 3) - Plant Mode Selector NAP Inconsistent with Procedure
24. SCR-DR-6159 (Part 3)1 - NAPS display issues 25. SCR-DR-6160 (Not part of aggregate study) – Component Cooling System (CCS) screen issue 26. SCR-DR-6164 (Parts 3, 4) – Wall Panel Information System (WPIS) Downscale Arrow Absent 27. SCR-DR-6165 (Not part of aggregate study) – WPIS Tavg scale 28. SCR-DR-6169 (Parts 1, 3, 5) – Redundant Sensors Algorithm (RSA) NAP Does Not Process Failed
Channels Correctly 29. SCR-DR-6170 (Not part of aggregate study) – Radioactive Waste Drain (WRS) graphic issue 30. SCR-DR-6180 (Not part of aggregate study) – Trend for Time to Boil unit indication 31. SCR-DR-6187 (Not part of aggregate study) - Rod sequence skips steps
32. SCR-DR-6259 (Part 3)1 - Unexpected Bank Sequence Out of Sequence Alarm
33. SCR-DR-6267 (Parts 3, 5, 6)1 - Urgent Alarm (Causes control rods to swap to manual and stop) Occurs During Case 2 CRE
34. SCR-DR-6278 (Not part of aggregate study) – Battery bank indications mislabeled for EDS1, EDS2, and EDS4
35. SCR-DR-6302 (Parts 2, 3, 5)1 - Improper Bank Overlap Occurs when Data Point OCB07CE00C_OUTAV Increments Improperly During Startup
36. SCR-DR-6315 (Parts 1, 2, 3, 4, 5)1 - Manual Reactor Trip Alarm Occurred without a Reactor Trip Request 37. SCR-DR-6398 (Part 4) - Excitation Transformer Graphic Issue
38. SCR-DR-6409 (Part 4) - Graphic 1805 has reversed rods
39. SCR-DR-6621 (Parts 3, 4, 5)1 - RSA NAP for Power Range Power does not Eliminate Erroneous Input
40. SCR-DR-6651 (Part 3)1 - Inconsistent Digital Processing Unit (DPU) Alarm Priority Levels 41. SCR-DR-6698 (Parts 3, 4) - Safety Mimic Display for SGS-V255A& B Indicates Bad Quality Following a
Startup Feedwater Isolation
42. Alarm Server update with alarm prioritizations. This update resolves HED issues such as the CCS surge
tank leak going unidentified due to excessive alarms1
1These SCR-DRs were identified as impacting 10 CFR 55.45 in the aggregate.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
14 | P a g e
10 CFR 55.45(a)(1) Perform pre-startup procedures for the facility, including operating of those controls associated with plant equipment that could affect reactivity.
Executive Summary
The team determined that there was not anything in this section that could not be handled by a standard three (3) person control room crew. The individual SCRs listed here do not influence each other with respect to responding to simulator scenarios. From a procedural standpoint, they could be evaluated effectively and do not impact responding to simulator scenarios.
The team determined there are no negative aggregate impacts from these issues that will affect operator training or operations in the simulator.
In terms of examination there is still enough of a representative sample that could test operator effectiveness. There remains a large population of JPM tasks which could be combined to evaluate this otherwise.
Reactor Coolant Pump (RCP) Net Positive Suction Head (NPSH) Curve has Inadequate Range for Operation
SCR-DR-5577 This issue impacts the following RO/SRO task: RO-PRI-RCS-005-00 Operate the RCS during shutdown/cooldown conditions
Disposition
This issue was dispositioned as acceptable by a Subject Matter Expert (SME). The current display requires
finer pressure control by the operators, but the procedures can still be used to successfully accomplish the task.
Description
The Reactor Coolant Pump (RCP) minimum Net Positive Suction Head (NPSH) display (60029) shows the required Reactor Coolant System (RCS) pressures for given RCS temperatures for starting RCPs. The display shows the limits based on the instruments used for indication, RCS Wide Range (WR) pressure (RCS-PT140A/B/C/D) or Normal Residual Heat Removal System (RNS) pump suction pressure (RNS- PT011A/B).
The current display only indicates the RNS limits below 275oF. Per RCS Component Control Requirements (APP-RCS-M3C-100 Rev. 9) logic sheets RCS-13 and RCS-18, the pressure indication used for determining NPSH should be [ ]a,c. Per procedure, RNS is placed on
service when [ ]a,c.
During the subsequent cooldown and depressurization using RNS, this display does not indicate the larger
margin allowed for NPSH using the RNS suction pressure above 275oF as the margin to NPSH limits using the RCS WR pressure instruments is very small at these lower temperatures and pressures.
Area of Impact
Decay heat removal (forced circulation) during startup/shutdown
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
15 | P a g e
Rod Withdrawal button deselects During Continuous Operation
SCR-DR-5584 This issue impacts the following RO/SRO task: RO-INC-PLS-003-02 Monitor the Control Rod Drive System
Disposition
Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description
While performing extended rod withdrawals during startups, depressing the rod withdrawal button (UP ARROW) may cause the UP ARROW button to un-highlight and momentarily flash gray even though still depressed. Rod motion will still occur.
Area of Impact
Reactivity Management
Issue with Automatic Control of Deaerator Storage Tank (DST) level and Auto Start of Standby Condensate Pump
SCR-DR-5655 This issue impacts the following RO/SRO task: RO-SEC-FWS-002-03 Establish level in the DST
Disposition
This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). The reasoning was that the current controller tuning is at the best estimated values. Hot functional testing would provide as- built tuning. However, this issue did contribute to the aggregate impact of sections 3 and 5 with regards to secondary plant control. Description
During plant startup from Mode 5 to 100% power, the Condensate system pressure would lower to the auto start setpoint of the standby Condensate pump. This pressure drop is due to Condensate System (CDS) valves, CDS-V022 and CDS-V025 modulating to maintain level in the Deaerated Storage Tank (DST). In accordance with reference plant procedures for normal operation, the second condensate pump is started at 40-45% power. However, the second condensate pump will have already auto started in the heatup and startup procedures, due to the slow response of CDS-V022 and CDS-V025.
Area of Impact
Plant design deficiency impacts operations during startup
Model Instability during pressurizer (PZR) Fill to Solid (no vapor bubble remaining)
SCR-DR-5698 This issue impacts the following RO/SRO task: RO-PRI-CVS-003-04 Operate the Chemical and Volume Control System to control the primary system pressure in water solid mode
Disposition
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
16 | P a g e
This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs
Analysis was performed under Condition Report (CR) 10000465. Training Needs Analysis determined training involving establishment of solid plant should not be performed until issue is corrected. A review of current training material did not reveal any scenarios where this was required.
Added to the DR Global Issues list, this will be briefed to the students at the beginning of the Simulator portion of training. Scenario AP-LT-I-SIM-GOPSDCD (Covering GOP-205, Plant Cooldown MODE 3 to
MODE 5) does not train on Solid Plant Operations.
Description
The Liquid Radwaste System (WLS) model is prone to failure during evolutions involving near solid pressurizer operations if the Effluent Holdup Tank is filled too rapidly.
Area of Impact
Difficulty in achieving solid plant operations continuously
OPDMS Rod Insertion Limit (RIL) Indication Does Not Align to Combined Operating Limits Report (COLR) Rev. 0
SCR-DR-5736 This issue impacts the following RO/SRO task: RO-INC-PLS-002-03 Control Rods: Verify Sequence and Overlap within COLR specifications
Disposition
This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs Analysis was performed under CR 10000474. The following is taken from the analysis performed. The Training Needs Analysis determined that the M1 Bank Insertion Low-2 alarm would be received prior to the M2 Bank Insertion Low-2 alarm due to rod sequencing and bank overlap. The M1 Bank alarm is set at the correct value.
Description
It is noted that the Core Operating Limits Report (COLR) Rev. 0 Rod Insertion Limit (RIL) for M2 bank at all power levels is [ ]a,c Steps (fully withdrawn). The Online Power Distribution Monitoring System (OPDMS) Rod Insertion Limit (RIL) display ([ ]a,c) indicates the RIL for M2 bank is [ ]a,c steps. Further investigation indicates the M2 RIL indication high limit is [ ]a,c steps and therefore cannot indicate above this level (determined using point information page instrumentation tab for RB-INSERT- M2LIM.SV3@NET0). All Shutdown (SD) bank indications are capable of indicating a maximum of [ ]a,c steps.
Area of Impact
Reactivity Management with regards to indication
Decay Heat Calculation Summary - Assembly Move NAP Function Not Functional
SCR-DR-6022
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
17 | P a g e
This issue impacts the following RO/SRO task: RO-INC-PLS-004-03 Perform decay time surveillance
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the minimal training impact this particular Nuclear Application (NAP) has.
Description When attempting to simulate fuel assemblies being moved from the core to the Spent Fuel Pool (SFP), it was noted that the Decay Heat Calculation (DHC) NAP to maintain the administrative location of fuel does not work correctly. On display 40203 the assembly move buttons on the lower right portion indicate they are only available when the light DDS-AP-DHC Status indicates it is ACTIVE. This light is driven by the automatic mode selector and is INACTIVE when in MODES 1&2 and ACTIVE in MODES 3- 6. However, when the light indicates INACTIVE the buttons for moving are raised and available. When the light changes status to ACTIVE the buttons for moving are grayed out and no longer available. The light being active or inactive is currently driven by the auto mode selector and becomes active in MODES 3-6. However, fuel cannot be moved from the core into the SFP in any MODE other than MODE 6. The light should be driven by the manual input of the Rx vessel head being removed or installed or upper internals position on display 40004.
Area of Impact Reactivity Management with regards to indication and administration
M Control Rod Banks B & C Reversed on DRPI Health Screen
SCR-DR-6030 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The DRPI Health Screen (1805) control rod banks M-B and M-C have the wrong rods listed as being in each bank. The rods listed as being M-B are actually M-C and the rods listed for M-C are the M-B rods. The correct arrangement of rods is shown in APP-RXS-M3-001 Rev 4 Figure 4-1 as well as on the DRPI M Bank screen (11172).
Area of Impact Plant indications
Liquid Radwaste System WLS-MP-08C improperly Pumps Monitor Tank C
SCR-DR-6068 This issue impacts the following RO/SRO task: RO-SUP-WLS-002-00 Operate the WLS
Disposition
Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
18 | P a g e
Description While performing a startup from Mode 6 it was discovered that the Liquid Waste System (WLS) WLS- MP-08C will not pump Monitor tank C around 37 inches. The pump will turn on and occasionally the downstream check valve will throttle open and shut but there is little or no evidence of flow. Also, discharge pressure never goes above 12-13psig. Normal discharge pressure for the other monitor tank pumps is around [ ]a,c.
Note that it does pump when level is above 37 inches as the tank has been pumped down to 37 inches successfully. It appears to exhibit strange behavior at 37 inches and below.
Area of Impact Correct operation of plant systems
Excessive Startup Feedwater (SFW) Control Valve Cycling
SCR-DR-6151 This issue impacts the following RO/SRO task: RO-SEC-FWS-002-01 Monitor SFWS and MFWS system and component parameters
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). The reasoning was that the current controller tuning is at the best estimated values. Hot functional testing would provide as- built tuning. However, this issue did contribute to the aggregate impact of sections 3 and 5 with regards to secondary plant control.
Description At low pressure conditions less than 350 psig, the operator often has to take manual control of Startup Feed Water (SFW) control valves due to excessive cycling of the valves. Indicated flow rates oscillate erratically between no flow and max flow every 10 to 15 seconds. This requires 100% of the operator’s attention until RNS can be placed into service removing cooldown function from the steam dumps.
Area of Impact Plant control during startup and shutdown
Redundant Sensors Algorithm Application NAP Does Not Process Failed Channels Correctly
SCR-DR-6169 This issue impacts the following RO/SRO task: RO-INC-PMS-003-00 Monitor System Indications
Disposition
Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The Redundant Sensors Algorithm Application (RSA) driven source range counts on the WPIS displays (main, trends, and safety functions) will still reflect an abnormally high value for source range power after a source range channel failure. The RSA NAP should account for the failure and remove it from the calculation.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
19 | P a g e
Area of Impact Reactivity management
Manual Reactor Trip Alarm Occurred without a Reactor Trip Request
SCR-DR-6315 This issue impacts the following RO/SRO task: RO-INC-PMS-005-02 Respond to a loss of a
Protection and Safety Monitoring System (PMS) division or failure of PMS components
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The below failure sequence was performed and the reactor did not trip, but a Reactor Trip alarm was received. The alarm was for a Manual Reactor Trip, but a manual Rx Trip was not inserted. A P-4 was not received.
• RCS TE122C -> Open Circuit: PMS indicates Low-2 Tcold Cold Leg 2 Bistable Trip with Bad Quality
& Maintenance Bypass for Division C & PMS Cabinet Fault Alarm for Division C • RCS TE122D -> Open Circuit: PMS indicates Low-2 Tcold Cold Leg 2 Bistable Trip with Bad Quality
& Maintenance Bypass for Division D with other alarms
• Cold Leg 2 Temperature Low-2 Bypass inserted for Division A.
• Open the circuit for ECS-TE121B
Manual Reactor Trip Alarm (PMS-RXTR-MA-X0) actuate though none of the PMS divisions indicated that a Manual Reactor Trip had been inserted. PMS-J3-308 shows that the PMS-RXTR-MA alarms should only be activated by the Reactor Trip Switches at the PDSP or the RSR.
Area of Impact Reactivity Management
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
20 | P a g e
10 CFR 55.45(a)(2) Manipulate the console controls as required to operate the facility between shutdown and designated power levels.
Executive Summary
Rod control issues during startup were grouped together for analysis (5736, 6089, 6102 and 6302). The letdown heat exchanger issue will impact startup due to its need to be manipulated by the Balance of Plant (BOP) operator at the same time (6158). Crews that plan ahead will be prepared for this combination of issues but the study also looked at a crew that isn’t planning ahead appropriately. There are some procedural controls in place to prevent these issues from manifesting, but the need to continue to strengthen our procedures remains.
All rod control issues in this section will apply during plant startup. Analysis took into account if the controls manipulated during startup with these rod control issues encroach on the ability to successfully manage simulator scenarios. The inverse count-rate ratio (1/M) plot may be performed manually by procedure so that wasn’t determined an issue. Issues with the Rod Insertion Limit screen are mitigated because the COLR has precedence over a graphic. Operators understand that the COLR is the definitive document on Rod Insertion Limits.
A concern with this particular combination of issues potentially disrupting operational analysis, decision making, and action was mitigated during the analysis by a belief that the crews will be able to handle these issues effectively (rod control group with letdown heat exchanger issue).
The issues with rods out of sequence alarms occurring were analyzed for operation impact during a startup. If a startup is occurring and a rod out of sequence alarm actuates, the operator may just stop and say that there won’t be a rods out of sequence during this start up and subsequently commence a shutdown and retry. Since this is an identified issue, the instructor would have to intervene to continue the startup.
For the aggregate study, an assumption was made that these issues will not manifest simultaneously or in a combination such that the students cannot dissipate the alarms in the proper order without putting rods to manual. This is backed up by nearly a year of simulator operation. These issues have been identified one at a time over a period of time. There is a need for additional reinforcement of the skills required for the task due to these issues. The instructors will need to provide additional information to the students in order to effectively deal with rods out of sequence alarms in the course of the licensed operator training program. The study recognized the risk, but determined additional skill reinforcement will allow crews to successfully manage simulator scenarios.
Rod Withdrawal button deselects During Continuous Operation
SCR-DR-5584 This issue impacts the following RO/SRO task: RO-INC-PLS-003-02 Monitor the Control Rod Drive System
Disposition
Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
21 | P a g e
Description
While performing extended rod withdrawals during startups, depressing the rod withdrawal button (UP ARROW) may cause the UP ARROW button to un-highlight and momentarily flash gray even though still depressed. Rod motion will still occur.
Area of Impact Reactivity Management
Issue with Automatic Control of DST level and Auto Start of Standby Condensate Pump
SCR-DR-5655 This issue impacts the following RO/SRO task: RO-SEC-FWS-002-03 Establish level in the DST
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). The reasoning was that the current controller tuning is at the best estimated values. Hot functional testing would provide as- built tuning. However, this issue did contribute to the aggregate impact of sections 3 and 5 with regards to secondary plant control.
Description During plant startup from Mode 5 to 100% power, the Condensate system pressure would lower to the auto start setpoint of the standby Condensate pump. This pressure drop is due to Condensate System (CDS) valves CDS-V022 and CDS-V025 modulating to maintain level in the Deaerated Storage Tank (DST). In accordance with reference plant procedures for normal operation, the second condensate pump is started at 40-45% power. However, the second condensate pump will have already auto started in the heatup and startup procedures, due to the slow response of CDS-V022 and CDS-V025.
Area of Impact Plant design deficiency impacts operations during startup.
OPDMS RIL Indication Does Not Align to COLR Rev. 0
SCR-DR-5736 This issue impacts the following RO/SRO task: RO-INC-PLS-002-03 Control Rods: Verify Sequence and Overlap within COLR specifications
Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs Analysis was performed under CR 10000474. The following is taken from the analysis performed. The Training Needs Analysis determined that the M1 Bank Insertion Low-2 alarm would be received prior to the M2 Bank Insertion Low-2 alarm due to rod sequencing and bank overlap. The M1 Bank alarm is set at the correct value.
Description It is noted that the Core Operating Limits Report (COLR) Rev. 0 Rod Insertion Limit (RIL) for M2 bank at all power levels is [ ]a,c Steps (fully withdrawn). The Online Power Distribution Monitoring System
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
22 | P a g e
(OPDMS) Rod Insertion Limit (RIL) display ([ ]a,c) indicates the RIL for M2 bank is [ ]a,c steps. Further investigation indicates the M2 RIL indication high limit is [ ]a,c steps and therefore cannot indicate above this level (determined using point information page instrumentation tab for RB-INSERT- M2LIM.SV3@NET0). All SD bank indications are capable of indicating a maximum of [ ]a,c steps.
Area of Impact Reactivity Management with regards to indication
NAP for 1/M Intermediate Range Not Functional
SCR-DR-6089 This issue impacts the following RO/SRO task: RO-PRO-GEN-014-00 Perform an Inverse Count Rate Plot using GOP-307
Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs Analysis was performed. It also was determined to impact the third section of 55.45 section 3 in the aggregate. The functionality of the Intermediate Range (IR) Inverse Count Rate Ratio Nuclear Application has been fixed such that it is usable by operators but additional work will be performed by WEC to restore this to full use. WEC RITS 38306 for tracking.
Description The intermediate range 1/M plot Nuclear Application (NAP) does not work. Once P-6 (Permissive 6) was blocked and source range de-energized, the operator no longer had a 1/M plot generated.
Area of Impact Reactivity Management
MA Bank Rods Sometimes Stop at 263 steps during a CRE
SCR-DR-6102 This issue impacts the following RO/SRO task: RO-INC-PLS-003-06 Perform grey rod exchange
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The plant control system operating procedure allows for a case 1 Control Rod Exchange in the event that Case 2 is not functioning. With MA rods at 234 steps and AO rods in manual at 218 steps, the AO rods were stepped into [ ]a,c steps, which will drive MA out. The MA rods stopped at 263 steps, with an outward demand still in. The audible rod clicking stopped as well.
If the CRE is still continued in accordance with the procedure, MD will step into the core and MA will remain at [ ]a,c. This generates a “Rods out of sequence alarm”.
Area of Impact
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
23 | P a g e
Reactivity Management
Excessive SFW Control Valve Cycling
SCR-DR-6151 This issue impacts the following RO/SRO task: RO-SEC-FWS-002-01 Monitor SFWS and MFWS system and component parameters
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). The reasoning was that the current controller tuning is at the best estimated values. Hot functional testing would provide as- built tuning. However, this issue did contribute to the aggregate impact of sections 3 and 5 with regards to secondary plant control.
Description At low pressure conditions less than 350 psig, the operator often has to take manual control of Startup Feed Water (SFW) control valves due to excessive cycling of the valves. Indicated flow rates range from 0 to greater than 600 gpm within 10 to 15 second cycles. This requires 100% of the operator’s attention until RNS can be placed into service removing cooldown function from the steam dumps.
Area of Impact Plant control during startup and shutdown
Audible Rod Step Skips
SCR-DR-6186 This issue impacts the following RO/SRO task: RO-INC-PLS-003-01 Monitor the reactor power control system
Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). A training needs analysis was performed under CR 10025690.
Explain/Brief students prior to beginning a simulator training phase or segment. Update a “SIMULATOR TRAINING STUDENT HANDOUT” (example is attached) and file in the “Operator Aids” notebook. Reference the SIMULATOR TRAINING STUDENT HANDOUT in each sum guide.
Description
During outward rod motion, the audible step counter randomly can have an extra second pause in it with rod motion continuing. The step counter indication does not update at the same rate as the audible cue occurs.
Area of Impact Reactivity management
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
24 | P a g e
Improper Bank Overlap Occurs when Data Point OCB07CE00C_OUTAV Increments Improperly During Startup
SCR-DR-6302 This issue impacts the following RO/SRO task: RO-INC-PLS-002-03 Control Rods: Verify Sequence and Overlap within COLR specifications
Disposition Subsequent to the performance of the Aggregate Study this item was determined to be invalid. This behavior is per current Westinghouse design.
Description An Ovation data point in rod control is initially set to 0 when the Digital Rod Control System (DRCS) is reset. However, during SD1 withdrawal, OCB07CE00C_OUTAV (the Ovation data point) will increment to a value of 2 and then stay at this value. This data point is only supposed to increment for inward rod motion during M bank rod movement. The end result is that bank overlap will be incorrect if not manually corrected in Ovation.
Area of Impact Reactivity Management
Manual Reactor Trip Alarm Occurred without a Reactor Trip Request
SCR-DR-6315 This issue impacts the following RO/SRO task: RO-INC-PMS-005-02 Respond to a loss of a PMS
division or failure of PMS components
Disposition
Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description The below failure sequence was performed and the reactor did not trip, but a Reactor Trip alarm was received. The alarm was for a Manual Reactor Trip, but a manual Rx Trip was not inserted. A P-4 was not received.
• RCS TE122C -> Open Circuit: PMS indicates Low-2 Tcold Cold Leg 2 Bistable Trip with Bad Quality &
Maintenance Bypass for Division C & PMS Cabinet Fault Alarm for Division C • RCS TE122D -> Open Circuit: PMS indicates Low-2 Tcold Cold Leg 2 Bistable Trip with Bad Quality &
Maintenance Bypass for Division D with other alarms • Cold Leg 2 Temperature Low-2 Bypass inserted for Division A. • Open the circuit for ECS-TE121B
Manual Reactor Trip Alarm (PMS-RXTR-MA-X0) actuate though none of the PMS divisions indicated that a Manual Reactor Trip had been inserted. PMS-J3-308 shows that the PMS-RXTR-MA alarms should only be activated by the Reactor Trip Switches at the Primary Dedicated Safety Panel (PDSP) or the Remote Shutdown Workstation (RSR).
Area of Impact
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
25 | P a g e
Reactivity Management
10 CFR 55.45(a)(3) Identify annunciators and condition-indicating signals and perform appropriate remedial actions where appropriate.
Executive Summary
The 1/M NAP issues (DR-6089) was determined to be mitigated by the manual performance option. The issues with NAP Surveillance screen (DR-6159) is not a screen any of the operators use or are trained to use.
Excessive valve modulating status alarms (DR-5813) challenges responding to simulator scenarios by the operators. There are overabundances of these alarms which come in that require operator attention. Additionally, there is a risk of desensitizing operators to alarms which may ultimately be important in the plant during testing, but are not important in the simulator during training. The Initial Test Program will occur in the plant and not the simulator.
The indication issues and the alarm issues combined are significant enough to conclude that they will impact operation in the aggregate. The magnitude and influx of alarms and indications (especially when the indications and the alarms are not aligned) is too great to mitigate. Training may be able steer the operators to properly prioritize the information, but ultimately this is not a preferred way to train. Operators can’t be permitted to operate differently in the plant than the simulator. This is unacceptable.
Containment Cooling System (VCS) fan response due to loss of power
SCR-DR-216 This issue impacts the following RO/SRO task: RO-VNT-VCS-002-01 Monitor the VCS parameters
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the existence of a design change CAPAL 100044029 and current documentation states that the behavior is correct.
Description During normal operations, the “A” and “B” Containment Cooling System (VCS) fans are running in fast speed. Under a loss of power condition, the “A” VCS fan will trip and the “C” VCS fan should automatically start in fast speed at a low flow setpoint on VCS-FT010C. This was not observed on the STS. The “A” VCS fan does trip, but the “C” VCS fan does not auto start at a low flow. Flow indication, VCS-FT010C reads 0 cfm.
Area of Impact Containment cooling operations
EDS Power Supply Assignments to PLS/DDS Cabinets Incomplete
SCR-DR-5546 This issue impacts the following RO/SRO tasks: RO-LT-R-EDS.001 Monitor the Non Class 1E DC and UPS system (EDS) for proper operation
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
26 | P a g e
RO-LT-R-EDS.004 Respond to a loss of EDS DC power Disposition
This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). SMEs determined that the current power supply arrangement was adequate to teach since it is per design documentation. Power supplies are an item that will be continuously taught as they are updated and changed.
Description A loss of individual Non Class 1E DC and UPS System (EDS) busses will result in incomplete system response. Some Ovation drops (computers and other equipment that are part of the plant computer system) are not dynamically powered by the EDS model but are powered by a permanently energized model constant (specifically DPU047, DPU048, and DPU044). The load lists for the STS do not assign a power supply to all the Ovation drops so there is no plant design data to insert into the simulator.
Area of Impact Effective plant response to loss of power
Modeled BEACON Data Cannot Determine Quality
SCR-DR-5583 This issue impacts the following RO/SRO task: RO-INC-IIS-004-00 Determine functionality of the On-line Power Distribution Monitoring System
Disposition SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. Operator training simulator guides inform the training instructors on whether BEACON is operable or not operable. (Note: This evaluation is actually based on whether or not BEACON if functional or not functional. BEACON is not safety-related and does not have applicable Technical Specifications.)
Description Failure of the BEACON Data Processing (BDP) NAP causes the manual override signal originated by BDP to have BAD quality as expected. The BAD indication is passed through the BEACON operability calculations in the Plant State Monitoring (PST) NAP and appears on the OPDMS displays as 'operable' with BAD quality. Failure of the BDP application will cause BEACON to indicate ‘inoperable’ in the reference unit. However, the current STS scope of limitation has the BEACON outputs driven by the core model. There is currently no ability to pass quality over the interface for outputs. Inputs are not taken from the BDP NAP but from other plant process data. The core model does not know the status of the BDP NAP. The core model does not pass operability information.
Area of Impact Plant control
Rod Withdrawal button deselects During Continuous Operation
SCR-DR-5584 This issue impacts the following RO/SRO task: RO-INC-PLS-003-02 Monitor the Control Rod Drive System
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
27 | P a g e
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description While performing extended rod withdrawals during startups, depressing the rod withdrawal button (UP ARROW) may cause the UP ARROW button to un-highlight and momentarily flash gray even though still depressed. Rod motion will still occur.
Area of Impact
Reactivity Management
Containment Radiation Alarm Reset Points Incorrect
SCR-DR-5597 This issue impacts the following RO/SRO task: RO-INC-RMS-003-07 Startup and operate a containment high-range area radiation monitor (safety related)
Disposition Acceptance criteria not fully met. SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. This is an alarm deadband issue that WEC must resolve. If operators encounter this condition, they will follow their procedures. The procedure provides the steps necessary for operators to respond to the condition.
Description The current high setpoint and deadband combinations for Passive Core Cooling System (PXS) PXS-RY160, RY161, RY162, and RY163 do not allow for the High-1 and High-2 alarms to clear.
Area of Impact Radiation control
Pressurizer Heater Current Indicates BAD Quality at Limits
SCR-DR-5598 This issue impacts the following RO/SRO task: RO-PRI-RCS-003-03 Operate the pressurizer level control system in manual and automatic
Disposition This issue was dispositioned as acceptable by the Subject Matter Expert (SME). This issue was dispositioned as acceptable because the SME determined that the issue did not impact any operator actions during training or examination.
Description RCS-EH-04A-1-AMPB indicates 0 and BAD quality when the PZR backup heaters are off. It indicates 100 and BAD quality when the backup heaters are on. This applies for 04B, 04C, 04D points as well. All points can be viewed on Ovation display 33001.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
28 | P a g e
The control heaters (RCS-EH-03-1-AMPB) will display BAD quality at a value of 0 when they are off. These heaters are typically at some value other than 0 or 100 and the quality is good, however, it seems to have the same issue as the backup heaters at the limits.
Area of Impact Plant control
Unidentified and Identified Leak Rate Always Indicates BAD Data
SCR-DR-5599 This issue impacts the following RO/SRO task: RO-PRO-AOP-053-00 Respond to a Reactor Coolant Leak using AOP-112 Disposition
After performing V&V testing, SNC determined that the update was not successful. The Leak Rate Monitoring Application is informational only and does not drive any alarms based upon the calculated leakage. For this reason, any leak rate calculations would have to be performed manually per plant procedures vice using the NAP calculated values.
Description The Leak Rate Monitor (LRM) has BAD quality point indication for the Identified and Unidentified leak rates. They never change to good quality and indicate BAD when using the on demand leak rate calculation.
Area of Impact Plant control
Low Flow Alarm on TCS-FT007 Occurs Earlier than Expected
SCR-DR-5603 This issue impacts the following RO/SRO task: AP-LT-R-TCS.003 Monitor the TCS
Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs Analysis was performed under CR 10000466. Information sharing of the discrepancy was selected as the solution to this issue.
Description A low flow alarm occurs on Turbine Building Closed Cooling Water System (TCS) TCS-FT007 at approximately 80% power during a down power evolution which is earlier than SMEs expected.
Area of Impact Plant control
MFP 'B' Alarm Response Differs For Identical Fault
SCR-DR-5613 This issue impacts the following RO/SRO task: RO-SEC-FWS-002-01 Monitor SFWS and MFWS system
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
29 | P a g e
and component parameters Disposition
Subsequent to the performance of the Aggregate Study, this item was reinvestigated at Westinghouse’ request and could not be replicated after multiple attempts. SNC performed additional investigation and determined that the original SCR entry was invalid (SNC was the entity from whom this issue originated). This item has been dispositioned as invalid and closed.
Description
A spurious trip of Main Feed Pump (MFP) 'B' responds differently than the on MFPs 'A' or 'C'. A spurious trip of MFP 'A' or 'C'’s respective supply breaker will cause three alarms: FWS-FT011A/C, FWS-FT012A/C and Feedwater Pump Control Status 1. The flow transmitter alarms are automatically taken to Cutout (CO) by the Alarm Presentation System (APS) and removed from the typical screens available to the operator but the control status alarm remains. When MFP 'B' has a spurious trip inserted on the supply breaker, 2 alarms are received: FWS-FT011B and FWS-FT012B. Since both of these alarms are automatically taken to CO the only indication is an audible noise from APS with no visual cue as to what caused the alarm. This results in a spurious trip of MFP 'B' supply breaker indicating exactly the same as if an operator took the controller to STOP.
Area of Impact
Plant control
Unexpected Response of Alarm Cutout of RWS Pressure Alarms
SCR-DR-5621 This issue impacts the following RO/SRO task: AP-RO-ADM.015 Respond to alarms using the Alarm Presentation System (APS)
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). The SMEs determined that adding this issue to the known issues list was adequate. The student would be aware of the issue’s existence if it ever manifested and there are a very few number of pumps which cause this to occur.
Description The Raw Water System (RWS) pump discharge pressures P003A/B/C (associated with pumps RWS-MP- 02A/B/C) each have a Low-1 alarm setpoint at 66 psig that is only active when the pump is running (i.e., alarm is cutout when the associated pump is off). However an audible nuisance alarm occurs whenever the operators take normal action on these pumps.
This “ghost alarm” occurs also during normal operation of CCS-MP-01A/B.
Area of Impact Plant control
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
30 | P a g e
Pressurizer Pressure Out of Range Indication Not Properly Displayed
SCR-DR-5623 This issue impacts the following RO/SRO task: RO-PRI-RCS-003-02 Operate the RCS/pressurizer pressure master controller in manual and automatic
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the isolated nature of the issue. However, this issue was found to impact operator indications in the aggregate.
Description The Primary Dedicated Safety Panel (PDSP) Reactor Coolant System (RCS) Parameters of Pressurizer Pressure (PT191A-D) stops lowering at [ ]a,c, but a low out of range arrow does not display indicating bottom of range.
Area of Impact Plant control
Subcriticality Indication on Critical Safety Function Screen Drops to Bad Quality
SCR-DR-5627 This issue impacts the following RO/SRO task: RO-PRO-EOP-031-00 Implement and evaluate Critical Safety Function Status trees using CSF-F-0
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The Nuclear Application (NAP) point DDS-SPD31-X0 randomly cycles to magenta bad quality due to DDS- RSA3J-J1 driving into bad quality. The result of the former point going bad quality is subcriticality and reactivity control display on the Critical Safety Function Wall Panel Information System (WPIS) screen goes magenta.
Area of Impact Reactivity Management with regards to indication
Issue with Automatic Control of DST level and Auto Start of Standby Condensate Pump
SCR-DR-5655 This issue impacts the following RO/SRO task: RO-SEC-FWS-002-03 Establish level in the DST
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). The reasoning was that the current controller tuning is at the best estimated values. Hot functional testing would provide as- built tuning. However, this issue did contribute to the aggregate impact of sections 3 and 5 with regards to secondary plant control.
Description During plant startup from Mode 5 to 100% power, the Condensate system pressure would lower to the
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
31 | P a g e
auto start setpoint of the standby Condensate pump. This pressure drop is due to Condensate System (CDS) valves CDS-V022 and CDS-V025 modulating to maintain level in the Deaerated Storage Tank (DST). In accordance with reference plant procedures for normal operation, the second condensate pump is started at 40-45% power. However, the second condensate pump will have already auto started in the heatup and startup procedures, due to the slow response of CDS-V022 and CDS-V025.
Area of Impact Plant design deficiency impacts operations during startup
Degasifier Level Alarm Limits
SCR-DR-5686 This issue impacts the following RO/SRO task: RO-SUP-WLS-002-00 Operate the WLS
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal impact on training. Description CVS-M3C-100 Rev 8 states that Liquid Radwaste System (WLS) WLS-LICA-016 High-3 is the degasifier level setpoint that controls the operation of letdown. The control circuit is working as described, however the Point Information Limits show only High-1 has a value of 85”.
Area of Impact Plant control
PMS Mimic Screens
SCR-DR-5689 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS (DDS is Data Display and Processing System)
Disposition
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. This is a simulator I&C issue. PMS mimic screens are used for verification of indications only. If a question arises regarding an indication on the PMS mimic screen, operators will use primary indications from the PMS displays on the Primary Dedicated Safety Panel (PDSP). No operator action is available through the PMS mimic screens. All actions must be taken from the division’s PMS PDSP. Description Protection and Safety Monitoring System (PMS) Mimic screens on Ovation do not reflect what is shown on the associated Primary Dedicated Safety Panel (PDSP). This is especially true when there is any FAULT condition shown on the Primary Dedicated Safety Panel (PDSP).
Area of Impact Plant control
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
32 | P a g e
OPDMS RIL Indication Does Not Align to COLR Rev. 0
SCR-DR-5736 This issue impacts the following RO/SRO task: RO-INC-PLS-002-03 Control Rods: Verify Sequence and Overlap within COLR specifications
Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs Analysis was performed under CR 10000474. The following is taken from the analysis performed. The Training Needs Analysis determined that the M1 Bank Insertion Low-2 alarm would be received prior to the M2 Bank Insertion Low-2 alarm due to rod sequencing and bank overlap. The M1 Bank alarm is set at the correct value.
Description It is noted that the Core Operating Limits Report (COLR) Rev. 0 Rod Insertion Limit (RIL) for M2 bank at all power levels is [ ]a,c Steps (fully withdrawn). The Online Power Distribution Monitoring System (OPDMS) Rod Insertion Limit (RIL) display ([ ]a,c) indicates the RIL for M2 bank is [ ]a,c steps. Further investigation indicates the M2 RIL indication high limit is [ ]a,c steps and therefore cannot indicate above this level (determined using point information page instrumentation tab for RB-INSERT- M2LIM.SV3@NET0). All SD bank indications are capable of indicating a maximum of [ ]a,c steps.
Area of Impact Reactivity Management with regards to indication
Nuisance Valve Modulating Status Alarms
SCR-DR-5813 This issue impacts the following RO/SRO task: AP-RO-ADM.015 Respond to alarms using the Alarm Presentation System (APS)
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The following valves modulations alarms cause an excessive nuisance.
• Pressurizer Spray Valves (RCS-V110A & B)
• Main Feedwater Control Valves (SGS-V250A & B)
• Startup Feedwater Control Valves (SGS-V255A & B)
Area of Impact Plant Control
Unexpected VRS High Rad Alarm
SCR-DR-5828
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
33 | P a g e
This issue impacts the following RO/SRO task: RO-VNT-VRS-005-01 Respond to Radwaste building HVAC system alarms
Disposition
Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description When ES-1 is re-energized by the diesel sequencer after a loss of power incident, the Radwaste Building HVAC System (VRS) VRS-RY023 high radiation alarm will actuate. During this actuation, process flow through VRS is unavailable per design. Only the detector receives power.
Area of Impact Radiation Control
VFD Transformer Temperature
SCR-DR-5910 This issue impacts the following RO/SRO task: RO-PRI-RCS-005-01 Operate RCP VFDs
Disposition
Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description ECS-EV-X1-TMPC points associated with the Variable Frequency Drive (VFD) hottest cell parameters are not being driven by the models. They are a constant value.
Area of Impact Plant Operations
Print Feature from NAP non-functional
SCR-DR-5913 This issue impacts the following RO/SRO task: AP-RO-ADM.020.07 Document surveillance test in log book
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal training impact as this simply means the automated system is not capable of being used to complete surveillance requirement testing, operators are still capable of using the paper copies.
Description While performing surveillance procedure "Incore Detector Comparison to Nuclear Instrument Channel Axial Flux Difference", the select PRINT SURVEILLANCE REPORT does not result in a printout.
Area of Impact Plant response – this prevents or impacts the performance of most Surveillance Tests
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
34 | P a g e
VHS Rad Monitor Response to Loss of Process Flow
SCR-DR-5914 This issue impacts the following RO/SRO task: RO-VNT-VHS-005-01 Respond to health physics and hot machine shop HVAC alarms
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description Health Physics and Hot Machine Shop HVAC System (VHS) VHS-RE001 goes up by 4 decades in 10 minutes on a loss of process flow. This gives a Priority 1 alarm on VHS and RADIATION MONITORING.
Area of Impact Radiation Control
Pressurizer Narrow Range Pressure Does Not Indicate Bottom of Scale
SCR-DR-5920 This issue impacts the following RO/SRO task: RO-PRI-RCS-003-02 Operate the RCS/pressurizer pressure master controller in manual and automatic
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description With RCS pressure less than [ ]a,c, the Wall Panel Information System (WPIS) for Mode 1-4 does not indicate the instrument is bottom of scale via graphical down arrow.
Area of Impact Plant Control
CDS-TE040A/B Range is Inadequate
SCR-DR-5921 This issue impacts the following RO/SRO task: RO-SEC-BDS-005-07 Verify automatic blowdown isolation upon high-2 temperature in heat exchanger shell outlet (CDS fluid) or high-2 DST level
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) since an open item exists in the design documentation.
Description When the condensate outlet of the blowdown heat exchanger temperature element is failed high the blowdown flow remains un-isolated. The high-2 temperature ([ ]a,c) should isolate blowdown flow in accordance with APP-CDS-M3C-101 Rev 3. However, the range of the instrument (CDS-T-040A/B) listed in APP-CDS-M3C-101 Rev 3 is [ ]a,c which would never allow blowdown isolation on high temperature.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
35 | P a g e
Area of Impact Plant Control
DRPI Health Screen Alarms for Data Cabinet A and B Crossed
SCR-DR-5924 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to not being a frequently used indication and the associated alarm points functioning properly.
Description The Digital Rod Position Indication (DRPI) Health screen (1805) has the following Data Cabinet alarms addressing the wrong point:
1) "A (-15V)" is addressing RM-DATAB4-ALM.SV3 and it should be addressing DATAA4 2) "A (+15V)" is addressing RM-DATAB3-ALM.SV3 and it should be addressing DATAA3
3) "B (-15V)" is addressing RM-DATAA4-ALM.SV3 and it should be addressing DATAB4 4) "B (+15V)" is addressing RM-DATAA3-ALM.SV3 and it should be addressing DATAB3
Area of Impact Plant Control
Digital Rod Position Indication (DRPI) Health Screen Incorrect Logic Cabinet Alarms
SCR-DR-5925 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS. Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description
The DRPI Health Screen (1805) should have the following Logic Cabinet Alarms with associated points:
1) "A (-15V)" point number RM-DCLON15VA-ALM.SV3 2) "A (+15V)" point number RM-DCLOP15VA-ALM.SV3 3) "B (-15V)" point number RM-DCLON15VB-ALM.SV3 4) "B (+15V)" point number RM-DCLOP15VB-ALM.SV3
Area of Impact Plant Control
Containment Recirculation Actuation Indication Issue
SCR-DR-5972
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
36 | P a g e
This issue impacts the following RO/SRO task: RO-INC-PMS-003-00 Monitor System Indications
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal impact on training.
Description Once containment recirculation is actuated, the actuation indication for Divisions C and D did not have the white box with an X on the ESF Act Status Screen for the divisional PDSPs or the Non-Safety Operational Overview screen (33020). The individual PMS division screen for CNMT Recirc actuation (IRWST/INJT Recirc) did show that it had been actuated on all 4 divisions.
Area of Impact Verifying plant response
Uncontrolled Heat-up (H/U) Indication Incorrect
SCR-DR-6009 This issue impacts the following RO/SRO task: RO-PRO-GEN-008-00 Perform Plant Cooldown from Mode 5 to Refueling Mode using GOP-206 Disposition
Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The "Uncontrolled HU or CD" (point DDS-SSF28-X0) indication on screen 60032 does not appear to change whether RCS temps are stable or changing. The only time the point driving the uncontrolled HU/CD indication would change state was when the Plant Mode Control (Screen 40003) was cycled to Manual Mode 4 during a RCS heatup IC in Mode 4. The controller was then cycled back to Auto and that uncontrolled HU/CD light energized. These same actions were repeated with a 100% steady-state IC, mode 5 IC and mode 3 IC without any changes in the indication.
Area of Impact Plant indications
DHC Summary - Assembly Move NAP Function Not Functional
SCR-DR-6022 This issue impacts the following RO/SRO task: RO-INC-PLS-004-03 Perform decay time surveillance
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the minimal training impact this particular Nuclear Application (NAP) has.
Description When attempting to simulate fuel assemblies being moved from the core to the Spent Fuel Pool (SFP), it was noted that the Decay Heat Calculation (DHC) NAP to maintain the administrative location of fuel does not work correctly. On display 40203 the assembly move buttons on the lower right portion
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
37 | P a g e
indicate they are only available when the light DDS-AP-DHC Status indicates it is ACTIVE. This light is driven off of the automatic mode selector and is INACTIVE when in MODES 1&2 and ACTIVE in MODES 3- 6. However, when the light indicates INACTIVE the buttons for moving are raised and available. When the light changes status to ACTIVE the buttons for moving are grayed out and no longer available. The light being active or inactive is currently driven by the auto mode selector and becomes active in MODES 3-6. However, fuel cannot be moved from the core into the SFP in any MODE other than MODE 6. The light should be driven by the manual input of the Rx vessel head being removed or installed or upper internals position on display 40004.
Area of Impact Reactivity Management with regards to indication and administration
RCP Vibration Alarm Naming
SCR-DR-6025 This issue impacts the following RO/SRO task: RO-PRI-RCS-008-00 ****Respond to RCS alarms
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal impact on training.
Description Reactor Coolant Pump (RCP) Vibration alarms are received at [ ]a,c. When looking at the individual RCP display (12105 for example) the bottom right corner has a light indication that displays "RCP 1A Vibration" and is grayed out when no alarming condition is met. When any vibration monitor goes above [ ]a,c mils the light will illuminate and change to "RCP 1A High-1". When the H2 setpoint is reached the light [ ]a,c alarm even though the H2 alarm is a Pri-2 alarm. The small button poke next to the light will also be available and will provide indication that both the H1 and H2 alarms are in. The point identifier is a good indication that these alarms are HIGH alarm (have H1 or H2 in the identifier). However, the Point information has L1 and L2 in the descriptions for the H1 and H2 alarms. This is an error likely naming convention as the L1 and L2 could mislead assumptions to LOW1 and LOW2.
Area of Impact Plant Control
HSS Display does not Include ESOP Discharge Pressure
SCR-DR-6078 This issue impacts the following RO/SRO task: RO-SUP-HSS-002-00 Operate the HSS
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description Hydrogen Seal Oil System (HSS) HSS-MP02 discharge pressure HSS-PT017 is not on the HSS display (15100). APP-HSS-M6-001 Rev 2 indicates that the transmitter should have an available point reference in ovation (PT-017 has a box with PIA - Pressure Indication/Alarm).
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
38 | P a g e
Area of Impact Plant Control
NAP for 1/M Intermediate Range Not Functional
SCR-DR-6089 This issue impacts the following RO/SRO task: RO-PRO-GEN-014-00 Perform an Inverse Count Rate Plot using GOP-307
Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs Analysis was performed. It also was determined to impact criterion (3) of 55.45(a) in the aggregate. The functionality of the Intermediate Range (IR) Inverse Count Rate Ratio Nuclear Application has been fixed such that it is usable by operators but additional work will be performed by WEC to restore this to full use. WEC RITS 38306 for tracking.
Description The intermediate range 1/M plot NAP does not work. Once P-6 (Permissive 6) was blocked and source range de-energized, the operator no longer had a 1/M plot generated.
Area of Impact Reactivity Management
ECS Penetration Temperature off Scale Low
SCR-DR-6103 This issue impacts the following RO/SRO task: AP-LT-R-ECS.005 Monitor the ECS
Disposition
Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description ECS penetration temperature reading is off scale low on display 22503. This is for the penetration to containment for the power cables for the RCPs as indicated on ECS-TE001A/B and TE002A/B which currently show the electrical penetration temperature as 0 degree F. The temperature should be reading something slightly higher than the ambient conditions.
Area of Impact Plant Control
Plant Mode Selector NAP Inconsistent with Procedure
SCR-DR-6144 This issue impacts the following RO/SRO task: RO-PRO-GEN-016-00 Perform Plant Power Escalation From 2% Power to 100% Power using GOP-306
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
39 | P a g e
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description In accordance with reference procedures, Mode 2 is entered when all Axial Offset (AO) bank rods are off the bottom. Currently Plant Control System (PLS) Auto Plant Mode selector changes from Mode 3 to Mode 2 when the RTBs are closed (P-3 is cleared).
Area of Impact Plant Startup
NAPS display issues
SCR-DR-6159 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description Several human factor related issues exist on the NAP surveillance screens. They involve unexplained acronyms, grammar errors, and inconsistent color coding.
Area of Impact Accessing information from DDS
WPIS Downscale Arrow Absent
SCR-DR-6164 This issue impacts the following RO/SRO task: RO-PRI-RCS-005-03 ****Cool down the pressurizer
Disposition This is a backup indication to alert the operator that the instrument is at its lower limit; the numeric indication is still available. This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the isolated issue and minimal impact on training once students are briefed on issue. Description No downscale arrow on Wall Panel Information System (WPIS) trend display (mode 3 / 4) exists for Tavg when bottom of scale.
Area of Impact Plant control
RSA NAP Does Not Process Failed Channels Correctly
SCR-DR-6169 This issue impacts the following RO/SRO task: RO-INC-PMS-003-00 Monitor System Indications
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
40 | P a g e
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The Redundant Sensors Algorithm Application (RSA) driven source range counts on the WPIS displays (main, trends, and safety functions) will still reflect an abnormally high value for source range power after a source range channel failure. The RSA NAP should account for the failure and remove it from the calculation.
Area of Impact Reactivity management
ZVS and ZBS Alarm Scaling Incorrect
SCR-DR-6171 This issue impacts the following RO/SRO task: RO-ELE-ZVS-001-00 Respond to Excitation and Voltage Regulation System (ZVS) abnormalities
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due the fact that there are other faults available that will result in the alarm notification to the operator.
Description The large APS tile EXCITATION VOLTAGE REG or any ZVS alarm will not come in on a regulator failure which causes voltage to peg high. The alarm setpoints in the database are maxed out so alarm will never come in on these parameters at this point.
Area of Impact Plant control
Flux doubling difference between divisions
SCR-DR-6175 This issue impacts the following RO/SRO task: RO-INC-PMS-003-00 Monitor System Indications Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) as the protective functions associated with these signals will still occur.
Description The alarms/alarm response for A/D Divisions differs significantly from B/C divisions. A/D divisions activates at 1.6 in 50 seconds whereas B/C at 2.2 in 10 seconds. J3 documents specify 2.2 in 10 sec.
Area of Impact Safety System Operation
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
41 | P a g e
Time to Boil Calculation
SCR-DR-6179 Disposition
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The Time to Boil NAP is a tool that is used for information only. The NAP is active when in Mode 5/6 conditions for RCS time to boil and when fuel is present in the spent fuel pool for spent fuel pool time to boil. The value is displayed in exponential units versus hh:mm. This means that operators will need to convert the scientific notation values into hours and minutes. Although this does take a short amount of time, the net effect is that it does not remove the ability to monitor the time to boil and it has no impact on actions the operator may, or may not, take in response to plant conditions. Description When core exit temperature was 300oF, Thot was > 212oF and TTB was >0 (15-25 min range). Either the NAP is calculating Time to Boil (TTB) incorrectly or the inputs to the NAP are wrong. TTB should reflect actual plant conditions.
Area of Impact NAP
Audible Rod Step Skips
SCR-DR-6186 This issue impacts the following RO/SRO task: RO-INC-PLS-003-01 Monitor the reactor power control system
Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). A training needs analysis was performed under CR 10025690.
Explain Brief students prior to beginning a simulator training phase or segment. Update a “SIMULATOR TRAINING STUDENT HANDOUT” (example is attached) and file in the “Operator Aids” notebook. Reference the SIMULATOR TRAINING STUDENT HANDOUT in each sim guide.
Description During outward rod motion, the audible step counter randomly can have an extra second pause in it with rod motion continuing. The step counter indication does not update at the same rate as the audible cue occurs.
Area of Impact Reactivity management
WPIS Display VARs
SCR-DR-6190
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
42 | P a g e
This issue impacts the following RO/SRO task: RO-ELE-ZAS-002-05 Maintain generator power factor and reactive load within acceptable ranges
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the fact that the numeric value displayed in all locations is the same while only the units of measurement change.
Description The Wall Panel Information System (WPIS) display has VARS indicated rather than Mega Volt-Amps Reactive (MVARS) for Generator Output.
Area of Impact Plant control
CMT WR Level Indications go Bad Quality
SCR-DR-6217 This issue impacts the following RO/SRO task: AP-LT-S-EOP.007 Direct implementation of E-1, AP1000 Loss of Reactor or Secondary Coolant
Disposition The Aggregate Study team determined this issue as not impacting simulator training due to only providing indication function. All protective functions are still available via the CMT Narrow Range level instruments.
Description The Wide Range (WR) Core Makeup Tank (CMT) level indications shift to Bad Quality once Automatic Depressurization System 1-3 (ADS 1-3) Actuate. Prior to this event, they would toggle to Bad Quality intermittently. The Bad Quality status is on indications PXS-LT009A/B & -LT010A/B (on Passive Core Cooling System (PXS) Supplemental Ind. Screen) and DDS-RSA11-L1 & DDS-RSA13-L1 (on WPIS screen 60017). The NAP driving the calculation of this indication drives them to bad quality whenever it determines voiding is occurring in the CMT (which is expected per design transients).
Area of Impact Plant control during decay heat removal
Unexpected Bank Sequence Out of Sequence Alarm
SCR-DR-6259 This issue impacts the following RO/SRO task: RO-INC-PLS-002-03 Control Rods: Verify Sequence and Overlap within COLR specifications
Disposition
The Bank Out of Sequence alarm and corresponding Alarm Response Procedure drives operators to perform AOP-104, “Rod Control Malfunction.” This will lead the crew to ensure all equipment is operating properly and to confirm that the rod alignment requirements of the Core Operating Limits Report are
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
43 | P a g e
met. This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). Description A Bank Sequence Out of Sequence (DDS-RSU01-X0) alarm actuates anytime banks M1 and MD (MA) are in overlap. When M1 and MD (MA) are in overlap the NAP generates an alarm showing bank M2 as being Out of Sequence (OOS). The NAP may be incorrectly calculating the OOS condition.
Area of Impact Reactivity control
Urgent Alarm Occurs During Case 2 CRE
SCR-DR-6267
This issue impacts the following RO/SRO task: RO-INC-PLS-003-06 Perform grey rod exchange
Disposition SNC evaluated this issue and determined that it does not impact the suitability of the simulator for the conduct of operating tests.
This is an AP1000 plant design issue. The simulator models the plant design. If operators encounter this condition, they will follow procedural guidance. The procedures provide the steps necessary for operators to respond to the event.
Description The Urgent Failure Alarm (UA) occurs when MA and MD banks are in motion and the Tavg-Tref deviation requires the AO rods to move to restore Tavg-Tref back into band. This only occurs if MA and MD rods are in motion. For plant conditions where only the MA or MD rods are in motion and the Tavg-Tref deviation requires AO rods to move, then an UA does not occur.
The UA appears to be a timing issue that occurs only when MA and MD banks are both in motion when the Tavg-Tref deviation occurs. Basically, the Ovation controllers briefly generate a RODS IN and a RODS OUT signal to the MA bank and a RODS IN and a RODS OUT signal to the MD bank which results in a UA from the Power Cabinets.
Area of Impact Reactivity control
Improper Bank Overlap Occurs when Data Point OCB07CE00C_OUTAV Increments Improperly During Startup
SCR-DR-6302 This issue impacts the following RO/SRO task: RO-INC-PLS-002-03 Control Rods: Verify Sequence and Overlap within COLR specifications
Disposition Subsequent to the performance of the Aggregate Study this item was determined to be invalid. This behavior is per current Westinghouse design.
Description
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
44 | P a g e
An Ovation data point in rod control is initially set to 0 when the Digital Rod Control System (DRCS) is reset. However, during SD1 withdrawal, OCB07CE00C_OUTAV (the Ovation data point) will increment to a value of 2 and then stay at this value. This data point is only supposed to increment for inward rod motion during M bank rod movement. The end result is that bank overlap will be incorrect if not manually corrected in Ovation.
Area of Impact Reactivity Management
Manual Reactor Trip Alarm Occurred without a Reactor Trip Request
SCR-DR-6315 This issue impacts the following RO/SRO task: RO-INC-PMS-005-02 Respond to a loss of a PMS
division or failure of PMS components
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The below failure sequence was performed and the reactor did not trip, but a Reactor Trip alarm was received. The alarm was for a Manual Reactor Trip, but a manual Rx Trip was not inserted. A P-4 was not received.
• RCS TE122C -> Open Circuit: PMS indicates Low-2 Tcold Cold Leg 2 Bistable Trip with Bad Quality
& Maintenance Bypass for Division C & PMS Cabinet Fault Alarm for Division C • RCS TE122D -> Open Circuit: PMS indicates Low-2 Tcold Cold Leg 2 Bistable Trip with Bad Quality
& Maintenance Bypass for Division D with other alarms
• Cold Leg 2 Temperature Low-2 Bypass inserted for Division A.
• Open the circuit for ECS-TE121B
Manual Reactor Trip Alarm (PMS-RXTR-MA-X0) actuate though none of the PMS divisions indicated that a Manual Reactor Trip had been inserted. PMS-J3-308 shows that the PMS-RXTR-MA alarms should only be activated by the Reactor Trip Switches at the PDSP or the RSR.
Area of Impact Reactivity Management
Controller Fault Alarms Received on Turbine Trip
SCR-DR-6366 This issue impacts the following RO/SRO task: AP-RO-ADM.015 Respond to alarms using the Alarm Presentation System (APS)
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
45 | P a g e
The Priority 4 alarms for Controller 34 (Drop 34) occur for Turbine Trips from 100%, 75% and 50% power. A Controller 21 (Drop 21) alarm occurs after the Turbine trip from 100% power.
Area of Impact Alarm Management
Diesel Fuel Oil Day Tank Level Transmitter Operation
SCR-DR-6491 This issue impacts the following RO/SRO task: RO-SUP-DOS-001-00 Operate Standby Diesel Fuel Oil System (DOS)
Disposition SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The initial tank levels established in the initial conditions provides suitable inventory for at least 2 hours of operation before refill of the tank would initiate at the incorrect level. Most scenarios are established such that the scenario would be complete prior to this refill level being achieved and the issue does not result in a loss of the DG. In addition, there is no procedural guidance that would direct an operator to verify the day tank level or proper operation of the day tank level control system. SNC Simulator Group continues to investigate the issue.
Description Diesel Fuel Oil System (DOS) level transmitters DOS-LT016A/017A and 016B/017B on the day tank control the refilling of the day tank based on level. The refilling should start when day tank level reaches low level ([ ]a,c) and stop at high level ([ ]a,c). The refilling of the day tank actually begins at 44.67% and stops at 100%. Additionally as level rises at ~85% the level indication jumps to 100%.
Area of Impact Plant Control
Inconsistent UAT Line Voltage Alarm Priorities
SCR-DR-6492 This issue impacts the following RO/SRO task: AP-RO-ADM.015 Respond to alarms using the Alarm Presentation System (APS)
Disposition Subsequent to the performance of the Aggregate Study this item was retested after the new APS was loaded. All UAT Undervoltage alarms except ES-7 had the proper priority assigned to them, ES-7 is still defined as a priority 1 alarm. The improper alarm priority being assigned to ES-7 does not impact the simulator’s suitability for the conduct of operating tests. The operator response to this alarm is consistent with ES-1 through ES-6 UAT Line Undervoltage alarm response procedures.
Description While performing a LOOP with Fast Bus Transfer to Reserve Auxiliary Transformer (RAT), it was noted that the Unit Auxiliary Transformer (UAT) Breaker Line Undervoltage Alarms for ES-2, 3 and 6 come in as Priority 2 alarms. Alarms for ES-4, 5, and 7 come in as Priority 1 alarms. These should be consistent.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
46 | P a g e
Area of Impact Plant Indication
Any Rods at Bottom Alarm
SCR-DR-6532 This issue impacts the following RO/SRO task: RO-INC-PLS-005-01 Respond to control rod position alarms
Disposition Subsequent to the performance of the Aggregate Study the simulator group developed a method for ensuring this issue is transparent to the to operators. Specific data points in the I&C architecture are configured in the Initial Conditions File prior to the start of a scenario such that the alarm does not appear. The deficiency remains open pending final correction of the underlying issue from WEC. This does not impact the simulator’s suitability for the conduct of operating tests. Description The "Any Rods at Bottom" alarm is actuating anytime rods are being driven through [ ]a,c steps. Based on APP-PLS-J1-023 Rev 2 3.1.2 Rev. 16 [
]a,c.
Area of Impact Alarm Management
WGS Sample Package Ovation Interface
SCR-DR-6612 This issue impacts the following RO/SRO task RO-SUP-WGS-003-00 Monitor WGS operation
Disposition
This issue does not impact the suitability of the simulator for the conduct of operating tests.
The functions associated with PS-001 are covered by APP-MS27-M6-001, APP-MS27-E5-001 and APP-WGS-M3C-101. These different design documents provide conflicting guidance as to what functions should and should not be present. The simulator modeling appears to be per APP-WGS-M3C-101 which does not include any functions based on PS-001. There is not any procedural guidance in place to take any actions based upon the indications that would be provided if PS-001 were modeled.
Description
Per drawing APP-MS27-E5-001 & APP-MS27-M6-001, PS-001 for monitoring N2 pressure to the Gaseous Radwaste System (WGS) Sample Package should be modeled to give an Ovation alarm when pressure decreases below 60 psig. PS-001 does not seem to be monitored in Ovation or the WGS model. Per the listed references, PS-001 will generate an Ovation alarm on low pressure or loss of power by de- energizing relay CR-3.
Area of Impact
WGS Indications
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
47 | P a g e
WGS Sample Package Digital Indication
SCR-DR-6613 This issue impacts the following RO/SRO task RO-SUP-WGS-003-00 Monitor WGS operation
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description Per drawing APP-MS27-E5-001, APP-MS27-M6-001, APP-MS27-VMM-004 pages 20 & 354, APP-WGS- MC3-101 page 16, H2 monitor AT032 (AE032) provides only a digital output.
Ovation drawing 16100 shows WGS-AT032 as having continuous indication. This continuous indication is inferred when looking at APP-MS27-M6-001, APP-WGS-M3C-101 page 23, and APP-WGS-M6-001. However per APP-MS27-VMM-004 page 354, AE032 provides only a digital output via a normally closed contact.
Area of Impact WGS Indications
RSA NAP for Power Range Power does not Eliminate Erroneous Input
SCR-DR-6621 This issue impacts the following RO/SRO task: RO-INC-PMS-003-00 Monitor System Indications
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description When the power range (PR) B lower detector fails high, the Redundant Sensors Algorithm Application (RSA) NAP for Power Range Power does not eliminate this input. This causes an erroneous PR power reading on the WPIS. Area of Impact Off Normal Event Response
Inconsistent DPU Alarm Priority Levels
SCR-DR-6651 This issue impacts the following RO/SRO task: AP-RO-ADM.015 Respond to alarms using the Alarm Presentation System (APS)
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description Controller 39 Alarm (DROP39_) is ranked as Priority 1 while the other Controller alarms are all Priority 4. The data process unit (DPU) alarms are all Priority 4 (including DPU 39).
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
48 | P a g e
Area of Impact Alarm Management
Safety Mimic Display for SGS-V255A& B Indicates Bad Quality Following a SFW Isolation SCR-DR-6698 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The PLS Safety Mimic display for all 4 divisions indicate bad quality for SGS-V255A&B following a Startup Feedwater System (SFW) Isolation Signal. The valve is closed as verified by the FW Components Status tab on the PDSP.
Area of Impact Plant Control
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
49 | P a g e
10 CFR 55.45(a)(4) Identify the instrumentation systems and the significance of facility instrument readings.
Executive Summary
The inability of operators to understand the operational significance of their indications due to the impact from these issues could cause a delay in tripping the unit or prevent them from tripping the unit. By the definition of safety, this would cause a failure and is not acceptable. Additional reinforcement during training of the operator skill set is required to mitigate this effect of particular issues. The issues affecting the operators are the Wall Panel Information System, Redundant Sensors Algorithm Application, Nuclear Application (WPIS RSA NAP) indications (especially Power Range (PR) and Intermediate Range (IR)).
The study team believes that the current issues impacting instrumentation will not preclude successful and safe operation. The team feels that the issues in this section have no aggregate impact. With proper training and reinforcement of skills, there is no overall impact on safety.
Stage 3 ADS Box Unused on Divisions C and D
SCR-DR-5619 This issue impacts the following RO/SRO task: RO-INC-PMS-003-00 Monitor System Indications
Disposition SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The simulator has N/A instead of a box for Stage 3 ADS. A review of current procedures confirms that the Division C and D Stage 3 ADS signal is not required. Therefore, there is no impact to the operator. Description The PMS divisions C and D indication for Stage 3 ADS status do not ever indicate actuation status.
Area of Impact Plant control
Subcriticality Indication on Critical Safety Function Screen Drops to Bad Quality
SCR-DR-5627 This issue impacts the following RO/SRO task: RO-PRO-EOP-031-00 Implement and evaluate Critical Safety Function Status trees using CSF-F-0
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The Nuclear Application (NAP) point DDS-SPD31-X0 randomly cycles to magenta bad quality due to DDS- RSA3J-J1 driving into bad quality. The result of the former point going bad quality is subcriticality and reactivity control display on the Critical Safety Function Wall Panel Information System (WPIS) screen goes magenta.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
50 | P a g e
Area of Impact Reactivity Management with regards to indication
VWS-TE079 Point Named Incorrectly
SCR-DR-5643
This issue impacts the following RO/SRO task: AP-LT-R-VWS.007 Monitor VWS parameters
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description Central Chilled Water System (VWS) VWS-TE079 is the inlet temperature for VWS Low Capacity Chiller #2. However, the point name is currently Low Cap Chiller 3 Inlet Temp.
Area of Impact Plant control
Calorimetric Data Precision
SCR-DR-5712 This issue impacts the following RO/SRO task: RO-INC-PMS-013-04 Perform MCR actions associated with calorimetric calibration of the excore NIS power range instrumentation
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The following points do not have the sufficient precision to perform the startup to 100% procedure steps regarding verification of power.
DDS-PPP08-J0 (SG FW RTO) DDS-PPP08-J0-AVP (SG RTO 1 Hr)
Area of Impact Plant control
Inconsistent OPDMS QPT Indications
SCR-DR-5903 This issue impacts the following RO/SRO task: RO-INC-IIS-005 Respond to On-line Power Distribution Monitoring System (OPDMS) malfunctions.
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the specific nature of the tasks associated with this screen.
Description During dropped control rods at core locations C7 and H8, it was noted that the indications provided by
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
51 | P a g e
the OPDMS Excore and Incore Quadrant Power Tilt (QPT) monitors were not consistent. The following observations were made:
• The Excore display lists the detectors as N41-N44. A detailed search of AP1000 documentation
found no reference material in which the PR excore detectors are referred to by N41-N44 except on page M-4 of APP-OCS-J4V-207, “Operation and Control Centers Display Design Document for Online Power Distribution Monitoring System.” Page M-4 has a table showing that OPDMS has N41, N42, N43, and N44 mapped to PR C, D, B, and A respectively, which appears to be consistent with what was seen in the PRS.
• The values for all of the excore detectors read 1.0 which appears to be just some sort of default. There are two problems with this. 1) One decimal point worth of data is not enough to adequately assess QPTR. 2) In this scenario, they should definitely not be reading 1.0.
• On the values displayed on this graphic, there are 8 labels titled “PR Power Upper Detector”, with no additional label as to what division.
• For the Incore QPT display, the letters in the corners do not match up with the data. For example, in this scenario the flux shifted towards PR A and C but the display shows it greatest near A and D and suppressed at C.
Area of Impact Reactivity Management
VFD Transformer Temperature
SCR-DR-5910 This issue impacts the following RO/SRO task: RO-PRI-RCS-005-01 Operate RCP VFDs
Disposition
Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description ECS-EV-X1-TMPC points associated with the Variable Frequency Drive (VFD) hottest cell parameters are not being driven by the models. They are a constant value.
Area of Impact Plant Operations
CDS-TE040A/B Range is Inadequate
SCR-DR-5921 This issue impacts the following RO/SRO task: RO-SEC-BDS-005-07 Verify automatic blowdown isolation upon high-2 temperature in heat exchanger shell outlet (CDS fluid) or high-2 DST level
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) since an open item exists in the design documentation.
Description
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
52 | P a g e
When the condensate outlet of the blowdown heat exchanger temperature element is failed high the blowdown flow remains un-isolated. The high-2 temperature ([ ]a,c) should isolate blowdown flow in accordance with APP-CDS-M3C-101 Rev 3. However, the range of the instrument (CDS-T-040A/B) listed in APP-CDS-M3C-101 Rev 3 is [ ]a,c which would never allow blowdown isolation on high temperature.
Area of Impact Plant Control
CVS-PT040 does not Provide Proper Protective Functions
SCR-DR-5968
This issue impacts the following RO/SRO task: RO-PRI-CVS-004-00 Monitor CVS operations
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description CVS-PT040 (pressure transmitter upstream of the letdown control valve) does not provide the proper protective functions for low pressure and high pressure protection in accordance with design documentation.
Per APP-CVS-M3C-101 Rev 6 Appendix C.3.2, at [ ]a,c when CVS-V047 is in automatic pressure control mode there is supposed to be a signal sent to “trip the CVS Makeup Pumps.” The high pressure signal is generated but does not trip the pumps; it presently feeds a Pump Auto Stop Demand signal. This signal will stop any pumps that are running in automatic only. When the plant is in water-solid mode, as determined in logic diagrams as having CVS-V047 in automatic pressure control mode, Chemical and Volume Control System (CVS) makeup pumps must be operated in manual; see APP-CVS-M3C-100 Rev 11 Logic Sheet CVS-4 Note10. Since the CVS makeup pumps are in manual the Auto Stop Demand signal will not shut the pump off to provide overpressure protection.
Per APP-CVS-M3C-101 Rev 6 Appendix C.3.2, at [ ]a,c when CVS-V047 is in automatic pressure control mode there is supposed to be a signal sent to “trip the Reactor Coolant Pumps…in order to protect them from reduced suction pressure.” This signal is also discussed in APP-RCS-M3C-100 Rev 9 Logic Sheet RCS-13 table and note 11. As presently designed there is no logic tie between CVS-PT040 and the Reactor Coolant Pumps (RCPs) to prevent damaging the RCPs upon a loss of Reactor Coolant System (RCS) pressure.
Area of Impact Plant Control
M Banks B & C Reversed on DRPI Health Screen
SCR-DR-6030 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
53 | P a g e
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The DRPI Health Screen ([ ]a,c) control rod banks M-B and M-C have the wrong rods listed as being in each bank. The rods listed as being M-B are actually M-C and the rods listed for M-C are the M-B rods. The correct arrangement of rods is shown in APP-RXS-M3-001 Rev 4 Figure 4-1 as well as on the DRPI M Bank screen ([ ]a,c).
Area of Impact Plant indications
Quality of RWS-V503 BAD at Limits
SCR-DR-6038 This issue impacts the following RO/SRO task: RO-SUP-RWS-005-00 Monitor the Raw Water System
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due minimal training impact based on still having other indications of the valve position available.
Description RWS-V503 and Circulating Water System (CWS) CWS-V514 are BAD quality at valve operating limits (open and closed).
Area of Impact Plant Control
Reactor Coolant Pump (RCP) Stator Temperature Indication off Scale Low at Lower Speeds
SCR-DR-6071 This issue impacts the following RO/SRO task: RO-PRI-RCS-003-01 Monitor the RCS during steady-state- power operation of the plant
Disposition
Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description RCP Stator Temperature indication is off scale low at lower speeds. With RCPs at 50% speed, the stator temperature (RCS-TE271, 272, 273, 274) indicates off scale low of 50F 'V'. This does not appear to be a valid temperature reading as the CCS temperature is 72F, SG cubicle temperature is 72F, and bearing temperature is 82F. With the ambient temperature above 50 degrees, it would be expected to have the motor at the same or slightly higher temperature.
Area of Impact Plant Control
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
54 | P a g e
HSS Display does not Include Emergency Seal Oil Pressure (ESOP) Discharge Pressure
SCR-DR-6078 This issue impacts the following RO/SRO task: RO-SUP-HSS-002-00 Operate the HSS
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description Hydrogen Seal Oil System (HSS) HSS-MP02 discharge pressure HSS-PT017 is not on the HSS display (15100). APP-HSS-M6-001 Rev 2 indicates that the transmitter should have an available point reference in ovation (PT-017 has a box with PIA - Pressure Indication/Alarm). Area of Impact Plant Control
DWS-LT006 has Insufficient Range
SCR-DR-6099 This issue impacts the following RO/SRO task: RO-SUP-DWS-002-00 Monitor DWS operation
Disposition The simulator is modeling the plant as currently designed; once the design has been updated the simulator model will be updated as well. This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal impact on training.
Description Demineralized Water Transfer and Storage System (DWS) DWS-LT006 is the level indication for the Condensate Storage Tank (CST). This level spans [ ]a,c per APP-DWS-M3C-101. However, the calculation note for DWS (APP-DWS-M3C-002) has the high 2 alarm at [ ]a,c. This is an important alarm because it is designed to give operators time to determine why the CST is full before it overflows to a drain. Overflow will occur at [ ]a,c.
Area of Impact Plant Control
MA Bank Rods Sometimes Stop at 263 steps during a CRE
SCR-DR-6102 This issue impacts the following RO/SRO task: RO-INC-PLS-003-06 Perform grey rod exchange
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The plant control system operating procedure allows for a case 1 Control Rod Exchange in the event that Case 2 is not functioning. With MA rods at 234 steps and AO rods in manual at 218 steps, the AO rods were stepped into [ ]a,c steps, which will drive MA out. The MA rods stopped at 263 steps, with an outward demand still in. The audible rod clicking stopped as well. If the CRE is still continued in
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
55 | P a g e
accordance with the procedure, MD will step into the core and MA will remain at [ ]a,c. This generates a “Rods out of sequence alarm”.
Area of Impact Reactivity Management
ECS Penetration Temperature off Scale Low
SCR-DR-6103 This issue impacts the following RO/SRO task: AP-LT-R-ECS.005 Monitor the ECS
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description
ECS penetration temperature reading is off scale low on display 22503. This is for the penetration to containment for the power cables for the RCPs as indicated on ECS-TE001A/B and TE002A/B which currently show the electrical penetration temperature as 0 deg F. The temperature should be reading something slightly higher than the ambient conditions.
Area of Impact Plant Control
Improper function of C-2 reactor power control interlock
SCR-DR-6122 This issue impacts the following RO/SRO task: RO-INC-PLS-008-00 Respond to PLS - related abnormalities
Disposition
Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description C-2 control interlock utilizes a non-conservative power input for operation. It currently uses BDP corrected power which is lower than NI power during over power scenarios.
Area of Impact Plant Control
WPIS RCS Inventory Issues
SCR-DR-6154 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS
Disposition This is a graphic issue that does not effect operational decisions and is representative of the information in GOP-114. This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
56 | P a g e
the fact the procedure will always be the preferred reference.
Description The WPIS RCS Inventory screen has reference level lines for hot leg top and bottom which appear incorrect. They are only 18 inches apart. It is, however a faithful reproduction of the chart in GOP-114. The procedure and display both need to be looked at. Issues found in calculation notes, procedure and display.
Area of Impact Plant control
WPIS Downscale Arrow Absent
SCR-DR-6164 This issue impacts the following RO/SRO task: RO-PRI-RCS-005-03 ****Cool down the pressurizer
Disposition This is a backup indication to alert the operator that the instrument is at its lower limit; the numeric indication is still available. This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the isolated issue and minimal impact on training once students are briefed on issue. Description No downscale arrow on WPIS trend display (mode 3 / 4) exists for Tavg when bottom of scale.
Area of Impact Plant control
Tuning of VBS Required for Stability
SCR-DR-6168 This issue impacts the following RO/SRO task: RO-VNT-VBS-001-15 Purge smoke from MCR/CSA
Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). A training needs analysis was performed under CR 10025689. SNC evaluated this issue and determined that even though it is unacceptable as a design issue, it does not impact actions taken by the operator.
Therefore, it does not impact the suitability of the simulator for the conduct of operating tests.
Description During Main Control Room (MCR) purge operations, Nuclear Island Nonradioactive Ventilation System (VBS) air handling unit trains cannot maintain stable flow and as a result, enter an indefinite cycling between two trains. The current tuning does not allow enough time to establish stable flow.
Area of Impact Plant control
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
57 | P a g e
Condensate Polisher Bypass Valve Control
SCR-DR-6172 This issue impacts the following RO/SRO task: RO-SEC-CPS-002-02 Place CPS in service
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal impact on training and examination as the chemistry levels in the secondary plant are transparent to operators.
Description CPS-V001 (CDS Polisher Bypass Valve) Setpoint controls are confusing. The procedure directs placing the controller in auto and never has a setpoint to control to. The current setpoint is set at the high end of the scale, so the bypass valve will never modulate closed. The calc note states that signals will be set based on CDS header and polisher flow. No setpoint is yet determined.
Area of Impact Plant control
Time to Boil Calculation
SCR-DR-6179 Disposition
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The Time to Boil NAP is a tool that is used for information only. The NAP is active when in Mode 5/6 conditions for RCS time to boil and when fuel is present in the spent fuel pool for spent fuel pool time to boil. The value is displayed in exponential units versus hh:mm. This means that operators will need to convert the scientific notation values into hours and minutes. Although this does take a short amount of time, the net effect is that it does not remove the ability to monitor the time to boil and it has no impact on actions the operator may, or may not, take in response to plant conditions.
Description
When core exit temperature was 300oF, Thot was > 212oF and TTB was >0 (15-25 min range). Either the NAP is calculating time to boil (TTB) incorrectly or the inputs to the NAP are wrong. TTB should reflect actual plant conditions.
Area of Impact NAP
CMT WR Level Indications go Bad Quality
SCR-DR-6217 This issue impacts the following RO/SRO task: AP-LT-S-EOP.007 Direct implementation of E-1, AP1000 Loss of Reactor or Secondary Coolant
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
58 | P a g e
Disposition The Aggregate Study team determined this issue as not impacting simulator training due to only providing indication function. All protective functions are still available via the CMT Narrow Range level instruments.
Description The WR Core Makeup Tank (CMT) level indications shift to Bad Quality once Automatic Depressurization System 1-3 (ADS 1-3) Actuate. Prior to this event, they would toggle to Bad Quality intermittently. The Bad Quality status is on level transmitter indications PXS-LT009A/B & -LT010A/B (on PXS Supplemental Ind. Screen) and DDS-RSA11-L1 & DDS-RSA13-L1 (on WPIS screen 60017). The NAP driving the calculation of this indication drives them to bad quality whenever it determines voiding is occurring in the CMT (which is expected per design transients).
Area of Impact Plant control during decay heat removal
Manual Reactor Trip Alarm Occurred without a Reactor Trip Request
SCR-DR-6315 This issue impacts the following RO/SRO task: RO-INC-PMS-005-02 Respond to a loss of a PMS division or failure of PMS components
Disposition
Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description The below failure sequence was performed and the reactor did not trip, but a Reactor Trip alarm was received. The alarm was for a Manual Reactor Trip, but a manual Rx Trip was not inserted. A P-4 was not received.
• RCS TE122C -> Open Circuit: PMS indicates Low-2 Tcold Cold Leg 2 Bistable Trip with Bad Quality
& Maintenance Bypass for Division C & PMS Cabinet Fault Alarm for Division C
• RCS TE122D -> Open Circuit: PMS indicates Low-2 Tcold Cold Leg 2 Bistable Trip with Bad Quality & Maintenance Bypass for Division D with other alarms
• Cold Leg 2 Temperature Low-2 Bypass inserted for Division A.
• Open the circuit for ECS-TE121B
Manual Reactor Trip Alarm (PMS-RXTR-MA-X0) actuate though none of the PMS divisions indicated that a Manual Reactor Trip had been inserted. PMS-J3-308 shows that the PMS-RXTR-MA alarms should only be activated by the Reactor Trip Switches at the PDSP or the RSR.
Area of Impact Reactivity Management
Main Generator Output breaker logic
SCR-DR-6392
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
59 | P a g e
This issue impacts the following RO/SRO task: RO-ELE-ZAS-002-04 Synchronize the generator in manual/auto modes
Disposition
SNC evaluated this issue and determined that since the MCR operator will not see this anomaly when the procedure is followed; it does not impact actions taken by the operator. Therefore, it does not impact the suitability of the simulator for the conduct of operating tests. Description A discrepancy in the turbine generator synchronization logic results in the following:
• IF you have the ‘acknowledge ready for auto-sync’ poke selected or not, the Generator breaker will close when ‘GEN’ is selected and operator action ceases for ~90 seconds. If you continue in a timely manner with the procedure before or after the edits one will think the plant response is correct due to the time it takes to auto sync. Apparently, selecting ‘GEN’ is the trigger for auto sync actuation regardless the state of the ‘sync check’ poke on 50212.
• IF you select Manual on the ZAS-EP-05 controller prior to depressing GEN (initially the controller comes up with neither selected), depressing GEN will NOT cause the ZAS-ES-01 breaker to close until ZAS-EP-05 controller is selected to AUTO. Again, the status of Acknowledge Ready for Auto Sync poke is irrelevant. The generator syncs.
Area of Impact Secondary plant management
Excitation Transformer Graphic Issue
SCR-DR-6398 This issue impacts the following RO/SRO task: RO-ELE-ZVS-001-00 Respond to Excitation and Voltage Regulation System (ZVS) abnormalities
Disposition
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. This is a simulator I&C issue related to a non-consequential graphic indication. Description On screen 22101 there is indication of downstream voltage of the excitation transformer. The indication displays ~26Kv when in fact this voltage should be around 900Vac. The drawing on the screen could be changed to connect to the upstream side of the transformer and then the operator would be able to see voltage on both sides of the main generator breaker when synchronizing to the grid.
Area of Impact Plant Control
IDS Charger Capacity and Design Float Voltage Requirement are Incompatible
SCR-DR-6400 This issue impacts the following RO/SRO task: RO-LT-R-IDS.002 Monitor the IDS
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
60 | P a g e
Disposition
SNC has determined that this issue does not impact the simulator’s suitability of the simulator for the conduct of operating tests. This discrepancy is due to partially implementing a forthcoming design change in different documents. The output voltage of the IDS chargers will be increased at a later date. The simulator is properly modeling the design documentation that it was built to. The lower output voltage indication will not drive operators to perform or not perform any actions.
Description Reference documentation for IDS currently states the rated voltage for the battery charger is limited to 250 VDC. Other documentation also states that the battery (IDS) should be normally on a float charge of 264 VDC. This cannot happen without a larger battery charger.
Area of Impact Plant Control
Graphic 1805 has reversed rods
SCR-DR-6409 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description MB and MC rods are reversed on Graphic 1805.
Area of Impact Plant Indication
Residual Bus Transfer Issues
SCR-DR-6481 This issue impacts the following RO/SRO task: AP-LT-R-ECS.012 Block fast bus transfer
Disposition
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The current simulator implementation does not provide the capability of the instructor to insert a malfunction that will result in the actuation of a Residual Bus Transfer. However, the Fast Bus Transfer and the Diesel Generator starting sequence function properly and provide the capability to examine the operators on electrical failures that would result in similar indications and Abnormal Operating Procedure entries. Description
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
61 | P a g e
For undervoltage conditions (loss of power) sensed by 27B2 (two-out-of-two or two-out-of-three logic) in conjunction with a source undervoltage condition sensed by 27S, the Unit Auxiliary Transformer (UAT) breaker will be tripped, leading to automatic closing of the RAT breaker completing residual bus transfer after establishing that the Reserve Auxiliary Transformer (RAT) source is live (59S1), the bus is dead (27B2), and all motor feeders are tripped.
The sequence of events for a residual bus transfer to occur is as follows:
• At 75% rated voltage (~ 3 sec time delay) the load shed occurs. The associated bus output breakers are tripped open. FOR ES-1 and 2 ONLY the associated DG starts.
• At 30% rated voltage the residual bus transfer occurs, the UAT source supply breaker opens and the RAT source supply breaker shuts for the associated bus.
This is not modeled currently.
Area of Impact Plant Control
Diesel Fuel Oil Day Tank Level Transmitter Operation
SCR-DR-6491 This issue impacts the following RO/SRO task: RO-SUP-DOS-001-00 Operate Standby Diesel Fuel Oil System (DOS)
Disposition SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The initial tank levels established in the initial conditions provides suitable inventory for at least 2 hours of operation before refill of the tank would initiate at the incorrect level. Most scenarios are established such that the scenario would be complete prior to this refill level being achieved and the issue does not result in a loss of the DG. In addition, there is no procedural guidance that would direct an operator to verify the day tank level or proper operation of the day tank level control system. SNC Simulator Group continues to investigate the issue.
Description Diesel Fuel Oil System (DOS) level transmitters DOS-LT016A/017A and 016B/017B on the day tank control the refilling of the day tank based on level. The refilling should start when day tank level reaches low level ([ ]a,c) and stop at high level ([ ]a,c). The refilling of the day tank actually begins at 44.67% and stops at 100%. Additionally as level rises at ~85% the level indication jumps to 100%.
Area of Impact Plant Control
WGS Sample Package Digital Indication
SCR-DR-6613 This issue impacts the following RO/SRO task RO-SUP-WGS-003-00 Monitor WGS operation
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
62 | P a g e
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description Per drawing APP-MS27-E5-001, APP-MS27-M6-001, APP-MS27-VMM-004 pages 20 & 354, APP-WGS- MC3-101 page 16, H2 monitor AT032 (AE032) provides only a digital output.
Ovation drawing 16100 shows WGS-AT032 as having continuous indication. This continuous indication is inferred when looking at APP-MS27-M6-001, APP-WGS-M3C-101 page 23, and APP-WGS-M6-001. However per APP-MS27-VMM-004 page 354, AE032 provides only a digital output via a normally closed contact.
Area of Impact WGS Indications
RSA NAP for Power Range Power does not Eliminate Erroneous Input
SCR-DR-6621 This issue impacts the following RO/SRO task: RO-INC-PMS-003-00 Monitor System Indications
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description When the power range (PR) B lower detector fails high, the RSA NAP for Power Range Power does not eliminate this input. This causes an erroneous PR power reading on the WPIS.
Area of Impact Off Normal Event Response
Safety Mimic Display for SGS-V255A& B Indicates Bad Quality Following a SFW Isolation SCR-DR-6698 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The PLS Safety Mimic display for all 4 divisions indicate bad quality for SGS-V255A&B following a SFW Isolation Signal. The valve is closed as verified by the FW Components Status tab on the PDSP.
Area of Impact Plant Control
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
63 | P a g e
Safety Mimic Display Navigation Issue
SCR-DR-6670 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS Disposition
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
PMS mimics in Ovation have several graphics that change division when using bottom links from a high level page. CVS for example, when in PMSC or PMSD and select CVS, then select Status, you will be changed to PMSA.
The PMS mimic in Ovation is an operator aid and not needed for plant operation or PMS actuations. Therefore, this issue does not impact the conduct of operating tests.
Operators are trained to always apply Human Performance (HU) tools when operating the plant, including changing from one Ovation screen to another.
CR 10070361 for WEC resolution.
Description PMS mimics in Ovation have several graphics than change division when using bottom links from a high level page. CVS for example, When in PMSC or PMSD and select CVS, then Status, you will changed to PMSA.
Area of Impact Plant Navigation
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
64 | P a g e
10 CFR 55.45(a)(5) Observe and safely control the operating behavior characteristics of the facility.
Executive Summary
Nine issues associated with rod control were assessed on an individual basis and deemed acceptable individually. The team believes operators can respond to simulator scenarios with these rod control issues and the NAPs issues. However, some of these issues require the operators to ignore the NAP indication or rod control behavior. There are some NAP issues having the potential to be resolved via a procedure update.
Issues with the RSA NAP concerned the team. Unnecessary reactor trips could happen due to misleading indications on the WPIS. This was deemed as a threat to quality examination.
The team discussed the numerous alarms (modulating status alarms in this case) and the potential to desensitize the operator to not pay attention to these alarms. The excessive alarm count coupled with the rod control and NAPs deficiencies led the team to determine that the issues in the section as an aggregate were unacceptable. Operations and Training could not deal with these issues and effectively train.
EDS Power Supply Assignments to PLS/DDS Cabinets Incomplete
SCR-DR-5546 This issue impacts the following RO/SRO tasks: RO-LT-R-EDS.001 Monitor the Non Class 1E DC and UPS system (EDS) for proper operation RO-LT-R-EDS.004 Respond to a loss of EDS DC power
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). SMEs determined that the current power supply arrangement was adequate to teach since it is per design documentation. Power supplies are an item that will be continuously taught as they are updated and changed.
Description A loss of individual Non Class 1E DC and UPS System (EDS) busses will result in incomplete system response. Some Ovation drops are not dynamically powered by the EDS model but are powered by a permanently energized model constant (specifically DPU047, DPU048, and DPU044). The load lists for the STS do not assign a power supply to all the Ovation drops so there is no plant design data to insert into the simulator.
Area of Impact Effective plant response to loss of power
Modeled BEACON Data Cannot Determine Quality
SCR-DR-5583 This issue impacts the following RO/SRO task: RO-INC-IIS-004-00 Determine functionality of the On-line Power Distribution Monitoring System
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
65 | P a g e
Disposition SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. Operator training simulator guides inform the training instructors on whether BEACON is operable or not operable. (Note: This evaluation is actually based on whether or not BEACON if functional or not functional. BEACON is not safety-related and does not have applicable Technical Specifications.)
Description Failure of the BEACON Data Processing (BDP) NAP causes the manual override signal originated by BDP to have BAD quality as expected. The BAD indication is passed through the BEACON operability calculations in the Plant State Monitoring (PST) NAP and appears on the OPDMS displays as 'operable' with BAD quality. Failure of the BDP application will cause BEACON to indicate ‘inoperable’ in the reference unit. However, the current STS scope of limitation has the BEACON outputs driven by the core model. There is currently no ability to pass quality over the interface for outputs. Inputs are not taken from the BDP NAP but from other plant process data. The core model does not know the status of the BDP NAP. The core model does not pass operability information.
Area of Impact Plant control
Repeatability issues involving CL 1B
SCR-DR-5594 This issue impacts the following RO/SRO task: RO-PRO-AOP-054-00 Respond to Reactor Coolant Pump Malfunctions using AOP-114
Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). This issue was further evaluated and determined to relate with a current design issue involving a lead/lag circuit associated with the main steam line pressure detection and input to the Safeguards ESF actuation. The pressure drop in the main steam line turns very close to the actuation set-point and the lead/lag circuit amplification of the rate of change may or may not cause the actuation. This is the expected plant response with the current actuation logic/software. Operators have sufficient procedure guidance directing them to respond to this event. For this reason, it was determined that this issue does not impact the suitability of the simulator for the conduct of operating tests.
Description The 1B RCP shaft shear malfunction may cause an unexpected safeguards depending on the initial conditions when the malfunction is inserted. Due to response caused by lead/lag filters in the steam line pressure logic, the signal is driven into the dead-band range of actuation at certain initial steam pressures. Elevating the initial steam pressure completely mitigates this occurrence.
Area of Impact Abnormal operation response
Unidentified and Identified Leak Rate Always Indicates BAD Data
SCR-DR-5599
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
66 | P a g e
This issue impacts the following RO/SRO task: RO-PRO-AOP-053-00 Respond to a Reactor Coolant Leak using AOP-112
Disposition
After performing V&V testing, SNC determined that the update was not successful. The Leak Rate Monitoring Application is informational only and does not drive any alarms based upon the calculated leakage. For this reason, any leak rate calculations would have to be performed manually per plant procedures vice using the NAP calculated values.
Description The Leak Rate Monitor (LRM) has BAD quality point indication for the Identified and Unidentified leak rates. They never change to good quality and indicate BAD when using the on demand leak rate calculation.
Area of Impact Plant control
Primary Dedicated Safety Panel Screens Do Not Update during MCR/RSR Transfer SCR-DR-5680 This issue impacts the following RO/SRO task: RO-INC-DDS-008-00 Transfer function from the Main Control Room to the Remote Shutdown Workstation
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) since the operators would have already evacuated the MCR and the indications would not been seen.
Description During the transfer between Main Control Room (MCR) and Remote Shutdown Room (RSR), it was noted that Protection and Safety Monitoring System (PMS) screens were not reflecting the position of the transfer switch correctly. Once the PDSP screen was refreshed, it displayed “correcting”.
Areas of Impact Impacts operation during unavailability of MCR for emergency safe shutdown
PMS Mimic Screens
SCR-DR-5689 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS
Disposition
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. This is a simulator I&C issue. PMS mimic screens are used for verification of indications only. If a question arises regarding an indication on the PMS mimic screen, operators will use primary indications from the PMS displays on the Primary Dedicated Safety Panel (PDSP). No operator action is available through the PMS mimic screens. All actions must be taken from the division’s PMS PDSP.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
67 | P a g e
Description PMS Mimic screens on Ovation do not reflect what is shown on the associated PDSP. This is especially true when there is any FAULT condition shown on the PDSP.
Area of Impact Plant control
Hot Leg Fluctuations at Mid-loop
SCR-DR-5707 This issue impacts the following RO/SRO task: RO-PRI-CVS-003-12 ****Fill the IRWST
Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). A training needs analysis was performed. The training needs analysis identified that this only occurs if CVS is used to refill the IRWST while isolated from the RCS. Since this is an abnormal lineup, the operator is unlikely to see this. For this reason, SNC has determined that the issue does not impact the suitability of the simulator for the conduct of operating tests.
Description While at mid-loop, the hot leg level and pressurizer wide range level began to fluctuate erratically. The hot leg was ~80% full.
Area of Impact
Plant control
OPDMS RIL Indication Does Not Align to COLR Rev. 0
SCR-DR-5736 This issue impacts the following RO/SRO task: RO-INC-PLS-002-03 Control Rods: Verify Sequence and Overlap within COLR specifications
Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs Analysis was performed under CR 10000474. The following is taken from the analysis performed. The Training Needs Analysis determined that the M1 Bank Insertion Low-2 alarm would be received prior to the M2 Bank Insertion Low-2 alarm due to rod sequencing and bank overlap. The M1 Bank alarm is set at the correct value.
Description It is noted that the Core Operating Limits Report (COLR) Rev. 0 Rod Insertion Limit (RIL) for M2 bank at all power levels is [ ]a,c Steps (fully withdrawn). The Online Power Distribution Monitoring System (OPDMS) Rod Insertion Limit (RIL) display ([ ]a,c) indicates the RIL for M2 bank is [ ]a,c steps. Further investigation indicates the M2 RIL indication high limit is [ ]a,c steps and therefore cannot indicate above this level (determined using point information page instrumentation tab for RB-INSERT- M2LIM.SV3@NET0). All SD bank indications are capable of indicating a maximum of [ ]a,c steps.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
68 | P a g e
Area of Impact Reactivity Management with regards to indication
Inconsistent OPDMS QPT Indications
SCR-DR-5903 This issue impacts the following RO/SRO task: RO-INC-IIS-005 Respond to On-line Power Distribution Monitoring System (OPDMS) malfunctions.
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the specific nature of the tasks associated with this screen.
Description During dropped control rods at core locations C7 and H8, it was noted that the indications provided by the OPDMS Excore and Incore Quadrant Power Tilt (QPT) monitors were not consistent. The following observations were made:
• The Excore display lists the detectors as N41-N44. A detailed search of AP1000 documentation
found no reference material in which the PR excore detectors are referred to by N41-N44 except on page M-4 of APP-OCS-J4V-207, “Operation and Control Centers Display Design Document for Online Power Distribution Monitoring System.” Page M-4 has a table showing that OPDMS has N41, N42, N43, and N44 mapped to PR C, D, B, and A respectively, which appears to be consistent with what was seen in the PRS.
• The values for all of the excore detectors read 1.0 which appears to be just some sort of default. There are two problems with this. 1) One decimal point worth of data is not enough to adequately assess QPTR. 2) In this scenario, they should definitely not be reading 1.0.
• On the values displayed, there are 8 titled “PR Power Upper Detector”, with no additional label as to what division.
• For the Incore QPT display, the letters in the corners do not match up with the data. For example, in this scenario the flux shifted towards PR A and C but the display shows it greatest near A and D and suppressed at C.
Area of Impact Reactivity Management
Print Feature from NAP non-functional
SCR-DR-5913 This issue impacts the following RO/SRO task: AP-RO-ADM.020.07 Document surveillance test in log book
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal training impact as this simply means the automated system is not capable of being used to complete surveillance requirement testing, operators are still capable of using the paper copies.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
69 | P a g e
Description While performing surveillance procedure "Incore Detector Comparison to Nuclear Instrument Channel Axial Flux Difference", the select PRINT SURVEILLANCE REPORT does not result in a printout.
Area of Impact Plant response – this prevents or impacts the performance of most Surveillance Tests
CDS-TE040A/B Range is Inadequate
SCR-DR-5921 This issue impacts the following RO/SRO task: RO-SEC-BDS-005-07 Verify automatic blowdown isolation upon high-2 temperature in heat exchanger shell outlet (CDS fluid) or high-2 DST level
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) since an open item exists in the design documentation.
Description When the condensate outlet of the blowdown heat exchanger temperature element is failed high the blowdown flow remains un-isolated. The high-2 temperature ([ ]a,c) should isolate blowdown flow in accordance with APP-CDS-M3C-101 Rev 3. However, the range of the instrument (CDS-T-040A/B) listed in APP-CDS-M3C-101 Rev 3 is [ ]a,c which would never allow blowdown isolation on high temperature.
Area of Impact Plant Control
CVS-PT040 does not Provide Proper Protective Functions
SCR-DR-5968 This issue impacts the following RO/SRO task: RO-PRI-CVS-004-00 Monitor CVS operations
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description CVS-PT040 (pressure transmitter upstream of the letdown control valve) does not provide the proper protective functions for low pressure and high pressure protection in accordance with design documentation.
Per APP-CVS-M3C-101 Rev 6 Appendix C.3.2, at [ ]a,c when CVS-V047 is in automatic pressure control mode there is supposed to be a signal sent to “trip the CVS Makeup Pumps.” The high pressure signal is generated but does not trip the pumps; it presently feeds a Pump Auto Stop Demand signal. This signal will stop any pumps that are running in automatic only. When the plant is in water-solid mode, as determined in logic diagrams as having CVS-V047 in automatic pressure control mode, Chemical and Volume Control System (CVS) makeup pumps must be operated in manual; see APP-CVS-M3C-100 Rev 11
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
70 | P a g e
Logic Sheet CVS-4 Note10. Since the CVS makeup pumps are in manual the Auto Stop Demand signal will not shut the pump off to provide overpressure protection.
Per APP-CVS-M3C-101 Rev 6 Appendix C.3.2, at [ ]a,c when CVS-V047 is in automatic pressure control mode there is supposed to be a signal sent to “trip the Reactor Coolant Pumps…in order to protect them from reduced suction pressure.” This signal is also discussed in APP-RCS-M3C-100 Rev 9 Logic Sheet RCS-13 table and note 11. As presently designed there is no logic tie between CVS-PT040 and the Reactor Coolant Pumps (RCPs) to prevent damaging the RCPs upon a loss of Reactor Coolant System (RCS) pressure. Area of Impact Plant Control
Containment Recirculation Actuation Indication Issue
SCR-DR-5972 This issue impacts the following RO/SRO task: RO-INC-PMS-003-00 Monitor System Indications
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal impact on training.
Description Once containment recirculation is actuated, the actuation indication for Divisions C and D did not have the white box with an X on the ESF Act Status Screen for the divisional PDSPs or the Non-Safety Operational Overview screen (33020). The individual PMS division screen for CNMT Recirc actuation (IRWST/INJT Recirc) did show that it had been actuated on all 4 divisions.
Area of Impact Verifying plant response
CVS-V094 Power Failure Response
SCR-DR-6019 This issue impacts the following RO/SRO task: RO-INC-PMS-005-02 Respond to a loss of a PMS division or failure of PMS components
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). When this issue manifests itself on the simulator, the operators will still comply with Technical Specifications requirements. An additional Tech Spec call would have to be made since there is an issue with Zinc addition in the CVS when power is lost. For this reason, this issue does not impact the suitability of the simulator for the conduct of operating tests.
Description CVS-V094 does not close upon a loss of power to ILCA02 as expected. It did close on loss of power to ILCA03, which is not in accordance with design documentation.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
71 | P a g e
Area of Impact Response to power loss
DHC Summary - Assembly Move NAP Function Not Functional
SCR-DR-6022 This issue impacts the following RO/SRO task: RO-INC-PLS-004-03 Perform decay time surveillance
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the minimal training impact this particular NAP has.
Description When attempting to simulate fuel assemblies being moved from the core to the Spent Fuel Pool (SFP), it was noted that the Decay Heat Calculation (DHC) NAP to maintain the administrative location of fuel does not work correctly. On display 40203 the assembly move buttons on the lower right portion indicate they are only available when the light DDS-AP-DHC Status indicates it is ACTIVE. This light is driven off of the automatic mode selector and is INACTIVE when in MODES 1&2 and ACTIVE in MODES 3-6. However, when the light indicates INACTIVE the buttons for moving are raised and available. When the light changes status to ACTIVE the buttons for moving are grayed out and no longer available. The light being active or inactive is currently driven by the auto mode selector and becomes active in MODES 3-6. However, fuel cannot be moved from the core into the SFP in any MODE other than MODE 6. The light should be driven by the manual input of the Rx vessel head being removed or installed or upper internals position on display 40004.
Area of Impact Reactivity Management with regards to indication and administration
HSS Display does not Include Emergency Seal Oil Pump (ESOP) Discharge Pressure
SCR-DR-6078 This issue impacts the following RO/SRO task: RO-SUP-HSS-002-00 Operate the HSS
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description Hydrogen Seal Oil System (HSS) HSS-MP02 discharge pressure HSS-PT017 is not on the HSS display (15100). APP-HSS-M6-001 Rev 2 indicates that the transmitter should have an available point reference in ovation (PT-017 has a box with PIA - Pressure Indication/Alarm).
Area of Impact Plant Control
DWS-LT006 has Insufficient Range
SCR-DR-6099 This issue impacts the following RO/SRO task: RO-SUP-DWS-002-00 Monitor DWS operation
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
72 | P a g e
Disposition The simulator is modeling the plant as currently designed; once the design has been updated the simulator model will be updated as well. This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal impact on training.
Description Demineralized Water Transfer and Storage System (DWS) DWS-LT006 is the level indication for the Condensate Storage Tank (CST). This level spans [ ]a,c per APP-DWS-M3C-101. However, the calculation note for DWS (APP-DWS-M3C-002) has the high 2 alarm at [ ]a,c. This is an important alarm because it is designed to give operators time to determine why the CST is full before it overflows to a drain. Overflow will occur at [ ]a,c.
Area of Impact Plant Control
MA Bank Rods Sometimes Stop at 263 steps during a CRE
SCR-DR-6102 This issue impacts the following RO/SRO task: RO-INC-PLS-003-06 Perform grey rod exchange Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The plant control system operating procedure allows for a case 1 Control Rod Exchange in the event that Case 2 is not functioning. With MA rods at 234 steps and AO rods in manual at 218 steps, the AO rods were stepped into [ ]a,c steps, which will drive MA out. The MA rods stopped at 263 steps, with an outward demand still in. The audible rod clicking stopped as well.
If the CRE is still continued in accordance with the procedure, MD will step into the core and MA will remain at [ ]a,c . This generates a “Rods out of sequence alarm”.
Area of Impact Reactivity Management
Improper function of C-2
SCR-DR-6122 This issue impacts the following RO/SRO task: RO-INC-PLS-008-00 Respond to PLS - related abnormalities
Disposition
Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description C-2 control interlock utilizes a non-conservative power input for operation. It currently uses BDP corrected power which is lower than NI power during over power scenarios.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
73 | P a g e
Area of Impact Plant Control
Excessive SFW Control Valve Cycling
SCR-DR-6151 This issue impacts the following RO/SRO task: RO-SEC-FWS-002-01 Monitor SFWS and MFWS system and component parameters
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). The reasoning was that the current controller tuning is at the best estimated values. Hot functional testing would provide as- built tuning. However, this issue did contribute to the aggregate impact of sections 3 and 5 with regards to secondary plant control.
Description At low pressure conditions less than 350 psig, the operator often has to take manual control of Startup Feed Water (SFW) control valves due to excessive cycling of the valves. Indicated flow rates range from 0 to greater than 600 gpm within 10 to 15 second cycles. This requires 100% of the operator’s attention until Normal Residual Heat Removal System (RNS) can be placed into service removing the cooldown function from the steam dumps.
Area of Impact
Plant control during startup and shutdown
SWS temperature control
SCR-DR-6152 This issue impacts the following RO/SRO task: RO-LT-R-SWS.019 Monitor the Service Water System
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal operator training impact. Operators might question the low SWS temperature reading, but at this reading, procedures require no operator actions be taken. Since this issue does not impact actions
taken by the operator, SNC determined it does not impact the suitability of the simulator for the conduct of operating tests.
Description SWS temperature got as low as about 51 deg F with an ambient air temp of 70 deg F. When the cooling fan kicked off, temperature returned to about 60 deg F. Cooling tower fan seems modeled more like an air conditioner than an evaporative cooler.
Area of Impact Plant performance
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
74 | P a g e
FWS-V037 Control Issue
SCR-DR-6156 This issue impacts the following RO/SRO task: RO-SEC-CDS-004-06 Respond to abnormal DST water level
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). The reasoning was that the current controller tuning is the current design. Hot functional testing would provide as-built tuning. However, this issue did contribute to the aggregate impact of section 5 with regards to secondary plant control. The team determined that there is procedural guidance already in place which mitigates the impact on operations. Therefore, it does not impact the suitability of the simulator for the conduct of operating tests.
Description The Deaerated Water Storage Tank (DST) level control presents an undue operator workload. DST level control often cannot maintain the DST above 95% to prevent Main Feedwater System (FWS) FWS-V037 rejecting to manual. Manual control of the input and output flow streams to the DST does not result in stable level control either due to the lag in valve response.
Area of Impact Plant control during startup and shutdown
SGS MSL drain pot erratic indication
SCR-6157 This issue impacts the following RO/SRO task: RO-SEC-SGS-006-00 Monitor SG system and component parameters
Disposition This issue does not impact the suitability of the simulator for the conduct of operating tests.
The drain pot level going high will cause operators to take the procedurally directed actions in the Alarm Response Procedure. The erratic indications will not solely drive the operators to perform plant maneuvers.
Description
SGS MSL drain pots became erratic and were flashing on both main steam lines at 53% Rx Power. By 90%, both pots filled with water, even with the drains open.
Area of Impact
Plant Indications
Stuck Rod Recovery Malfunction
SCR-DR-6162 This issue impacts the following RO/SRO task: RO-INC-PLS-005-00 Respond to DRCS related abnormalities
Disposition
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
75 | P a g e
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This SCR is specifically related to rod K10 and an individual fault. This fault is functional for all other rods and the reset capability described in the SCR is functional for all other rods. Training and exam scenarios are written to avoid this specific rod with no impact on the ability to train or examine on this fault.
WEC has created a tracking RITS (42516).
Description Rod K10’s keyboard reset does not go to the K10 algorithm in rod control, but K6. Rod K10 can never get an individual reset via the operator keyboard.
Area of Impact Rod control abnormality recovery
Tuning of VBS Required for Stability
SCR-DR-6168 This issue impacts the following RO/SRO task: RO-VNT-VBS-001-15 Purge smoke from MCR/CSA
Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). A training needs analysis was performed under CR 10025689. SNC evaluated this issue and determined that even though it is unacceptable as a design issue, it does not impact actions taken by the operator.
Therefore, it does not impact the suitability of the simulator for the conduct of operating tests.
Description During Main Control Room (MCR) purge operations, Nuclear Island Nonradioactive Ventilation System (VBS) air handling unit trains cannot maintain stable flow and as a result, enter an indefinite cycling between two trains. The current tuning does not allow enough time to establish stable flow.
Area of Impact Plant control
RSA NAP Does Not Process Failed Channels Correctly
SCR-DR-6169 This issue impacts the following RO/SRO task: RO-INC-PMS-003-00 Monitor System Indications
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The Redundant Sensors Algorithm Application (RSA) driven source range counts on the WPIS displays (main, trends, and safety functions) will still reflect an abnormally high value for source range power after a source range channel failure. The RSA NAP should account for the failure and remove it from the calculation.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
76 | P a g e
Area of Impact Reactivity management
Flux doubling difference between divisions
SCR-DR-6175 This issue impacts the following RO/SRO task: RO-INC-PMS-003-00 Monitor System Indications Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) as the protective functions associated with these signals will still occur.
Description The alarms/alarm response for A/D Divisions differs significantly from B/C divisions. A/D divisions activates at 1.6 in 50 seconds whereas B/C at 2.2 in 10 seconds. J3 documents specify 2.2 in 10 sec.
Area of Impact Safety System Operation
Time to Boil Calculation
SCR-DR-6179
Disposition
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The Time to Boil NAP is a tool that is used for information only. The NAP is active when in Mode 5/6 conditions for RCS time to boil and when fuel is present in the spent fuel pool for spent fuel pool time to boil. The value is displayed in exponential units versus hh:mm. This means that operators will need to convert the scientific notation values into hours and minutes. Although this does take a short amount of time, the net effect is that it does not remove the ability to monitor the time to boil and it has no impact on actions the operator may, or may not, take in response to plant conditions. Description When core exit temperature was 300oF, Thot was > 212oF and TTB was >0 (15-25 min range). Either the NAP is calculating TTB incorrectly or the inputs to the NAP are wrong. TTB should reflect actual plant conditions.
Area of Impact NAP
Audible Rod Step Skips
SCR-DR-6186 This issue impacts the following RO/SRO task: RO-INC-PLS-003-01 Monitor the reactor power control system
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
77 | P a g e
Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). A training needs analysis was performed under CR 10025690.
Explain Brief students prior to beginning a simulator training phase or segment. Update a “SIMULATOR TRAINING STUDENT HANDOUT” (example is attached) and file in the “Operator Aids” notebook. Reference the SIMULATOR TRAINING STUDENT HANDOUT in each sim guide.
Description During outward rod motion, the audible step counter randomly can have an extra second pause in it with rod motion continuing. The step counter indication does not update at the same rate as the audible cue occurs.
Area of Impact Reactivity management
VFS Radiation Monitoring Issue
SCR-DR-6192 This issue impacts the following RO/SRO task: RO-VNT-VFS-002-01 Monitor VFS parameters
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description VFS-RY102 indicates off scale high on a loss of Shutdown Cooling. The alarm appears to be related to the cycling flows in the Containment Air Filtration System (VFS) system. Alarms will also occur during containment purge activities with no RCS leakage or increased activity providing a false radiation alarm.
Area of Impact Radiation management
CMT WR Level Indications go Bad Quality
SCR-DR-6217 This issue impacts the following RO/SRO task: AP-LT-S-EOP.007 Direct implementation of E-1, AP1000 Loss of Reactor or Secondary Coolant
Disposition The Aggregate Study team determined this issue as not impacting simulator training due to only providing indication function. All protective functions are still available via the CMT Narrow Range level instruments.
Description The WR Core Makeup Tank (CMT) level indications shift to Bad Quality once Automatic Depressurization System 1-3 (ADS 1-3) Actuate. Prior to this event, they would toggle to Bad Quality intermittently. The Bad Quality status is on indications PXS-LT009A/B & -LT010A/B (on PXS Supplemental Ind. Screen) and
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
78 | P a g e
DDS-RSA11-L1 & DDS-RSA13-L1 (on WPIS screen 60017). The NAP driving the calculation of this indication drives them to bad quality whenever it determines voiding is occurring in the CMT (which is expected per design transients).
Area of Impact Plant control during decay heat removal
Urgent Alarm Occurs During Case 2 CRE
SCR-DR-6267 This issue impacts the following RO/SRO task: RO-INC-PLS-003-06 Perform grey rod exchange
Disposition SNC evaluated this issue and determined that it does not impact the suitability of the simulator for the conduct of operating tests.
This is an AP1000 plant design issue. The simulator models the plant design. If operators encounter this condition, they will follow procedural guidance. The procedures provide the steps necessary for operators to respond to the event.
Description The Urgent Failure Alarm (UA) occurs when MA and MD banks are in motion and the Tavg-Tref deviation requires the AO rods to move to restore Tavg-Tref back into band. This only occurs if MA and MD rods are in motion. For plant conditions where only the MA or MD rods are in motion and the Tavg-Tref deviation requires AO rods to move, then an UA does not occur.
The UA appears to be a timing issue that occurs only when MA and MD banks are both in motion when the Tavg-Tref deviation occurs. Basically, the Ovation controllers briefly generate a RODS IN and a RODS OUT signal to the MA bank and a RODS IN and a RODS OUT signal to the MD bank which results in a UA from the Power Cabinets.
Area of Impact Reactivity control
Improper Bank Overlap Occurs when Data Point OCB07CE00C_OUTAV Increments Improperly During Startup
SCR-DR-6302 This issue impacts the following RO/SRO task: RO-INC-PLS-002-03 Control Rods: Verify Sequence and Overlap within COLR specifications
Disposition Subsequent to the performance of the Aggregate Study this item was determined to be invalid. This behavior is per current Westinghouse design.
Description An Ovation data point in rod control is initially set to 0 when the Digital Rod Control System (DRCS) is reset. However, during SD1 withdrawal, OCB07CE00C_OUTAV (the Ovation data point) will increment to a value of 2 and then stay at this value. This data point is only supposed to increment for inward rod motion during
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
79 | P a g e
M bank rod movement. The end result is that bank overlap will be incorrect if not manually corrected in Ovation.
Area of Impact Reactivity Management
Manual Reactor Trip Alarm Occurred without a Reactor Trip Request
SCR-DR-6315 This issue impacts the following RO/SRO task: RO-INC-PMS-005-02 Respond to a loss of a PMS
division or failure of PMS components
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The below failure sequence was performed and the reactor did not trip, but a Reactor Trip alarm was received. The alarm was for a Manual Reactor Trip, but a manual Rx Trip was not inserted. A P-4 was not received.
• RCS TE122C -> Open Circuit: PMS indicates Low-2 Tcold Cold Leg 2 Bistable Trip with Bad Quality
& Maintenance Bypass for Division C & PMS Cabinet Fault Alarm for Division C
• RCS TE122D -> Open Circuit: PMS indicates Low-2 Tcold Cold Leg 2 Bistable Trip with Bad Quality & Maintenance Bypass for Division D with other alarms
• Cold Leg 2 Temperature Low-2 Bypass inserted for Division A.
• Open the circuit for ECS-TE121B
Manual Reactor Trip Alarm (PMS-RXTR-MA-X0) actuate though none of the PMS divisions indicated that a Manual Reactor Trip had been inserted. PMS-J3-308 shows that the PMS-RXTR-MA alarms should only be activated by the Reactor Trip Switches at the PDSP or the RSR.
Area of Impact Reactivity Management
Main Generator Output breaker logic
SCR-DR-6392 This issue impacts the following RO/SRO task: RO-ELE-ZAS-002-04 Synchronize the generator in manual/auto modes
Disposition
SNC evaluated this issue and determined that since the MCR operator will not see this anomaly when the procedure is followed; it does not impact actions taken by the operator. Therefore, it does not impact the suitability of the simulator for the conduct of operating tests. Description A discrepancy in the turbine generator synchronization logic results in the following:
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
80 | P a g e
• IF you have the ‘acknowledge ready for auto-sync’ poke selected or not, the Generator breaker will close when ‘GEN’ is selected and operator action ceases for ~90 seconds. If you continue in a timely manner with the procedure before or after the edits one will think the plant response is correct due to the time it takes to auto sync. Apparently, selecting ‘GEN’ is the trigger for auto sync actuation regardless the state of the ‘sync check’ poke on 50212.
• IF you select Manual on the ZAS-EP-05 controller prior to depressing GEN (initially the controller comes up with neither selected), depressing GEN will NOT cause the ZAS-ES-01 breaker to close until ZAS-EP-05 controller is selected to AUTO. Again, the status of Acknowledge Ready for Auto Sync poke is irrelevant. The generator syncs.
Area of Impact Secondary plant management
Residual Bus Transfer Issues
SCR-DR-6481 This issue impacts the following RO/SRO task: AP-LT-R-ECS.012 Block fast bus transfer
Disposition
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The current simulator implementation does not provide the capability of the instructor to insert a malfunction that will result in the actuation of a Residual Bus Transfer. However, the Fast Bus Transfer and the Diesel Generator starting sequence function properly and provide the capability to examine the operators on electrical failures that would result in similar indications and Abnormal Operating Procedure entries. Description For undervoltage conditions (loss of power) sensed by 27B2 (two-out-of-two or two-out-of-three logic) in conjunction with a source undervoltage condition sensed by 27S, the Unit Auxiliary Transformer (UAT) breaker will be tripped, leading to automatic closing of the Reserve Auxiliary Transformer (RAT) breaker completing residual bus transfer after establishing that the RAT source is live (59S1), the bus is dead (27B2), and all motor feeders are tripped.
The sequence of events for a residual bus transfer to occur is as follows:
• [ ]a,c
• [ ]a,c
This is not modeled currently.
Area of Impact Plant Control
Diesel Fuel Oil Day Tank Level Transmitter Operation
SCR-DR-6491
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
81 | P a g e
This issue impacts the following RO/SRO task: RO-SUP-DOS-001-00 Operate Standby Diesel Fuel Oil System (DOS)
Disposition SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The initial tank levels established in the initial conditions provides suitable inventory for at least 2 hours of operation before refill of the tank would initiate at the incorrect level. Most scenarios are established such that the scenario would be complete prior to this refill level being achieved and the issue does not result in a loss of the DG. In addition, there is no procedural guidance that would direct an operator to verify the day tank level or proper operation of the day tank level control system. SNC Simulator Group continues to investigate the issue.
Description Diesel Fuel Oil System (DOS) level transmitters DOS-LT016A/017A and 016B/017B on the day tank control the refilling of the day tank based on level. The refilling should start when day tank level reaches low level ([ ]a,c) and stop at high level ([ ]a,c). The refilling of the day tank actually begins at 44.67% and stops at 100%. Additionally as level rises at ~85% the level indication jumps to 100%.
Area of Impact Plant Control
VES Supply Header Pressure Response to Temperature Changes
SCR-DR-6547 This issue impacts the following RO/SRO task: RO-VNT-VBS-001-11 Monitor the main control room/control support area HVAC subsystem parameters
Disposition
Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description When the Radiological Controlled Area Ventilation System (VAS) ventilation is secured to the Main Control Room Emergency Habitability System (VES) Air Storage Area (Rm 12555), room temperature will rise as expected, as indicated on VAS-TE080A/B. This rise in temperature should cause VES Supply Header Pressure (VES-PT001A and B) to rise, since the volume of air in the VES tanks is not changing. As depicted in the attached trend, this does not happen in the current model. Since VES air storage tank pressure does not change in the model, this causes the calculated air quantity (VES-QIY008A and B) to lower, which should not be the case. The VES air quantity should only change if you are filling VES or depressurizing it.
Area of Impact Observation of fundamental parameters
ECS-EC-313 Loads not modeled
SCR-DR-6593 This issue impacts the following RO/SRO task: AP-LT-R-ECS.005 Monitor the ECS
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
82 | P a g e
Disposition The Aggregate Study team determined this issue does not affect the operators because they will be fairly transparent. The students will most likely not notice these issues during training. The team determined this issue in the aggregate is acceptable. While there will be manual actions that occur, it is still acceptable and should be addressed in a training moment.
Description APP-ECS-E3-EC31302 Rev 1 depicts ECS-ET-3131. Model requirements document SV3-STS-J4-119 Rev 1 required WEC/GSE to model the loads of EC-313. Work in SCR-DR-6568 determined they neglected to meet this requirement. SV3-STS-J1-119 Rev 0 shows loads off of EC-312 but nothing on EC-313.
Area of Impact Plant response
D/G Sequencer Operation
SCR-DR-6610 This issue impacts the following RO/SRO task: RO-ELE-ZOS-002-06 Perform priority load AC load sequencing and verify proper 6.9 kV priority load system voltage
Disposition The Aggregate Study team determined this issue does not affect the operators because they will be fairly transparent. The students will most likely not notice these issues during training. The team determined this issue in the aggregate is acceptable. While there will be manual actions that occur, it is still acceptable and should be addressed in a training moment.
Description According to APP-ZOS-E0C-001 (Onsite Standby Power System Diesel Generator Sizing Calculation) Rev 1, the following busses should get re-energized by the sequencer at Load Step [
]a,c
[ ]a,c [ ]a,c [ ]a,c [ ]a,c [ ]a,c [ ]a,c
These busses are not currently sequenced once ECS-ES-1/2 gets re-energized from its respective DG. Area of Impact Effective plant response to loss of power
RSA NAP for Power Range Power does not Eliminate Erroneous Input
SCR-DR-6621
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
83 | P a g e
This issue impacts the following RO/SRO task: RO-INC-PMS-003-00 Monitor System Indications
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description When the power range (PR) B lower detector fails high, the RSA NAP for Power Range Power does not eliminate this input. This causes an erroneous PR power reading on the WPIS.
Area of Impact Off Normal Event Response
VZS Dampers do not Fail As-Is after Loss of Power
SCR-DR-6623 This issue impacts the following RO/SRO task: RO-VNT-VZS-005-00 Respond to Abnormal VZS Conditions
Disposition This issue does not impact the suitability of the simulator for the conduct of operating tests. This incorrect modeling issue is transparent to the operators and therefore has no impact on operator actions. The specific dampers in question are verified to be in the proper position by AOP-302 Attachment 1. The only time the positions of the dampers in question are checked is when the associated diesel is operating. During this scenario the dampers will be operating properly as they are energized. If the diesel generators are not powering their respective buses then the steps directing the operators to verify the position of these dampers will not be performed. This procedural guidance will cause the incorrect modeling of these dampers to not be observed.
Description Diesel Generator Building Heating and Ventilation dampers VZS-D014A/B and VZS-D015A/B do not fail as is on loss of power.
Area of Impact Plant Control during loss of power
Turbine Bypass Control Valve Control Logic cannot Support Design Power Supplies
SCR-DR-6634 This issue impacts the following RO/SRO task: RO-SEC-MSS-003-01 Monitor MSS system and component parameters
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The turbine bypass control valves (MSS-PL-V001, MSS-PL-V002, MSS-PL-V003, MSS-PL-V004, MSS-PL- V005, MSS-PL-V006) have four solenoid valves that are designed to share two different power supplies. However, the current design control logic) uses one power supply for all 4 solenoids.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
84 | P a g e
Area of Impact Plant operations during loss of power
Battery Temperature does not change
SCR-DR-6645 This issue impacts the following RO/SRO task: RO-LT-R-EDS.001 Monitor the Non Class 1E DC and UPS system (EDS) for proper operation
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description During a loss of all AC sources discharge test it was identified that the battery temperature did not change during the 15 hour run. Further research showed the battery room temperature does not change during loss of ventilation
Area of Impact
Effective battery temperature indication
Fire Protection System is not modeled in Containment
SCR-DR-6657 This issue impacts the following RO/SRO task: RO-SUP-FPS-002-00 Operate the FPS
Disposition This issue does not impact the suitability of the simulator for the conduct of operating tests. Due to the low importance values in NUREG-2103 of the items associated with this failure no simulator scenarios have been developed that would make use of this malfunction. The modeling of this malfunction will be investigated as time permits and corrected if possible.
Description While attempting to create a FR-Z.2 (Response to High Containment Level) scenario, it was discovered that a leak from the Fire Protection System (FPS) header in containment had no effect on any containment parameters (ex. Containment Sump Level, Containment Humidity). FPS Containment Spray also had no effect on any containment parameters.
Area of Impact Plant control
IRWST Temperature Response
SCR-DR-6701 This issue impacts the following RO/SRO task: RO-PRO-EOP-003-00 Respond to a loss of Reactor or Secondary Coolant using EEP-E-1
Disposition
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
85 | P a g e
The Aggregate Study team determined this issue does not have an aggregate impact on operations. Additionally, this modeling of the differential temperatures has no impact on operator actions because, by procedure, there are no operator actions related to IRWST temperature. For this reason, it was determined that this issue does not impact the suitability of the simulator for the conduct of operating tests.
Description [
]a,c
Area of Impact Plant response
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
86 | P a g e
10 CFR 55.45(a)(6) Perform control manipulations required to obtain desired operating results during normal, abnormal, and emergency situations.
Executive Summary
Issues 5655, 5926, 6151, 6156, and 6172 were grouped together for analysis by the team. The team acknowledged that there is procedural guidance already in place which mitigates some of these items impact on operation. The team discussed the potential burden they do place but also discussed examples of successful navigation through these items during ISV.
The issues involving having to take manual control to enhance stability were discussed by a team. For example, Nuclear Island Nonradioactive Ventilation System (VBS) issue will require manual control (6168) during VBS operations. In discussing if these issues presented too much of an aggregate challenge, the team determined a manual reactor trip would likely not be initiated. The issues described in this section are always in, thus making them a part of the standard training session. The Initial Conditions (IC) or scenario guide would be setup to already have these issues placed in a status where the scenario wouldn’t be adversely affected. The team justified the first group of issues being acceptable. Issues like having to manually control FWS-V037 are part of the training process. Each operator candidate is already familiar with how to handle them. Issues with VBS and VFS can be handled because they are out of the way items during training and operations. These systems are rarely operated as a normal process.
Issues 6610 and 6593 were grouped together and determined not to affect the operators because they will be fairly transparent. The students will most likely not notice these issues during training. Issue 5609 has been addressed as a procedure change.
The team determined that the combination of these issues in the aggregate were acceptable. While there will be manual actions that occur, it is still acceptable and should be addressed in a training moment.
These controllers are expected to work in automatic, and when manual control is required reinforcement to the operator’s skills will be required to establish proficiency. The absence of this ability will not preclude safety, but additional training will provide additional proficiency. Simulator JIT will be a valuable tool in preparing the operators for encountering these issues in the plant.
Rod Withdrawal button deselects During Continuous Operation
SCR-DR-5584 This issue impacts the following RO/SRO task: RO-INC-PLS-003-02 Monitor the Control Rod Drive System
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description While performing extended rod withdrawals during startups, depressing the rod withdrawal button (UP ARROW) may cause the UP ARROW button to un-highlight and momentarily flash gray
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
87 | P a g e
even though still depressed. Rod motion will still occur. Area of Impact Reactivity Management
Unstable VFS Containment Exhaust Flow
SCR-DR-5593 This issue impacts the following RO/SRO task: RO-VNT-VFS-002-05 Regulate Containment Air Pressure
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to low impact on training and the ability to take some controllers to manual to control flow.
Description When Containment Air Filtration System (VFS)-MS-02A/B are placed in service to establish containment exhaust flow in AUTO with a setpoint of 4000 scfm, a stable flow of 4000 scfm cannot be obtained on VFS-FT011A/B. Flow gradually increases above 4000 scfm and oscillates. 4000 scfm can be achieved in MANUAL control, but when placed back into AUTO, the oscillations continue. This point is on screen 20102.
Area of Impact Plant response
GSS Header Pressure Response
SCR-DR-5609 This issue impacts the following RO/SRO task: RO-SEC-GSS-002-00 Monitor the GSS
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description In both at-power ICs with Main Steam System (MSS) supplying, GSS-V002 was able to maintain 4 psig. Once the shutdown IC, 32, was loaded with Auxiliary Steam Supply System (ASS) supplying the Gland Seal System (GSS) header GSS-V002 was unable to maintain downstream pressure. If GSS-V004 (bypass for GSS-V002) was opened to > 50% there was sufficient pressure/flow to allow GSS-V002 to control pressure at 4 psig.
To determine if flow was the root of the problem aligned MSS to supply GSS while still using IC 32. Once steam load was on MSS, the GSS header pressure lowered until GSS-V002 was fully open (GSS pressure was ~ 1.8psig); MSS-V022A was ~ 28% open. Adjustment was required to the automatic pressure setpoint of MSS-V022A to 240 psig to increase the steam pressure upstream of GSS-V002. Once MSS-PT015A was ~ 240 psig GSS-V002 closed to 98% to control at 4 psig.
Area of Impact Plant response
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
88 | P a g e
Issue with Automatic Control of DST level and Auto Start of Standby Condensate Pump
SCR-DR-5655
This issue impacts the following RO/SRO task: RO-SEC-FWS-002-03 Establish level in the DST
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). The reasoning was that the current controller tuning is at the best estimated values. Hot functional testing would provide as- built tuning. However, this issue did contribute to the aggregate impact of sections 3 and 5 with regards to secondary plant control.
Description During plant startup from Mode 5 to 100% power, the Condensate system pressure would lower to the auto start setpoint of the standby Condensate pump. This pressure drop is due to Condensate System (CDS) CDS-V022 and CDS-V025 modulating to maintain level in the Deaerated Storage Tank (DST). In accordance with reference plant procedures for normal operation, the second condensate pump is started at 40-45% power. However, the second condensate pump will have already auto started in the heatup and startup procedures, due to the slow response of CDS-V022 and CDS-V025.
Area of Impact Plant design deficiency impacts operations during startup
Model Instability during PZR Fill to Solid
SCR-DR-5698 This issue impacts the following RO/SRO task: RO-PRI-CVS-003-04 Operate the Chemical and Volume Control System to control the primary system pressure in water solid mode
Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs Analysis was performed under CR 10000465. Training Needs Analysis determined training involving establishment of solid plant should not be performed until issue is corrected. A review of current training material did not reveal any scenarios where this was required. Wording from analysis follows.
Added to the DR Global Issues list, this will be briefed to the students at the beginning of the Simulator portion of training. Scenario AP-LT-I-SIM-GOPSDCD (Covering GOP-205, Plant Cooldown MODE 3 to MODE 5) does not train on Solid Plant Operations.
Description The Liquid Radwaste System (WLS) model is prone to failure during evolutions involving near solid pressurizer operations if the Effluent Holdup Tank is filled too rapidly.
Area of Impact Difficulty in achieving solid plant operations continuously
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
89 | P a g e
Steam Generator Level Instability with Control Valves Shut
SCR-DR-5926 This issue impacts the following RO/SRO task: RO-SEC-SGS-006-00 Monitor SG system and component parameters
Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs Analysis was performed under CR 10000465. Further analysis determined that this condition has no impact on operator actions. Under this condition, the Steam Generator has been isolated and there are no further actions required by operators. For this reason and because the level stabilized after a relatively short period of time, it was determined that this issue does not impact the suitability of the simulator for the conduct of operating tests. Description Rapid steam generator level oscillations can be observed with the steam generator isolated and all control valves shut for a period of three to five minutes before stabilizing. Area of Impact Plant Response
CVS-V094 Power Failure Response
SCR-DR-6019 This issue impacts the following RO/SRO task: RO-INC-PMS-005-02 Respond to a loss of a Protection and Safety Monitoring System (PMS) division or failure of PMS components
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). When this issue manifests itself on the simulator, the operators will still comply with Technical Specifications requirements. An additional Tech Spec call would have to be made since there is an issue with Zinc addition in the CVS when power is lost. For this reason, this issue does not impact the suitability of the simulator for the conduct of operating tests.
Description CVS-V094 does not close upon a loss of power to ILCA02 as expected. It did close on loss of power to ILCA03, which is not in accordance with design documentation.
Area of Impact Response to power loss
DHC Summary - Assembly Move NAP Function Not Functional
SCR-DR-6022 This issue impacts the following RO/SRO task: RO-INC-PLS-004-03 Perform decay time surveillance
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the minimal
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
90 | P a g e
training impact this particular NAP has.
Description When attempting to simulate fuel assemblies being moved from the core to the Spent Fuel Pool (SFP), it was noted that the Decay Heat Calculation (DHC) NAP to maintain the administrative location of fuel does not work correctly. On display 40203 the assembly move buttons on the lower right portion indicate they are only available when the light DDS-AP-DHC Status indicates it is ACTIVE. This light is driven off of the automatic mode selector and is INACTIVE when in MODES 1&2 and ACTIVE in MODES 3-6. However, when the light indicates INACTIVE the buttons for moving are raised and available. When the light changes status to ACTIVE the buttons for moving are grayed out and no longer available. The light being active or inactive is currently driven by the auto mode selector and becomes active in MODES 3-6. However, fuel cannot be moved from the core into the SFP in any MODE other than MODE 6. The light should be driven by the manual input of the Rx vessel head being removed or installed or upper internals position on display 40004.
Area of Impact Reactivity Management with regards to indication and administration
WLS-MP-08C improperly Pumps Monitor Tank C
SCR-DR-6068 This issue impacts the following RO/SRO task: RO-SUP-WLS-002-00 Operate the WLS
Disposition The cause of this issue was determined to be a modeling issue with the variables associated with piping arrangements. SNC Simulator Group has corrected this deficiency and resolved the issue.
Description
While performing a startup from Mode 6 it was discovered that WLS-MP-08C will not pump Monitor tank C around 37 inches. The pump will turn on and occasionally the downstream check valve will throttle open and shut but there is little or no evidence of flow. Also, discharge pressure never goes above 12-13psig. Normal discharge pressure for the other monitor tank pumps is around [ ]a,c.
Note that it does pump when level is above 37 inches as the tank has been pumped down to 37 inches successfully. It appears to exhibit strange behavior at 37 inches and below.
Area of Impact
Correct operation of plant systems
MA Bank Rods Sometimes Stop at 263 steps during a CRE
SCR-DR-6102 This issue impacts the following RO/SRO task: RO-INC-PLS-003-06 Perform grey rod exchange
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
91 | P a g e
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The plant control system operating procedure allows for a case 1 Control Rod Exchange (CRE) in the event that Case 2 is not functioning. With MA rods at 234 steps and AO rods in manual at 218 steps, the AO rods were stepped into [ ]a,c steps, which will drive MA out. The MA rods stopped at 263 steps, with an outward demand still in. The audible rod clicking stopped as well.
If the CRE is still continued in accordance with the procedure, MD will step into the core and MA will remain at [ ]a,c. This generates a “Rods out of sequence alarm.”
Area of Impact Reactivity Management
WRS Sump Pump B Discharge Pressure Inadequate
SCR-DR-6126 This issue impacts the following RO/SRO task: RO-SUP-WRS-002-00 Operate WRS
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description Radioactive Waste Drain System (WRS) sump pump WRS-MP-01B indicates a low pressure when pumping with an automatic start signal. This is evident when a leak is inserted that fills the WRS sump (such as a RNS leak). The "A" pump has proper discharge pressure, but "B" does not indicating low pressure in alarm. If the "A" pump is taken to manual and secured, the "B" pump still does not develop proper pressure.
Area of Impact Plant control
Excessive SFW Control Valve Cycling
SCR-DR-6151 This issue impacts the following RO/SRO task: RO-SEC-FWS-002-01 Monitor SFWS and MFWS system and component parameters
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). The reasoning was that the current controller tuning is at the best estimated values. Hot functional testing would provide as- built tuning. However, this issue did contribute to the aggregate impact of sections 3 and 5 with regards to secondary plant control.
Description At low pressure conditions less than 350 psig, the operator often has to take manual control of Startup Feed Water (SFW) control valves due to excessive cycling of the valves. Indicated flow rates range from 0 to
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
92 | P a g e
greater than 600 gpm within 10 to 15 second cycles. This requires 100% of the operator’s attention until RNS can be placed into service removing cooldown function from the steam dumps.
Area of Impact Plant control during startup and shutdown
Stuck Rod Recovery Malfunction
SCR-DR-6162 This issue impacts the following RO/SRO task: RO-INC-PLS-005-00 Respond to DRCS related abnormalities
Disposition SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This SCR is specifically related to rod K10 and an individual fault. This fault is functional for all other rods and the reset capability described in the SCR is functional for all other rods. Training and exam scenarios are written to avoid this specific rod with no impact on the ability to train or examine on this fault.
WEC has created a tracking RITS (42516).
Description Rod K10’s keyboard reset does not go to the K10 algorithm in rod control, but K6. Rod K10 can never get an individual reset via the operator keyboard.
Area of Impact Rod control abnormality recovery
Polisher Bypass Valve Control
SCR-DR-6172 This issue impacts the following RO/SRO task: RO-SEC-CPS-002-02 Place CPS in service
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal impact on training and examination as the chemistry levels in the secondary plant are transparent to operators.
Description CPS-V001 (CDS Polisher Bypass Valve) Setpoint controls are confusing. The procedure directs placing the controller in auto and never has a setpoint to control to. The current setpoint is set at the high end of the scale, so the bypass valve will never modulate closed. The calc note states that signals will be set based on CDS header and polisher flow. No setpoint is yet determined.
Area of Impact Plant control
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
93 | P a g e
Urgent Alarm Occurs During Case 2 CRE
SCR-DR-6267 This issue impacts the following RO/SRO task: RO-INC-PLS-003-06 Perform grey rod exchange
Disposition SNC evaluated this issue and determined that it does not impact the suitability of the simulator for the conduct of operating tests.
This is an AP1000 plant design issue. The simulator models the plant design. If operators encounter this condition, they will follow procedural guidance. The procedures provide the steps necessary for operators to respond to the event.
Description The Urgent Failure Alarm (UA) occurs when MA and MD banks are in motion and the Tavg-Tref deviation requires the AO rods to move to restore Tavg-Tref back into band. This only occurs if MA and MD rods are in motion. For plant conditions where only the MA or MD rods are in motion and the Tavg-Tref deviation requires AO rods to move, then an UA does not occur.
The UA appears to be a timing issue that occurs only when MA and MD banks are both in motion when the Tavg-Tref deviation occurs. Basically, the Ovation controllers briefly generate a RODS IN and a RODS OUT signal to the MA bank and a RODS IN and a RODS OUT signal to the MD bank which results in a UA from the Power Cabinets.
Area of Impact Reactivity control
IDS Charger Capacity and Design Float Voltage Requirement are Incompatible
SCR-DR-6400 This issue impacts the following RO/SRO task: RO-LT-R-IDS.002 Monitor the IDS
Disposition
SNC has determined that this issue does not impact the simulator’s suitability of the simulator for the conduct of operating tests. This discrepancy is due to partially implementing a forthcoming design change in different documents. The output voltage of the IDS chargers will be increased at a later date. The simulator is properly modeling the design documentation that it was built to. The lower output voltage indication will not drive operators to perform or not perform any actions.
Description Reference documentation for IDS currently states the rated voltage for the battery charger is limited to 250 VDC. Other documentation also states that the battery (IDS) should be normally on a float charge of 264 VDC. This cannot happen without a larger battery charger.
Area of Impact Plant Control
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
94 | P a g e
Residual Bus Transfer Issues
SCR-DR-6481 This issue impacts the following RO/SRO task: AP-LT-R-ECS.012 Block fast bus transfer
Disposition
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The current simulator implementation does not provide the capability of the instructor to insert a malfunction that will result in the actuation of a Residual Bus Transfer. However, the Fast Bus Transfer and the Diesel Generator starting sequence function properly and provide the capability to examine the operators on electrical failures that would result in similar indications and Abnormal Operating Procedure entries. Description For undervoltage conditions (loss of power) sensed by 27B2 (two-out-of-two or two-out-of-three logic) in conjunction with a source undervoltage condition sensed by 27S, the UAT breaker will be tripped, leading to automatic closing of the RAT breaker completing residual bus transfer after establishing that the RAT source is live (59S1), the bus is dead (27B2), and all motor feeders are tripped.
The sequence of events for a residual bus transfer to occur is as follows:
• At 75% rated voltage (~ 3 sec time delay) the load shed occurs. The associated bus output breakers are tripped open. FOR ES-1 and 2 ONLY the associated DG starts.
• At 30% rated voltage the residual bus transfer occurs, the UAT source supply breaker opens and the RAT source supply breaker shuts for the associated bus.
This is not modeled currently.
Area of Impact Plant Control
ECS-EC-313 Loads not modeled
SCR-DR-6593 This issue impacts the following RO/SRO task: AP-LT-R-ECS.005 Monitor the ECS
Disposition The Aggregate Study team determined this issue does not affect the operators because they will be fairly transparent. The students will most likely not notice these issues during training. The team determined this issue in the aggregate is acceptable. While there will be manual actions that occur, it is still acceptable and should be addressed in a training moment.
Description APP-ECS-E3-EC31302 Rev 1 depicts ECS-ET-3131. Model requirements document SV3-STS-J4-119 Rev 1 required WEC/GSE to model the loads of EC-313. Work in SCR-DR-6568 determined they neglected to meet this requirement. SV3-STS-J1-119 Rev 0 shows loads off of EC-312 but nothing on EC-313.
Area of Impact Plant response
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
95 | P a g e
D/G Sequencer Operation
SCR-DR-6610 This issue impacts the following RO/SRO task: RO-ELE-ZOS-002-06 Perform priority load AC load sequencing and verify proper 6.9 kV priority load system voltage
Disposition The Aggregate Study team determined this issue does not affect the operators because they will be fairly transparent. The students will most likely not notice these issues during training. The team determined this issue in the aggregate is acceptable. While there will be manual actions that occur, it is still acceptable and should be addressed in a training moment.
Description According to APP-ZOS-E0C-001 (Onsite Standby Power System Diesel Generator Sizing Calculation) Rev 1, the following busses should get re-energized by the sequencer at Load Step [
]a,c
[ ]a,c [ ]a,c [ ]a,c [ ]a,c [ ]a,c [ ]a,c
These busses are not currently sequenced once ECS-ES-1/2 gets re-energized from its respective DG.
Area of Impact Effective plant response to loss of power
Turbine Bypass Control Valve Control Logic cannot Support Design Power Supplies
SCR-DR-6634 This issue impacts the following RO/SRO task: RO-SEC-MSS-003-01 Monitor MSS system and component parameters
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The turbine bypass control valves (MSS-PL-V001, MSS-PL-V002, MSS-PL-V003, MSS-PL-V004, MSS-PL- V005, MSS-PL-V006) have four solenoid valves that are designed to share two different power supplies. However, the current design control logic) uses one power supply for all 4 solenoids.
Area of Impact Plant operations during loss of power
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
96 | P a g e
Fire Protection System is not modeled in Containment
SCR-DR-6657 This issue impacts the following RO/SRO task: RO-SUP-FPS-002-00 Operate the FPS
Disposition This issue does not impact the suitability of the simulator for the conduct of operating tests. Due to the low importance values in NUREG-2103 of the items associated with this failure no simulator scenarios have been developed that would make use of this malfunction. The modeling of this malfunction will be investigated as time permits and corrected if possible.
Description While attempting to create a FR-Z.2 (Response to High Containment Level) scenario, it was discovered that a leak from the Fire Protection System (FPS) header in containment had no effect on any containment parameters (ex. Containment Sump Level, Containment Humidity). FPS Containment Spray also had no effect on any containment parameters.
Area of Impact Plant control
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
97 | P a g e
10 CFR 55.45(a)(7) Safely operate the facility’s heat removal systems, including primary coolant, emergency coolant, and decay heat removal systems, and identify the relations of the proper operation of these systems to the operation of the facility.
Executive Summary
Issues 6701, 6217, and 6179 are not issues impacting simulator training. For example, Time to boil is not a decision-making factor.
Issue 6022 currently has a work-around in place. Numerous work-around solutions become problematic, but there are currently few work-around solutions in place. The instructors make the present work-around solutions transparent. Continued work-around solutions are also tedious for the operators and it causes the NAPs to lose credibility.
Issue 5594 was determined not to impact reactor safety because it causes safeguards when it is required.
Issue 5968 only occurs during solid plant conditions so the impact to training is minimal. With operator action, simulator scenarios can still be successful. Also, if pressure control wasn’t a focus for one operator when the plant is solid, then this could be an issue, but that isn’t the concern.
Issue 6151 is mitigated by the manual operation of these valves when this issue manifests.
Issue 6634 was evaluated as not having an aggregate impact on operations.
The team determined that no additional reinforcements were required for completion of these tasks. Therefore, the team evaluated the collection of these issues as acceptable in the aggregate.
Repeatability issues involving CL 1B
SCR-DR-5594 This issue impacts the following RO/SRO task: RO-PRO-AOP-054-00 Respond to Reactor Coolant Pump Malfunctions using AOP-114 Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). This issue was further evaluated and determined to relate with a current design issue involving a lead/lag circuit associated with the main steam line pressure detection and input to the Safeguards ESF actuation. The pressure drop in the main steam line turns very close to the actuation set-point and the lead/lag circuit amplification of the rate of change may or may not cause the actuation. This is the expected plant response with the current actuation logic/software. Operators have sufficient procedure guidance directing them to respond to this event. For this reason, it was determined that this issue does not impact the suitability of the simulator for the conduct of operating tests.
Description
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
98 | P a g e
The 1B RCP shaft shear malfunction may cause an unexpected safeguards depending on the initial conditions the malfunction is inserted. Due to response caused by the lead/lag filters in the steam line pressure logic, the signal is driven into the dead band range of actuation at certain initial steam pressures. Elevating the initial steam pressure completely mitigates this occurrence.
Area of Impact Abnormal operation response
CVS-PT040 does not Provide Proper Protective Functions
SCR-DR-5968
This issue impacts the following RO/SRO task: RO-PRI-CVS-004-00 Monitor CVS operations
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description CVS-PT040 (pressure transmitter upstream of the letdown control valve) does not provide the proper protective functions for low pressure and high pressure protection in accordance with design documentation.
Per APP-CVS-M3C-101 Rev 6 Appendix C.3.2, at [ ]a,c when CVS-V047 is in automatic pressure control mode there is supposed to be a signal sent to “trip the CVS Makeup Pumps.” The high pressure signal is generated but does not trip the pumps; it presently feeds a Pump Auto Stop Demand signal. This signal will stop any pumps that are running in automatic only. When the plant is in water-solid mode, as determined in logic diagrams as having CVS-V047 in automatic pressure control mode, Chemical and Volume Control System (CVS) makeup pumps must be operated in manual; see APP-CVS-M3C-100 Rev 11 Logic Sheet CVS-4 Note10. Since the CVS makeup pumps are in manual the Auto Stop Demand signal will not shut the pump off to provide overpressure protection.
Per APP-CVS-M3C-101 Rev 6 Appendix C.3.2, at [ ]a,c when CVS-V047 is in automatic pressure control mode there is supposed to be a signal sent to “trip the Reactor Coolant Pumps…in order to protect them from reduced suction pressure.” This signal is also discussed in APP-RCS-M3C-100 Rev 9 Logic Sheet RCS-13 table and note 11. As presently designed there is no logic tie between CVS-PT040 and the Reactor Coolant Pumps (RCPs) to prevent damaging the RCPs upon a loss of Reactor Coolant System (RCS) pressure.
Area of Impact Plant Control
DHC Summary - Assembly Move NAP Function Not Functional
SCR-DR-6022 This issue impacts the following RO/SRO task: RO-INC-PLS-004-03 Perform decay time surveillance
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the minimal training impact this particular NAP has.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
99 | P a g e
Description When attempting to simulate fuel assemblies being moved from the core to the Spent Fuel Pool (SFP), it was noted that the Decay Heat Calculation (DHC) NAP to maintain the administrative location of fuel does not work correctly. On display 40203 the assembly move buttons on the lower right portion indicate they are only available when the light DDS-AP-DHC Status indicates it is ACTIVE. This light is driven off of the automatic mode selector and is INACTIVE when in MODES 1&2 and ACTIVE in MODES 3- 6. However, when the light indicates INACTIVE the buttons for moving are raised and available. When the light changes status to ACTIVE the buttons for moving are grayed out and no longer available. The light being active or inactive is currently driven by the auto mode selector and becomes active in MODES 3-6. However, fuel cannot be moved from the core into the SFP in any MODE other than MODE 6. The light should be driven by the manual input of the Rx vessel head being removed or installed or upper internals position on display 40004.
Area of Impact Reactivity Management with regards to indication and administration
Excessive SFW Control Valve Cycling
SCR-DR-6151 This issue impacts the following RO/SRO task: RO-SEC-FWS-002-01 Monitor SFWS and MFWS system and component parameters
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). The reasoning was that the current controller tuning is at the best estimated values. Hot functional testing would provide as- built tuning. However, this issue did contribute to the aggregate impact of sections 3 and 5 with regards to secondary plant control.
Description At low pressure conditions less than 350 psig, the operator often has to take manual control of Startup Feed Water (SFW) control valves due to excessive cycling of the valves. Indicated flow rates range from 0 to greater than 600 gpm within 10 to 15 second cycles. This requires 100% of the operator’s attention until RNS can be placed into service removing cooldown function from the steam dumps.
Area of Impact Plant control during startup and shutdown
Time to Boil Calculation
SCR-DR-6179
Disposition
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The Time to Boil NAP is a tool that is used for information only. The NAP is active when in Mode 5/6
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
100 | P a g e
conditions for RCS time to boil and when fuel is present in the spent fuel pool for spent fuel pool time to boil. The value is displayed in exponential units versus hh:mm. This means that operators will need to convert the scientific notation values into hours and minutes. Although this does take a short amount of time, the net effect is that it does not remove the ability to monitor the time to boil and it has no impact on actions the operator may, or may not, take in response to plant conditions. Description When core exit temperature was 300oF, Thot was > 212oF and TTB was >0 (15-25 min range). Either the NAP is calculating TTB incorrectly or the inputs to the NAP are wrong. TTB should reflect actual plant conditions.
Area of Impact NAP
CMT WR Level Indications go Bad Quality
SCR-DR-6217 This issue impacts the following RO/SRO task: AP-LT-S-EOP.007 Direct implementation of E-1, AP1000 Loss of Reactor or Secondary Coolant
Disposition The Aggregate Study team determined this issue as not impacting simulator training due to only providing indication function. All protective functions are still available via the CMT Narrow Range level instruments.
Description The WR Core Makeup Tank (CMT) level indications shift to Bad Quality once Automatic Depressurization System 1-3 (ADS 1-3) Actuate. Prior to this event, they would toggle to Bad Quality intermittently. The Bad Quality status is on indications PXS-LT009A/B & -LT010A/B (on PXS Supplemental Ind. Screen) and DDS-RSA11-L1 & DDS-RSA13-L1 (on WPIS screen 60017). The NAP driving the calculation of this indication drives them to bad quality whenever it determines voiding is occurring in the CMT (which is expected per design transients).
Area of Impact Plant control during decay heat removal
Turbine Bypass Control Valve Control Logic cannot Support Design Power Supplies
SCR-DR-6634 This issue impacts the following RO/SRO task: RO-SEC-MSS-003-01 Monitor MSS system and component parameters
Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.
Description The turbine bypass control valves (MSS-PL-V001, MSS-PL-V002, MSS-PL-V003, MSS-PL-V004, MSS-PL- V005, MSS-PL-V006) have four solenoid valves that are designed to share two different power supplies.
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
101 | P a g e
However, the current design control logic) uses one power supply for all 4 solenoids.
Area of Impact Plant operations during loss of power
IRWST Temperature Response
SCR-DR-6701 This issue impacts the following RO/SRO task: RO-PRO-EOP-003-00 Respond to a loss of Reactor or Secondary Coolant using EEP-E-1
Disposition The Aggregate Study team determined this issue does not have an aggregate impact on operations. Additionally, this modeling of the differential temperatures has no impact on operator actions because, by procedure, there are no operator actions related to IRWST temperature. For this reason, it was determined that this issue does not impact the suitability of the simulator for the conduct of operating tests.
Description [
]a,c
Area of Impact Plant response
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
102 | P a g e
10 CFR 55.45(a)(8) Safely operate the facility’s auxiliary and emergency systems, including operation of those controls associated with plant equipment that could affect reactivity or the release of radioactive materials to the environment.
Executive Summary
It would be desirable to repair the Core Makeup Tank Nuclear Application (CMT NAP) issue in this section (6217), but it doesn’t affect the section in aggregate. When issues 5546 and 6019 manifest on the simulator, the operators will still comply with Technical Specifications requirements. An additional Tech Spec call would have to be made with 6019 since there is an issue with Zinc addition in the CVS when power is lost. Ultimately, the containment isolation is the issue, but the team believes that this can be dealt with safely.
Issue 5546 will be transparent with most operators without a reference diagram and detailed exploration on the Ovation status screens. Ultimately, this is not an issue for the aggregate study.
EDS Power Supply Assignments to PLS/DDS Cabinets Incomplete
SCR-DR-5546 This issue impacts the following RO/SRO tasks: RO-LT-R-EDS.001 Monitor the Non Class 1E DC and UPS system (EDS) for proper operation RO-LT-R-EDS.004 Respond to a loss of EDS DC power
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). SMEs determined that the current power supply arrangement was adequate to teach since it is per design documentation. Power supplies are an item that will be continuously taught as they are updated and changed.
Description A loss of individual Non Class 1E DC and UPS System (EDS) busses will result in incomplete system response. Some Ovation drops are not dynamically powered by the EDS model but are powered by a permanently energized model constant (specifically DPU047, DPU048, and DPU044). The load lists for the STS do not assign a power supply to all the Ovation drops so there is no plant design data to insert into the simulator.
Area of Impact Effective plant response to loss of power
CVS-V094 Power Failure Response
SCR-DR-6019 This issue impacts the following RO/SRO task: RO-INC-PMS-005-02 Respond to a loss of a PMS division or failure of PMS components
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). When this issue manifests itself on the simulator, the operators will still comply with Technical
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
103 | P a g e
Specifications requirements. An additional Tech Spec call would have to be made since there is an issue with Zinc addition in the CVS when power is lost. For this reason, this issue does not impact the suitability of the simulator for the conduct of operating tests.
Description CVS-V094 does not close upon a loss of power to ILCA02 as expected. It did close on loss of power to ILCA03, which is not in accordance with design documentation.
Area of Impact Response to power loss
CMT WR Level Indications go Bad Quality
SCR-DR-6217 This issue impacts the following RO/SRO task: AP-LT-S-EOP.007 Direct implementation of E-1, AP1000 Loss of Reactor or Secondary Coolant
Disposition The Aggregate Study team determined this issue as not impacting simulator training due to only providing indication function. All protective functions are still available via the CMT Narrow Range level instruments.
Description The WR Core Makeup Tank (CMT) level indications shift to Bad Quality once Automatic Depressurization System 1-3 (ADS 1-3) Actuate. Prior to this event, they would toggle to Bad Quality intermittently. The Bad Quality status is on indications PXS-LT009A/B & -LT010A/B (on PXS Supplemental Ind. Screen) and DDS-RSA11-L1 & DDS-RSA13-L1 (on WPIS screen 60017). The NAP driving the calculation of this indication drives them to bad quality whenever it determines voiding is occurring in the CMT (which is expected per design transients).
Area of Impact Plant control during decay heat removal
Fire Protection System is not modeled in Containment
SCR-DR-6657 This issue impacts the following RO/SRO task: RO-SUP-FPS-002-00 Operate the FPS
Disposition This issue does not impact the suitability of the simulator for the conduct of operating tests. Due to the low importance values in NUREG-2103 of the items associated with this failure no simulator scenarios have been developed that would make use of this malfunction. The modeling of this malfunction will be investigated as time permits and corrected if possible.
Description While attempting to create a FR-Z.2 (Response to High Containment Level) scenario, it was discovered that a leak from the Fire Protection System (FPS) header in containment had no effect on any containment
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
104 | P a g e
parameters (ex. Containment Sump Level, Containment Humidity). FPS Containment Spray also had no effect on any containment parameters.
Area of Impact Plant control
IRWST Temperature Response
SCR-DR-6701 This issue impacts the following RO/SRO task: RO-PRO-EOP-003-00 Respond to a loss of Reactor or Secondary Coolant using EEP-E-1
Disposition The Aggregate Study team determined this issue does not have an aggregate impact on operations. Additionally, this modeling of the differential temperatures has no impact on operator actions because, by procedure, there are no operator actions related to IRWST temperature. For this reason, it was determined that this issue does not impact the suitability of the simulator for the conduct of operating tests.
Description [
]a,c
Area of Impact Plant response
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
105 | P a g e
10 CFR 55.45(a)(9) Demonstrate or describe the use and function of the facility’s radiation monitoring systems, including fixed radiation monitors and alarms, portable survey instruments, and personnel monitoring equipment.
Executive Summary
There is only a single issue categorized here and therefore the team did not deem this an aggregate issue.
Simulator MCR missing Rad Monitoring Panel
SCR-DR-237 This issue impacts the following RO/SRO task: RO-INC-RMS-003-00 Startup and operate radiation monitors
Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). There is little training value with this panel at this time. The indications that would otherwise be provided by this panel are available in Ovation. Therefore, operator actions are not impacted and it does not impact the suitability of the simulator for the conduct of operating tests.
Description Simulator MCR does not have the radiation monitoring panel on the back wall as depicted in the design reference.
Area of Impact Physical fidelity of simulator MCR and radiation control
ND-15-1333
Enclosure 6
VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Southern Nuclear Operating Company
Vogtle Electric Generating Plant (VEGP) Units 3 and 4
ND-15-1333
Enclosure 7
List of Westinghouse Simulator Corrective Actions
(This Enclosure consists of 8 pages, including this cover page)
ND-15-1333 Enclosure 7, Page 2 of 8 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 1.0 Summary Description of Westinghouse’ Simulator Corrective Actions
In July, SNC commissioned a team to perform an aggregate study of all open
simulation discrepancies.
As of May 15, 2015, SNC had 166 open discrepancies associated with its simulator.
SNC evaluated these discrepancies and their impact on the suitability of the
simulation facility for the conduct of operating tests. Out of the 166 discrepancies,
101 were determined to be relevant to 9 of the 13 criteria listed under 10 CFR
55.45(a). SNC evaluated each of these 101 discrepancies and determined that no
singular issue challenged the simulation facility’s suitability for the conduct of
operating tests. SNC then commissioned a team with representatives from
Operations, Training, and Human Factors Engineering to perform a study to
determine if, in aggregate, these 101 discrepancies presented a challenge to the
simulation facility’s suitability for the conduct of operating tests. The team
determined that, in the aggregate, the discrepancies could challenge the suitability
of the simulator for the conduct of operating tests. Specifically, 10 CFR 55.45(a)
criterions (3) and (5) were impacted. It was further determined that these criteria
would no longer be challenged if a specified subset of discrepancies could be
corrected by WEC.
On July 7 h, SNC confirmed that the scope of this subset satisfied the near term
request to support the VEGP Units 3&4 STS CAS.
2.0 Itemized List of Westinghouse’ Simulator Corrective Actions
Table E7-1 lists the items that Westinghouse included in a patch delivered to SNC
on August 14, 2015. Verification and Validation (V&V) testing was performed and
the patch was deployed in simulator load V3.R1.7F8.1.1.0.
ND-15-1333 Enclosure 7, Page 3 of 8 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E7-1
List of Westinghouse’ Simulator Corrective Actions
A.S.
# {1}
NRC
UI # {*}
WEC RITS/CAPA
I:
SNC SCR
#
Other Ref #:
Summary V&V Test Results
1 RITS 37569 5584 Rod Withdrawal button un-highlights during continuous operation
Pass
2
RITS 39523
RITS 39409
CAPAL-100025685
5597 TO-117
Containment Radiation Alarm Reset Points
Acceptance criteria not fully met.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This is an alarm deadband issue that WEC must resolve.
If operators encounter this condition, they will follow their procedures. The procedure provides the steps necessary for operators to respond to the condition.
3 RITS 39633 5599 TO-09
Unidentified and Identified Leak Rate always
indicates BAD quality
Acceptance criteria not fully met.
See Table E5-1 Item 40.
4 1 5627 TO-40
Sub criticality on Critical Safety Function Screen bad quality
Pass
5 RITS 37623 5643 1503-
02 VWS-TE079 Labeled incorrectly Pass
6 5644 SSS Display # 17600 incorrect
Acceptance criteria not fully met.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This is a simulator I&C issue related to a non-consequential graphic indication.
7 RITS 39605 RITS 42463
5688 TO-25
Graphic 50308 Issue Pass
ND-15-1333 Enclosure 7. Page 4 of 8 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E7-1 (continued)
A.S.
# {1}
NRC
UI # {*}
WEC RITS/CAPA
I:
SNC SCR
#
Other Ref #:
Summary V&V Test Results
8 RITS 39466 5689 TO-28
PMS Mimic Screens
Acceptance criteria not fully met.
(This item was not included in the patch received from WEC.)
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This is a simulator I&C issue.
PMS mimic screens are used for verification of indications only. If a question arises regarding an indication on the PMS mimic screen, operators will use primary indications from the PMS displays on the Primary Dedicated Safety Panel (PDSP). No operator action is available through the PMS mimic screens. All actions must be taken from the division’s PMS PDSP.
9 RITS 42477 5702 1501-
07 IDS screens in simulator show inaccurate Power supplies
Pass
10 5712 1411-
06 Calorimetric Power Data Points do not have required precision
Pass
11
14
5813
1503-16
ISV Pri-1 HED-14 Alarm Overload Pass
12 RITS 42461 5909 TO-59
Digital Rod Control System (DRCS) M bank rod control graphic
Pass
13 RITS 38522 5920 1410-
6
Pressurizer narrow range pressure does not indicate bottom of scale on WPIS 2 for Mode 1-4 screen
Pass
ND-15-1333 Enclosure 7. Page 5 of 8 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E7-1 (continued)
A.S.
# {1}
NRC
UI # {*}
WEC RITS/CAPA
I:
SNC SCR
#
Other Ref #:
Summary V&V Test Results
14 5924 1410-
2 DRPI Health Screen has alarms for Data Cabinet A and B crossed
Acceptance criteria not fully met.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This is a simulator I&C issue related to a non-consequential graphic indication.
15 5925 1410-
3 DRPI Health Screen (1805) Incorrect Logic Cabinet Alarms
Pass
16 26 RITS 38825 5968 1504-
09 CVS-PT040 do not function as described Pass
17 RITS 40030 6009 1504-
3
Uncontrolled H/U or C/D light on Mode 5/6 CSFST WPIS Display does not indicate correctly
Pass
18 6030 M Banks B & C Reversed on DRPI Health Screen
Pass
19 6078 HSS Display does not include ESOP Disch Pressure
Pass
20 RITS 38306 6089 NAP for 1/M Intermediate Range does not work
Acceptance criteria not fully met.
See Table E9-2.
21
RITS 43153 6102
MA Bank Rods Sometimes Stop at 263 steps during a CRE (Control Rod Exchange)
Pass
(This item was corrected locally by SNC. It was not included in the patch received from WEC.)
22 RITS 41846 6129 Display 40023 units issue Pass
23 3 6144 TO-52
PLS Auto Plant Mode Selector not consistent with Mode 3 to Mode 2 entry
Pass
24 40 RITS 39434 6159 NAPS display issues Pass
ND-15-1333 Enclosure 7. Page 6 of 8 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E7-1 (continued)
A.S.
# {1}
NRC
UI # {*}
WEC RITS/CAPA
I:
SNC SCR
#
Other Ref #:
Summary V&V Test Results
25 RITS 39470 6160 CCS Screen issue Pass
26 RITS 39588 6164 WPIS downscale arrow issue
Acceptance criteria not fully met.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This is a simulator I&C issue.
This serves as backup indication only and has no impact on operator decision making.
27 RITS 39403 6165 WPIS Tavg Scale Pass
28 RITS 40059 6169 Erroneous NAP RSA behavior Pass
29 RITS 39783 6170 WRS graphic issue Pass
30 RITS 38601 6180 TTB unit indication Acceptance criteria not fully met.
See Table E5-1 Item 40.
31 RITS 40314 RITS 41693
6187 Rod stop logic Pass
32 RITS 41193 6259 Received Bank Sequence Out of Sequence Alarm
Pass
33
6267
Urgent Alarm (Causes control rods to swap to manual and stop) Occurs During Case 2 CRE
Acceptance criteria not fully met.
(This item was not included in the patch received from WEC.)
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This is an AP1000 plant design issue. The simulator models plant design.
If operators encounter this condition, they will follow their procedures. The procedure provides the
ND-15-1333 Enclosure 7. Page 7 of 8 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E7-1 (continued)
A.S.
# {1}
NRC
UI # {*}
WEC RITS/CAPA
I:
SNC SCR
#
Other Ref #:
Summary V&V Test Results
steps necessary for operators to respond to the condition.
34 6278 Battery bank indications are mislabeled for EDS1, EDS2, and EDS4
Pass
35
6302
Improper Bank Overlap Occurs when Data Point OCB07CE00C_OUTAV Increments Improperly During Startup
Closed. Invalid.
Subsequent to the Aggregate Study team forwarding this issue to WEC for correction, WEC evaluated the issue and determined that this is per the AP1000 design and that the overlap is not improper. Therefore, no correction was required.
36 6315 Manual reactor trip alarm when one is not requested
Pass
37 6398 Screen 22101 has incorrect indication
Acceptance criteria not fully met.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This is a simulator I&C issue related to a non-consequential graphic indication.
38 6409 1501-
05 Graphic 1805 has reversed rods Pass
6483 1502-
13 During Load Rejection events, Load Unbalance response is inconsistent
Pass
39 6621 PR B lower detector failure is not compensated for by the RSA NAP
Pass
40 14 6651 Inconsistent priority levels of Data Processing Unit alarms.
Pass
41 RITS 45989 6698 Safety Mimic Display for SGS-V255A and B indicates BQ following a SFW Isolation
Pass
ND-15-1333 Enclosure 7. Page 8 of 8 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E7-1 (continued)
A.S.
# {1}
NRC
UI # {*}
WEC RITS/CAPA
I:
SNC SCR
#
Other Ref #:
Summary V&V Test Results
35 None 1503-
15 Failure to identify CCS leak Pass
28 6830 1410-
07 Steam Dump Capacity Pass
37 RITS 40480 5679 TO-04
EDS performance on Battery Pass
5 5987 TO-89
Condensate Makeup flow rate Pass
10 RITS 42360 5722 1411-
03 MTS Alarm at 75% Power Pass
9 RITS 39748 5609 TO-131
Gland Seal Steam system pressure discrepancy between procedure and actuals
Pass
Notes: {*}
Numbers and descriptions correspond to the table “Summary of Unresolved Items as of 06-30-2015” as it appeared in an NRC letter dated July 2, 2015,
(Reference 1). {1}
Numbers correspond to the items as they appear in the “List of Proposed Actions” on page 13 of the Aggregate Study, Enclosure 6.
3.0 References
1. Virgil C. Summer Nuclear Station Units 2 and 3 - Request For A Commission-Approved Simulation Facility dated July 2, 2015 -
ML15182A097
Southern Nuclear Operating Company
Vogtle Electric Generating Plant (VEGP) Units 3 and 4
ND-15-1333
Enclosure 8
Evaluation of Priority One (1) Potential Human Engineering Discrepancies (PHEDs) from Integrated Systems Validation (ISV) Daily Assessments
(Non-Proprietary)
(This Enclosure consists of 6 pages, including this cover page)
ND-15-1333 Enclosure 8, Page 2 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 1.0 Summary of Priority One (1) Potential Human Engineering Discrepancies (PHEDs)
from Integrated Systems Validation (ISV) Daily Assessment
Integrated System Validation (ISV) was conducted as part of the Human Factors Engineering
(HFE) Verification and Validation process. Conduct of the AP1000 ISV resulted in a number
of Potential HEDs (PHEDs). These PHEDs have the potential to impact the simulator, plant
design, operator training, and/or procedures.
In order to elevate simulator fidelity to a level suitable for the conduct of operating tests
commensurate with the requirements of 10 CFR 55.45(b), Priority 1 PHEDs specific to the
operation of the simulator have been evaluated and corrected. A list of all ISV Priority 1
PHEDs and how they were resolved appears in Table E8-1. Table E8-1 also includes
PHEDs related to other elements of the Integrated System Validation, specifically procedures
and training.
2.0 PHED Assessment
Westinghouse performed a preliminary assessment of ISV issues during the conduct of the
ISV to identify those scenario failures that could have benefited from a fourth trial run while
the Utility Crews were present. These assessments identified 15 Priority 1 PHEDs.
Westinghouse’ resolution of these 15 PHEDs identified four items requiring simulator model
or HSI changes and 11 PHEDs specific to procedures and training. The changes resolved
the Alarm Workload PHED.
ND-15-1333 Enclosure 8, Page 3 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E8-1
Priority 1 PHEDs
# PHED Description - (# trial failures of # trials
performed) Resolution
1
[
]a,c
([ ]a,c)
[ ]a,c
[ ]a,c
[ ]a,c
[ ]a,c
2
[
]a,c
([ ]a,c)
[ ]a,c
[ ]a,c
[ ]a,c
[ ]a,c
[ ]a,c
[ ]a,c
[ ]a,c
[ ]a,c
3
[
]a,c
([ ]a,c)
[ ]a,c
[ ]a,c
[ ]a,c
4
[
]a,c
([ ]a,c)
[ ]a,c
[ ]a,c
[
]a,c
ND-15-1333 Enclosure 8, Page 4 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E8-1 (continued)
# PHED Description - (# trial failures of # trials
performed) Resolution
[ ]a,c
[ ]a,c
5
[
]a,c.
([ ]a,c)
[ ]a,c
[
]a,c
6
[
]a,c
([ ]a,c)
[ ]a,c
7
[
]a,c
([ ]a,c)
[ ]a,c
8
[
]a,c
([ ]a,c)
[ ]a,c
9
[
]a,c
([ ]a,c)
[ ]a,c
[ ]a,c
[
ND-15-1333 Enclosure 8, Page 5 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E8-1 (continued)
# PHED Description - (# trial failures of # trials
performed) Resolution
]a,c
[
]a,c
10
[ ]a,c
([ ]a,c)
[ ]a,c
[ ]a,c
[
]a,c
11
[
]a,c
([ ]a,c)
[ ]a,c
[ ]a,c
[ ]a,c
[ ]a,c
[
]a,c
12
[
]a,c
([ ]a,c)
[ ]a,c
[ ]a,c
[ ]a,c
[ ]a,c
[
]a,c
13
[
]a,c
([ ]a,c)
[ ]a,c
[ ]a,c
[ ]a,c
[ ]a,c
ND-15-1333 Enclosure 8, Page 6 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E8-1 (continued)
# PHED Description - (# trial failures of # trials
performed) Resolution
14
[
]a,c
[ ]a,c
[ ]a,c
[ ]a,c
[ ]a,c
[
]a,c
[
]a,c
Using the above criteria, WEC provided an update that enhances alarm prioritization, greatly lowering the number of alarms that have to be addressed by the operator post-event. In addition, SNC enabled the [
]a,c that is part of the Alarm Presentation System design. As can be seen from the results below, the number of alarms are now in an acceptable range and the operator workload has been greatly reduced.
Event Alarms
Turbine Trip [ ]a c
Reactor Trip [ ]a c
ES1 [ ]a,c
ES2 [ ]a c
ES3 [ ]a,c
ES4 [ ]a,c
ES5 [ ]a c
ES6 [ ]a,c
LOOP with TT [ ]a,c
LOOP= loss of offsite power TT= turbine trip
15
[
]a,c
([ ]a,c)
[ ]a,c
[
]a,c
Southern Nuclear Operating Company Vogtle Electric Generating Plant (VEGP) Units 3 and 4
ND-15-1333
Enclosure 9
List of Open Simulator Discrepancies (Non-Proprietary)
(This Enclosure consists of 23 pages, including this cover page)
ND-15-1333 Enclosure 9, Page 2 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility Summary Description of Open Simulator Discrepancies
There were 166 open simulator discrepancies as of May 15, 2015. Each discrepancy was
evaluated to determine if it impacted any of the criteria listed under 10 CFR 55.45(a). Out of the
166 discrepancies, 101 were determined to be relevant to the 13 criteria listed under 10 CFR
55.45(a). Since SNC’s evaluations and conclusions for these items are reflected and/or
imbedded in the Aggregate Study contained in Enclosure 6, they are not repeated here. Table
E9-1 lists these 101 items.
The remaining 65 discrepancies that were determined not to pose a challenge to the criteria
listed under 10 CFR 55.45(a) are listed in Table E9-2. A Training Needs Assessment as
defined in ANSI/ANS-3.5-1998, Section 4.2.1.4, was performed for these 65 discrepancies It
was determined that none of these issues impacted any of the six criteria listed under
ANSI/ANS-3.5-1998, Section 4.2.1.4. or any of the 13 criteria listed under 10 CFR 55.45(a).
ND-15-1333 Enclosure 9, Page 3 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-1
SCRs Impacting 55.45(a) Criteria Evaluated In Aggregate
SCR# Description Current Status
216 VCS fan response due to loss of power Open
237 Simulator MCR missing Rad Monitoring Panel Open
5546 EDS Power Supply to PLS/DDS cabinets not IAW EDS Load List Open
5577 RCP Net Positive Suction Head Curve - Display 60029 needs extension Open
5583 BEACON inoperability calculation in reactor core model Open
5584 Rod Withdrawal button un-highlights during continuous operation Closed
5593 Unstable VFS Containment Exhaust Flow Open
5594 Repeatability issues involving CL 1B Open
5597 Containment radiation alarm reset points Open
5598 PZR heater current indicates BAD quality at limits Open
5599 Unidentified and Identified Leak Rate always indicates BAD data Open
5603 Investigate validity of low flow alarm on TCS-FT007 Open
5609 GSS Header pressure will not maintain pressure as required Closed
5613 MFP 'B' Alarm Response Differs For Identical Fault Closed
5619 Stage 3 ADS box unused on Divisions C and D Open
5621 Problems with Alarm Cutout and RWS pressure alarms Open
5623 PZR pressure out of range indication not properly displayed Open
5627 Subcriticality on Critical Safety Function Screen bad quality Closed
5643 VWS-TE079 Labeled incorrectly Closed
5655 Plant issue with automatic control of DST level and autostart of standby Condensate pump
Open
5680 PMS Screens not updating during MCR/RSR transfer Open
5686 Degasifier Level Alarm Limits Open
5689 PMS Mimic Screens Open
5698 ISS freeze during PZR fill to solid Open
5707 HL Level and PZR WR Level Fluctuations during Midloop when filling In-Containment Refueling Water Storage Tank (IRWST) with CVS
Open
5712 Calorimetric Power Data Points do not have required precision Closed
5736 OPDMS RIL for M2 does not match COLR Rev. 0 Open
5813 Valve Modulating Status Alarms are a nuisance Closed
5828 VRS High Rad alarm when ES1 de-energizes and is re-energized from DG Closed
5903 Inconsistent OPDMS QPTR Indications Open
5910 VFD Transformer temperature Closed
5913 Print from NAP SRM not working Open
5914 VHS rad monitor response on loss of flow Closed
5920 Pressurizer narrow range pressure does not indicate bottom of scale on WPIS 2 for Mode 1-4 screen
Closed
5921 During simulator scenario validation CDS-TE040A/B range found to be inadequate. Open
5924 DRPI Health Screen has alarms for Data Cabinet A and B crossed Open
5925 DRPI Health Screen (1805) Incorrect Logic Cabinet Alarms Closed
5926 Startup Feedwater Control Valve is opening and closing with level high is associate Open
ND-15-1333 Enclosure 9, Page 4 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-1 (continued)
SCR# Description Current Status
steam generator
5968 CVS-PT040 Does not provide the proper protective functions Closed
5972 CNMT Recirc Actuation for Div C and D not visible on PMS ESF Screen Open
6009 Uncontrolled H/U or C/D light on Mode 5/6 CSFST WPIS Display does not indicate correctly
Closed
6019 CVS-V094 Power Failure Open
6022 DHC Summary - Assembly Move NAP function doesn't work Open
6025 RCP Vibration Alarms Open
6030 M Banks B & C Reversed on DRPI Health Screen Closed
6038 Quality of RWS-V503 BAD at limits Open
6068 WLS-MP-08C will not pump monitor tank C Closed
6071 RCP stator temperature indication off scale low at lower speeds Closed
6078 HSS Display does not include ESOP Disch Pressure Closed
6089 NAP for 1/M Intermediate Range does not work Open
6099 DWS-LT006 has insufficient range Open
6102 M Bank A Rods would not step out beyond 263 steps during a CRE Closed
6103 ECS penetration temperature reading off scale low Closed
6122 Improper function of C-2 Closed
6126 WRS sump pump B does not indicate proper discharge pressure Closed
6144 PLS Auto Plant Mode Selector not consistent with Mode 3 to Mode 2 entry Closed
6151 SFW control valve cycling Open
6152 SWS temperature control Open
6154 WPIS RCS inventory screen issues Open
6156 FWS-V037 Control issue Open
6157 SGS MSL drain pot erratic indication Open
6159 NAPS display issues Closed
6162 Stuck Rod Recovery Malfunction Open
6164 WPIS downscale arrow issue Open
6168 Tuning of VBS required Open
6169 NAP RSA behavior Closed
6171 APS ZVS and ZBS alarm scaling Open
6172 Polisher bypass valve control Open
6175 Flux doubling difference between divisions Closed
6179 TTB calculation Open
6186 Tracking issue for rod step sound problems Open
6190 WPIS display for VARs Open
6192 VFS radiation monitoring Closed
6217 CMT WR Level Indications go Bad Quality Open
6259 Received Bank Sequence Out of Sequence Alarm Closed
6267 Urgent Alarm during Case 2 CRE at 90% Power Open
6302 Improper bank overlap occurs when data point OCB07CE00C_OUTAV is incremented during Rx startup
Closed
ND-15-1333 Enclosure 9, Page 5 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-1 (continued)
SCR# Description Current Status
6315 Manual reactor trip alarm when one is not requested Closed
6366 During Turbine Trips, Pri 4 controller faults for Drop 21 and 34 are received Closed
6392 Main Generator output breaker closes before it is supposed to when syncing the main generator to the grid.
Open
6398 Screen 22101 has incorrect indication Open
6400 IDS charger and documented float voltage are incompatible Open
6409 Graphic 1805 has reversed rods Closed
6481 Currently there is no way to cause a residual bus transfer of ECS-ES-1 thru 6 Open
6491 Diesel Fuel oil day tank level transmitters not operating correctly Open
6492 UAT Bkr Line Undervoltage Priorities are Wrong Open
6532 Any Rods at Bottom Alarm Open
6547 VES Supply Header Pressure not modeled correctly for a change in temperature w/o a change in mass
Closed
6593 Model ECS-EC-313 loads Open
6610 Potential issue with DG Sequencer Open
6612 Possible modeling and/or Ovation issues with WGS Sample Package MS-01 PS-001 Open
6613 Possible modeling and/or Ovation issues with WGS Sample Package MS-01 AT-032 (AE032)
Closed
6621 PR B lower detector was failed high, the RSA for Power Range Power did not eliminate this input causing an erroneous PR PWR read
Closed
6623 VZS-D014A/B and VZS-D015A/B do not fail as-is after loss of power. Open
6634 The Turbine Bypass Control Valve control scheme does not support multiple power supplies
Closed
6645 Battery Temperature does not trend during battery operations Closed
6651 Inconsistent priority levels of Data Processing Unit alarms. Closed
6657 Fire Protection System is not modeled in Containment Open
6670 PMS mimic screen navigation issues Open
6698 Safety Mimic Display for SGS-V255A and B indicates BQ following a SFW Isolation Closed
6701 Examine IRWST temperature Change Open
Total Remaining Open 62
Total Closed 39
ND-15-1333 Enclosure 9, Page 6 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2
SCRs Not Impacting 55.45(a) Criteria
SCR# Summary Description Evaluation Basis
233
Simulator MCR missing cooling fins:
Simulator MCR is missing cooling fins as designated by the Unit 3 design documentation
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
Missing [ ]a,c was determined to be a physical fidelity issue that imposes no operational restrictions on the indications available to or actions taken by the operator. A Training Needs Assessment was performed and it was determined that this issue does not impact any of the six criteria listed under ANSI/ANS-3.5-1998, Section 4.2.1.4 or any of the 13 criteria listed under 10 CFR 55.45(a).
235
Simulator MCR lights not hanging from chains:
Simulator Main Control Room (MCR) lights are not located as designated by the Unit 3 Design Document APP-1242-EL-001.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
The MCR lights not hanging from chains in the MCR was determined to be a minor physical difference between MCR and simulation facility that imposes no operational restrictions on the indications available to or actions taken by the operator. The current MCR lighting responds properly to loss of power scenarios. A Training Needs Assessment was performed and it was determined that this issue does not impact any of the six criteria listed under ANSI/ANS-3.5-1998, Section 4.2.1.4 or any of the 13 criteria listed under 10 CFR 55.45(a).
5540
ECS (1) Components Tagout Graphic display wrong:
When the Main AC Power System (ECS) ‘ECS (1) Components’ poke is selected from the Tagout Navigation Display, the graphic that appears is the CWS Components Tagout interface.
This issue has been corrected.
The display graphic was corrected with the recent simulator software update. The change has been tested and the graphics found to be functioning correctly.
5588
Tagged Components are not being captured in Snap:
When tagging out components, the current Initial Condition (IC) is snapped to incorporate this change in component status. When the system is reset to the same IC (newly snapped) the tagged out component does not reflect being tagged out. It seems the graphics are overriding the bits that allow the tagged-out component to remain in that state. After a second reset to the same IC, the graphics allow the component to be reflected as tagged-out (i.e., the correct status).
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This issue is of the same nature as that described in SCR 5590 and 5905. Many simulators allow the booth operator to save a scenario’s initial conditions. This gives the booth operator the ability to efficiently reset the simulator to the same conditions when a scenario is to be repeated. [
]a,c and the booth operator must perform this task manually. The
scenario guide used by the booth operator contains guidance to this effect.
ND-15-1333 Enclosure 9, Page 7 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2 (continued)
SCR# Summary Description Evaluation Basis
In either case, this activity is performed prior to trainees being admitted to the simulator and is completely transparent to the trainee.
A Training Needs Assessment was performed and it was determined that this issue does not impact any of the six criteria listed under ANSI/ANS-3.5-1998, Section 4.2.1.4 or any of the 13 criteria listed under 10 CFR 55.45(a).
5590
Clearance Tags Remaining after Simulator Reset:
Several components were tagged-out while running a scenario; the current Initial Condition (IC) was not snapped. The system was reset to a different IC and the components tagged-out in the previous IC were still reflecting a tagged-out status. The half of the components were manually cleared out and the system then was reset to a different IC, the components that were manually cleared remained clear and the other half remained tagged-out.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This issue is of the same nature as that described in SCR 5588 and 5905. The scenario guide used by the booth operator directs him/her to perform a check for clearance tags when resetting the simulator to a different initial condition.
A Training Needs Assessment was performed and it was determined that this issue does not impact any of the six criteria listed under ANSI/ANS-3.5-1998, Section 4.2.1.4 or any of the 13 criteria listed under 10 CFR 55.45(a).
5608
Simulator Operations (SIMOPS) datalink alarms incorrect:
When the Map and Mitigation Tools are setup to assign each individual datalink a unique entry in the configuration, the digital alarm points respond incorrectly when datalink failures are inserted.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
[ ]a,c These alarms
are inconsequential to the MCR operator as the only action that would be required would be to notify the associated building operator and I&C to investigate the local equipment. This issue does not impact indications available to or actions taken by the operator.
Westinghouse stated that this is a software conversion tool issue.
5614
Ovation indicating incorrect time after backtrack:
After placing the simulator in FREEZE a BACKTRACK was performed. Once the simulator was reset into the backtrack IC and placed in Run, it was observed: the real time on Ovation was incorrect.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
The BACKTRACK feature is a tool used during training scenarios to return to a given point in a scenario or evolution. The instructor uses this feature as a training tool to reinforce a particular point of interest relative to operator actions and plant transient responses. The existence of this fault is an inconvenience to the instructor and, worse case, deprives the instructor of an additional training tool. This issue is transparent to the trainees and does not impact the suitability of the simulator for the conduct of operating tests.
ND-15-1333 Enclosure 9, Page 8 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2 (continued)
SCR# Summary Description Evaluation Basis
It is the expectation during exam scenarios that the entire scenario would be reset and re-performed or a different scenario would be used for evaluation if a simulator fault occurred. This fault does not affect the indications available to or actions taken by operators.
A Training Needs Assessment was performed and it was determined that this issue does not impact any of the six criteria listed under ANSI/ANS-3.5-1998, Section 4.2.1.4 or any of the 13 criteria listed under 10 CFR 55.45(a).
WEC is aware of the issue and a RITS number was provided. Since this issue does not impact the suitability of the simulator for conducting operating tests, no further action is planned to be taken by SNC simulator staff and the issue will be closed only after a WEC solution is made available.
5618
MSS-V016A/B Response during reactor shutdown:
During a reactor shutdown, it is instructed to shut down the MSR. Based on Document APP-MSS-GJP-101 Attachment 4, step 4.3.3 when the turbine load is 10%, control valves should remain closed but they open 5% causing an undesired RCS temperature transient.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This issue is related to a controller design implementation error. [
]a,c which is not consistent with plant design.
SNC has developed an APP file to override the controller and close the valves at the required turbine load. This results in the controller error being transparent to the operator.
SNC will continue to use this APP file until a permanent correction is provided by WEC.
5644
SSS Display # 17600 incorrect:
The Secondary Sampling System (SSS) display in Ovation indicates incorrect flow passing these valves SSS-V920 or SSS-V921.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This issue is related to the system mimic presented on the graphical display. [
]a,c The error does not change the operation of the valves or the system, just improperly represents the process fluid flow paths that will be diverted when the valves reposition.
These displays are infrequently used during normal routine operations.
ND-15-1333 Enclosure 9, Page 9 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2 (continued)
SCR# Summary Description Evaluation Basis
5649
Reference plant procedure and limitations of CVS design to support plant heat-up are incompatible:
During a scenario it was discovered that the design of the CVS purification loop and letdown system was unable to handle a plant heat-up in accordance with GOP-107.
This is an AP1000 plant design issue. The simulator models plant design.
This issue does not impact the suitability of the simulator for the conduct of operating tests.
The AP1000 plant design allows for a heat-up rate that the letdown system cannot keep up with. [
]a,c
SNC operators are trained to follow their procedures. [
]a,c
A Training Needs Assessment was performed and it was determined that this issue does not impact any of the six criteria listed under ANSI/ANS-3.5-1998, Section 4.2.1.4 or any of the 13 criteria listed under 10 CFR 55.45(a).
SNC has determined that the guidance contained in the procedures is adequate to address the design issue and maintain safe operation of the plant.
This issue is similar to SCR-6158.
5656
Failed test-AP-OPS-EVO-003:
During ANSI ANS-3.5 test AP-OPS-EVO-003, Reactor Trip Recovery, an issue was identified resulting in the inability of the operator to recover the plant using reference plant procedures. Reference plant procedures do not provide guidance for proper plant alignment to perform a plant startup following an exit from the E-network. The test began with the plant at 100% full power and steady state. A manual reactor trip was directed and entry into E-0 commenced. From E-0, the operator [
]a,c When performing the Rx start-up using GOP-108, shutdown [
]a,c The Rapid
This issue has been corrected.
The Nuclear Application Program was designed [ ]a,c The
simulator modeling reflects the design. The issue here is that when the procedure was developed, it failed to include a step directing the operator to perform this action.
CR 898848 was written to identify the issue and GOP-108 was revised [ ]a,c A retest
of this evolution was performed and this issue was verified to be resolved by this procedural change.
ND-15-1333 Enclosure 9, Page 10 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2 (continued)
SCR# Summary Description Evaluation Basis
power reduction latched signal was found still locked in following the manual reactor trip blocking the withdrawal of [ ]a,c The reference plant procedure (GOP-108) does not direct resetting the rapid power reduction latched signal prior to rod withdrawal. The plant is unable to recover to 100% full power from a reactor trip using reference plant procedures.
5688
Graphic 50308 Issue:
Reactor Cooling System (RCS) – TASK RCS Heatup/Cooldown Primary Side, Ovation screen shows the outlet of RCS-V007C going to IRWST. APP-RCS-M6-002 Rev 13, shows this line going to RCDT.
This issue has been corrected.
The display graphic was corrected with the recent simulator software update. The change has been tested and the graphics found to be functioning correctly.
5702
IDS screens in simulator show inaccurate Power supplies:
IDS Screens (22701, 22702 and 22703) in Simulator show inaccurate Power Supplies for 24 Hr and 72 Hr Battery Chargers and Voltage regulating Transformers.
This issue has been corrected.
The display graphic was corrected with the recent simulator software update. The change has been tested and the graphics found to be functioning correctly.
5708
Plant Radiation Detector response to entering Mode 6:
While establishing Midloop and Mode 6, plant radiation detectors steadily rose throughout entire evolution. Containment radiation shows high levels with the Refueling Cavity at its normal refueling level.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
Determined that the required RCS cleanup and degasification were not completed prior to establishing the MODE 6 ICs. [
]a,c Therefore no effect on the indications available to or actions taken by an operator.
WEC is tracking RITS (41660) as the radiation systems are in preliminary development and expected to have setpoint and range changes during system testing/startup.
5905 User defined alarm limits not clearing on simulator reset:
Defined Alarm Limits (user) do not clear on RESET of
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This issue is of the same nature as that described in SCR 5588 and 5590. The
ND-15-1333 Enclosure 9, Page 11 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2 (continued)
SCR# Summary Description Evaluation Basis
an IC. They stay at the same value of the previous training session.
scenario guide used by the booth operator directs him/her to perform a check for any User Defined Alarm Limits that did not reset when resetting the simulator to a different initial condition.
A Training Needs Assessment was performed and it was determined that this issue does not impact any of the six criteria listed under ANSI/ANS-3.5-1998, Section 4.2.1.4 or any of the 13 criteria listed under 10 CFR 55.45(a).
This issue will remain open pending WEC correction.
5908
Shadow trails on Inverse Count Rate Ratio (ICRR) graphic 40052:
ICCR vs. Control Bank Withdrawal (Source Range) graphic 40052 is leaving "shadow" trails of previous calculations on WPIS monitors but not on any workstation screen. If the screen is refreshed the shadows disappear.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This affects the WPIS display ONLY. The operating station displays update properly. In addition, the “shadow” trail is similar to performing a manual ICRR plot with pencil and paper. The previous plot is not erased each time the plot is performed. Therefore, the “shadow” is an inconvenience, but all the proper information is still available to and will not lead to incorrect action by the operator.
5909
DRCS M bank rod control graphic:
Graphic 11181 – M and SD Bank Control, there is a “n” showing after an instance where the Rod Control rejects to Man (manual). Once the issue is cleared and Rods are returned to AUTO the “n” shows until the screen is refreshed.
The display graphic was corrected with the August 2015 update and tested to function correctly.
5911
MTS control valve position indication:
GOP-109, Rev. 1, checks the turbine control valves to be partially OPEN, the control valves are open (using signal diagrams) but indicated close on graphic 50211 of ovation.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This issue is observed during a turbine start-up where the control valves were previously closed, to perform the low speed rub checks, and the valves are now being re-opened to continue the turbine start-up. [
]a,c The issue has been tested at multiple different times in the evolution and proper plant response was observed in all instances.
ND-15-1333 Enclosure 9, Page 12 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2 (continued)
SCR# Summary Description Evaluation Basis
As part of the training process, the operators are trained to monitor the valve positions as the step is performed and monitor turbine RPM. Although valve position is difficult to use, the turbine RPM rising and stabilizing at the required RPM verifies the control valves are performing the required function and the step can be considered to be met.
This does not require an operator to remember a specific detail about the Simulator, but enforces the need to use multiple indications to verify proper plant response to operator actions.
This is current plant design in regards to the graphics indications. The normal process for plant design change recommendations is in progress to consider changing the accuracy provided on the valve position graphic (add additional decimal places to indication to see smaller changes in valve positions).
CR 10072985 was created to provide recommendations for procedure changes to more accurately portray all the indications the operator will need to use.
5915
Xenon graphic response:
OPDMS XENON CONTROL graphic is not responding. Using NRFE-07-100, Revision 4 information, all the data points that feed the "tail" of the DOT never change off of zero. All we have seen is a DOT which never moves or has a tail.
The display graphic was corrected with the Aug 2015 update and tested to function correctly.
5919
Investigate cause of MOV malfunction discrepancy on PMS division B RNS-V002A/B:
During ANSI testing for a plant shutdown when the procedure directed the operator to open RNS-V002A/B from Division B PMS PDSP, a MOV malfunction discrepancy alert was highlighted on the PMS screen.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
Per reference documentation V002A/B [
]a,c Therefore this is an expected indication for the operator and should not affect any operator actions while manipulating these components.
5958
Unexpected rise in indicated radiation on loss of process flow:
Ventilation System, including VHS, VRS and VFS radiation monitors are indicating high radiation after loss of process flow.
Closed. Fixed with patch WEC provided to SNC on August 14, 2015.
Two separate V&V tests were performed successfully.
ND-15-1333 Enclosure 9, Page 13 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2 (continued)
SCR# Summary Description Evaluation Basis
5969
Inconsistent response of alarm cutout operation:
During a scenario an alarm on the annunciator board came in briefly. The alarm was promptly placed in the cutout alarm list for APS and did not appear on the current alarms.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This SCR is the parent (tracking) of two other SCRs (5606 & 5621) which are associated with [
]a,c It was determined that the evolutions performed that lead to these anomalies are infrequent and while it is not expected to get an audible alarm [
]a,c
5973
RCPs indicate 5 RPM when secured:
While performing a shutdown and cooldown, after RCPs secured, RCS-SI263/264 indicated they are at 5 rpm.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This issue is consistent with current plant design.
The operating procedures [
]a,c
5984
JStation will freeze randomly:
Identical to DR 5891, this issue will track final resolution of the overall problem of AOI sims timing out.
This issue is related to the instructor station randomly freezing such that the booth instructor was unable to provide the required simulations. A patch was provided to SNC in October 2014 which corrected the issue.
SNC is maintaining this SCR in an open/pending status to ensure any future baseline updates do not re-create this issue.
5995
JStation remains open when Stop Trainer is performed in the ISS in some occasions.
Trainer does stop but JStation windows do not close.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This is a training instructor inconvenience during shutdown of the simulator during the post training evolution. This process is transparent to the operator and does not affect indications available to or actions taken by an operator.
5996 JStation RUN button enabled when DCIS not-ready.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This is a training instructor inconvenience during ‘RESET’ of the simulator. The operators are not actively being trained or examined during this process and therefore
ND-15-1333 Enclosure 9, Page 14 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2 (continued)
SCR# Summary Description Evaluation Basis
this issue is transparent to the operator.
5997
Turbine Runback Signal drop assignments:
The Load Correction Runback signal should originates in Drop 5 but the Simulator has it from Drop 6. The DeltaT Runback signal should originates in Drop 6 but the simulator has it from Drop 5.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
The issue is related to what Ovation cabinet controller generates the signal. The Simulator performs a turbine runback as required when a demand is generated. All expected responses occur and therefore the origin of the runback signal is transparent to the operator.
6058
Simulation of VES Airflow Noise is not currently modeled.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
The noise discussed being present or not present does not change the actions taken by an operator.
6081 CPS failure needs to respond to the simulator requirements for CPS as specified in design documentation.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
Although the available fault from the instructor station is not working as expected, the ability still remains for the instructor to prevent proper CPS operation. The method of the fault is transparent to the MCR operator and therefore does not affect the indication available to or actions taken by the operator.
6091 Kirk Key interlock not operable on spare battery LOAs
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
The Kirk Key indication in question is associated [
]a,c the booth operator can make the appropriate reports to the MCR and this issue is transparent to the MCR operators.
6092 EDSS-DF-1 nomenclature and switch operability issue on simulator diagram for EDS system.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This issue is similar to SCR 6091. The spare battery is still able to be placed in service when required. The nomenclature and operability issues are related to the simulator instructor station and are transparent to the operator.
ND-15-1333 Enclosure 9, Page 15 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2 (continued)
SCR# Summary Description Evaluation Basis
6101 Control Rod Drive Mechanism (CRDM) Fan Vibrations units in in/s instead of units in mils to be in alignment with plant reference procedures.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This issue was discovered while verifying surveillance procedures and determining which ones could be used for operational exams. The need to check these indications and the expected units is only provided in the System Operating Procedure (SOP) but not the Abnormal Operating Procedures (AOPs) and Emergency Operating Procedures (EOPs) this issue will be transparent to the MCR operators..
6128 Ovation 40023 graphic issue for the calibration offset AFD for quadrant A point name incorrect.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
The graphics display error is that the nomenclature dropped off the final number; DDS-AP-BDP-01959 is displayed with the final ‘9’ dropped off as DDS-AP-BDP-0195. The graphic in question is used during the performance of the calorimetric surveillance. The graphic problem does not impact the operation of the simulator model and only requires instructor intervention during the performance of the surveillance to identify the incorrect graphic nomenclature and has been captured in the training material.
6129
Display 40023 units issue for points associated with Nuclear Power Correction Factor and Axial Flux Difference Correction factor:
Request units of %power instead of %.
This issue was corrected subsequent to the Aggregate Study.
V&V testing was completed satisfactorily.
6153 Letdown control setpoints do not respond accurately in automatic control to changing plant parameters during a depressurization.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
These indications match current plant design.
The conditions at the time of the observation were that the plant was being depressurized using Aux Spray. [
]a,c This item is transparent to the operator.
6158 Letdown tuning:
CVS letdown heat exchanger alarm comes in when
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
SNC evaluated this item and determined that the simulator is modeling the AP1000
ND-15-1333 Enclosure 9, Page 16 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2 (continued)
SCR# Summary Description Evaluation Basis
CVS makeup secured and letdown is in progress. plant design. This issue represents an AP1000 design inadequacy.
Current plant design [
]a,c Since the actions operators are expected to take in response to this condition are identical to those they would otherwise take, the issue was determined to have no impact on the simulator’s suitability for the conduct of operating tests.
This issue is similar to SCR-5649.
6160 CCS Screen issue for CCS header flow in scientific notation.
This issue was corrected subsequent to the Aggregate Study.
V&V testing was completed satisfactorily.
6165 WPIS Tavg Scale not precise for use by operator. This issue was corrected subsequent to the Aggregate Study.
V&V testing was completed satisfactorily.
6166 Turbine Control Protection System (TCPS) rotor stress units are metric. Request change to standard units.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
The rotor stress inputs into the maximum RPM rate of change during a turbine startup based on the temperature variants across the rotor. [
]a,c Therefore, this does not change the available indications the operator will evaluate or affect the actions taken by an operator during turbine startup.
6167 ASS display units currently in psia. Request change to psig.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
The pressure indication in question resides on the graphic associated with the Aux Boiler and the header pressure just downstream of the Aux Boiler outlet. [
ND-15-1333 Enclosure 9, Page 17 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2 (continued)
SCR# Summary Description Evaluation Basis
]a,c The
display in question does indicate the correct pressure in ‘psia’. However, this graphic is not used by operators to verify header pressure and does not affect operator actions.
6170
WRS graphic shows the WHT room sump drain lines routed to Auxiliary Building sump. This is incorrect. The WHT room sump drain lines are routed to pump suction.
This issue was corrected subsequent to the Aggregate Study.
V&V testing was completed satisfactorily.
6173 WPIS nomenclature issue regarding Containment Floodup Level.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This issue stems from the procedure nomenclature differing from the graphic display nomenclature (i.e. procedure directs verification of CTMT FLOODUP LVL vice CTMT WR LEVEL). The two names are interchangeable. The graphic that was viewed was a wall panel display for Critical Safety Functions (CSFs). [
]a,c Therefore, these indications are the same and do
not affect actions taken by an operator.
6180 Time to Boil units on Mode 5/6 WPIS is in exponential units. Request change to standard units
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
Updates to the Time to Boil indications were included as part of the August 14, 2015, patch. After performing V&V testing, SNC personnel determined that the update was not successful. Time to Boil indication on the Mode 5/6 Primary trend WPIS was observed to be displayed in exponential minutes for the RCS time to boil. [
]a,c V&V testing was conducted under similar initial conditions as when the issue was first identified. The values indicated by each of these displays should be in hours and minutes. This V&V test failed.
The Time to Boil NAP is a tool that is used for information only. [
]a,c The value is displayed in exponential units versus hh:mm. This means that operators will need to convert the scientific notation values
into hours and minutes. Although this does take a short amount of time, the net effect is that it
ND-15-1333 Enclosure 9, Page 18 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2 (continued)
SCR# Summary Description Evaluation Basis
does not remove the ability to monitor the time to boil and it has no impact on actions the operator may, or may not, take in response to plant conditions.
6181 Tuning of EHC HX is requested to prevent outlet temperature high alarm being received during normal system operation.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This has been identified as a plant design issue.
The applicable HX has been identified as being too large. Therefore, the temperature control valve is not able to throttle cooling flow sufficiently to prevent the process variable flow to reach a temperature which drives the temperature control valve closed. When temperature reaches a point to demand the temperature valve to open, [
]a,c This cycle has been observed to occur over an approximate 2 hour time period. If the alarm were to be received the operator is expected to take the required actions.
6182 Primary trend screen rendering issue for RCS Tcold and RNS flow when called up from two locations at the same time
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
[
]a,c The only observed anomaly was a fluctuation of the graphic display when selecting “Print.” This fluctuation has no impact on actions the operator may, or may not, take in response to plant conditions.
6185 Default trend screen color for ovation trend screen is yellow and makes reading text difficult
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
The ovation trend program was opened and the settings associated with trend color were observed. The first two points that are selected to be trended default to red and light blue. If more points than this are required the operator is able to change the trend color using the properties menu. The operator is also able to shift to a tabular view vice a graphical view to see the information. Although, inconvenient, the capability is available to change the trend color if sufficient trends are added to a single trend window that results in one of them being yellow and therefore does not affect the indications available to or actions taken by an operator.
6187 Rod stop logic . When Shutdown Bank 1 fully SNC has determined that this issue does not impact the simulator’s suitability for the
ND-15-1333 Enclosure 9, Page 19 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2 (continued)
SCR# Summary Description Evaluation Basis
withdrawn, step counter indicates 265 steps. conduct of operating tests.
SNC evaluated this item and determined that the simulator is modeling the AP1000 plant design.
SNC was notified by WEC that the actual rod position will stop at 264 steps based on rod withdrawal limits. [
]a,c
WEC is updating [
]a,c The need for training on the revised procedure will be evaluated through
the Training Needs Analysis (TNA) process.
6188 Plant Mode NAP Temperature Input for the Automatic plant mode calculation uses average cold leg temperature instead of RCS average temperature.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
The plant conditions which have [
]a,c and the input that caused the MODE change is transparent to the operator.
6197 OCS Wall Panel Navigation System (WPNS) and Reactor Operator Peer Check System (ROPCS) Rebuild Required
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
The OCS WPNS and ROPCS have deficiencies in the software resulting in a loss of communication between the various WPIS displays and operator stations. The frequency with which this occurs is low. When it does occur, operators respond as they would to any problem with DDS (perform actions of the Abnormal Operating Procedure) and therefore are taking the actions they would be expected to take if the same event were to occur in the plant.
6207 Can't open signal diagrams from APS
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
The inability to open signal diagrams specifically from APS does not hinder an operator’s ability to view the desired diagram and has no impact on the actions an operator.
6241 SBT/Replay tool issues SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
ND-15-1333 Enclosure 9, Page 20 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2 (continued)
SCR# Summary Description Evaluation Basis
The replay tool is a capability of the simulator to be reset to a point in the past in order to re-run a scenario. This is a training tool.
6246
SMS Detector ranges not consistent with design statements:
Maximum indication on Ovation screens is [ ]a,c. Design documents list peak values of [
]a,c
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
Current training material directs the booth operator to insert user defined alarms to generate the alarms to allow operators to respond accordingly prior to the initiation of the training event, thus making this issue transparent to the operators.
6278 Battery bank indications are mislabeled for EDS1, EDS2, and EDS4
This issue was corrected subsequent to the Aggregate Study.
V&V completed satisfactorily.
6306 EHS Power Supplies not modeled
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
EHS is the system associated with heat trace for outside components that require cold weather protection. The impact of this issue would only be associated with scenarios specifically designed for cold weather mitigation. There are no training scenarios that require the use of EHS at this time. Therefore has no effect on the indications provided to or actions taken by an operator.
Although some portions of EHS were modeled, modeling of this portion was considered to be beyond the scope necessary to be simulated.
The SCR will continue to be tracked for possible model changes in the future to allow scenarios that result in temperature swings in plant areas.
6410 VFS/VAS performance during outside containment loss of cooling accident scenarios.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This issue was determined to be a plant design issue. The simulator is functioning consistent with current plant design.
WEC has corrected, but not implemented, a plant design change to address this issue.
SNC is keeping this SCR open for tracking purposes until the change has been implemented.
6418 Same points not available for similar equipment in the radiation monitoring system.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
The identified points are associated with power availability to the radiation detectors. These points were identified as being additional information that would be beneficial for
ND-15-1333 Enclosure 9, Page 21 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2 (continued)
SCR# Summary Description Evaluation Basis
incorporation into CPS logics for entry conditions into the associated Abnormal Operating Procedure. However, these specific indications are not required by any procedure to be viewed and a power loss would also be indicated by the detector values indicating ‘Bad Quality’. Therefore this does not impact the indications available to or actions taken by an operator.
6447 PCS Indications - Inconsistent Naming for containment pressure
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
The issue is associated with the nomenclature associated with CTMT pressure on the Critical Safety Function (CSF) WPIS display. On this display the CTMT pressures are labeled as ‘ExtR’ and ‘NormR’, meaning ‘Extended Range’ and ‘Normal Range’. Although the naming of the points on the graphics is not consistent, the information being provided by the graphics is readily identifiable. On the PCS graphic, which contains the instruments that provide the input to the CSF display, the instruments are labeled as ‘Wide Range’ and ‘Narrow Range’. For this reason it does not affect the indications provided to or actions taken by an operator.
6470
APS allows priority 1 alarm suppression via consequence logic:
This was discovered during Alarm Avalanche evaluation.
Closed. Invalid.
This issue has been determined to be invalid based upon updated information received from WEC. The AP1000 plant design permits the [
]a,c
6471 RCS flow during loss of all RCP Verification – SCR is to investigate a [ ]a,c level rise during the loss of all RCPs
SNC has determined that this issue is invalid. The issue is closed.
SNC investigated this issue and determined that this is in accordance with the AP1000
design and that this is the expected plant response.
SCR resolution [ ]a,c This issue was closed because this is
the expected simulator response.
6482 Steam generator parameters oscillating during a main steamline break outside containment.
Closed, could not duplicate.
This scenario was re-performed twice by SNC simulator staff under the conditions under which the issue was first identified and was unable to elicit the same response on its simulator. Additionally, the domestic AP1000 simulator that identified this issue was only able to reproduce the issue on one of its two simulators. Because SNC was unable to reproduce the issue and because the discovering organization was unable to reproduce the issue on their second simulator, the issue was closed as invalid.
ND-15-1333 Enclosure 9, Page 22 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2 (continued)
SCR# Summary Description Evaluation Basis
6498
Demineralized Water Feed Pump A and B (DWS-MP-01A/B) do not have a poke for operator use nor are they modeled in instructor station for the instructor to locally control.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
Design change APP-GW-GEE-2907 replaced the motors for the DWS pumps AND moved the control of the pumps to the Local Control Station. Therefore, operation of these components from the MCR is no longer in the plant design and is consistent with the controls currently available to the MCR operators.
6626
TCS-V025, Main Turbine Lube Oil Coolers Flow Control Valve, should switch setpoint values (SP) in AUTO after transition from normal operations to turbine startup
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This issue is related to a design document that infers an automatic change to the setpoint based [ ]a,c However, only one design document infers this with no discussion of this feature anywhere else in the document. Also, the instruments stated as providing an input to this automatic feature do not demonstrate having this output to the controller. This may be a design change that has not fully been implemented in all documentation. WEC has been tasked with determining the actual plant response required. Procedures call for operators to “Ensure” the controller setpoint. When this step is performed, the operator manually adjusts the setpoint to the correct value. The system then maintains the correct temperature in AUTO.
Finally, the system training material does not contain any reference to this automatic setpoint change and therefore is currently not an automatic function the operators would identify as not operating correctly.
6635 All Buildings Drain System (WWS) not modeled correctly.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This modeling deficiency was identified and is an ongoing simulator group project to ensure proper building drain indications. Significant items, ones that involve current training material (and the stem of this SCR being created), were prioritized and were corrected first. Therefore, the current training material will not be affected and there will be no effect on the indications available to, or the actions taken by, the operator.
6646 ECS 120V Bus voltage not properly modeled. Instead of 120V, they are modeled to 208V.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
The 120V bus indications are not available to the student on any graphic. Therefore this issue does not have an effect on indications available to or actions taken by an operator. When a component powered by these busses is required it will energize and
ND-15-1333 Enclosure 9, Page 23 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
Table E9-2 (continued)
SCR# Summary Description Evaluation Basis
function properly.
6668
“Show all points” tiles not printing to CSV (Comma-Separated Value). The points can be displayed on the APS screen, but will not print to a CSV file. The points can be printed to paper.
SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.
This feature is used during simulator testing and data collection. In the normal case of training or exam scenarios the operators would not use or be required to use this function of APS. This inability to create these files in a .csv file format does not impact the indications available to, or actions taken by, an operator in the MCR.
Southern Nuclear Operating Company
Vogtle Electric Generating Plant (VEGP) Units 3 and 4
ND-15-1333
Enclosure 10
BEACON
(This Enclosure consists of 2 pages, including this cover page)
ND-15-1333 Enclosure 10, Page 2 of 2 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 1.0 Summary Description of BEACON
Definition
BEACON (Best Estimate Analyzer for Core Operations – Nuclear) is a Westinghouse reactor core monitoring application. BEACON converts plant instrument readings and performs calculations in order to provide measurement and analysis of core performance.
Description
The AP1000 seamlessly integrates BEACON into everyday plant operations; all calculations and programming are transparent to the Control Room operators. BEACON directly interfaces with the plant data network. This allows operators to monitor core performance using the same workstations and software used for daily operations. This data (originating from BEACON) is literally one click from their top-level display. Unlike the AP1000, BEACON was a post-design modification for legacy Westinghouse plants. Those plants were originally designed with other means to monitor in-core reactor performance. Starting in approximately 1990, legacy plants began slowly incorporating BEACON into their systems in varying degrees, requiring additional hardware and software in the plant and simulator.
2.0 BEACON Simulation
Core power distribution data in the AP1000 simulator is generated by complex algorithms and code to accurately represent reactor behavior. Because the AP1000 core monitoring interface is fully integrated with the normal control system (and simulator), there is no need for any additional hardware or software systems in the simulator. The simulator has the ability to model a BEACON failure. With this level of functionality, the core parameters modeled and displayed in the simulator are indistinguishable from the actual plant.
Southern Nuclear Operating Company
Vogtle Electric Generating Plant (VEGP) Units 3 and 4
ND-15-1333
Enclosure 11
Acronyms & Definitions
(This Enclosure consists of 6 pages, including this cover page)
ND-15-1333 Enclosure 11, Page 2 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 1.0 Acronyms
ADS Slow Primary System Depressurization to Saturated Condition
ANS American National Standards
AO Axial Offset
APP Application
APS Alarm Presentation System
ATWS Anticipated Transient Without SCRAM
BEACON Best Estimate Analyzer for Core Operations – Nuclear
BL7 Baseline 7
BL8 Baseline 8
CAP Corrective Action Program
CAPAL Corrective Action, Prevention and Learning (WEC electronic document
control)
CAS Commission Approved Simulator
CB&I Chicago Bridge and Iron
CET Core Exit Thermocouple
CLFs Component Level Failures
CFR Code of Federal Regulations
CMS Configuration Management System
CMT Core Makeup Tank
CR Condition Report
CSF Critical Safety Function
CSV Comma-Separated Value
CTMT Containment
DCP Design Change Package
DG Diesel Generator
ND-15-1333 Enclosure 11, Page 3 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
DR Discrepancy Report
DRCS Digital Rod Control System
EHC Electro-Hydraulic Control
FW Feedwater
FRP Functional Restoration Procedure
GOP General Operating Procedure
GSE GSE Systems. This is the name of the simulator vendor contracted by
Westinghouse
HEDs Human Engineering Discrepancies
HFE Human Factors Engineering
HSI Human-System Interface
HX Heat Exchanger
I&C Instrumentation and Controls
IC Initial Condition
ICRR Inverse Count Rate Ratio
IIS Incore Instrumentation System
ISV Integrated Systems Validation
ITAAC Inspection, Test, Analysis and Acceptance Criteria
IR Intermediate Range Power
IRWST In-Containment Refueling Water Storage Tank
IVR In-Vessel Retention
LOAs Local Operator Actions
LOCA Loss of Coolant Accident
MS Main Steam
MSL Main Steam Line
MSR Moisture Separator Reheater
ND-15-1333 Enclosure 11, Page 4 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
NAPS Nuclear Application Programs
NRC Nuclear Regulatory Commission
OPDMS Online Power Distribution Monitoring System
P1 Priority 1
P2 Priority 2
PBX Private Branch Exchange
PDSP Primary Dedicated Safety Panel
PHED Potential Human Engineering Discrepancy
PORVs Power-Operated Relief Valves
PRA Probabilistic Risk Assessment
PRS Plant Referenced Simulator
PZR Pressurizer
RAIs Requests for Additional Information
RCPs Reactor Cooling Pumps
REN-MAN03 Reference to a specific RIHA
RIHA Risk Informed Human Actions
RITS RRAS Issue Tracking System (WEC CAP)
ROPCS Reactor Operator Peer Check System
Rx Reactor
SAT Systematic Approach to Training
SBT Simulator Scenario-Based Testing
SCANA SCANA is not an acronym, but is taken from the letters in South Carolina
SCR Simulator Change Request
SFCVs Start-up Feedwater Control Valves
SFW Startup Feedwater
SG Steam Generator
ND-15-1333 Enclosure 11, Page 5 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
SGWL Steam Generator Water Level
SNC Southern Nuclear Operating Company
SOP System Operating Procedure
STGR Steam Generator
STS Simulator Training System
TB Turbine Building
TNA Training Needs Analysis
UIs Unresolved Items
UFSAR Updated Final Safety Analysis
VCS Virgil C. Summer Nuclear Station
V&V Verification and Validation
VEGP Vogtle Electric Generating Plant
VHS Health Physics and Hot Machine Shop HVAC System
WEC Westinghouse Electric Company
WPIS Wall Panel Information System
WPNS Wall Panel Navigation System
WR Wide Range
2.0 Definitions
Mantis An SNC database program used for tracking, software changes,
hardware changes, administrative issues, modeling, I&C and Design
Change Packages (DCPs) that affect the simulator
Priority 1 Westinghouse Alarm Criteria (based on App-DDS-J4-010, Appendix B,
Rev 2) - Less than five minutes to respond, consequence can be a plant
trip or ESF actuation. This also includes radiation release or protection of
personnel.
Priority 2 Westinghouse Alarm Criteria (based on App-DDS-J4-010, Appendix B,
Rev 2) - five to 20 minutes to respond prior to degradation to a P1
condition. This may also include alarms that are important to operability
ND-15-1333 Enclosure 11, Page 6 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility
requirements with time-sensitive actions. Examples include bistable trips
that result in a P1 condition.
Simulator Simulator Training System
Training Needs Assessment
An appraisal by a subject matter expert of a simulator deviation,
deficiency, or modification, and its relative importance to the operator as
required tasks are performed.
Southern Nuclear Operating Company
Vogtle Electric Generating Plant (VEGP) Units 3 and 4
ND-15-1333
Enclosure 12
Westinghouse Authorization Letter CAW-15-4260, Application for Withholding Proprietary Information From Public Disclosure, Accompanying Affidavit, Proprietary
Information Notice and Copyright Notice
(This Enclosure consists of 11 pages, including this cover page)
@ Westinghouse
Document Control Desk U S Nuclear Regulatory Commission Washington, DC 20852-2738
Westinghouse Electric Company New Plants and Major Projects 1000 Westinghouse Drive, Building 1 Cranberry Township, Pennsylvania 16066 USA
Direct tel: (412) 374-3382 Direct fax: (724) 940-8519
e-mail: russpa@westinghouse.com Proj letter: SVP _SV0_003472
CAW-15-4260 9/17/2015
APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE
Subject: Transmittal of Evaluation of APlOOO Simulation Facility Summary of Umesolved Items (Uis) Issued By the NRC, Commission Approved Simulator Aggregate Study - Simulator Training System Deficiency Impact on 10 CRF 55.45(a) Compliance, Evaluation of Priority One (1) Potential Human Engineering Discrepancies (PHEDs) from Integrated Systems Validation (ISV) Daily Assessments, and List of Open Simulator Discrepancies
The proprietary information for which withholding is being requested in the above-referenced reports is further identified in Affidavit CAW -15-4260 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The Mfidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)( 4) of 10 CFR Section 2.390 of the Commission's regulations.
Accordingly, this letter authorizes the utilization of the accompanying Mfidavit by Southern Nuclear Company (SNC).
Correspondence with respect to the proprietary aspects of the Application for Withholding or the Westinghouse Affidavit should reference CAW-15-4260, and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066.
Very truly yours,
(]~a_~ Paul A. Russ, Director
U.S. Licensing & Regulatory Support
cc: Richard Paese Sarah DiTommaso Gerry Couture Steven Radomski Mark Chitty Mark Crosby David Midlik Wes Sparkman
Westinghouse Westinghouse Westinghouse Westinghouse SNC SNC SNC SNC
CAW-15-4260 September 17, 2015
Page 2 of2
COMMONWEALTH OF PENNSYL V ANlA:
COUNTY OF BUTLER:
AFFIDAVIT
ss
CAW-15-4260
September 17, 2015
I, Paul A. Russ, am authorized to execute tbis Affidavit on behalf of Westinghouse Electric
Company LLC (Westinghouse), and tbat tbe averments of fact set forth in tbis Affidavit are true and
correct to tbe best of my knowledge, information, and belief.
Paul A. Russ, Director
U.S. Licensing & Regulatory Support
2 CAW-15-4260
September 1 7, 2015
(1) I am Director, U.S. Licensing &Regulatory Support, Westinghouse Electric Company LLC
(Westinghouse), and as such, I have been specifically delegated the function of reviewing the
proprietary information sought to be withheld from public disclosure in connection with nuclear
power plant licensing and rule making proceedings, and am authorized to apply for its
withholding on behalf of Westinghouse.
(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the
Commission's regulations and in conjunction with the Westinghouse Application for Withholding
Proprietary Information from Public Disclosure accompanying this Affidavit.
(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating
information as a trade secret, privileged or as confidential commercial or financial information.
( 4) Pursuant to the provisions of paragraph (b)( 4) of Section 2.390 of the Commission's regulations,
the following is furnished for consideration by the Commission in determining whether the
information sought to be withheld from public disclosure should be withheld.
(i) The information sought to be withheld from public disclosure is owned and has been held
in confidence by Westinghouse.
(ii) The information is of a type customarily held in confidence by Westinghouse and not
customarily disclosed to the public. Westinghouse has a rational basis for determining
the types of information customarily held in confidence by it and, in that connection,
utilizes a system to determine when and whether to hold certain types of information in
confidence. The application of that system and the substance of that system constitute
Westinghouse policy and provide the rational basis required.
Under that system, information is held in confidence if it falls in one or more of several
types, the release of which might result in the loss of an existing or potential competitive
advantage, as follows:
(a) The information reveals the distinguishing aspects of a process (or component,
structure, tool, method, etc.) where prevention of its use by any of
3 CAW-15-4260
September 17, 2015
Westinghouse's competitors without license from Westinghouse constitutes a
competitive economic advantage over other companies.
(b) It consists of supporting data, including test data, relative to a process (or
component, structure, tool, method, etc.), the application of which data secures a
competitive economic advantage, e.g., by optimization or improved
marketability.
(c) Its use by a competitor would reduce his expenditure of resources or improve his
competitive position in the design, manufacture, shipment, installation, assurance
of quality, or licensing a similar product.
(d) It reveals cost or price information, production capacities, budget levels, or
commercial strategies of Westinghouse, its customers or suppliers.
(e) It reveals aspects of past, present, or future Westinghouse or customer funded
development plans and programs of potential commercial value to Westinghouse.
(f) It contains patentable ideas, for which patent protection may be desirable.
(iii) There are sound policy reasons behind the Westinghouse system which include the
following:
(a) The use of such information by Westinghouse gives Westinghouse a competitive
advantage over its competitors. It is, therefore, withheld from disclosure to
protect the Westinghouse competitive position.
(b) It is information that is marketable in many ways. The extent to which such
information is available to competitors diminishes the Westinghouse ability to
sell products and services involving the use of the information.
(c) Use by our competitor would put Westinghouse at a competitive disadvantage by
reducing his expenditure of resources at our expense.
4 CAW-15-4260
September 17, 2015
(d) Each component of proprietary information pertinent to a particular competitive
advantage is potentially as valuable as the total competitive advantage. 1f
competitors acquire components of proprietary information, any one component
may be the key to the entire puzzle, thereby depriving Westinghouse of a
competitive advantage.
(e) Unrestricted disclosure would jeopardize the position of prominence of
Westinghouse in the world market, and thereby give a market advantage to the
competition of those countries.
(f) The Westinghouse capacity to invest corporate assets in research and
development depends upon the success in obtaining and maintaining a
competitive advantage.
(iv) The information is being transmitted to the Commission in confidence and, under the
provisions of 10 CFR Section 2.390, it is to be received in confidence by the
Commission.
(v) The information sought to be protected is not available in public sources or available
information has not been previously employed in the same original manner or method to
the best of our knowledge and belief.
(vi) The proprietary information sought to be withheld in this submittal is that which is
appropriately marked in ND-15-1333, Enclosure 5P "Evaluation of AP1000 Simulation
Facility Summary of Unresolved Items (U1s) Issued By the NRC" (Proprietary),
"Commission Approved Simulator Aggregate Study - Simulator Training System
Deficiency Impact on 10 CRF 55.45" (Proprietary), ND-15-1333, Enclosure 8P
"Evaluation of Priority One (1) Potential Human Engineering Discrepancies (PHEDs)
from Integrated Systems Validation (ISV) Daily Assessments" (Proprietary), and
ND-15-1333, Enclosure 9P "List of Open Simulator Discrepancies" (Proprietary), for
submittal to the Commission, being transmitted by Southern Nuclear Company (SNC)
letter and Application for Withholding Proprietary Information from Public Disclosure,
5 CAW-15-4260
September 17, 2015
to the Document Control Desk. The proprietary infonnation is submitted to support the
review of the Southern Nuclear Company Vogtle commission approved simulator.
(a) This information is part of that which will enable Westinghouse to:
(i) Manufacture and deliver products to utilities based on proprietary
designs
(b) Further this infonnation has substantial commercial value as follows:
(i) Westinghouse plans to sell the use of similar information to its customers
for the purpose of licensing new nuclear power stations.
(ii) Westinghouse can sell support and defense of industry guidelines and
acceptance criteria for plant-specific applications.
(iii) The information requested to be withheld reveals the distinguishing
aspects of a methodology which was developed by Westinghouse.
Public disclosure of this proprietary information is likely to cause substantial harm to the
competitive position of Westinghouse because it would enhance the ability of
competitors to provide similar technical evaluation justifications and licensing defense
services for commercial power reactors without commensurate expenses. Also, public
disclosure of the information would enable others to use the information to meet NRC
requirements for licensing documentation without purchasing the right to use the
information.
The development of the technology described in part by the information is the result of
applying the results of many years of experience in an intensive Westinghouse effort and
the expenditure of a considerable sum of money.
6 CAW-15-4260
September 17, 2015
In order for competitors of Westinghouse to duplicate this information, similar teclmical
programs would have to be performed and a significant manpower effort, having the
requisite talent and experience, would have to be expended.
Further the deponent sayeth not.
PROPRIETARY INFORMATION NOTICE
CAW-15-4260 September 17, 2015
Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant -specific review and approval.
In orderto conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (t) located as a subscript innnediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(t) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(l).
COPYRIGHT NOTICE
The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.
Southern Nuclear Company (SNC) Letter for Transmittal to the NRC
CAW-15-4260 September 17, 2015
The following paragraphs should be included in your letter to the NRC Document Control Desk:
Enclosed are:
l. One copy of ND-15-1333, Enclosure 5P "Evaluation of APlOOO Simulation Facility Summary of Unresolved Items (Uis) Issued By the NRC" (Proprietary)
2. One copy of ND-15-1333, Enclosure 5NP "Evaluation of APlOOO Simulation Facility Summary of Unresolved Items (Uis) Issued By the NRC" (Non-Proprietary)
3. One copy of "Commission Approved Simulator Aggregate Study -Simulator Training System Deficiency Impact on 10 CRF 55.45(a) Compliance" (Proprietary)
4. One copy of "Commission Approved Simulator Aggregate Study - Simulator Training System Deficiency Impact on 10 CRF 55.45(a) Compliance" (Non-Proprietary)
5. One copy of ND-15-1333, Enclosure 8P "Evaluation of Priority One (1) Potential Human Engineering Discrepancies (PHEDs) from Integrated Systems Validation (ISV) Daily Assessments" (Proprietary)
6. One copy ofND-15-1333, Enclosure 8NP "Evaluation of Priority One (1) Potential Human Engineering Discrepancies (PHEDs) from Integrated Systems Validation (ISV) Daily Assessments" (Non-Proprietary)
7. One copy of ND-15-1333, Enclosure 9P "List of Open Simulator Discrepancies" (Proprietary)
8. One copy ofND-15-1333, Enclosure 9NP "List of Open Simulator Discrepancies" (Non-Proprietary)
Also enclosed is the Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-15-4260, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice.
As Item 1 contains information proprietary to Westinghouse Electric Company LLC, it is supported by an Affidavit signed by Westinghouse, the owner of the information. The Affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)( 4) of Section 2.390 of the Commission's regulations.
Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the Commission's regulations.
Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse Affidavit should reference CAW-15-4260 and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066.
top related