mark dominion® dominion - nrc.gov · dominion resources, inc. 5000 dominion boulevard, glen allen,...

Mark D. Mitchell Dominion®Vice President - Generation Construction DominionDominion Generation

An operating segment ofDominion Resources, Inc.5000 Dominion Boulevard, Glen Allen, VA 23060

dora.corn

January 23, 2015

U. S. Nuclear Regulatory Commission Serial No. NA3-15-001Attention: Document Control Desk Docket No. 52-017Washington, D. C. 20555 COL/BCB

DOMINION VIRGINIA POWERNORTH ANNA UNIT 3 COMBINED LICENSE APPLICATIONCOLA MARKUPS TO ALIGN WITH FERMI 3 OCTOBER 2014 COLA SUBMISSIONS

On October 15 and 31, 2014, DTE Energy submitted Fermi 3 Combined LicenseApplication (COLA) updates to the NRC (ML14295A354 and ML14308A337). Followingthe design center approach, Dominion reviewed the Fermi 3 COLA updates andidentified those changes that are applicable to the North Anna Unit 3 (NA3) COLA. As aresult of the Fermi 3 COLA revisions, Dominion is proposing similar changes to the NA3COLA as summarized in Enclosure 1. Corresponding NA3 COLA markups are providedin Enclosure 2.

In addition to the changes resulting from the Fermi 3 COLA revisions, Dominion isproposing minor formatting changes on certain pages of COLA Part 4, TechnicalSpecifications. These changes were previously discussed with the NRC projectmanager and are included in Enclosures 1 and 2.

The proposed changes described in the enclosures will be incorporated into a futuresubmission of the NA3 COLA.

Please contact Regina Borsh at (804) 273-2247 ([email protected]) if you havequestions.

Very truly yours,

Mark D. Mitchell

Serial No. NA3-15-001COLA Markups to Align with EF3

of 2

Enclosures:

1. Summary of Proposed North Anna Unit 3 COLA Changes

2. Markups for Proposed North Anna Unit 3 COLA Changes

Commitments made by this letter:

1. The proposed changes described in the enclosures will be incorporated into afuture submission of the NA3 COLA.

COMMONWEALTH OF VIRGINIA

COUNTY OF HENRICO

The foregoing document was acknowledged before me, in and for the County andCommonwealth aforesaid, today by Mark D. Mitchell, who is Vice President-GenerationConstruction of Virginia Electric and Power Company (Dominion Virginia Power). Hehas affirmed before me that he is duly authorized to execute and file the foregoingdocument on behalf of the Company, and that the statements in the document are trueto the best of his knowledge and belief.

Acknowledged before me this2. day of-TA A).ug , U-tls-

My registration number is 25 3 1 '3 . and my

Commission expires: 5EPTa•r•-j 30) 2.i b

P I 4I'S

cc: U. S. Nuclear Regulatory Commission, Region II •OMN .P. H. Buckberg, NRC m" s:

T. S. Dozier, NRCG. J. Kolcum, NRC '".,,D. Paylor, VDEQ ' ,,W. T. Lough, SCCP. W. Smith, DTEM. K. Brandon, DTER. J. Bell, NEI

Serial No. NA3-15-001Docket No. 52-017

ENCLOSURE 1

Summary of Proposed

North Anna Unit 3 COLA Changes


Enclosure 1

COLA Section, Table Proposed Change Reason forPart or Figure Change

Part 2 Section 1.1.1.1 Added statement that incorporation of Consistency withthe ESBWR certified design is discussed Fermi 3 COLAin Section 1.1.1.7

Part 2 Section 1.1.1.7 Added statement to incorporate 10 CFR Consistency withPart 52, Appendix E by reference Fermi 3 COLA

Part 2 Section 1.1.2.2 Added statement that incorporation of Consistency withthe ESBWR certified design is discussed Fermi 3 COLAin Section 1.1.1.7

Part 2 Section 1.5.1.1 Deleted "for Unit 3" Consistency withFermi 3 COLA

Part 2 Section 1.5.1.1.1 Corrected description of the ESBWR Consistency withstrategy for coping with extended loss of Fermi 3 COLAAC power events as a two-phaseapproach

Part 2 Table 1.9-203 Revised conformance evaluation for RG Consistency with1.206 Section C.I11.1 3.9.5.4 to state, Fermi 3 COLA"Classification of the reactor internals isdescribed in Section 3.9.2.4."

Part 2 Table 1.11-202 Revised LMA to "NAPS SUP 1.11-1" Consistency withFermi 3 COLA

Part 2 Table 13.4-201 Revised the description of the Consistency withEmergency Planning milestone Fermi 3 COLA

Part 2 Section Revised LMA to "NAPS SUP 13.5-37" SUP 13.5-37 is site-13.5.2.2.8 specific.

Part 4 B 2.1.2 Added a line break after the Safety Consistency withLimits section TSTF-GG-05-01

Part 4 B 3.4.4 Corrected indentation for Reference 6 Consistency withTSTF-GG-05-01

Part 4 Table B 3.7.21 Added a space after the comma Consistency withbetween room numbers 3277 and 3274 TSTF-GG-05-01

Page 2 of 2


ENCLOSURE 2

Markups for Proposed

North Anna Unit 3 COLA Changes

Serial No. NA3-15-001Docket No. 52-017Page 1 of 15

North Anna 3Combined License Application

Markup of North Anna COLA

The attached markup represents Dominion's good faith effort to show how the COLA will be revisedin a future COLA submittal. However, the same COLA content may be impacted by responses to

COLA RAIs, other COLA changes, plant design changes, editorial or typographical corrections, etc.

As a result, the final COLA content that appears in a future submittal may be somewhat different

than as presented herein.



Part 2: Final Safety Analysis Report

FINAL SAFETY ANALYSIS REPORT

Chapter 1 Introduction and General Description of Plant

1.1 Introduction

This section of the ESBWR Design Control Document (DCD), i.e., thereferenced DCD, is incorporated by reference with the followingdepartures and/or supplements.

1.1.1 Format and Content

NAPS SUP 1.1-1 1.1.1.1 10 CFR 52 and Regulatory Guide 1.206

This FSAR was developed to comply with the content requirements of

10 CFR 52.79, and to the extent feasible, the content and format

requirements contained in Regulatory Guide (RG) 1.206, "Combined

License Applications for Nuclear Power Plants (LWR Edition)." See

Table 1.9-203, Conformance With the FSAR Content Guidance InRG 1.206. If the information requested by RG 1.206 is not needed (e.g.,

because it is already provided in the DCD or is located elsewhere in the

FSAR), the table specifies the location of the information.

Section C.111.6 of RG 1.206 addresses referencing a design certification

(DC) application rather than a certified design. The existing DC rules(10 CFR 52 appendices) require that a Combined Operating License

Application (COLA) that references a certified design include a

plant-specific DCD containing the same type of information and using the

same organization and numbering as the generic DCD for the ESBWR

design, as modified and supplemented by the applicant's exemptions and

departures. Where necessary to present additional information, new

sections were added following the logical structure of the ESBWR

generic DCD.

Incorporation of the ESBWR certified design is discussed in

Section 1.1.1.7.

1.1.1.2 Standard Review Plan

As required by 10 CFR 52.79(a)(41), an evaluation of the facility for

conformance with the acceptance criteria contained in NUREG-0800,

"Standard Review Plan for the Review of Safety Analysis Reports for

Nuclear Power Plants LWR Edition," in effect six months prior to submittal

of the COLA was performed. This evaluation determined that this FSAR

I

1-1 Revision 9 (Draft (01/1 3/1 5)1-1 Revision 9 (Draft (01/13/15)

Serial No. NA3-15-001 North Anna 3Docket No. 52-017Page 3 of 15 Combined License Application


1.1.1.5 Proprietary and Security-Related Sensitive UnclassifiedNon-Safeguards Information (SUNSI)

Proprietary information and SUNSI 1 is withheld from public disclosure

and therefore not included in the public version of the FSAR. SUNSI

included in the non-public version of the FSAR is appropriately indicated.

1.1.1.6 Acronyms

In addition to the summary list of acronyms in the FSAR frontmatter,

acronyms are defined at their first occurrence in FSAR text.

1.1.1.7 Incorporation by Reference

10 CFR 52.79 states in part that, "The final safety analysis report need

not contain information or analyses submitted to the Commission in

connection with the design certification, provided, however, that the final

safety analysis report must either include or incorporate by reference the

standard design certification final safety analysis report and must contain,

in addition to the information and analyses otherwise required,

information sufficient to demonstrate that the site characteristics fall

within the site parameters specified in the design certification." Therefore,

because this COLA references the ESBWR DC application, this FSAR

incorporates the ESBWR DCD by reference, with the departures

presented in COLA Part 7, and with supplemental information, as

appropriate (see Section 1.1.1.10). References in this FSAR to the DCD

should be understood to mean the ESBWR DCD, Tier 2, submitted by

GE-Hitachi Nuclear Energy Americas LLC (GEH), as Revision 10.

The Design Certification Rule for the ESBWR Design, 10 CFR Part 52,

Appendix E, references Revision 10 of the ESBWR DCD (79 FR 61944).

1. Any information which, if lost, misused, modified, or accessed withoutauthorization, can reasonably be foreseen as causing harm to the publicinterest, the commercial or financial interest of the entity or individual towhom the information pertains, the conduct of NRC and Federal programs,or the personal privacy of individuals. SUNSI has been organized into thefollowing seven groups:" Allegation information" Investigation information" Security-related information" Proprietary information" Privacy Act information" Federal, State, Foreign Government, and international agency

information" Sensitive internal information

1-3 Revision 9 (Draft (01/13/15)




Appendix E of 10 CFR Part 52 is hereby incorporated by reference into

the COL application.

1.1.1.8 Departures from the Standard Design Certification (orApplication)

A departure is a plant-specific "deviation" from design information in a

standard DC rule or, consistent with Section C.111.6 of RG 1.206, from

design information in a DC application.

10 CFR 52 clarifies that Tier 2 information in a standard DC rule does not

include conceptual design information (CDI) and per Section C.I1I.6 of

RG 1.206, Tier 2 information in a standard DC application does not

include CDI. Therefore, replacement or revision of CDI does not

constitute a departure. Additionally, information addressing combined

license (COL) information/holder items and supplemental information

(see Section 1.1.1.10) that does not change the intent or meaning of the

ESBWR DCD text is not considered a departure from the ESBWR DCD.

I

NAPS SUP 1.1-2 1.1.1.9 Referencing of ESPA Information

As with the DCD, the FSAR incorporates by reference the North Anna

ESPA SSAR, Revision 9, with certain variances and/or supplements (see

Section 1.1.1.10). A variance is a plant-specific deviation from one or

more of the site characteristics, design parameters, or terms andconditions of an ESP or from the SSAR. A variance to an ESP is

analogous to a departure from a standard DC.

SSAR Chapter 1 is incorporated by reference for historical purposes as

an appendix to this chapter.

1.1.1.10 Supplements

Supplements fall into one of the following categories (see Table 1.1-201

for definitions of categories unless noted otherwise):

" COL Item

" Conceptual Design Information

" ESP COL Action Item

" ESP Permit Condition

" ESPA SSAR Correction

" Supplemental Information (see definition below)





sewage treatment plant, water treatment facilities, storage tanks for waterand fuel oil, a switchyard and other site support systems and structuresnecessary to support the operation and maintenance of the facility.

1.1.2.2 Type of License Request

Add the following at the end of this section.

NAPS SUP 1.1-3 Virginia Electric and Power Company (Dominion) is the applicant for a

combined construction permit and operating license (COL) under

Section 103 of the Atomic Energy Act, for the third nuclear power plant to

be located on the existing North Anna Power Station (NAPS) site inLouisa County, Virginia. This COLA references a DC application for an

ESBWR (consistent with Section C.I11.6 of RG 1.206) and the Early Site

Permit (ESP) for the NAPS site. The third unit is designated North Anna

Unit 3 (Unit 3).

Incorporation of the ESBWR certified design is discussed in

Section 1.1.1.7.

1.1.2.4 Description of Location


NAPS SUP 1.1-4 SSAR Section 2.1.1.1 is incorporated by reference with no departures or

supplements.

1.1.2.7 Rated Core Thermal Power

Replace the last four sentences of this section with the following.

NAPS COL 1.1-1-A Unit 3 operates at an estimated gross electrical power output at rated

power of approximately 1594 MWe (as shown in DCD Section 10.1). The

estimated net electrical power output, which is dependent on site ambientconditions, the normal plant heat sink (NPHS) operation controls, and

station electrical loads, is between approximately 1468 MWe and

1523 MWe.

I


Serial No. NA3-15-001Docket No. 52-017 North Anna 3

Page 6 of 15 Combined License ApplicationPart 2: Final Safety Analysis Report

1.5 Requirements for Further Technical Information

This section of the referenced DCD is incorporated by reference with thefollowing departures and/or supplements.

1.5.1 Evolutionary Design


CWR SUP 1.5-1 1.5.1.1 Post-Fukushima Near-Term Task ForceRecommendations

Following the March 11, 2011, Great Tohoku Earthquake and subsequenttsunami at the Fukushima Dai-ichi nuclear power plant, the NRC issuedOrders to licensees for implementing recommendations of the Near-TermTask Force Report (Reference 1.5-201). The following subsectionsdescribe how the recommendations applicable to the ESBWR areaddressed fer 'kit 3.

1.5.1.1.1 Recommendation 4.2, Mitigating Strategies forBeyond-Design-Basis External Events

Following the March 2011 events in Japan at the Fukushima Dai-ichinuclear power plant, the NRC issued to licensees Order EA-12-049,"Order Modifying Licenses with Regard to Requirements for MitigationStrategies for Beyond-Design-Basis External Events"(Reference 1.5-202). This Order was for implementingRecommendation 4.2 of the NRC Near-Term Task Force Report(Reference 1.5-201). Order EA-12-049 specifies a three-phase approachfor mitigating beyond-design-basis external events. The initial phaserequires the use of installed equipment and resources to maintain orrestore core, containment, and Spent Fuel Pool (SFP) coolingcapabilities. The transition phase requires providing sufficient, portable,on-site equipment and consumables to maintain or restore thesefunctions until they can be accomplished with resources brought from offsite. The final phase requires obtaining sufficient off-site resources tosustain those functions indefinitely. Interim Staff GuidanceJLD-ISG-2012-01, "Compliance with Order EA-12-049, Order ModifyingLicenses with Regard to Requirements for Mitigation Strategies forBeyond-Design-Basis External Events" (Reference 1.5-203), endorses,with clarifications, the methodologies described in Nuclear EnergyInstitute (NEI) 12-06, "Diverse and Flexible Coping Strategies (FLEX)Implementation Guide," (Reference 1.5-204). Although the guidance

1-24 Revision 9 (Draft (01/13/15)

Serial No. NA3-15-001Docket No. 52-017 North Anna 3Page 7 of 15 Combined License Application


does not specifically address the ESBWR design, which employs passive

design features, this subsection describes how ESBWR design features

for beyond-design-basis external events meet the intent of the guidance.

For the ESBWR, the underlying strategies for coping with extended loss

of AC power events involve a t-hee-two-phase approach as follows:

I. Initial Phase: Initial coping is implemented through installed plant

equipment, without any AC power or makeup to the ultimate heat

sink (i.e., safety-related Isolation Condenser System (ICS) and

Passive Containment Cooling System (PCCS) pools or

Gravity-Driven Cooling System (GDCS)). For the ESBWR, this

phase-initial coping is covered by the existing licensing basis (i.e.,

72-hr period for passive systems performance for core,

containment, and spent fuel storage pools cooling).

44, Trnc..itc.O. Phasc: Following the 72-hr passive system initial coping

time, support is required to continue passive system cooling and

makeup to the Isolation Condenser/Passive Containment Cooling

System (IC/PCCS) pools and spent fuel storage pools. This support

is-can be provided by installed plant ancillary equipment. The

installed ancillary equipment is designed with the capacity to support

passive system cooling from 3 to 7 days. As described in DCD

Sections 9.1.3 and 19A.3.1, makeup water can be provided to the

IC/PCCS or spent fuel pools through installed safety-related

connections to the Fire Protection System (FPS). Between 72 hours

and seven days, the resources for performing these safety functions

are available on site.

W. I1. Final Phase: In order to extend the passive system cooling and

IC/PCCS pools and spent fuel storage peels--pool cooling time te

beyond -7-dey -the initial phase (to an indefinite time), some off-site

assistance is required. Specifically, for the ESBWR design, diesel

fuel for the ancillary diesel generator or diesel fire pump must be

replenished. Also, mitigation strategies including procedures,

guidance, training, and acquisition, staging, or installation of

equipment needed for the strategies to maintain core, containment,

and spent fuel storage pools cooling for an extended period of time

will be fully implemented prior to initial fuel load.

1-25 Revision 9 (Draft (01/13/15)




NAPS COL 1.9-3-A Table 1.9-203 Conformance With the FSAR Content Guidance InRG 1.206

Section Section Title Conformance Evaluation

C.111.1 Pre-Operational Conforms. There are no BWR pressure3.9.2.4 Flow-Induced Vibration vessel internals that the referenced

Testing of Reactor certified design does not cover.Internals DCD Sections 3.9.2.3 and 3.9.2.4

adequately cover the analysis ofpotential adverse flow effects that couldimpact BWR vessel internals.

C.III.1 Dynamic System Conforms. Addressed in3.9.2.5 Analysis of the Reactor DCD Section 3.9.3.1 and

Internals Under Faulted DCD Table 3.9-2.Condition

C.111.1 Correlations of Reactor Conforms. Addressed in3.9.2.6 Internals Vibration Tests DCD Section 3.9.2.6.

with the AnalyticalResults

C.111.1 ASME Code Class 1, 2, Conforms. There are no3.9.3 and 3 Components and pressure-retaining components or

Component Supports, component supports designed orand Core Support constructed in accordance with ASMEStructures Code Class 1, 2, or 3, or GDC 1, 2, 4,

14, or 15, beyond those evaluated inthe referenced certified design.

C.111.1 Control Rod Drive Conforms3.9.4 Systems

C.1ll.1 Design Arrangements Conforms3.9.5.1

C.III.1 Loading Conditions Conforms3.9.5.2

C.111.1 Design Bases Conforms3.9.5.3

C.111.1 BWR Reactor Pressure Conforms. There are no reactor3.9.5.4 Vessel Internals pressure vessel intemals (including the

Including Steam Dryer steam dryer) or other main steamsystem components that are notcovered by the referenced certifieddesign. The rcootBFr *6 clasified ac,-,,e9e8ttYe- Classification of thereactor internals is described inSection 3.9.2.4.

C.111.1 Functional Design and Conforms. There is no safety-related3.9.6.1 Qualification of Pumps, equipment beyond the scope of the

Valves, and Dynamic referenced certified design.Restraints

1-164 Revision 9 (Draft (01/13/15)1-164 Revision 9 (Draft (01/13/15)




NAPS SUP44 41.11-1

Table 1.11-202 Supplementary Resolutions Related to NUREG-0933Table II TMI Action Plan Items and Human FactorsIssues

IAction

Plan Item/Issue

Number

Associated Location(s) WhereDiscussed and/or Technical

Description Resolution

TMI Action Plan Items

1.A.1.1 Shift Technical Advisor Sections 13.1.2.1.2.9 andDCD Section 18.6

1.A.1.2 Shift Supervisor Administrative Sections 13.1.2.1.2.5 andDuties 13.1.2.1.2.6

1.A.1.3 Shift Manning Section 13.1.2.1.4, Table 13.1-202,Figure 13.1-203, andDCD Section 18.6

1.A.2.1(1) Qualifications- Experience Section 13.1.3.1, Table 13.1-201,Section 17.5, and DCD Section 18.6

1.C.3 Shift Supervisor Sections 13.1.2.1.2.5 andResponsibilities 13.1.2.1.2.6

1.F.2(6) Increase the Size of Licensees' Table 13.1-201 and Section 17.5QA Staff

1.F.2(9) Clarify Organizational Section 13.1.1.2.7, Table 13.1-201,Reporting Levels for the QA and Section 17.5Organization

ll.B.3 Post Accident Sampling Appendix 12BB

II1.D.3.3 In-Plant Radiation Monitoring Appendix 12BB

Human Factors Issues

HF1.1 Shift Staffing Table 13.1-202 andSection 13.1.2.1.4

1-230 Revision 9 (Draft (01/13/15)

1-230 Revision 9 (Draft (01/13/15)


STD COL 13.4-1-ASTD COL 13.4-2-A

Table 13.4-201 Operational Programs Required by NRC Regulations

Program Source(Required by)

Implementation

Item Program Title Section Milestone Requirement

11. Non Licensed Plant Staff 10 CFR 50.120 13.2.2 18 months prior to 10 CFR 50.120(b)Training Program scheduled fuel load

(portions applicable to 10 CFR 30.32 Prior to initial receipt of 10 CFR 30.32(a)radioactive material) 10 CFR 40.31 byproduct source, or 10 CFR 40.31(a)

10 CFR 70.22 special nuclear materials 10 CFR 70.22(a)(excluding ExemptQuantities as described in10 CFR 30.18)

12. Reactor Operator Training 10 CFR 55.13 13.2.1 18 months priorto License ConditionProgram 10 CFR 55.31 scheduled fuel load

10 CFR 55.4110 CFR 55.4310 CFR 55.45

13. Reactor Operator 10 CFR 50.34(b) 13.2 Within 3 months after 10 CFR 50.54(i-1)Requalification Program 10 CFR 50.54(i) issuance of an operating

10 CFR 55.59 license or the date theCommission makes thefinding under 10 CFR52.103(g)

14. Emergency Planning 10 CFR 50.47 13.3 Full participation exercise 10 CFR Part 50,

10 CFR 50, Appendix E conducted within 2 years Appendix E,prior to scheduled date for Section IV.F.2.a(ii)initial loading of fuel

Onsite exercise conducted 10 CFR 50, Appendix E,within 1 year prior to the Section IV.F.2.a(ii)schedule date for initialloading of fuel

Part 2: Final Safety Analysis ReportNorth Anna 3 Combined License Application

Revision 9 (Draft (01/13/15)13-51


STD COL 13.4-1-ASTD COL 13.4-2-A

Table 13.4-201 Operational Programs Required by NRC Regulations

ImplementationProgram Source(Required by)Item Program Title Section Milestone Requirement

14. Emergency Planning(continued)

Licensee's detailedimplementing proceduresfor its emergency plansubmitted at least 180 daysprior to scheduled date forinitial loading of fuel

The licensee shall submit afully developed set ofsite-specific EmergencyAction Levels (EALs) to theNRC in accordance withthe NRC-endorsed versionof NEI 07-01, Rev. 0, withno deviations. The EALscheme shall have beendiscussed and agreedupon with State and localofficials. The fullydeveloped site-specificEAL scheme shall besubmitted to the NRC fee-eeRfimtmteR-at least180 days pFir. t8 initia fie-lead before the datescheduled for initial fuelload as set forth in thenotification submitted inaccordance with10 CFR 52.103a.

10 CFR 50, Appendix E,Section V

License Condition

Part 2: Final Safety Analysis ReportNorth Anna 3 Combined License Application

Revision 9 (Draft (01/13/15)13-52

Serial No. NA3-15-001 North Anna 3Docket No. 52-017Page 12 of 15 Combined License Application


................................................................................

,%-I;-NAPS SUP 13.5-37 13.5.2.2.8 Security Procedures

A discussion of security procedures is provided in the Security Plan.

The Special Nuclear Material (SNM) Physical Protection Program is the

10 CFR 70 required protection program in effect for the period duringwhich SNM is received and stored in a controlled access area (CAA), in

accordance with the requirements of 10 CFR 73.67.

The New Fuel Shipping Plan addresses the applicable 10 CFR 73.67

requirements in the event that unirradiated new fuel assemblies or

components are returned to the supplying fuel manufacturer(s) facility.................................................................................

STD SUP 13.5-38 13.5.2.2.9 Refueling and Outage Planning Procedures

Procedures provide guidance for the development of refueling andoutage plans, and as a minimum address the following elements:

" An outage philosophy which includes safety as a primary

consideration in outage planning and implementation

" Separate organizations responsible for scheduling and overseeing the

outage and provisions for an independent safety review team thatwould be assigned to perform final review and grant approval for

outage activities

" Control procedures, which address both the initial outage plan andsafety-significant changes to schedule

" Provisions that activities receive adequate resources

" Provisions that defense-in-depth during shutdown and margins arenot reduced or provisions that an alternate or backup system must be

available if a safety system or a defense-in-depth system is removed

from service

" Provisions that personnel involved in outage activities are adequately

trained including operator simulator training to the extent practicable,

and training of other plant personnel, including temporary personnel,commensurate with the outage tasks they are to perform

" The guidance described in NUMARC 91-06, "Guidelines for Industry

Actions to Assess Shutdown Management," to reduce the potential forloss of reactor coolant system boundary and inventory during

shutdown conditions (Reference 13.5-203)

13-70 Revision 9 (Draft (01/1 3/1 5)13-70 Revision 9 (Draft (01/13/15)


RCS Pressure SLB 2.1.2

BASES

APPLICABLESAFETY ANALYSES

(continued)

The RCS pressure SL has been selected such that it is at apressure below which it can be shown that the integrity ofthe system is not endangered. The reactor pressure vessel isdesigned to ASME, Boiler and Pressure Vessel Code,Section III, 2001 Edition, including Addenda through 2003(Ref. 5), which permits a maximum pressure transient of 110%of the design pressure of 8.618 MPaG (1250 psig). Therefore,the SL is 9.481 MPaG (1375 psig) at the lowest elevation ofthe RCS. The RCS pressure SL is selected to be the lowesttransient overpressure allowed by the applicable codes.

SAFETY LIMITS The maximum transient pressure allowable in the RCS pressurevessel under the ASME Code, Section III, is 110% of designpressure. The maximum transient pressure allowable in theRCS piping, valves, and fittings is 110% of design pressuresof 8.618 MPaG (1250 psig). The most limiting of theseallowances is the 110% of the RCS design pressure;therefore, the SL on maximum allowable RCS pressure isestablished at 9.481 MPaG (1375 psig) at the lowestelevation of the RCS.

APPLICABILITY SL 2.1.2 applies in all MODES.

SAFETY LIMIT Exceeding the RCS pressure SL may cause immediate RCSVIOLATIONS failure and create a potential for radioactive releases in

excess of 10 CFR 52.47(a)(2)(iv) (Ref. 4). Therefore, it isrequired to insert all insertable control rods and restorecompliance with the SL within 2 hours. The 2 hour CompletionTime ensures that the operators take prompt remedial actionand also assures that the probability of an accidentoccurring during this period is minimal.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 14 and GDC 15.

2. ASME, Boiler and Pressure Vessel Code, Section III,Article NB-7000.

3. ASME, Boiler and Pressure Vessel Code, Section XI,Article IW-5000.

4. 10 CFR 52.47(a) (2) (iv).

5. ASME, Boiler and Pressure Vessel Code, 2001 Edition,Addenda, 2003.

North Anna Unit 3 B 2. 1.2-2 Revision 7 (Draft 01/13/15)


RCS P/T LimitsB 3.4.4

BASES

SURVEILLANCEREQUIREMENTS

(continued)

The 30-minute Frequency reflects the urgency of maintainingthe temperatures within limits, and also limits the timethat the temperature limits could be exceeded. The 12-hourFrequency is reasonable based on the rate of temperaturechange possible at these temperatures.

REFERENCES 1. 10 CFR 50, Appendix G.

2. ASME, Boiler and Pressure Vessel Code, Section III,Appendix G.

3. ASTM E 185-82.

4. 10 CFR 50, Appendix H.

5. Regulatory Guide 1.99, Revision 2, May 1988.

6. NEDC-33441P, "GE Hitachi Nuclear Energy Methodologyfor the Development of ESBWR Reactor Pressure VesselPressure-Temperature Curves," Revision 6,November 2013.

CWR COL 16.0-1-A3.4.4-3

North Anna Unit 3 B 3.4.4-7 Revision 7 (Draft 01/13/15)


CRHAVSB 3.7.2

BASES

Table B 3.7.2-1

ESTABLISHEDHEAT SINK GROUP DESIGN TEMPERATURE

CRHA Heat Sink Group 1

Control Room Habitability Area: 23.3oC.(74°F)Main control room panel Rooms: No 3270, 3272, 3271,3201, 3202, 3273, 3206, 3205, 3204, 3275, 3207, 3208

Corridors: 1 25.60C (78 0 F)Rooms 3100, 3101 and Rooms 3200, 3203, 3277, 3274

HVAC chases: 1 25.6 0 C (780F)Rooms 3251, 3260


Q-DCIS equipment rooms: 25.6 0 C (780 F)Rooms No 3110, 3120, 3130 and 3140

N-DCIS equipment rooms: 25.6 0 C (780 F)Rooms 3301, 3302, 3303, 3300

Electrical chases: 1 25.6 0 C (78 0 F)Rooms 3250, 3261


HVAC equipment rooms: 40 0 C (104 0 F)Rooms 3401, 3402, 3403 and 3404

Safety Portions of CRHAVS: 40 0 C (104 0 F)Rooms 3406, 3407

1. Access corridors, electrical chases, and HVAC chases, although part ofthe CRHA heat sink, are not monitored because these areas do not containheat sources and their temperatures are assumed to match the average ofthe associated group.

North Anna Unit 3 B 3.7.2-13 Revision 7 (Draft 01/13/15)

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