,~

6212 DIN 8-013-523

PUBLIC SERVICE COMPANY OF OKLAHOMAA CENTRAL AND SOUTH WEST COMPANY

P O BOX 201/ TULSA. OKLAHOMA 74102 / (918) 583-3611

November 17, 1978,,,,,,,,g,,j,,

File: 6212.125.3500.32 LExecutie Vice Presk/ent

Public Service Company of OklahomaBlack Fox StationPSAR Amendment 13 gDocket STN 50-556 and STN 50-557 M

Office of Nuclear Reactor RegulationDivision of Project ManagementLight Water Reactors Branch No. 4U. S. Nuclear Regulatory CommissionWashington, D. C. 20555

Attn: Nteven A. Varga, Chief

Gentlemen:

Public Service Company of Oklahoma hereby submits sixty (60) copies plus threesigned originals of Amendment 13 to the Black Fox Station, Units One and Two,Preliminary Safety Analysis Report.

This Amendment is being submitted to update various areas of the docket as com-mitted to provide information to you. This information was previously submittedto you by letters dated November 7, 1978 (our 6212 DIN 8-013-291 and 6212 DIN 8-013-292) and November 10,1978 (our 6212 DIN 8-013-391) . Subjects covered are TheAbility of Ultimate Heat Sink to Withstand Tornado Missiles, LPCI Diversion EffectsOn ECCS Performance and revised PSAR Figure 14.16a showing the containment elevatorpit framing.

In addrion, Amendment 13 contains information on the educational and experiencerequirements for key personnel participating in the initial test program. Thisinformation is provided to update our response to staff question Q413.3A andreflects our present status in this area.

Finally, a copy of correspondence between General Electric Company and PSO regardingCalculated Peak Cladding Temperature From ECCS Analysis Following a Design Basis

N Accident For Black Fox Station is included as a part of this Amendment for yourinformation.

V ruly your

7811%80RTO , j_

~nMEF:so

Attacgmentsttached Service List w/o Attachmentsxc:

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k *$ hCENTRAL AND SOUTH WEST SYSTEM

Ei"% 'cgyg,nppght gc SppgCornpany of OMahorna g wgef,n}gjricPower gt}egs U':htiesPo

'b

* BLACK FOX STATION SERVICE LIST

XC: L. Dow Davis, Esquire Joseph R. Farris, EsquireWilliam D. Paton, Esquire John R. Woodard, III, EsquireColleen Woodhear, Esquire Green, Feldman, Hall & WoodardCounsel for NRC Staff 816 Enterprise BuildingU. S. Nuclear Regulatory Commission Tulsa, Oklahoma 74103Washington, D. C. 20555

Andrew T. Dalton, EsquireMr. Cecil Thomas 1437 South Main Street, Suite 302U. S. Nuclear Regulatory Commission Tulsa, Oklahoma 74119Phillips Building7920 Norfolk Avenue Mrs. Ilene H. YoungheinBethesda, Maryland 20014 3900 Cashion Place

Oklahoma City, Oklahoma 73112Mr. Jan A. NorrisEnvironmental Projects Branch 3 Mr. Lawrence BurrellU.S. Nuclear Regulatory Commission Route 1, Box 197Phillips Building Fairview, Oklahoma 737377920 Norfolk AvenueBethesda, Maryland 20014 Mrs. Carrie Dickerson

Citizens Action for Safe Energy, Inc.Mr. William G. Hubacek P. O. Box 924U.S. Nuclear Regulatory Commission Claremore, Oklahoma 74107Office of Inspection and EnforcementRegion IV611 Ryan Plaza Drive, Suite 1000Arlington, Texas 76012

Mr. Gerald F. DiddleGeneral ManagerAssociated Electric Cooperative, Inc.P. O. Box 754Springfield, Missouri 65801

Mr. Maynard HumanGeneral ManagerWestern Farmers Electric CooperativeP. O. Box 429Anadarko, Oklahoma 73005

Michael I. Miller, Esq.Isham, Lincoln & BealeOne 1st National PlazaSuite 4200Chicago, Illinois 60603

Mr. Joseph GalloIsham, Lincoln & Beale1050 17th Street N.W.Washington, D. C. 20036

+

.

$

.

.

RLACK & VEATCHTEL. (913) 967 2000

CONSULTING ENGINEERS TELEX 42-6263

1500 M E A DOW LAKE PAR KW AY

M AILING ADDRESS: P O. BOX No. 8405

KANSAS CITY. M ISSOU RI 64114

In the Matter of the Application ) Docket Nos. STE 50-556of PUBLIC SERVICE COMPANY OF ) STN 50-557OKLAHOMA Black Fox Station ) 6212.125.2000.21Units 1 and 2 )

CERTIFICATE OF TRANSMITTAL

I, Edwin L. Cox, Project Licensing Engineer for Black Fox Station, Black& Veatch Consulting Engineers, certify that a copy of Amendment 13 toPublic Service Company of Oklahoma's Black Fox Station (STN 50-556 andSTN 50-557) Preliminary Safety Analysis Report has been transmitted tothe following by United States Mail, postage prepaid, this 21st day ofNovember, 1978.

The Honorable Tommy Dyer Mr. Dale McHardMayor of Inola Oklahoma State Department of HealthCity Hall N. E. 10th St. & StonewallInola, Oklahoma 74036 P. O. Box 53551

Oklahoma City, Oklahoma 73105DirectorState Grand-in-Aid Clearinghouse Mr. Gerald Diddle4901 N. Lincoln Blvd. Associated Electric Coop, Inc.Oklahoma City, Oklahoma 73105 P. O. Box 754

Springfield, Missouri 65801Environmental Impact CoordinatorEncironmental Protection Agency Mr. Paul M. MurphyFirst International Building Isham, Lincoln, and Beale1201 Elm Street One First National PlazaDallas, Texas 75270 42nd Floor

Chicago, Illinois 60690Mr. Maynard HumanWestern Farmers Electric Coop Mrs. Ilene YoungheinP. O. Drawer 429 3900 Cashion PlaceAnada.-ko, Oklahoma 73005 Oklahoma City, Oklahoma 73112

Robert A. Franden, Esq. Chief, Docketing and Service SectionGreen, Feldman, Hall & Woodard Off.'.ce of the Secretary816 Enterprise Building U. S. Nuclear Regulatory CommissionTulsa, Oklahoma 74103 Washington, D. C. 20555

BLACK & VEATCH

.

.

Mr. Harold R. Denton Mr. Lawrence BurrellOffice of Nuclear Reactor Regulation Route #1U. S. Nuclear Regulatory Commission Box 197Washington, D. C. 20555 Fairview, Oklahoma 73737

Andrew T. Dalton, Esq. Joseph Farris, Esq.1437 South Main St., Suite 302 Green, Feldman, Hall & WoodardTulsa, Oklahoma 74103 816 Enterprise Building

Tulsa, Oklahoma 74103John R. Woodard, Esq.Green, Feldman, Hall & Woodard Atomic Safety and Licensing Appeal816 Enterprise Building Board PanelTulsa, Oklahoma 74103 U. S. Nuclear Regulatory Commission

Washington, D. C. 20555Ms. Carric DickersonP. O. Box 924Claremore, Oklahoma 74107

Copies of PSAR Amendment 13 were transmitted to PSO Legal Counsel, Isham,Lincoln and Beale, Washington, D. C., for service to:

Sheldon J. Wolfe, Esq.

Mr. Frederick J. Shon

Dr. Paul M. Purdom

#Edwin L. Cox

.

'

STATE OF OKLAHOMA

COUNTY OF TULSA

Martin E. Fate, Jr., being first duly sworn, deposes and states;That he is Executive Vice President of PUBLIC SERVICE COMPANY OF OKLAHOMA, the

Applicant herein; that he has read the following Amendment 13 to the Black FoxStation Units One and Two Preliminary Safety Analysis Report and knows thecontents thercof; that the same is true as he verily believes.

DATED: This /66 day of 'l') mm./uA) ,1978.

//f /

Si ed m'

,

f Mart [E.'TMJr. h)Executive Vice President

Subscribed and sworn to before methis /// 6 day of %c<am/A/ ,1978.

k &[Lthfu ahkro'

. Notary Public in and for the# dounty ofTulsa, State of Oklahoma

.

My Commission expires b(etum f /9f Lj'' /

-

.

STATE OF OKLAH0MA

COUNTY OF TULSA

Martin E. Fate, Jr., being first duly sworn, deposes and states;That he is Executive Vice President of PUBLIC SERVICE COMPANY OF OKLAHOMA, the

Applicant herein; that he has read the following Amendment 13 to the Black FoxStation Units One and Two Preliminary Safety Analysis Report and knows thecontents thereof; that the same is true as he verily believes.

DATED: This /[o #3 day of '7) m _ A A ) ,1978.

.

Sign Ob Mar,t47I~E M , Jr. (

Executive Vice Presidenf

Subscribed and sworn to before me

this /(, M day of 'h u v d o) ,1978.

k94 L4/tJY akA'n_'

Notary Public in and for the County ofTulsa, State of Oklahoma

My Lomission expires h>M uuo f / 91 L/

'

.

BFSPSAR

AMENDMENT 13DOCKET STN 50-556 and STN 50-557

.

STATE OF OKLAHOMA

COUNTY OF TULSA

Martin E. Fate, Jr. , being first duly sworn, deposes and states;That he is Executive Vice President of PUBLIC SERVICE COMPANY OF OKLAHOMA, the

Applicant herein; that he has read the following Amendment 13 to the Black FoxStation Units One and Two Preliminary Safety Analysis Report and knows'thecontents thereof; that the same is true as he verily believes.

DATED: This 16th day of November , 1978.

Signed s/ Martin E. Fate, Jr.

Martin E. Fate, Jr.Executive Vice President

Subscribed and sworn to before methis 16th day of _, November , 1978.

Roberta L. Parkey

Notary Public in and for the County ofTulsa, State of Oklahoma

My Commission expires February 5, 1982

BFS

ERRATA AND ADDENDA SHEETAMENDMENT 13, NOVEMBER 17, 1978

Remove Page Insert Page

Chapter One

1.9-22 1.9.221.9-23 1.9-23

Chapter Three

3C.14-17a 3C.14-17a

Chapter Six

6.3-1 through 6.4-1 6.3-1 through 6.4-1

6.3-la through 6.4-la 6.4-la

Chapter Nine

9.2-11 9.2-11

Chapter Thirteen

13.1-9 13.1-9

13.2-6 13.2-6

13.4-la 13. 4-la

Chapter Fourteen

14.1-1 14.1-1

14.1-2 14.1-2

14.1-3a 14.1-3a

14.1-3b 14.1-3b--- 14.1-6

Amendment 7

413-9 4 13 - 9

Place the Amendment 13 page

tab, notarization letter, and

Amendment 13 instructions

behind the Amendment 12 Questions

i

13-111778

BFS

ERRATA AND ADDENDA SHEET (Continued)

Remove Page Insert Page

and Responses. The remainder of

Amendment 13, consisting of the

LPCI Diversion Ef fects Analysis

and GE correspondence, is to be

placed behind the Amendment 13

instructions.

11

13-111778

BFS

TABLE 1.9-1 (Continued)Regulatory

Guide Title and Applicant's Position

1.111 Methods for Estimating AtmospLaric Transport and Dispersionof Gr aeous Ef fluents in Routine Releases from Light-Water-Cooled Reactors (Rev. O, 3/76 and Errata, 1/77)

Iodine dose pathway models follow this guide. A less con-servative analysis (General Electric's model) was also used.See the response to Regulatory Guide 1.109 for additionalCoEments.

1.112 Calculation of Releases of Radioactive Materials in Gaseousand Liquid Effluents from Light-Water-Cooled Fower Reactors(Rev. O, 4/76)Calculations utilized iodine and liquid radwaste sourceterms similar to those in this guide. In addition, calcu-lations were also performed using source terms from GeneralElectric's report, NED0-21159, for the above parameters.See PSAR Sections 11.2 and 11.3 for additional details.

1.113 Estimating Aquatic Dispersion of Effluents from Accidentaland Routine Reactor Releases for the Purpose of ImplementingAppendix I (Rev. O, 5/76)The aquatic dispersion model from WASH-1258 was used in thepreparation of the PSAR. That model is similar to that pre-sented in this guide.

1.114 Guidance on Being Operator at the Controls of a Nuclear PowerPlant (Rev. O, 2/76)

PS0 intends to meet the provisions of this guide.

1.115 Protection Against Low-Trajectory Turbine Missiles (Rev. O,

3/76)PS0 will meet the provisions of this guide in the design ofBFS.

1.116 Quality Assurance Requirements for Installation, Inspectionand Testing of Mechanical Equipment and Systems (Rev. O,6/76)PSO intends to meet the provisions of this guide.

1.117 Tornado Design Classification (Rev. O, 6/76)All systems, structures, and components required by thisguide to be protected against the effects of tornados areprotected by being housed in Category I structures. Theexception to this protection is the Off-Gas System which 13is located in the Turbine Building. For this system,PS0 has adopted the GE position that the release ofradioactivity as a result of damage to the system by atornado or tornado-generated missiles will not exceed

,

1.9-22 13-111778

BFS

TABLE 1.9-1 (Continued)Regulatory

Guide Title and Applicant's Position

10 CFR 100 guidelines.

13

1.118 Periodic Testing of Electric Power and Protection Systems(Rev. O, 6/76)

The PS0 supplied scope will conform to IEEE 338-71 as statedin PSAR Section 8.1. GE supplied scope will also meet thecriteria of IEEE 338-71 as stated in GESSAR Table 7.1-1.

1.119 Surveillance Program for New Fuel Assembly Designs (Rev. O,6/76)

This guide is not applicable to BFS since a previously provenfuel design will be utilized.

1.120 Fire Protection Guidelines for Nuclear Power Plants (Rev. O,6/76)

In lieu of following Regulatory Guide 1.120, PSO intends tocomply to the extent practicable with Branch TechnicalIosition APCSB 9.5-1, " Guidelines for Fire Protection for

Nuclear Power Plants," as modified by Appendix A to BranchTechnical Position APCSB 9.5-1, entitled " Guidelines for

Fire Protection for Nuclear Power Plants Docketed Prior toJuly 1, 1976."

In November, 1977 PS0 will submit a detailed fire hazardanalysis for Black Fox Station.

1.121 Bases for Plugging Degraded PWR Steam Generator Tubes(Rev. O, 8/76)

Not applicable to BFS.

1.122 Development of Floor Design Response Spectra for SeismicDesign of Floor-Supported Equipment or Components (Rev. O,9/76)

PSO will meet the requirements of this guide in the designof BFS with the exception of C.1, Boundary Peaks, which isdiscussed in Subsection 3.7.2.8.

1.9-23 13-111778

I BFS

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PARTIAL PLAN EL 592'-7*TOS

EL 592'-7*

EL 590'-9%**- s

EL 589'-8T 1 11

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SECTION I

; 13-111778 3C.14-17a

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CONTAINMENT

EL 59T-7'. NOTES

L STRUCTURAL STEEL FRAMING AT EL 589' 8 SH ALL BE' '

EL 58T- MS/g* I i LIMITED TO SECTIONS FOR WHICH THE WIDTH DOESNOT EXCEED 20~.

2. SUPPRESSION POOL HWL CL 570'-8*.

3. STRUCTURAL STEEL AT OR A80VE EL 589*-s* WILL BEP LA T E- DESIGNED IN ACCORDANCE WITH SECTION 12.0 OF

BFS PSAR APPENDIX 3C, SPECIFICALLY FIGURE 12 2SECTION 2 TO PREVENT THE ELEVATOR OR ANY PARTS THEREOF

FROM FALLING INTO THE SUPPRE$$ ION POOL.

6' It' 2' 0 5' 10'

3/16" = l'-0*

PUBLIC SERVICE COMPANY OF OKLAHOMA" " " *"

OI ACE FOR ST ATION - Unit t

REACTOR BUILDING-PLATFORM FRAMING 14-16 oELEVATOR PIT FRAM!NG

REACTOR BUILDING-PLATFORM FRAMINGELEVATOR PIT FRAMING

\u

FIGURE 14.16a k

BFS

6.3 EMERCENCY CORE COOLING SYSTEMS (CESSAR)

6.3.1.4 Capability to Meet Functional Requirements (CESSAR)

Addition: (CESSAR) page 6.3.2)

(8) Additional design conservatism for the pump suction lines

passing through containment is provided as described in

Subsection 3.9.3.5.

Justification: To identify design commitment described in

Subsection 3.9.3.5.

6.3.2.2.1 High Pressure Core Spray System (HPCS). (GESSAR Confirmed)

6.3.2.2.7 ECCS Pump Suppression Pool Screens (GESSAR).

Addition: The capability of ECCS and RCIC equipment to function

as needed in the presence of particles which can pass through the

Supression Pool suction line inlet screens will be reviewed and

the appropriate requirements will be included in specific equipment

specifications, as required, to ensure the proper performance of

the ECCS and RCIC systems.

6.3.3.10 Conformance with ECCS Acceptance Criteria of 10CFR50.46. (CESSAR)

Addition: Since the issuance of the 238 NSSS GESSAR changes have

occurred which affect the calculated peak cladding temperature (PCT).

Correspondance received by PS0 from GE summarizing these changes is

incorporated into this PSAR. A copy of the correspondance is located

in Volume X of this PSAR along with the LPCI Diversion report behind

the index tab for Amendment 13.13

6.3.3.12 Use of Dual Function Components (GESSAR)

Addition: Analysis of EPCl diversion on ECCS performance for Black

Fox Station Units 1 and 2 supplies specific information pertaining to

ECCS performance during diversion of flow from two of the RHR pumps

to containment spray. This analysis is included behind the question

tab for Amendment 13.

6.3.3.14 Thermal Shock Considerations. GESSAR Confirmed)

6.3.4 Tests and Inspections (GESSAR C6nfirmed)

6.4 HABITABILITY SYSTEMS

6.3-1 through 6.4 13-111778

.

6.4.1 Habitability Systems Functional Design

The habitability area design employs several systems and provisions

to assure habitability for a suitable period following postulated design

basis accidents. These habitability systems and provisions include the

following.

(1) Shielding

(2) Habitability Area HVAC System (which includes the EmergencyAir Cleanup Filtration System)

(3) Food and water storage(4) Kitchen and sanitary facilities

6.4.1.1 Design Bases. The functional design of the habitability systems

and provisions are established by the following design bases.

(1) Shielding

(a) The whole bcdy gamma radiation dose due to direct shine

from internal and external sources, will not exceed 5 rem

for 30 days following a DBA.

(2) Habitability Area HVAC System(a) The thyroid dose from the inhalation of radioactive idione

for the 30 day period following a LOCA will not exceed

30 rem to the thyroid.

(b) The beta skin dose from airborne activity within the habi-tability area for the 30 day period following a IOCA will

not exceed 30 rem.

(c) The habitability area environment will be maintained at a

minimum drybulb temperature of 70 F and a aaximum drybulbtemperature of 80 F. Maximum relative huuidity will be

60 per cent. For other design bases relative to the

Habitability Area HVAC System, refer to Subsection 9.4.1.

(d) Carbon dioxide and oxygen concentration in the habitabilityarea atmosphere will be maintained within safe levels, both

during complete airpatch isolation of the habitability area

and when outside air intakes are open.

(c) The habitability area envelope will be designed for minimal

leakage such that airborne radioactivity inleakage is within

acceptable limits following a design basis accident.

(f) Habitability requirements are based upon the assumptions of

6.4-la 13-11177813

BFS

9.2.5.3 Safety Evaluation. The Ultimate Heat Sink is capable of providing

sufficient cooling for more than 30 days:

(1) To permit simultaneous safe shutdown of both nuclear reactor units

and to maintain them in a safe shutdown condition and

(2) In the event of an accident in one unit, to permit safe control of

the accident and also permit simultaneous safe shutdown of the

other unit and to maintain it in a safe shutdown condition.

The UHS, consisting of cooling towers, fans, basin, pump house, and makeupbasin will be designed to withstand, without a loss of functional capability to

meet the requirements of items (1) and (?.) above, the following natural

phenomena; safe shutdown earthquake, probable maximum flood, and tornado wind

forces and tornado borne missiles.

The UHS cooling towers, basin, and the pump house will be constructed of

concrete walls and roofs.

The Ultimate Heat Sink is capable of withstanding the effects of other

applicable site-related events, reasonably probable combinations of less

severe phenomena, and any single creditable failure of any active component

without loss of the sink capability to provide the heat rejection necessary

to meet requirements of items 1 and 2 above. A single nonmechanistic failure

of a man-made structural feature of the UHS is considered to be incredible

since the UHS is designed to Seismic Category I requirements. Refer to

Tables 9.2-2 and 9.2-3 for a failure analysis of the Standby Service Water

System and the UHS.

The Ultimate Heat Sink in conjunction with the SSWS is designed to

withstand, without loss of safety function, the disabling of any one cooling

tower fan simultaneous with a single active failure in any one Unit 1 or

Unit 2 division during a loss of preferred power.

13

9.2-11 13-111778

BFS

13.1.2 Operating Organization

This section describes the organization structure, functions, and

responsibi'ities of the PSO operating organization for Black Fox Station.

13.1.2.1 Plant Organization. The organizational structure of Black Fox

Station is shown on Figure 13.1-4. This structure is based upon information

presented in ANSI N18.1-1971, WASH 1130, and as a result of review of the

organization of other nuclear power stations. The number of persons in each

position is shown on Figure 13.1-4 as is the license requirement to comply

with 10 CFR 55. These requirements also comply with the information

presented in 10 CFR 50.54.

13.1.2.2 Personnel Functions, Responsibilities, and Authorities

13.1.2.2.1 Station Manager. The Station Manager has overall responsi-

bility and authority for all phases of operation of the staff. He will

report to the Vice President, Power Generation in the PS0 General Office.

He is directly responsible for the safe, orderly, and efficient operation

of the station including administration of the BFS Emergency Plan and is

responsible to see that the station is operated in accordance with the

license and applicable regulations. He serves as the station's NRC

liaison for all communications concerning station operation. He is

responsible to maintain a qualified staff and assure that they are properlytrained. By supervisiag the activities of the Training Coordinator, he

shall assure that an effective training and retraining program is maintained.

The Station Manager shall also approve all administrative procedures and

policies regarding station operation and maintenance. Through the Securityand Office Supervisor, the Station Manager shall direct a security forceand the office staff.

13.1.2.2.2 Station Superintendent. The Station Superintendent reports

directly to the Station Manager and acts on behalf of the Station Manager

in his absence. He will be responsible for coordinating and directing the

activities of the Technical Supervisor, Radiation Safety and Chemistry (RS&C)

Supervisory, Operations Supervisor, and the Maintenance Supervisor. He will

3be chairman of the Operations Committee and the Test Working Group.13.1.2.2.3 Security and Office Supervisor. The Security and Office

Supervisor will be responsible for implementing and supervising the station

security.

13.1-9 13-111778

lJ

BLAC

1979 1980 1981

^BLACK FOX STATION

STATION MANAGER ESTATION SUPT 3 4 ESECURITY & OFFICE SUPV

LEAD SECRETARY *

CLER KS & SECR ETARIES *(2)1 13 1 3 ETRAINING COORD. *

,

'

OPERATIONS SUPV. 1 13 1 3 4 E* '

"

SHIFT SUPV. '(6) 1 13 1 4 E,

ASST SHIFT SUPV.

REACTOR OPERATOR '(13) 1 j 13 31 3 4 3ASST REACTOR OPERATOR 1 13 1 3 4 3'

EQUIPMENT OPERATOR *(10) 3@ 13MAINTENANCE SUPV. * 6 12 |

ELECTRICAL SUPV. * e 12 613 dSTATION ELECTRICIAN *

ELECTRICIAN '(2) 1ASST ELECTRICIAN '(2)

b13 6 12MECHANICAL SUPV. * '

MACHINIST *(1)-13 4MECHANIC *(3) '(3)' ;ilNTENANCE HELPER *(1) 1STORES CLERK

NON-SKILLED LABOR

RAD.SAF ETY & CHEM. SUPV. * 14 to 9 11 '

HP & CHEM. TECH. * (1) 14 10 9 13 *(8) 14 1

RADWASTE OPERATOR *(1) 14*

TECHNICA L SUPV. bREACTOR ENGINEER *(1)STATION ENGINEER *(3)

gOC TF.CH. *

*I & C SUPV. 14 8 12 |

SR.1 & C TECH. * 14 8 12 |

1 & C TECH. * (5) 14AMT I & C TECH. *(3)

*DESIGNA TES APPROXIMA TE ON-BOA RD DA TE.

() DESIGNA TES NUMBER OF INDIVIDUA LS APPOINTED OR HIRED IF OTHER THAN ONE.

DESIGNATES TIME ALLOTTED FOR INITIAL TRAINING. (THENUMBER REFERS TO THE COURSENUMBER SHOW"

- 13 - DESIGNA TES TIME ALLOTTED FOR ON-SITE TRAINING.

1S

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K FOX STATION

1982 1983 1984

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PERSONNEL INITIAL ASSIGNMENT& TRAINING SCHEDULE

N TABLE 13.2-1)

"O FIGURC 13.2-1

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).

BFS

The Station Superintendent will serve as the chairman of the TestWorking Group.

Members will be appointed as required to ensure competent representationon the TWG from the BFS project organization, B&V site organization,GE (NSSS) site organization, and the project field management organization.Other parties may be assigned by the chairman for advice and consultationas appropriate.

Education and experience for the members of the TWG are as follows:

(1) For the Chairman refer to the Station Superintendent undersubsection 13.1.3.2.2.

(2) Black & Veatch representative shall have a BS in Engineeringor the Physical Sciences and four years power plant design oroperation experience. He shall have general familiarity withthe design of BFS which is the responsibility of B&V.

(3) General Electric representative shall have a BS in Engineering 13

or the Physical Sciences and four years power plant design,construction, or operations experience at least two of which

shall be nuclear. He shall be familiar with the GE designedportion of the plant.

(4) BFS Project Engineering representative shall have a BS inEngineering or the Physical Sciences and four years powerplant design or operations experience. He shall have general

familiarity with the design of BFS.

(5) Proj ect field management member shall have a high schooleducation plus four years power plant construction experience.He shall be familiar with BFS construction procedures.

13.4-la 13-111778

BFS

14.0 INITIAL TESTS AND OPERATION

14.1 TEST PROGRAM (CESSAR)

14.1.1 Administrative Procedures (Testing)The preoperational testing and initial startup programs for all

safety-related systems and components will be coordinated by the stationTest Working Group (TWG) and executed under the direction of the StationManager as described in Subsection 13.4.2. Overall responsibility will

rest on the Station Manager for the performance of the tests by the BFSstaf f augmented as necessary by other PSO personnel.

The plant operating and engineering personnel will have the responsibilityfor the preparation of the preoperational and initial startup test procedureswith technical direction as needed from B&V and GE or other consultants.Scoping documents, prepared by B&V and GE, which describe the testobjectives and acceptance criteria for each test, will be used in the

development of detailed test procedures. The procedures will identify alltesting required to demonstrate that structures, systems, and componentswill perform satisfactorily in service. The procedures will insure that the 13

testing is done within the acceptance-rejection criteria for performancelisted in the appropriate design documents. The procedures will assurethat all prerequisites for the given test have been met, adequateinstrumentation and equipment are available, and that the tests are performedunder suitable environmental conditions adhering to approved test methods.The test procedures are reviewed by TWG and the plant operating and engineeringpersonnel to assure incorporation of scoping documents, establishment oftest prerequisites and compliance with applicable regulatory requirements.The appropriate station staff supervisors shall approve each procedure.Final authorization for use shall be the responsibility of the StationManager. Selected preoperational test procedures for system important tosafety will be reviewed by the Review and Audit Committee to ascertain

whether the procedures will satisfy the test objectives, meet acceptancecriteria established by the system designer, require documentation sufficientto verify results, and establish base line data for use during plant life.

Execution of the test procedures will be carried out by the operatingstaff of PSO with technical direction as needed from GE and Black & Veatch

14.1-1 13-111778

BFS

or other consultants. The test results are prepared in final form by theoperating staff or designates and i viewed by the station staff supervisorsto determine that test objectives have been satisfied and that the establishedacceptance criteria have been not. The prropriate system design organizationwill review the test results to determine that the system design objectiveshave been met.

The procedure approval and the test execution, review, and approval willbe documented according to established QA procedures. The approved testprocedure, results report, and data are retained at the station. These

documents will be retrievaole throughout the plant life or until they are nolonger required in accordance with ANSI N45.2.9-1974. Personnel qualifications

13are discussed in Chapter 13 and subsection 14.1.6.7. Organization of the

review committees is discussed in Chapters 13 and 16.14.1.2 Administrative Procedures (Modifications)

The appropriate system design organization shall review all test-resultsfor safety-related systems and components. If test results, test data, or

system performance are unsatisfactory, the system design organization shallrecommend appropriate corrective action.

If the deficiency is a result of equipment performance due to improperinstallation and does net require a design change, appropriate project fieldmanagement shall assure the deficiency is corrected by the appropriatecontractor. The documented corrective action will be reviewed and verified byan appropriate member of project field management, and the testing resumed.

Modifications which require a design or hardware change or which alterthe intent of the procedure must be reviewed and approved by B&V or GEengineering support personnel as appropriate.

Modifications which require technical specification changes or concernan unreviewed safety item must be reviewed and approved by tne Review andAudit Committee prior to submittal to the NRC as required by 10 CFR 50.59.

Minor modifications to test procedures which do not change the intentof the procedure, do not require design or equipment changes, and clearlydo not sacrifice plant or personnel safety may be made by the appropriatestation staff supervisor with postapproval by the Station Manager.

The station Test Working Group shall serve to interpret the type ofdeficiency for other than minor modifications and identify responsibilityfor correction. All modifications to systems or procedures must be soprovedby the Station Manager.

14.1-2 13-111778__

BFS

with station procedures which are formulated using the guidance of ANSIN18. 7-1972, " Administrative Controls for Nuclear Power Plants," as describedin Subsection 13.5.1.14.1.6.4 Utilization of Plant Operating and Testing Experiences at Other

Reactor Facilities. During the preparation of detailed preoperationaltesting procedures and operation procedures licensee event reports for similarreactor plants, and operating experiences will be reviewed to identifypotential problens. Precautionary measures will be factored into procedural 13steps to preclude or minimize their occurrence. The schedule for this

evaluation activity will be coincident with the schedule for preparation.14.1.6.5 Test Program Schedule. The test program will follow the scheduleindicated on Figure 14-1 and will follow the typical NSSS test sequencepresented on Figure 14.1-1 (GESSAR). The scheduled time period for developmentof detailed test procedures will commence approximately 6 months prior toperforming the initial tests which is consistant with the schedule for hiringand training of plant operating and engineering personnel presented onFigure 13.2-1. Staffing and training for the Black Fox Station is discussed

in Chapter 13. Plant operating and emergency procedures will be developedduring the initial testing phase concurrent with the preparation of pre-operational test procedures and the performance of preoperational tests.14.1.6.6 Trial Use of Plant Operating and Emergency Procedures. During thepreoperational testing program, the operating and emergency procedures for theplant will be placed in trial use, to the extent practical, to verify theiradequacy end appropriater.ess. The experiences gained during the trial use

,

period will be factored into the final operating and emergency procedures asappropriate.

14.1.6.7 Staff for Conduct of Test Progran. As described in Subsection 14.1.1,PSO's staff will be augmented as needed with technical direction provided byB&V and GE or other consultants for the preparation of preoperation.initial startup test procedures. Execution of the test procedures will oe

carried out by the operating staff of PS0 with technical direction as needed 13

from CE & B&V or other consultants. Such technical direction will consist of

providing scoping documents containing test objectives, performance requirements,and acceptance criteria and providing consultation as needed. Augmentation of

14.1-3a 13-111778

BFS

the normal BFS operating staf f during the performance of the preoperationaltests will be provided as needed by members of the PSO Power Generation staffas mentioned in subsection 13.4.2., B&V, GE and other consultants. Theinterrelationships, interfaces and the general qualifications of theparticipants, PSO, B&V, and CE, in the test program are described in thischapter and chapter 13. The schedule for augmenting PSO's staff will beconsistent with the schedule for preparation and execution of the testprocedutes.

The person in responsible charge of preparing preoperational testprocedures shall have a bachelor of science degree in engineering or thephysical sciences and four years experience in design and/or operation ofpower plants at least one of which will be in the preoperational phase of anuclear plant test program. An additional four years of experience shall 13

be required without the degree. This individual may also be in responsiblecharge of the conduct of preoperational tests, but if another is selectedthat person will meet the same requirements.

The person (s) assigned to be in responsible charge of developing post-fuel loading test procedures and of conducting the post-fuel loading testsshall meet the "American National Standard for Selection and Training ofNuclear Power Plant Personnel" ANS 3.1-1978 for the responsible person incharge of reactor engineering.

For both the preoperational testing and startup testing phases,personnel developing procedures and supervising the conduct of tests shallbe qualified using a program which will comply with " Qualifications ofInspection, Examinations, and Testing Personnel for the Construction Phaseof Nuclear Power Plants", ANSI N45.2.6-1973.

14-1-3b 13-111778

TEST PROGRAM SCHEDULE *

FIGURE 14-1

FUEL-24 MOS. - 18 MO S. -12 MOS. -6 MOS. LOAD +6 NOS.

PREPARE ADMIN. AND PERFORM COMPONENT, COMPONENT TEST PROCEDut,3 IESTS

I

Ah0 US PERFORM PRE-0P TESTS

FUELLOAD

2 PRECRITICAL, INITIAL CRIT.7 LOW PWR. PHYSICSe

POWERASCEN SION

< PRE-OP PHASE ; ; STARTUP TEST PHASE *

* AFPLIES TO UNIT ONE. THE DURATION OF PROCEDURE PREPARATION FOR UNIT TWO IS EXPECTED TO BE SHORTER.

BFS

Question 413.3A.

Question: We requested in Question 413.3 additional clarifying informationon the responsibilities of organizations participating in thedevelopment, conduct and review of test results for each phaseof the initial test program. Additionally, we requested thatminimum educational and experience requirements for key personnelparticipating in *he initial test program be described. To date,

this information has not been provided. We will require thatsufficient information be provided to demonstrate that adequateplans and minimum requirements have been established to assurethat adequate testing will be conducted.

Response: The Applicant is fully committed to properly performing aaastaffing an initial test program - including the preoperational,fuel load, low power testing, and power ascension phases. TheApplicant is also full; cognizant of the need for early planningfor the initial test program to avoid staffing and proceduralproblems.

Preliminary plans for conducting the initial test programare described in Section 14.1 of the Black Fox Station PSAR.Commitments to Regulatory Guide 1.68 and ANSI N18.7-1972 are

described in GESSAR NSSS Docket STN 50-550 which is referencedby PSAR Section 14.1.

The Applicant further commits to provide the remainder ofthe information concerning personnel qualifications in thelatter part of 1978.

13The detailed description of the initial tes: program willbe included in the FSAR. PSO excepts submittal of the FSAR to

be at least 24 months prior to the date scheduled for loadingfuel in Unit 1.

413-9 13-111778

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ANALYSIS OF LPCI DIVERSIONEFFECTS ON ECCS PERFORMANCE FOR

BLACK FOX STATIONUNITS 1 AND 2

NOVEMBER 2, 1978

DOCKETS STN 50-556STN 50-557

13-111778

BFS

This report was previously transmitted to the Nuclear Regulatory

Commission by letter from Public Service Company of Oklahoma on November 10,

1978. Two telephone conversations, November 16 and November 17, have taken

place between the NRC Staff and PSO relative to information in the report

and information requested in the NRC lette- dated October 11, 1978. This

report incorporated in the PSAR as part of t_endment 13 incorporates one

correction on page two, paragraph number ', where the previous word " increases"has been changed to " decreases". Public Service Company of Oklahoma will

respond fully to the balance of the NRC Staff concerns on this matter in

PSAR Amendment 14.

i 13-111778

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Purpose

This analysis was performed to investigate the effect on the ECCS

analysis for the Black Fox Nuclear Power Station of diverting low pressurecoolant injection (LPCI) pumps to the containment spray mode ten minutes

after a loss-of-coolant accident (LOCA) initiation.Automatic diversion of LPCI flow to containment spray has been provided

in response to an NRC requirement to assure containment integrity forpostulated high steam flow bypassing the suppression pool. Such flow

diversion would occur only if a high containment pressure (>9 psig)signal is present after ten minutes. The assumption of sufficient bypassingto cause such a pressure has been shown by CE to be extremely conservative_nd unrealistic.1Conclusion

The results show that the worst single failure / break type combinationis the high pressure core spray (HPCS) line break (approximately .02 ft )assuming the failure of the low pressure and core spray (LPCS) dieselgenerator (D/G) which powers one LPCS pump and one LPCI pump. This single

failure / break type combinat'on yields the highest peak cladding temperature(approximately 1985 F) o' '1 the cases affected by LPCI diversiv.: at ten

minutes. The peak c'. adding temperatures experienced by the cases affectedby LPCI diversio- re below the limits established in 10 CFR 50.46 (2200 F).This temperate s also below the peak clad temperature (PCT) calculated

for the break of a recirculation line (2038 F) which is not adverselyaffected by LPCI diversion at ten minutes.

Assumptions

(1) A maximum of two LPCI pumps (specifically LPCI "A" and LPCI "B")can be fully diverted at ten minutes to the containment spraymode. (NOTE: LPCI "A" shares an emergency diesel generator with

the LPCS: LPCI "B" and "C" share an emergency diesel generator.

The pump associated with LPCI "C" cannot be diverted to containment

sprays.)

1NEDO-10977 Drywell Integrity 'udy: Investigation of Potential.

Cracking in BWR/6 Mark III Containment.

1 13-111778

BFS

(2) The standard FSAR assumption of one automatic depressurization

system (ADS) valve failure combined with the worst additional

single failure was retained because this assumption is built into

the present model. This bounding assumption yields conservatively

higher calculated peak cladding temperatures (PCTs) by approxi-

mately 100 F. The PCT reported on Page 1 does not include this

. assumption.

(3) Approved Appendix K analysis models were used, except that someLPCI flow to the reactor vessel was stopped ten minutes after the

accident.

General Observations from the Analyses

only those accident cases which are not reflooded to the hot node

before ten minutes are affected by the assumed LPCI diversion. Once the

. core has been reflooded, only one ECCS pump is necessary to keep the core

covered. Thus, the breaks affected include small breaks less than

approximately 0.2 ft (depending on the break location) and outside steam

line breaks (OSLB). The effect of the assumed LPCI diversion on the OSLBis small and is discussed in a later section of this report.

Af ter reviewing the effect of diversion on the rest of the small breaks,

general statements can be made to describe the results in the area of

interest:

1. The calculated PCTs (no LPCI diversion) in the small breakregions affected by LPCI diversion generally decrease with

decreasing break size. This follows from the fact that the core

is uncovered for shorter periods and that the decay heat is lower

at the time of uncovery as the break size decreases.

2. The maximum temperature for the assumed LPCI diversion case for

any given break location occurs at approximately that break size

where the LPCI system would normally inject flow into the vr el

starting at 600 seconds (i.e. the assumed LPCI diversion time).

Bigger breaks get some reflooding benefit from the LPCI pumps

before diversion. Smaller breaks have the same ECC systems

available as this maximum break, but the smaller break area has

2 13-111778

BFS

a lower calcularca PCT, as discussed previously. As an exa aple,this worst break is ind' ated on Figure 1. A longer LPCI diversion

time muld have correspo.adingly smaller breaks where the maximumtemperature would occur and hence lower calculated PCT.

3. Diverting LPCI from its ECCS flooding function does not alwaysresult in higher PCTs. When compared to no LPCI diversion, a

reduction in PCT can be observed as a result of diverting LPCIif the LPCS is available. The reduction of subcooled LPCI waterresults in a reflooding mixture (due largely to LPCS flow) ofsteam and water which has higher voids. Thus, in the case where

little LPCI flow is available for re looding, even though lessECCS flow is entering the vessel, the swollen level inside the

lower plenum is higher and reflooding can occur sooner. In such

cases the calculated PCTs are extremely low and changes in PCTin either direction are insignificant.

4. Because this investigation is primarily concerned with smallbreaks, the failure of the HPCS, for non-core spray line breaks,is the worst single failure for this study. If the HPCS wereoperable, the break sizes being analyzed would reflood earlierthan ten minutes with the very small break sizes never uncovering.

he following break locations were considered: A) core spray line,

B) recirculation line, C0 feedwater line, D) the steam line, and E) LPCIline. A brief summary of each analysis is provided below.

A. Core Spray Line Break (HPCS Line) - It is conservatively assumedthat no flow enters the vessel through the broken line independentof the break size. For this case, the failure of the diesel

generator associated with LPCS and LPCI "A" is the worst single

failure since all credit for core spray cooling is eliminated.

The ECC systems remaining before diversions are 2 LPCI + ADS

after diversion at ten minutes. Because in both cases thereflooding the vessel, there is a longer ~eflooding time associatedwith the diverted case with reduced ECCS flow. The results ofthis investigation are shown in Figure 1. Because the temperature

3 13-111778

BFS

increase from the non-divered case is a result of a loss ofreflooding flow from 1 LPCI pump, intermediate cases (loss of partof the flow) will experience intermediate (lower) temperature

increases.

This particular failure / break type combination was the most

adversely affected by the assumed LPCI diversion. However,the peak cladding temperatures are still below the limit of

2200 F.

B. .eirculation Line Break - For this break, the worst single

failure is the HPCS failure, as described previously. The ECCSremaining before diversion are 3 LPCI + LPCS + ADS and, after

diversion, 1 LPCI + LPCS + ADS. Since in the diverted case the

remaining LPCI flow is not enough to significantly quench the

voids in the lower plenum, the mixture in the lower plenum willreflood with a higher voided mixture. This higher void fraction

for the diverted case more than offsets the reduction in ECCS flowentering the vessel due to this div .... of LPCI. Hence, there

is a net reduction in PCT due to a shorter reflooding time andthe recirculation line break without diversion which has alreadybeen reported is bounding relative to a line break with diversion.

A representative break (.01 ft ) was analyzed which confirmedthese results. The results of this investigation are shown in

Table 1. Intermediate cases (diversion of less than the fullflow from two pumps) should result in smaller temperr.uredecreases.

C. & D. Feedwater and Steam Line Breaks - For these breaks, the worst

single failure is the HPCS failure, as described previously.The ECCS remaining before diversion are 3 LPCI + LPCS + ADS

and after diversion 1 iPCI . LPCS + ADS. For the diverted case,

there will be a reduct!.on in calculated PCT for the samereasons discussed for :he recirculation line break. A representa-

tive break (i.e. 01 fc ) was again analyzed which confirmed the

anticipated results. The results of this investigation are shown

4 13-111778

BFS

in Table 1. For both cases, insignificant decreases in calculated

PCT result from LPCI diversion. Tae outside (isolated) steamline break was also considered with similar results.

E. LPCI Line Break - As in the case of the core spray line break,

it is conservatively assumed that no flow enters the vessel

through the broken line independent of the size. For this

break, the worst single failure is the HPCS failure, as described

previously. The ECCS remaining before diversion are 2 LPCI +

LPCS + ADS ano, tfter diversion, LPCS + ADS (if the break is in

line "C") or LPCS + LPCI + ADS (if the break is in lire "A"/ or "B") .In either case there is insuffic'-nt LPCI flow to significantli

quench the voids in the lower rienum. Therefore, the core

will reflood with a voided mixture. This higher void fraction

more than offsets the reduction in ECCS flow entering the vessel

due to diversion of LPCI. Hence, there is a net reduction in PCT

due to a shorter reflooding time.

As above, the .01 ft break was analyzed which confirmed the

anticipated results. The results of both diverted cases are

shown in Table 1.

Response to Question (1)

The system provided for diversion of LPCI flow is a safety grade system.

Consequently, it has a high reliability in performing its intended function.

Postulation of a failure of this system to perform its function in combination

with another single failure is not required under CDC 35 or 10 CFR 50.46.

Response to Question on Operator Action

The operation of the ECC systems including diversion of LPCI to contain-

ment sprays requires no operator action for at least 10 minutes following

accident initiation. Ten minutes is the present licensing basis for

operator manual action time following automatic actuation of the ECC system.

There is no requirement either in 10CFR50.46 or GDC 35 for assuming nooperator action 20 minutes after the initiation of the accident. Ten minutes

continues to be the licensing basis used and supported by General Electric.

It is also the basis for the containment- performance evaluation as it has

been for other BWR plants.

5 13-111778

TABLE 1

THE EFFECT ON THE PCT OF DIVERTING LPCI FLOW AT10 MINUTES FOR VARIOUS .01 FT2 BREAK TYPES

PCT PCTBreak Type No Diversion With Diversion

Recirculation Line 948 F 877 F

Feedwater Line 917 F 836 F

Inside Steam Line 920 F 831 F

LPCI Line 834 F 804 F(964 F( }

NOTE: (1) PCT if break occurs in LPCI line "A" or "B"(2) PCT if break occurs in LPCI line "C"

.

(

6 13-111778

2000

REFLOODING TIME <10 MINUTES

\ '-

2000 - /p \/mm.m. # \

- .

E,un

g 3500 -

4mE1w DI

hin

g 3000 -

BREAK SIZE FOR WHICH LPCI INJECTS$ FLOW INTO THE VESSEL AT 10 MINUTESU

kE BASE CASE (NO DIVERSION)

- ==== ammme i LPCI REMAINING AFTER

1r DIVERSION AT to MINUTESSoo i

I I I I Ioo.002 0.005 0.0 t 0.02 0.05 0.1 0.2 0.3g

BREAK AREA (f t2)

O$ Figure 1. Peak Cladding Tenperature versus Break Area for an HPCS Line

Break Assuming an LPCS Diesel-Generator Failure. SystemsRemaining: 2 LPCI + ADS

60 -

40 -

;

wJ

b TAF

a Mm 5 M

YJ ~

BAF

g I I l00 3m 600 1200 1500

C TIME isec)LCD Figure 4a. Water Level Inside the Shroud Following a HIGH PRESSURE Core* Spray Line Break, LPCS DG Failure, Break Area = 0.02 ft2 (SBM).

1200 -

_

_i .00 _

ua

5awEdNw w>e ns

t/2

400

| | |*

O 300 000 900 1200 1500mu TIME (sec)i

ZC Figure 4b. Reactor Vessel Pressure Following a IIIGli PRESSURE Core Spray5 Line Break, LPCS DC Failure; Break Area = 0.02 ft2 (SBM).

-

E0

N.ot

43E$e9i8U

O ,

2<

ai

10,000

25

I | | |

0 300 000 900 1200 1500Cb TIME (sec)

'

CZ Figure 4c. Convective lleat Transfer Coefficient Following a llIGil PRESSURE

Core Spray Li.ne Break, LPCS DG Failure, Break Area = 0.02 ft2(SBM).

.

3000 -

E*_us

3 2000 -

45t#-

90 a !a=

ed4

som -

t

8 | I0

0 600 900 1200 1500

C TIME (setl,L

Z Figure 4d. Peak Cladding Temperature Following a llIGli PRESSURE Core SprayLine Break, LPCS DG Failure, Break Area = 0.02 ft2 (sgM),*

2000

t

i-

eD

aS+ i000 -

5 .*i N~

w *

-

| | |-

tw 0$ 0 500 1000 1500

TIME (sec)N*

Figure 4e. LPCI Flow Rate Following a llIGli PRESSURE Core Spray Line Break,LPCS DG Failure, Break Area = 0.02 ft2 (SBM).

BFS

GEN ER AL @ ELECTRIC NUCLEAR ENERGY

PROJECTS DIVISION

GENERAL ELECTRIC COMPANY,175 CURTNER AVE., SAN JOSE, CALIFORNIA 95125

MC 392, (408) 925-3217

DIN # 6212 DIN 8-013-528November 14, 1978FILE: 6212.312.5560.21L

Mr. T. N. EwingProject ManagerBlack Fox Station Nuclear ProjectPublic Service Company of OklahomaP. O. Box 201Tulsa, OK 74102

Attention: Mr. Vaughn Conrad

Gentlemen:

SUBJECT: CALCULATED PEAK CLADDING TEMPERATURE (PCT) FROM ECCSANALYSIS FOLLOWING A DESIGN BASIS ACCIDENT (DBA) FORBLACK FOX PLANT '

It has become apparent recently that there is some confusion as to whatthe appropriate PCT following a DBA at Black Fox should be. This letterdescribes the changes which have occurred in the PCT from the issuanceof 238 NSSS GESSAR which is referenced by Black Fox, to the present,along with the reason (s) for each change.

Table 1 illustrates the ghronology of changes to the PCT starting withthe GESSAR value gf 2180 F and ending with the current value for BlackFox which is 2038 F. The reason (s) for the changes are also noted. Itis of interest to note that the trend in the PCT since GESSAR due tothese changes has been decreasing. It should also be noted that therecent ECCS analysis with low pressure coolant injection (LPCI) diversiondoes not change the current limiting calculated PCT of 2038 F for BlackFox.

Sincerely

Q*

-

A. J. Levine, ManagerProject Licensing Unit 1

AJL:daj/873

cc: Dr. M. J. RobinsonL. R. Cannard

13 13-111778'

BFS

CHANGES IN PEAK CLADDING TEMPERATURE (PCT)SINCE GESSAR

PCT (*F) Reason for Change

2180 GESSAR Table 6.3-3

2129 1. Change from REFLOOD 03 to REFLOOD 04 to incorpor-ate a correction to the calculation of vaporflow split between the fuel bundles and jetpumps.

2042 1. Correct the water level setpoint in SAFEwhich signals initiation of the high pressurecore spray system.

2. Modify the core power in REFLOOD to beconsistent with requirements of 10CFR50Appendix K (102% licensed power level)

3. Correct initial core average quality toimprove code accuracy in representing bubblerise velocity.

4. Correct recirculation line break area for thesuction end of the recirculation suction linebreak in SAFE calculation. The limiting flowarea is based on the vessel nozzle safe endinside diameter.

5. The guide tube thermal resistance was increaseda factor of 10 due to deciaml point error.

2038 1. Implement additional 1 psi core pressure dropin REFLOOD as required by the NRC.

2. Use of more accurate pressure table,

h 14 13-111778