ML17275B199

Forwards App B to FSAR Response to Regulatory Issues Resulting from TMI-2. App Is Part of Amend 17,which Will Be Sent within 1 Month

ML17275B199
Person / Time
Site:	Columbia
Issue date:	07/31/1981
From:	Bouchey G WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Schwencer A Office of Nuclear Reactor Regulation
References
GO2-81-219, NUDOCS 8108250601
Download: ML17275B199 (237)

Text

~ e1 J% REGULAT INFORMATION DISTRIBUTIO YSTEM (RIDS)

ACCESSION- NBR o 81 0825060 f DOC 0 DATE't 81/07/3 f NOTARIZED:: NO DOCKEiTI ¹ FACIL{:50: 397 NPPSS Nuclear ProJect~ Unit 2'r Hash'ington Public Powe 05090397 AUTH{,NAMEl AUTHOR AFF ILI ATION BOUCHEIY~G.D, Nasheington Public Power Supply System REC IP ~ NAME{ REC'IPIKNT AFF ILIATION-SCHWENCERgA ~ L{icensing, Branch 2

SUBJECT:

Forwards" App. 8{ to FSAR "Response" to Regulatory Issues-Resultlng From, TMI ?.": App is part" of Amend 17~which.'ill bel sent w,ithin 1 month, DISTRIBUTION CODEi: BOOIS COPIES RECEIVED:LiTR J 'NCL'PP SIZE ~:

PSAR/FSAR AMDTS and Rel ated Correspondence gg'iITlE{

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'1 Washington Public Power Supply System P.O. Box 968 3000 George Washington Way Richland, Washington 99352 (509) 372-5000 Docket No. 50-397 July 31, 1981 G02-81-'219 NS-L-02-CDT-81-018 Director, Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington D.C. 20555 Attention: Mr. A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing Gentlemen:

Subject:

SUPPLY SYSTEM NUCLEAR PROJECT NO. 2 FSAR APPENDIX B - WNP-2.'RESPONSE TO REGULATORY ISSUES RESUL'T'ING FROM. TMI-2 Enclosed are sixty (60) copies of Appendix B to the WNP-2 FSAR, "WNP-2 Response to Regulatory Issues. Resulting From TMI-2".

This appendix is a part of FSAR Amendment 17 which will be distributed to you within one month.

Very truly G. D. BOUCHEY yours,'DB:CDT:ct

~g Director, Nuclear Safety

'I Enclosure

. cc: WS Chin, BPA V. Stello, NRC AD Toth, NRC, Resident Inspector J'. Plunkett, NUS Corporation WNP-2 Files NS Reynolds, D8L .

SSOSa5OSOi, 05000397 S~07iit'DR ADOCK '

l PDR

WU ~

'mraumhNT NO. 17

.July 1981 810825pgpy APPENDIX B WNP-2 RESPONSE TO REGULATORY ISSUES sO RESULTING FROM TMI-2

WNP-2 AMENDMENT NO ~ 17 July 1981 APPENDIX B WNP-2 RESPONSE TO REGULATORY ZSSUES .-

RESULTING FROM TMI-2 TABLE OF CONTENTS Pacae I.A. 1. 1 Shi ft Technical Advisor B. 1-1 Z.A.1.2 ~

Shift Supervisor Responsibilities ~ B.1-4 I.A.1.3 Shift Manning B.1-8 Z.A.2.1 Immediate Upgrading of Operator and Senior Operator Training and Qualification B~ 1-13 Z.A.2.3 ~ Administration of Training Programs

~

~

for Licensed Operators B.1-15

'I.A.3.1 Revise Scope and Criteria for Licensing Examinations - Simulator for Exams B~ 1-17 Z.BE 1.2 Xndependent Safety Engineering Group B.1-19 I.C.1 GUXDANCE FOR THE EVALUATION AND DEVELOP-MENT OF PROCEDURES FOR TRANSXENTS AND ACCIDENTS B~ 1-23 Z.C.2 SHIFT AND RELIEF TURNOVER PROCEDURES B~ 1-28 X.C.3 SHIFT SUPERVISOR 'RESPONSIBILITY B.1-30 I.C.4 CONTROL ROOM ACCESS B.1-33 I.C.5I PROCEDURES FOR FEEDBACK OF OPERATING EXPERIENCE TO PLANT STAFF B 1-34 I.C.6 GUIDANCE ON, PROCEDURES FOR VERIFYING CORRECT PERFORMANCE OF OPERATING ACTIVITIES B.1-36 I.C.7 NSSS VENDOR REVIEW OF PROCEDURES B.1-38 Z.C.8 PXZ OT MONITORING OF SELECTED EMERGENCY PROCEDURES FOR NEAR-TERM OPERATING LICENSE APPLICANTS B.1-39

N WNP-2 AMENDMENT NO. 17 July 1981 TABLE OF CONTENTS (Continued)

Pacae I.D.1 CONTROL ROOM DESIGN REVIEWS B.1-40 I.D.2 PLANT SAFETY PARAMETER- DISPLAY CONSOLE B.1-43 I.G. 1 PREOPERATIONAL AND LOW-POWER TESTING B.1-45 II.B 1 REACTOR COOLANT SYSTEM VENTS B 2-1 IZ.B.2 DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION OF EQUIP-MENT FOR SPACES/SYSTEMS WHICH MAY BE USED XN POST-ACCIDENT OPERATIONS B. 2-6 ZI.B.3 POST-ACCIDENT SAMPLING CAPABXLITY B. 2-12 ZI'.4 TRAINING FOR MZTXGATZNG CORE DAMAGE B.2-17 II.D. I RELXEF AND'AFETY VALVE TEST REQUIREMENTS B.2-18 ZI.D 3 REZZEF AND SAFETY VALVE POSITION INDICATION B. 2-19 II.E.4.1 Dedicated Hydrogen Penetrations B.2-21 ZX=.E.4.2 Containment Isolation Dependability BE 2-23 ZI.F. 1 ADDITIONAL ACCIDENT-MONXTORING INSTRUMENTATXON (NUREG-0737) B.2-35 XZ.F.1.1 Noble Gas Effluent Monitor B.2-36 ZX.F.1.2 Sampling and Analysis of Plant Effluents BE 2-40 IZ.F.1.3 Containment High-Range Radiation Monitor B.2-44 II.F.1.4 Containment Pressure Monitor B.2-47 XI F.1.5

~ ~ ~ Containment Water Level Monitor B.2-48 II.F.1.6 Containment Hydrogen Monitor B.2-50

WNP-2 AMENDMENT NO. 17 July 1981 TABLE OF CONTENTS (Continued)

Pacae II.F.2 INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING B.2-51 II.K.1.5 Assurance of Proper ESF Functioning B.2-52 II.K. 1. 10 Safety-Related System Operability Status Assurance B.2-54 II.K.1.22 Proper Functioning of Heat Removal Systems B.2-55 II.K.1.23 Reactor Vessel Level Instrumentation B.2-58 II.K.3.3 Failure of PORV or Safety Valve to Close B.2-.60 II.K.3.13 Separation of HPCI and RCIC System Initiation Levels B~

Modify Break-Detection Logic to Prevent 2-61'I.K.3.15 Spurious Isolation of High-Pressure Coolant Injection and Reactor Core Isolation Cooling B.2-63 II.K.3. 16 Reduction of Challenges and Failures of Relief Valves Feasibility Study and System Modification B.2-64 II.K.3.17 Report on Outages of Emergency Core Cooling Systems Licensee Report and Proposed Technical Specification Changes B.2-67 II.K.3.18 Modification of Automatic Depressurization System Logic Feasibility for Increased Diversity for Some Event Sequences B.2-69 II.K.3.21 Restart of Core Spray and Low Pressure Coolant Injection Systems B.2-70 II.K'.3.22 Automatic Switchover of Reactor Core Isolation Cooling System Suction Verify Procedures and Modify Design B 2-71 II.K.3.24 Confirm Adequacy of Space Cooling for High-Pressure Coolant Injection and Reactor Core Isolation Cooling Systems B.2-72

WNP-2 AMENDMENT NO. 29 March 1983 TABLE OF CONTENTS (Continued)

II.K. 3. 25 Fffect of Loss of Alternating-Current Power on Pump Seals B~ 2-73 IX. K. 3. 27 Provide Common Reference Level for Vessel Level Instrumentation B. 2-74 II K. 3. 28

~ Ver i fy Qual i fication of Accumulator s on Automatic Depressurization System Valves B. 2-75 l

II.K. 3. 30 Revised Small-Break Loss-of-Coolant Accident Methods to Show Compliance with 10 CFR Part 50, Appendix K B. 2-76 II.K. 3. 31 Plant-Specific Calculations to Show Compliane with 10 CFR Part 50.46 B. 2-78 II.K. 3. 44 Adequate Core Cooling for Transients with a Single Failure B. 2-79 II. K. 3. 45 Evaluation of Depressur ization with Other Than Automatic Depressur ization System B. 2-90 XI. K. 3. 46 Response to List of Concerns from ACRS Consultant (Michelson Concerns) B. 2-103 IIX.A.l.l Upgrade Emergency Preparedness B. 3-1 XIX.A.1.2 Upgr ade Emergency Suppor t Faci 1 i ties B~ 3-2 XII.D.1.1 Integr i ty of Systems Outside Contain-ment Likely to Contain Radioactive Material for Pressurized Water Reac-tors and Boiling Water Reactors B. 3-5 III.D. 3. 3 Improved Inplant Iodine Instrumenta-tion Under Accident Conditions 1

B~ 3-8 IX. D. 3. 4 Control Room -Habitability Requirements B~ 3-10 B-iv

WNP-2 AMENDMENT NO ~ 17 July .1981 TABZE OF CONTENTS (Continued)

~acae II.K.3.25 Effect of Loss of Alternating-Current

~wee on Pump Seals BE 2-73

/o~cr-II.K.3. 27 Provide Common Re ference Level for Vessel Level Instrumentation B.2-74 XX.K.3.28 Verify Qualification of Accumulators on Automatic Depressurization System Valves B.2-75 XI.K.3 ~ 30 Revised Small-Break Zoss-of-Coolant Accident Methods to Show Compliance

'ith 10 CFR Part 50, Appendix K B.2-76 Plant>>Specific Calculations to Show Compliance with 10 CFR Part 50.46 BE 2-78 II.K.3.44 Adequate Core Cooling for Transients with a Single Failure B.2-79 IX.K 3.45 Evaluation of Depressurization with Other Than Automatic Depressurization System BE 2-90 II.K.3.46 Response to Zist of Concerns from ACRS Co'nsultant (Michelson Concerns) B.2-103 XII.A.1.1 Upgrade Emergency Preparedness .B 3-1 IXI.A.1. 2 Upgrade Emergency Support Facilities B 3-2 IIXiD.1 1 Xntegrity of Systems Outside Containment Likely to Contain Radioactive Material for Pressurized Water Reactors and Boiling Water Reactors B.3-5 IXI.D.3.3 Improved Inplant Iodine Instrumentation Under Accident Conditions BE 2>>8 III.DE 3.4 Control Room Habitability Requirements BE 2-10 B-iv

AMENDMENT NO. 17 July 1981 Z.A.1.1 Shift Technical Advisor Position (NUREG-0737)

Each licensee shall provide an on-shift technical advisor to the shift supervisor. The shift technical advisor (STA) may s'erve more than one unit at a multi-unit site if qualified to perform the advisor function for the various units'he STA shall have .a bachelor's degree or equivalent in a scientific or engineering discipline and have received speci-fic training in the response and analysis of the plant for transients and accidents. The STA shall also receive training in plant design and layout, including the capabilities of instrumentation and controls in the control room. The licen-see shall assign normal duties to the STAs that pertain to the engineering aspects of assuring safe operations of the plant<

including the review and'valuation of operating experience.

Clarification The letter of October 30, 1979 clarified the short-term STA requirements. That letter indicated that the STAs must have completed all training by January 1, 1981. This paper con-firms these requirements and requests additional information.

The need for the STA position may be eliminated when the qualifications of the shift supervisors and senior operators have been upgraded and the man-machine interface in the control room has been acceptably upgraded. However, until those long-term improvements are attained, the need for an STA program will continue.

The staff has not yet established the detailed elements of the academic and training requirements of the STA beyond the guidance given in its October 30, 1979 letter. Nor has the staff made a decision on the level of upgrading required for licensed operating personnel and the man-machine interface in the control room that would be acceptable for eliminating the need of an STA. Until these requirements for eliminating the STA position have been established, the staff continues to require that, in addition to the staffing requirements speci-fied in its July 31, 1980 letter (as revised by item I.A.1.3 of this enclosure), an STA be available for duty on each operating shift when a plant is being operated in Modes 1-4 for a PWR and Modes 1-3 for a BWR. At other times, an STA is not required to be on duty.

Since the October 30, 1979 letter was issued, several efforts have been made to establish, for the longer term, the minimum

WNP-2 AMENDMENT NO 17 July 1981 level of experience, education, and training for STAs. These efforts include work on the revision to AN8-3.1, work by the Institute of Nuclear Power Operations (XNPO), and efforts.

internal'taff INPO recently made available a document entitled "Nuclear Power Plant Shift Technical Advisor Recommendations for Position Description,= Qualifications, Education and Training."

A copy of'evision 0 of" this document, dated April 30, 1980, is attached as Appendix C. Sections 5 and 6 of the XNPO docu-require-ment describe the education, training, and experience ments for STAs. The NRC staff finds that the descriptions as set forth in Sections 5 and 6 of Revision 0 to the XNPO docu-ment are an acceptable approach for the selection and training of personnel to staff the STA positions. (Note: This should not be interpreted to mean that this is an NRC requirement at this time. The intent is to refer to the INPO document as acceptable for interim guidance for a utility in planning its STA program over the long term (i.e., beyond the January 1, 1981 requirement to have STAs in place in accordance with the qualification requirements specified in the staff's October 30, 1979 letter.)

No later than January 1, 1981, all licensees of operating reactors shall provide this office with a description of their STA training program and their plans for requalification training. This description shall indicate the level of training attained by STAs by January 1, 1981 and demonstrate conformance with the qualification and training requirements in the October 30, 1979 letter. Applicants for operating licenses shall provide the same information in their -applica-tion, or amendments thereto, on a schedule consistent with the NRC licensing review schedule.

No later than January 1, 1981, all licensees of operating reactors shall provide this office with a description of their long-term STA program, including qualification, selection cri-teria, training plans, and plans, if any, for the eventual phase-out of the STA program. (Note: The description shall include a -comparison of the licensee/applicant program with the above mentioned INPO document. This request solicits industry views to assist NRC in .establishing long-term improv-ments in the STA program. Applicants for operating licenses shall provide the same information in their application, or amendments thereto, on a schedule consistent with the NRC licensing review schedule.)

B. 1-2

AMENDMENT NO 17 July 1981 WNP-2 Position For the short term, WNP-2 will provide qualified and trained engineers who will be assigned on shift (as required) to per-form the STA function. The engineers will be from the Plant Technical staff and will meet the intent of the qualifica-tions, educati'on, experience, and training requirements pre-sented in the INPO documents, dated April-18, 1980, titled "Nuclear Power Plant Shift Technical Advisor."

\

For the long term, WNP-2 plans to upgrade the qualifications and training of the shift supervisors and senior reactor operators and upgrade the man-machine interface .in the control room, thereby eliminating the requirement for providing the STA function by Technical staff engineers.

A detailed description of the training program will be pro-vided by April 1982.

B. 1-3

WNP-2 AMENDMENT NO ~ 17 July 1981 I.A.1.2

~ Shift Supervisor Responsibilities

~

Position- (NUREG-0578, 2.2.1.A)

~

a. The-highest level of corporate management of each licensee shall issue and periodically reissue a management directive that emphasizes the primary management responsibility of the shift supervisor for safe operation of the plant under all conditions .on his shift and that clearly establishes his command duties.

b. Plant procedures shall be reviewed to assure that the duties, responsibilities, and authority of the shift supervisor and control room operators are properly defined to effect the establishment of a definite line of command and clear delin-.

eation of the command decision authority of the shift supervisor in 'the control room relative to other plant management personnel. Particular emphasis shall be placed on the following:

1. The responsibility arid authority of the shift supervisor shall be to maintain the broadest perspective of operational conditions affecting the safety of the plant as a matter of highest priority at.all times when on duty in the control room. The idea shall be rein-forced that the shift supervisor should not become totally involved in any single opera>>

. tion in times of emergency when multiple operations are required in the control room.

2. The shift supervisor, until properly relieved, shall remain in the control room at all times during accident situations to direct the activities of control room opera-tors. Persons authorized to relieve the shift- supervisor shall be specified.

3. If the shift supervisor is temporarily absent from the control room during routine opera-tions, a lead control room operator shall be designated to assume the control room command function. These temporary duties, respon-sibilities, and authority shall be clearly specified.

c. Training programs for shift supervisors shall emphasize and reinforce the responsibility for

WNP-2 "

AMENDMENT NO. 17 July 1981 safe operation and the management function of the shift supervisor is to provide for assuring safety.

d. The administrative duties of the shift supervisor shall be xeviewed by the senior officer of each utility responsible for plant opex'ations.

Administrative functions that detract from or are subordinate to the management responsibility for assuring the safe, operation of the plant shall be delegated to other operations personnel not on duty in the control room.

Clarification The table attached provides clarification to the above posi-tion.

WNP-2.Position The administrative duties of the shift supervisox'ill be reviewed. Xnappropriate functions will be delegated to other personnel to meet the intent of this position. Appropriate documentation will be available on site for review by NRC X&E by July 1982.

B~ 1-5

AMENDMENT NO ~ 17:

July 1981 TABLE I.A 1.2-1 SHIFT SUPERVISOR RESPONSIBILITY (2 '.1.A)

NUREG-0578 POSITION (POSITION NO.) CLARIFICATION, Highest Level of Corporate V. P. For Operations Management (1 '

Periodically Reissue (1. ) Annual Reinforcement of Company Policy Management Direction (1.) Formal Documentation of Shift Personnel, All Plant Management; Copy to IE Region Properly Defined (2.0) Defined in Nriting in a Plant Procedure Until Properly Relieved (2.B)

~ ~

~ Formal Transfer of Authority, Valid SRO License, Recorded in Plant Log Temporarily Absent (2.C) Any Absence Control Room Defined (2.C) Includes Shift Supervisor Office Adjacent to the Control Room Designated (2.C) In Administrative Procedures Clearly Specified Defined in .

Administrative Procedures SRO Training Specified in ANS 3.1 (Draft) Section 5.2.1 8~

Administrative Duties (4.) Not Affecting Plant Safety

WNP-2 AMENDMENT NQ 17 July 1981 TABZ E X.A. 1. 2-1 (Continued)

,NUREG-0578 POSITION (POSXTION NO ) CLARIFICATION Administrative Duties Reviewed (4.) On Same Xnterval as Reinforcement: i.e.,

Annual by U. P. for Operations This requirement shall be met before fuel loading. See NUREG-0578, Section 22.la, Xtem 4 and NRC letters of September 27, and November 9, 1979.

B~ 1-7

WNP-2 AMENDMENT NO. 17 July 1981 I.A.1.3 Shift Manning Position This position defines shift manning requirements for normal operation. The letter of July 31, 1980 from D. G. Eisenhut to all power reactor licensees and applicants sets forth the interim criteria for shift staffing (to be effective pending general criteria that will be the subject of future rulemaking). Overtime restrictions were also included in the July 31, 1980 letter.

Clarification Page 3 of'he July 31, 1980 letter is superseded in its entirety by the following:

Licensees of operating plants and applicants for operating licenses 'shall include in their administrative procedures (required by license conditions) provisions governing required shift staffing and movement of key individuals about the plant. These provisions are required to assure that qualified plant personnel to man the operational shifts are readily available in the event of an abnormal or emergency situation.

These administrative procedures shall also set forth a policy, the objective of which is to operate the plant with the required staff and develop working schedules such that use of overtime is, avoided, to the extent practicable, for the plant staff who perform safety-related functions (e.g., senior reac-tor operators, reactor operators, health physicistsi auxiliary operators, I&C technicians and key maintenance personnel).

IE Circular No. 80-02, "Nuclear Power Plant Staff cwork Hours,"

dated February 1, 1980 discusses the concern of overtime work for members of the plant staff who perform safety-related functions (see WNP-2 position).

The staff recognizes that there are 'diverse opinions on the amount of overtime that would be considered permissible and that there is a lack of hard data on the effects of overtime beyond the generally recognized normal 8-hour working day, the effects of shift rotation, and other factors. The NRC has initiated studies in this area. Until a firmer basis is deve-loped on working hours, the administrative procedures shall include as an interim measure the following guidance< which generally follows that of IE Circular No. 80-02.

WNP-2 AMENDMENT NO. 17 July 1981 In the event that overtime must be used (excluding extended periods of shutdown for refueling, major maintenance or major plant modifications), the following overtime restrictions should be followed:

a. An individual should not be permitted to work more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> straight (not including shift turnover time).

b. There should be a break of at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (which can include shift t'urnover time) between all work periods.

c. An individual should not work more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7-day period.

d. An individual should not be required to work more than 14 consecutive days without having 2 con-secutive days off.

However, recognizing that circumstances may arise requiring deviation from the above restrictions, such deviation shall be authorized by the plant manager or his deputy, or higher levels of management in accordance'ith published procedures and with appropriate documentation of the cause.

If a reactor operator or senior reactor operator has been working more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during periods of extended shutdown (e.g., at duties away from the control board), such indivi-duals shall not be assigned shift duty in the control room without at least a 12-hour break preceding such an assignment.

I The NRC encourages the development of a staffing policy permit the licensed reactor operators and senior reactor that'ould operators to be periodically assigned to other duties away from the control board during their normal tours of duty.

If a reactor operator is required to work in excess of 8 con-tinuous hours, he shall be periodicaLly relieved of primary duties at the control board, such that periods of duty at the board do not exceed about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at a time.

The guidelines on overtime do not apply to the shift technical advisor provided he or she is provided sleeping accommodations and a 10-minute availability is assured.

Operating license applicants shall complete these administra-tive procedures before fuel loading. Development and imple-mentation of the administrative procedures at operating plants will be reviewed by the Office of Inspection and Enforcement, beginning 90 days after July 31, 1980.

I WNP-2 AMENDMENT NO ~ 17 July 'l981 "

See Section III.A.1.2 for minimum staffing and augment capa-bilities for emergencies.

WNP-2 Position Minimum Shift Crew:

Minimum shift manning for WNP-2 will consist, of the following:

a. One Shift Manager (shift supervisor) with a senior reactor operator's license (SRO) on site at all times when the reactor contains fuel.

b. A Control Room Supervisor with a senior reactor operator's license (SRO) in the control room at all times when the reactor is in power operation, startup, or hot shutdown (conditions 1, 2, and 3). This Control Room Supervisor may, from. time to time, be relieved by the Shift Manager (item a, above) or by any other licensed senior reactor operator.

c. A licensed reactor operator (RO) in the control room at all. times when the reactor contains fuel.

d. One additional licensed reactor operator (RO) shall be on site at all times when the reactor is in power operation, startup, or hot shutdown (conditions 1, 2 and 3). This individual may serve as relief operator for the control room, when the reactor zs operating.

e. Two equipment (non-licensed) operators shall be on site at all times when the reactor is in power operation, startup, or hot shutdown. At least one equipment (non-licensed) operator shall be on site at all times when the reactor contains fuel.

f.. During core alterations, an additional licensed senior reactor operator (SRO) or limited senior reactor operator (SROL) to directly supervise the core alterations. The SRO or SROL may have fuel handling duties but shall not have other con-c urrent operational duties.

Staffing Plan:

Shift coverage is provided by utilizing a rotating shift schedule depending on operating needs and bargaining unit contractual requirements. The schedules are based on a 40-

AMENDMENT NO ~ 17 July 1981 hour0.0229 days <br />0.55 hours <br />0.00328 weeks <br />7.537705e-4 months <br /> work week and shifts are normally 8-hour duration (excluding shift turnover time) .

To assure that sufficient SRO and RO licensed individuals are available as required for plant operation, we are preparing, for cold license exams, an adequate number of SRO and RO shift personnel to support a six crew rotation plus additional man-agement and training SRO candidates. We anticipate a high suc-cess rate and therefore expect no problem 'in maintaining a sufficient number of licensed individuals to meet the minimum manning requirements.

Overtime and Work Hours:

It shall be WNP-2 policy to maintain an adequate number of personnel in the Shift Manager, Control Room Supervisor< Shift Technical Advisor (if required), Control Room Operator, and Equipment Operator positions such that the use of overtime is not routinely required to compensate for inadequite staffing.

Administrative procedures, prepared by July 1982, will docu-ment our policy concerning the use of 'overtime work.

The administrative procedures will also stipulate that work schedules for .the Shift Manager, Control Room Supervisor, Shift Technical Advisor and Equipment Operators (if required), Control Room Operators, shall be established in advance to ensure that the potential for exceeding the following guide-lines is minimized when filling the minimum shift manning requirements previously defined; that is:

a. No individual should work more than 12 con-secutive hours. This does not include time necessary for shift turnover;

b. No individual should work more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period.

c. No individual should work more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7-day period.

d. No individual should work more than 14 con-secutive days without having two consecutive days off.

It should be noted that vacancies due to resignation, promo-tion, unexpected illness, time off for personal business, or other uncontrollable factors may create situations requiring extended overtime outside these guidelines. Such deviations shall be corrected as soon as possible. Those instances resulting in deviations will be documented and reviewed by the

WNP-2 AMENDMENT NO ~ 17 July 1981 Plant Manager or his designee as soon as practicable following the occurrence.

B.1-12

0 "WNP-2 AMENDMENT NO ~ ,17 July 1981 I.A.2.1 Immediate Upgrading of Operator and Senior

. Operator Training and Qualification Position (NUREG-0737) l Ef fective December 1, 1980, an applicant for a senior reactor operator (SRO) license will be required to have been 'a

.licensed operator for one year.

8 Applicants for SRO either come through the operations chain (C operator to B operator to A operator, etc.) or are degree-holding staff engineers who ob'tain licenses for backup purposes.

In the past, many individuals who came through the operator ranks were administered SRO examinations without first being an operator.

of March 28, 1980 requires reactor operator experience letter This was clearly a poor practice and the for SRO applicants.

However< the NRC does.not wish to discourage 'staff engineers from becoming licensed SRO's., This effort is encouraged because it forces engineers to broaden their kriowledge about the plant and its operation.

In addition, in order to attract degree-holding engineers to' consider the shift super'visor's job as part of their career development, the NRC should provide an alternate. path to holding an operator's license for one year.

The track followed by a high school graduate (a non-degreed to become an SRO would 'be four years as a control 'ndividual) room'operator, at least one of which would be as .a licensed operator, and participation in an SRO training program that includes three months on shift as an extra person.

The track followed by a degree-holding engineer would be, at a minimum, two years of responsible nuclear power plant experience as a .staff engineer, participation. in an SRO training program equivalent .to a cold applicant training program, and three months on shift as an extra. person in training'or an SRO position.

direct the licensed activities of licensed operators will Holding these positions -assures that individuals who have. had the necessary combination of education,. training, a'nd actual operating experience prior to assuming a supervisory role at the facility.

WNP-2 AMENDMENT NO ~ 17 July 1981 The staff realizes that the necessary knowledge and experience can be gained in a

~

variety of ways. Consequently, credit for

~

equivalent experience should be given to applicants for SRO

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licenses.

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Applicants for SRO licenses at a facility may obtain their one year operating experience in a licensed capacity (operator or senior operator) at another nuclear power plant. In addition, actual operating experience in a position that is equivalent to a licensed operator or senior operator at military pro-pulsion reactors will be acceptable on a one-for-one basis.

Individual applicants must .document this experience in their individual applications in sufficient detail so that the staff can make a finding regarding equivalency.

Applicants for SRO licenses who possess a degree in engi-neering or applicable sciences are deemed to meet the-above requirement, provided they meet the requirements set forth in Sections A.1.a and A.2 in enclosure 1 in the letter from H. R.

Denton and all power reactor applicants and licensees, dated March 28, 1980, and have participated in a training program equivalent to that of a cold senior operator applicant.

The NRC has not imposed that one year. experience requirement on cold applicants for SRO licenses. Cold applicants are to work on a facility not yet in operation; their training programs are designed to supply the equivalent of the experience not available to them.

Clarification None WNP-2 Position The intent of the above position will be implemented for WNP-2 when operator license submittals are made. Xf i.t should become necessary or desirable to deviate from the experience levels identified as prerequisite for SRO licensing, this deviation shall be identified and justified as a part of the individual's license application.

B~ 1-14

WNP-2 AMENDMENT NO. 17 July 1981 I.A.2.3 Administration of Training Programs for Licensed-Operators Position Pending accreditation of training institutions, licensees and applicants for operating licenses will assure that training center and facility instructors who teach systems, integrated responses, transient, and simulator courses demonstrate SRO qualifications and be enrolled in. appropriate requalification programs.

Clarification The above position is a short-term position. In the future, accreditation of training institutions will include review of the procedure for certification of instructors. The cer-tification of instructors may, or may not, include successful completion of a senior operator examination.

The purpose of the examination is to provide the NRC with reasonable assurance during the interim period that instruc-tors are technically competent.

The requirement is directed to permanent members of the training staff that teach the subjects enumerated above, including members of other organizations who routinely conduct training at the facility. There is no intention to require guest lecturers who are experts in particular subjects (reactor theory, instrumentation, thermo-dynamics, health phy-sics, chemistry, etc.) to successfully complete a senior operator examination. Nor do we intend to require a system expert, such as the Instrument and Control Supervisor teaching the rod control drive system to sit for a senior operator exa-mination. The- use of guest lecturers should be limited.

WNP-2 Position Applications for SRO examinations of Training Engineers who teach license candidates and/or licensed Operators will be submitted prior to July 1982. These instructors will have p'articipated in the Cold License Training Programs and will continue to participate in appropriate retraining or requali-fication programs as either instructor or student.

The requirement is directed to permanent members of the Training Department that teach the subjects enumerated above, including members of other organizations who routinely conduct extensive training at the facility. WNP-2 does not intend to require guest lecturers who are experts in particular subjects

WNP-2 AMENDMENT NO. 17 July 1981 (reactor theory, electrical theory, instrumentation, thermo-dynamics, health physics, chemistry, etc.) to successfully complete a Senior Operator Exam. Nor does WNP-2 intend to require a system or component expert, such as the Instrument and Electrical Supervisor teaching the rod drive control system, to sit for a Senior Operator Examination. Use of guest lecturers will be limited so that program continuity can be maintained.

B~ 1-16

WNP-2 AMENDMENT NO ~ 17 July 1981 I.A.3.1 Revise Scope and Criteria for Licensing Examinations Simulator Exams Position Simulator examinations will be included as part of the licensing examinations.

Changes to Previous Requirements and Guidance: The admi-nistration of simulator examinations will be deferred for applicants whose facilities do not have simulators on site as of October 1, 'l 980. These deferred simulator ex'aminations will be initiated by October 1,,1981.

Clarification The clarification does not alter the staff's position regarding simulator examinations.

The clarification does provide additional preparation time for utility companies and the NRC. to meet examination requirements as stated. A study is under way to consider how similar a nonidentical simulator should be for a valid examination. In addition, present simulators are fully booked months in advance.

.Application of this requirement was stated on June 1, 1980 to applicants where a simulator is located at the facility'.

Starting October 1, 1981 simulator examinations will be con-ducted for appli.cants of facilities that do not have simula-tors at the site.

NRC simulator examinations normally require 2 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

Normally, two applicants are examined during this time period by two examiners.

Utility companies should make the necessary arrangements with an appropriate simulator training center to provide time for these examinations. Preferably these examinations should be scheduled consecutively with the balance of the examination.

However, they may be scheduled no sooner than 2 weeks prior to and not later than 2 weeks after the balance of the examination.

WNP-2 Position The new subject matter and grading criteria have been imple-mented in the Cold License Simulator Training classes being conducted for WNP-'2 by the General Physics Corporation.

B 1-17

WNP-2 AMENDMENT NO. 17 July .1981 The'equirement for applicants for operator licenses to grant permission for the NRC to inform facility management regarding results of the examinations will be implemented just prior to the administration of these exams.

The simulator for WNP-2 may, not be available to support simu-lator examinations for the initial operating staff. In this event, WNP-2 has arranged to conduct these examinations at another simulator facility. Scheduling with the'RC for the simulator examinations will be made as part of the normal scheduling of license. examinations.

B.1-18

WNP-2 AMENDMENT NO. 17 July 1981 I.B.1.2 Independent Safety Engineering Group Position Each applicant for an operating license shall establish an onsite independent safety engineering group (ISEG) to perform independent reviews of plant operations.

The principal function of the ISEG is to examine plant operating characteristics, NRC issuances, Licensing Information Service advisories, and other appropriate sources of plant design and operating experience information that may indicate areas for improving plant safety. The ISEG is to perform independent review and audits of plant activities, including maintenance, modifications, operational problems, and operational analysis, and aid in the establishment of programmatic requirements for plant activities. Where useful improvements can be achieved, it is expected that this group will develop and present detailed recommendations to corporate management for such things as revised procedures or equipment modifications.

Another function of the ISEG is to maintain surveillance of plant operations and maintenance activities to provide inde-pendent verification that these activities are performed correctly and that human errors are reduced as far as prac-ticable. ISEG will then be in a position to advise utility management on the overall quality and safety of operations.

ISEG need not perform detailed audits of plant operations and shall not be responsible for signoff functions such that it becomes involved in the operating organization.

Clarification The new ISEG shall not replace the plant operations review committee (PORC) and the utility's independent review and audit group as specified by current staff guidelines (Standard Review Plan, Regulatory Guide 1.33, Standard Technical Specifications). Rather, it is an additional inde-pendent group of a minimum of five dedicated, full-time engi-neers, located onsite, but reporting offsite to a corporate official who holds a highlevel, technically-oriented position that is not in the management chain for power production. The ISEG will increase the available technical expertise located onsite and will provide continuing, systematic, and indepen-dent assessment of plant activities. Integrating the shift technical advisors (STAs) into the ISEG in some way would be desirable in that it could enhance the group's contact with the knowledge of day-to-day plant operations to provide addi-tional expertise. However, the STA on shift is necessarily a member of the operating staff and cannot be independent of it.

WNP-2 AMENDMENT NO. 30 June 1983 It is expected that the ISEG may interface with the quality assurance (QA) organization, but perferably should not be an integral part of the QA organization.

The functions of the ISEG require daily contact with the operating personnel and continued access to plant facilities and records'he ISEG review functions can, therefore, best be carried out by a group physically located onsite.

However, for utilities with mulitple sites, it may be possi--

ble to perform portions of the independent safety assessment function in a centralized location for all the utility's plants. In such cases, an onsite group still is required, but it may be slightly smaller than would be the case were performing the entire independent safety assessment if it function. Such cases will be reviewed on a case-by-case basis't this time, the requirement for establishing an ISEG is being applied only to applicants for operating licenses in accordance with Acton Plan Item I.B.1.2. The staff intends to review this activity in about a year to determine its effectiveness and to see whether changes are required.

Applicability to operating plants will be considered in implementing long-term improvements in organization and management for operating plants (Action Plan Item I.B.1 1) ~ ~

WNP-2 Position The Supply System has established a Nuclear Safety Assurance group for WNP-2 within the Licensing and Assurance Director-ate as shown in Figure I.B.1.2-1. The onsite Nuclear Safety Assurance group (comprised of a minimum of one supervisor and two engineers) is supplemented by offsite technical expertise from within the Licensing and Assurance Directorate as required with a minimum of two qualified engineers available to support the WNP-2 assurance group. The WNP-2 Nuclear Safety Assurance group is independent of the line management responsible for power production and chartered with ensuring and improving operational nuclear safety of the WNP-2 plant.

The functions of the WNP-2 Nuclear Safety Assurance group include the followingc Evaluation of procedures important to safe operation of WNP-2 for technical adequacy and clarity.

BE 1-20

WNP-2 AMENDMENT NO. 30 June 1983

b. Evaluation of plant operations from a safety perspective.

C ~ Evaluation of the operating experience of WNP-2 to provide recommendations on safety-related concerns. Xn this regard operating experience of other plants of similar design is assessed for applicability to WNP-2.

d ~ Overall assessment of WNP-2 plant performance regarding conformance to safety requirements.

e. Other matters relating to safe operation of WNP-2 that independent review deems appropriate for consideration.

The qualification and training requirements for the Nuclear Safety Assurance Manager are comparable to that described in Section 4.2 of ANS 3.1, Draft Revision, dated March 13, 1981 'ther qualifications and training requirements meet ANS 3.1, Draft R vision, dated March 13, 1981, Section 4.2 or 4.4 or equivalent as described in Section 4.1 ~

BE 1-21

WNP-7

,WNP-2 Posit ion The Supply System plans to provide an on-site Safety Engineer-ing Group for WNP-2 consisting of four. dedicated full-time engineers. This group will be part of the Nuclear Safety Organization shown in Figure I.B.1.2-2 that repor ts directly to the Nanag ing Director. The Nuclear Safety Org'anizat ion is chartered with ensuring and improving overaLL nuclear safety of Supply System nuclear facilities and has no direct responsibility for day"to-day power production. The on-site safety engineers will be supported by off-site independent engineering and safety expertise as required to accomplish their functions. The WNP 2 Safety Engineering Group wilL be established and staffed sufficiently in advance of fuel Load-ing to allow orientation of the staff and review of plant operating procedures prior to their use.

The general functions of the WNP-2 Safety Engineering Group will incLude the following:

a ~ Evaluation of procedures important to safe operation of WNP-2 for technical adequacy and clarity.

be Evaluation of plant operations from a safety perspective.

c. Evaluation of the operating experience of WNP-2 to

.provide recommendations on safety-related concerns. In this regard operating experience of other plants of similar design wiLL be assessed for applicability to WNP-2.

d. Overall assessment o f WNP-? per f ormance at WNP-2 regarding conformance to safet'y requirements.

e. Other matters relating to safe operation of WNP-2 that independent review seems appropriate for considerations

f. Assessment of plant safety programs..

The qualification 'and training of the SEG Manager wiLL be comparable to that described in Section 4.2 of ANS 3..1r draft revisions dated Narch 13.. 1981. Other SEG member

'qualifications and training will meet ANS 3.1r draft revisions dated Narch 13m 1981m Section 4.2r 4.4 or 4-5r or equivalents as described in Section 4.1 ~

WNP-2.

Considering the number of Supply System nuclear plants now under constructions the Supply System considers that the safety rev,.iew function can be best served by having highly qualified experts in disciplines (which would n'o't be ful ly utilized at one. site) availa'ble in the, independent Nuclear Safety Organization and other off-site centralized technical or ganizations to support, all sites on an ad hoc basis. This would be particularly true for: personnel'n f ields where there are few qualified people available. The technical assets of the company- wi(L .be availabLe as needed and when necessary specialists. wiLL be assigned to the Director of Nucl'ear Safety to support 'plant safety evaluations.

The concept and specific role of a dedicated on-site safety, group is expected to evolve as'xperience is obtained with this approach. 'he Supply System will continue to evaluate this approach and it is expected that 'we may need to modify the funct'ionsi role or organizational approach of thisgroup consistent with effeative utilization of resources and improving overall safety and efficiency of our plants.

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WNP-2 AMENDMENT NO. 17 July 1981 I.C.1 Guidance for the Evaluation and Development of Pro-cedures for Transients and Accidents Position (NUREG-0737)

In the letters of September 13 and 27, October 10 and 30, and November 9, 1979, the Office of Nuclear Reactor Regulation required licensees of operating plants, applicants for operating licenses and licensees of plants under construction to perform analyses of transients and accidents, prepare emergency procedure guidelines, upgrade emergency procedures, including procedures for operating with natural circulation conditions, and to conduct operator retraining (see also item I.A.2.1). Emergency procedures are required to be consistent with the actions necessary to cope with the transients and accidents analyzed. Analyses of transients and accidents were to be completed in early 1980 and implementation of procedures and retraining were to be completed 3 months after emergency procedure guidelines were established; however, some dif-ficulty in completing these requirements has been experienced.

Clarification of the scope of the task and appropriate sche-dule revisions are being developed. In the 'course of review of these matters on Babcock and Wilcox (B&W) designed plants, the staff will follow up on the bulletin and orders matters relating to'nalysis methods and results, as listed in NUREG-0660, Appendix C (see Table C.1, items 3, 4, 16, 18, 24/

25, 26, 27; Table C. 2, items 4, 12, 17, 18, 19, 20; and Table C.3, items 6, 35, 37, 38, 39, 41, 47, 55, 57).

Changes to Previous Requirements and Guidance:

a. Modification to Clarification

1. Addresses owners'roup and vendor submittals.

2. References to task action plan items I.C.8 and I.C.9.

3. Scope of procedures review is explained.

4. Establishes configuration control of guide-lines for emergency procedures.

b. Modification to Implementation

1. Deleted reference to NUREG-0578, Recommendation 2.1.9 for item I.C.1(a)2, in-adequate core cooling.

B. 1-23

WNP-2 AMENDMENT NO ~ 30 June 1983 The complete NRC position description and clarification is contained in NUREG-0737 Task I.C.l.

This requirement is to be completed by fuel loads Clarification None WNP-2 Position WNP-2 has participated, and continues to participate, in the BWR Owners'roup program to develop Emergency Procedure Guidelines for General Electric Boiling Water Reactors Fol-lowing are a brief description of the submittals to date, and a justification of their adequacy to support guidelines development.

a. Description of Submittals 1 ~ NEDO-24708, "Additional Information NRC Staff Generic Report on Boiling Required'or Water Reactors," August 1979> including additional sections submitted in prepublica-tion form since August 1979.

(a) Section 3.1.1 (Small Break LOCA) ~

Description and analysis of small break loss-of-coolant events, considering a range of break sizes, location, and con-ditions, including equipment failures and operator errorsg description and justification of analysis methods'b)

Section 3.2.1 (Loss of Feedwater) revised and resubmitted in prepublica-tion from March 31, 1980 and analysis of loss of feed

'escription water events, including cases involving stuck-open relief valves, and including equipment failures and operator errors>

description and justification of analysis methods.

(c) Section 3.2.2 (Other Operational Tran-sients) submitted in prepublication form March 31, 1980'evised and resub-mitted in prepublication form August 22, 1980.

B.l-24

WNP-2 AMENDMENT NO. 17 July 1981 Description and analysis of each FSAR Chapter 15 event resulting in a reactor system transient; demonstration of appli-cability of analyses of PSAR 3.1.1, 3.2.1, and 3.5.2.1 to each event; demonstration of applicability of Emergency Procedure Guidelines to each event.

(d) Section 3.3 (BWR Natural and Forced Circulation).

Description of natural and forced cir-culation cooling; factors influencing natural circulation, including noncondensibles; re-establishment of forced circulation under transient and accident conditions.

(e) Section 3.5.2.1 (Analyses to Demonstrate Adequate Core Cooling) - submitted in prepublication form November 30, 1979; revised and resubmitted in prepublication form September 16, 1980.

S Description and analysis of loss-of-coolant events, loss of feedwater events, and stuck-open relief valve events, including severe multiple equipment, failures and operator errors which, if not mitigated, could result in conditions of inadequate core cooling.

(f) Section 3.5.2.3 (Diverse Methods of Detecting Adequate Core Cooling) - sub-mitted in prepublication form December 28, 1979.

Description of indications available to the BWR operator for the detection of adequate core cooling (detailed instru-ment responses are described in PSAR 3.1.1, 3.2.1, and 3.5.2.1).

(g) Section 3.5.2.4 (Justification of Analysis Methods) - submitted in pre-publication form September 16, 1980.

Description and justification of analysis methods for extremely degraded cases treated in FSAR 3.5.2.1.

B.1-25

WNP AMENDMENT NO ~ 17 July 1981

2. BWR Emergency'rocedure Guidelines (Revision

1) submitted on January 31,'981.

Guidel'ines for BWR Emergency Procedures based on identification and response to plant symptoms; including a range of equipment failures and operator er'rors; including severe multiple equipment failures and opera-tor errors which, if not mitigated, would result in conditions of inadequate core cooling; including conditions when core cooling status is uncertain or unknown.

3. NED0-24708A, Revision 1, December 1980.

b. Adequacy of Submittals:

The submittals described in paragraph a have been discussed and reviewed extensively among the BWR Owners'roup,-the General Electric Company, and the NRC staff. The NRC staff has found (NUREG-0737 p. I.C. 1-3 ) that "the analysis and, guidelines submi t ted by General Electr ic Company (GE) Owners'roup...comply with the requirements (of the NUREG-0737 clarification)." Zn Reference 1, the Director of the Division of Licensing states, "we find the Emergency Procedure.

Guidelines, acceptable for trial. implementation (on six plants with applications for operating licenses pending) ."

WNP-2 believes that in view of these findings, no further detailed justification of the analyses or guidelines is necessary at this time.

Reference 1 further, states, "(during the course of implementation we m'ay identify areas that require modification or further analysis and justification." The enclosure to Reference 1 identif ies several such areas. WNP-2 will work with the BWR Owners'roup in responding to such requests.

By our commitment to work with the Owners'roup on such requests, on schedules mutually agreed to by the NRC and the Owners'roup, and by reference to the BWR Owners'roup ana-lyses and guidelines already submitted, our response to the NUREG-0737 requirement "for reanalysis of transients and acci-dents and inadequate 'core cooling and preparation of guide-lines for development of emergency procedures" is complete.

B~ 1-26

WNP-2 AMENDMENT NO.. 17 July 1981 References (1) Letter, D. G. Eisenhut (NRC) to S. T. Rogers (BWR regarding Emergency Procedure Guidelines> Owners'roup),

October 21, 1980.

B.1-27

WNP-2 AMENDMENT NO ~ 17 July 1981

'.CD 2 SHZFT AND RELIEF TURNOVER PROCEDURES Position The licensees shall review and revise as necessary the plant procedure for shift and relief turnover to assure the following:

a. A checklist shall be provided for the oncoming and offgoing control room operators and the oncoming shift supervisor to complete and sign.

The following items, as a minimum, shall be included in the checklist.

1. Assurance that critical plant parameters are within allowable limits (parameters and allowable limits shall be listed on the checklist) .

2. Assurance of the availability and proper alignment of all systems essential to the pr'evention and mitigation of operational transients and accidents by a check of the control console (what to check and criteria for acceptable status shall be included in the checklist) .

3. Zdentification of systems and components that are in a degraded mode of 'operation permitted by the Technical Specifications. For such systems and components, the length of time in, the degraded mode shall be compared with the Technical Specifications action statement (this shall be recorded as a separate entry on the checklist) .,

b. Checklists or logs shall be provided for comple-tion by the offgoing and ongoing auxiliary opera-tors and technicians. Such checklists or logs shall include any equipment under maintenance or test that by themselves could degrade a system critical to the prevention and mitigation of operational transients and accidents or initiate an operational transient (what to check and cri-teria for acceptable status shall'e included on the checklist) .

C ~ A system shall be established to evaluate the effectiveness of the shift and relief turnover B.1-28

WNP-2 AMENDMENT NO. 17 July 1981 procedure (for example, periodic independent verification of system alignments).

Clarification

'one WNP-2 Position The directives and procedures necessary to meet the intent of the position are being prepared and will be implemented prior to July 1982. These procedures will be made available on site for review by NRC X&E.

B.1-29

WNP-2 AMENDMENT NO ~ 17 July 1981 I.CD 3 SHIFT SUPERVISOR RESPONSIBILITY Position (NUREG-0578, 2.2.1.A)

a. The highest level of corporate management of each licensee shall issue and periodically reissue a management directive that emphasizes the primary management responsibility of the shift supervisor for safe operation of the plant under all con-ditions on his shift .and that clearly establishes his command duties.

b. Plant procedures shall be reviewed to assure that the duties, responsibilities, and authority of the shift supervisor and control room operatoxs are properly 'defined to effect the establishment of a definite line of command and clear delin-eation of the command decision authority of the shift supervisor in the control room relative to other plant management personnel. Particular emphasis shall be placed on the following:

1. The responsibility and authority of the shift supervisor shall be to maintain the broadest perspective of operational conditions affecting the safety of. the plant as a matter of highest px'iority at all times when on duty in the control room. The idea shall be reinforced that the shift supervisor should not become totally involved in any single operation in times of emergency when multiple operations are required in the control x'oom.

2. The shift supervisor, until properly relieved shall remain in the control room at all times during accident situations to direct the activities of control room operators.

Persons authox'ized to relieve the shift supex'visor shall be specified.

3. If the shift supervisor is temporarily absent from the control room during routine operations, a lead control room operator shall be designated to assume the control room command function. These temporary duties, responsibilities, and authority shall be clearly specified.

c. ,Training programs for shift supervisors shall emphasize and reinforce the responsibility for B. 1-30

WNP-2 AMENDMENT NO. 30 June 1983 safe operation and the management function the shift supervisor is to provide for assuring safety.

de The administrative duties of the shift supervisor shall be reviewed by the senior officer of each utility responsible for plant operations. Administrative functions that detract from or are subordinate to the the management responsibility for assuring safe operation of the plant shall be dele-gated to other operations personnel not on duty in the control room.

Clarification The attachment provides clarification to the above positions WNP-2 Position The administrative duties of the shift supervisor have been reviewed. Inappropriate functions were delegated to other personnel.

WNP-2 procedures were reviewed to ensure that the shift manager, control room supervisor, and operator funcions are defined adequately to establish the shift manager in the control room as the commanding authority for plant operations relative to other plant management.

This principle is reinforced by a management directive that issued annually from the office of the Director of Generation emphasies that the shift manager's primary responsibility is the safe operation'f the plant under all conditions'he shift manager's administrative duties are reviewed annually by the plant manager to ensure that administrative responsibilities do not interfere with the primary responsi-bility.

Appropriate documentation is available onsite for review by NRC IS E Branch.

B.l-31

WNP-2 AMENDMENT NO. 17 July 1981 TABLE I.C.3-1 SHIFT SUPERVISOR RESPONSIBILITY (2.2.1.A)

NUREG-0578 POSITION (POSITION NO.) CLARIFICATION Highest Level of Corporate V. P. For Operations Management (1.)

Periodically Reissue (1.) Annual Reinforcement of Company Policy Management Direction (1.) Formal Documentation of Shift Personnel, All Plant Management, Copy to IE Region Properly Defined (2.0) Defined in Writing in a Plant Procedure Until Properly Relieved (2.8) Formal Transfer of Authority, Valid SRO License, Recorded in Plant Log Temporarily Absent (2.C) Any Absence Control Room Defined (2.C) Includes Shift Super-visor Office Adjacent to the Control Room Designated (2.C) In Administrative Procedures Clearly Specified Defined in Administra-tive Procedures SRO Training Specified in ANS 3.1 (Draft) Section 5.2.1.8 Administrative Duties (4.) Not Affecting Plant Safety Administrative Duties Reviewed (4.) On Same Interval as Reinforcement: i.e.,

Annual by V. P. for Operations This requirement shall be met prior to July 1982. See NUREG-0578, Section 22.1a, Item 4 and NRC letters of September 27 and November 9, 1979.

B.1-32

WNP-2 AMENDMENT NO. 30 June 1983 I ~ C. 4 CONTROL ROOM ACCESS Position (NUREG-0578 2 '.2.A)

The licensee shall make provisions for limiting access to the control room to those individuals responsible for the direct operation of the nuclear power plant (e.g., operations supervisor, shift supervisor, and control room operators), to technical advisors who,may be requested or required to sup-port the operation, and to predesignated NRC shall include the followinga personnel'rovisions

a. Develop and implement an administrative proce-dure that establishes the authority and respon-sibility of the person andin charge of the control room to limit access, be Develop and implement procedures that establish a clear line of authority and responsibility in the control room in the event of an line of succession for the person in charge emergency'he of the control room shall be established and limited to persons possessing a current senior reactor operator's license. The plan shall clearly define the lines of communication and authority for plant management personnel not in direct command of operations, including those who report to stations outside of the control room.

Clarification None WNP-2 Position A WWP-2 procedure bas been prepared and implemented to estab-lish the shi ft manager (SRO) and, in his absence, the control room supervisor (SRO) as the authority, and responsibility for limiting access to the control room. Nonessential personnel are excluded from the control room when their presence is hampering operations. Nonessential personnel are defined as those not required by the shift manager to assist in safe plant operation and may include anyone not normally assigned a shift control room position. If required, plant security can be utilized to enforce the policy' B. 1-33

WNP-2 AMENDMENT NO. 30 June 1983 Additionally, procedures establish the same line of succes-sion for control room authority and responsibility in the event of an emergency. The procedures specifically address lines of communication and authority for management personnel not in direct command of operations and assigned responsibil-ities outside the control room. Instructions or orders impacting operations are reviewed by the operations manager and transmitted to the shift manager ~

B.1-33a

WNP-2 AMENDMENT NO. 17 July 1981

~ I.C.5 Position PROCEDURES PLANT STAFF FOR FEEDBACK OF OPERATING EXPERIENCE TO In accordance with Task Action Plan I.C.5, Procedures for Feedback of Operating Experience to. Plant Staff (NUREG-0660),

.each applicant for an operating license shall prepare proce-dures to assure that operating information pertinent to plant safety originating both within and outside the utility organi-zation is continually supplied to operators and other person-nel and is incorporated into training and retraining programs.

These procedures shall:

a. Clearly identify organizational responsibilities for review of operating experience, the feedback of pertinent information to operators and other personnel, and the incorporation of such infor-mation into training and retraining programsg

b. Identify the administrative and technical review steps necessary in translating recommendations by the operating experience assessment group into plant actions (e.g., changes to procedures, operating orders);

c. Identify the recipients of various categories of information from operating experience (i'.e.g supervisory personnel, shift technical advisors, operators, maintenance personnel, health physics technician's) or otherwise provide means through which such information can be readily related to the job functions of the recipients;

d. Provide means to assure that affected personnel become aware of and understand information of sufficient importance that should not wait for emphasis through routine training and retraining programs;

e. Assure that plant personnel do not. routinely receive extraneous and, unimportant information on operating experience in such volume that it would obscure priority'nformation or otherwise detract from overall job performance and proficiency;

f. Provide suitable checks to assure that conflicting or contradictory information is not conveyed to operators and other personnel until resolution is reached; and, B.1-34

i WNP-2 AMENDMENT NO. 23 February 1982

g. Provide periodic internal audit to assure that the feedback program functions effectively at all levels.

Clarification Each utility shall carry out an. operating experience assessment function that will involve utility personnel having collective competence in all areas important to plant safety.

In connection with this assessment function, it is important that procedures exist to assure that important information on operating experience originating both within and outside the organization is continually provided to operators and other personnel and that it is incorporated into plant operating procedures and training and retraining programs.

Those involved in the assessment of operating experience will review information from a variety of sources. These include operating information from the licensee's own plant(s), publi-cations such as IE Bulletins, Circulars, and Notices, and per-tinent NRC or industrial assessments of operating experience.

In some cases, information may be of sufficient importance that it must be dealt with promptly (through instructions, changes to operating and emergency procedures, issuance of special changes to operating and emergency procedures, issuance of special precautions, etc.) and must be handled in such a manner to assure that operations management personnel would be directly involved in the process. In many other cases, however, important information will become available which should be brought to the attention of operators and other personnel for their general information to assure con-tinued safe plant operation. Since the total volume of infor-mation handled by the assessment group may be large, it is important that assurance be provided that high-priority mat-ters are dealt with promptly and that discrimination is used in the feedback of other information so that personnel are not deluged with unimportant and extraneous information to the detriment of their overall proficiency. It is important, also, that technical reviews be conducted to preclude prema-ture dissemination of conflicting or contradictory information.

WNP-2 Position The WNP-2 plant will have procedures covering the review and feedback of operating experiences. These procedures will be in effect at the time of fuel load and will be made available for onsite review by NRC I&E.

B.1-35

WNP-2 AMENDMENT NO. 30 June 1983 The following summary presents the essence of the final procedure(s)i a ~ The WNP-2 Nuclear Safety Assurance group initi-ates a review of operating experience material utilizing technical expertise within the Supply System as appropriate. Extraneous and unimport-ant information is sorted out and conflicting opinions resolved.

b. Following this review, the WNP-2 Nuclear Safety Assurance Manager supplies detailed information to the appropriate section manager for dissemin-ation to identified recipients (supervisory personnel, STAs, operators, maintenance person-nel, health physics technicians, etc.).

c. Xt is the responsibility of the appropriate section manager receiving the operating experi-ence information to provide the information to the training section for inclusion in the retraining program and to disseminate the infor-mation directly to the identified recipients a more rapid transmittal is appropriate.

if I

The WNP-2 Nuclear Safety Assurace Manager, with assistance from the appropriate section manager, makes recommendations to the Plant Manager on the need for procedure changes or plant modifi-cations identified from the operating experience review. Those changes deemed appropriate by th' Plant Manager are then processed as described in the Plant Administrativ Procedures.

e. The WNP-2 Training Manager is responsible for incorporating into the training program the information received from the operating experi-ence review program. To pr vent conflicting or contradictory information being conv yed to plant personnel, the Training Manager clears through the WNP-2 Nuclear Safety Assurance Manager any operating expirence information received f rom other sources ~

B.1-35a

WNP-2 AMENDMENT NO. 30 June 1983 The WNP-2 Nuclear Safety Assurance Manager is responsible for periodic surveillance of the entire operating experience review process. and Particular attention is paid to appropriate timely incorporation of procedure change and plant modification recommendations resulting from the review of operating QA periodically audits the operating experiences'perational experiences review program.

B. 1-35b

WNP-2 AMENDMENT NO. 17 July 1981 I.C.6 GUIDANCE ON PROCEDURES FOR VERIFYING CORRECT PERFORMANCE OF OPERATING ACTIVITIES Position It is 'required (from NUREG-0660) that licensees'rocedures be reviewed a'nd revised, as necessary, to assure that an effec-tive system of verifying the correct performance of operating activities is provided as' means of reducing human errors and improving the quality of normal operations. This will reduce the frequency of occurrence of situations that could result in or contribute to accidents. Such a verification system may include automatic system status monitoring, human verification of operations and maintenance activities independent of the people performing the activity (see NUREG-0585, Recommendation 5), or both.

Implementation of automatic status monitoring if required will reduce the extent of human verification of operations and maintenance activities but will not eliminate the need for such verification in all instances. The procedures adopted by the licensees may consist of two phases one before and one after installation of automatic status monitoring equipment, if required, in accordance with item I.D.3.

Clarification Item I.C.6 of the U.S. Nuclear Regulatory Commission Task Action Plan (NUREG-0660) and Recommendaton 5 of NUREG-0585 propose requiring that licensees'rocedures be reviewed and revised, as necessary, to assure that an effective system of verifying the correct performance of operating activities is provided. An acceptable program for verification of operating activities is described below.

The American Nuclear Society has prepared a draft revision to ANSI Standard N18.7-1972 (ANS 3.2), "Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants." A second proposed revision to Regulatory Guide 1.33, "Quality Assurance Program Requiremments (Operation),"

which is. to be issued for public comment in the near future, will endorse the latest draft revision to ANS 3.2 subject to the following supplemental provisions:

a. Applicability of the guidance of Section 5.2.6 should be extended to cover surveillance testing in addition to maintenance.

b. In lieu of any designated senior reactor operator (SRO), the authority to release systems and B. 1-36

WNP-2 AMENDMENT NO ~ 17 July 1981 P

equipment for maintenance or surveillance testing or return-to-service may be delegated to an on-shift SRO, provided provisions are made to ensure that the shift supervisor is kept fully informed of system status.

c. Except in cases of significant radiation exposure, a second qualified person should verify correct implementation of equipment control measures such as tagging of equipment.

d. Equipment control procedures should include assurance that control room operators are informed of changes in equipment status and the effects of such changes.

e. For the return-to-service of equipment important to safety, a second qualified operator should verify proper systems alignment unless functional testing can be performed without compromising plant safety, and can prove that all equipment, valves, and switches involved in the activity are correctly aligned.

NOTE: A licensed operator possessing knowledge of the systems involved and the relationship of the systems to plant safety would be a "qualified" person. The staff is investigating the level of qualification necessary for other operators to perform these functions.

For plants that have or will have automatic system status monitoring as discussed in Task Action Plan item I.D.3, NUREG-0660, the extent of human verification of operations and maintenance activities will be reduced. However, the need for such verification will not be eliminated in all instances.

WNP-2 Position WNP-2 will prepare the directives and procedures necessary to implement an effective system of verification of operating activities important to safety prior to July 1982. This program will include both automatic status monitoring and human verification by a qualified second person. We do not, however, consider it necessary to verify implementation of equipment by use-of a licensed operator. Proper return-to-servi e of safety-related equipment will be empha-sized and will e accomplished either by functional testing, automatic stat s monitoring, or by verification by a second qualified per on.

B ~ 1-,37

WNP-2 AMENDMENT NO. 17 Ju3,y 1981 I.C 7 NSSS VENDOR REVIEW OF PROCEDURES Position Obtain nuclear steam supply system (NSSS) vendor review of low power testing procedures to further verify their adequacy.

This requirement must be met befoxe fuel loading (NUREG-0694).

Clarification None WNP-2 Position The NSSS vendor (General Electric Company) will review and document the low power testing, power ascension test, and emergency procedures by July 1982. This review will consider the BWR Emergency Procedure guide-lines submitted to the NRC on behalf of a BWR Owners'roup on June 30, 1980, by letter from R. H. Buchholz to D. G. Eisenhut.

B~ 1-38

WNP-2 AMENDMENT NO ~ 17 July 1981 C.8 PILOT MONITORING OF SELECTED EMERGENCY PROCEDURES FOR NEAR-TERM OPERATING LICENSE APPLICANTS Position Correct emergency procedures, as necessary, based on the NRC audit of selected plant emergency operating procedures (e.g.,

small-break LOCA, loss of feedwater, restart of engineered safety features following a loss of ac power, steamline break, or steam-generator tube rupture).

This action will be completed prior to issuance of a fullpower license (NUREG-0694).

Clarification None WNP-2 Position.

WNP-2 will have procedures based on the BWR Emergency Procedure Guidelines by April 1982. These procedures are further addressed in response to I.C.1, Short-Term Accident Analysis and Procedure Revision.

, B.1-39

0 WNP-2 AMENDMENT NO ~ 17 July 1981 I AD.1 CONTROL ROOM DESIGN REVIEWS Position In accordance with Task Action Plan I.D.1, Control Room Design Reviews (NUREG-0660), all licensees and applicants for operating licenses will be required to conduct a detailed control room design review to identify and correct design deficiencies.'his detailed control room design review is expected to take about a year. Therefore, the Office of Nuclear Reactor Regulation (NRR) requires that those appli-cants for operating licenses who are unable to complete this review prior to issuance of a license make preliminary assessments of their control rooms to identify significant human factors and inst'rumentation problems and establish a schedule approved by NRC for correct'ing deficiencies. These applicants will be required to complete the more detailed control room reviews on the same schedule as licensees with operating plants (NUREG-0737).

Clarification NRR is presently developing human engineering guidelines to assist each licensee and applicant in performing detailed control room review. A draft of the guidelines has been published for public comment as NUREG/CR-1580i Human Engineering Guide to Control Room Evaluation." The due date for comments on this draft document was September 29, 1980.

NRR will issue the final version of the guidelines as NUREG-0700, by February 1981, after receiving, reviewing, and incorporating substantive public comments from reac-tor licensees, applicants for operating licenses,operating human fac-tors engineering experts, and other interested parties.. NRR will issue evaluation criteria, by July 'f981, which will be used to judge the acceptability of the detailed reviews per-formed and the design modifications implemented.

Applicants for operating licenses who will be unable to complete the detailed control room design review prior to issuance of a license are required to perform a preliminary control room design assessment to identify significant human factors problems. Applicants will find it of value to refer to the draft document NUREG/CR-1580, "Human Engineering Guide to Control Room Evaluation<" in performing the preliminary assessment. NRR will evaluate the applicants'reliminary assessments including the performance by NRR of onsite review/audit. The NRR onsite review/audit will be on a sche-dule consistent with licensing needs and will emphasize the following aspects of the control room:

B.1-40

WNP-2 AMENDMENT NO. 23 February 1982

a. The adequacy of information presented to the operator to reflect plant status for normal operation, anticipated operational occurrences,"

and accident conditions;

b. The groupings of displays and the layout of panels;

c. Improvements in the safety monitoring and human factors enhancement of controls and control displays;

d. The communications from the control room to points outside the control room, such as the onsite technical support center, remote shutdown panel, offsite telephone lines, and to other areas within the plant for normal and emergency operation;

e. The use of direct rather than derived signals for the presentation of process and safety infor-mation to the operator; The operability of the plant from the control room with multiple failures of nonsafety-grade and nonseismic systems;

g. The adequacy of operating procedures and operator training with respect to limitations of instru-mentation displays in the control room;

h. The categorization of alarms, with unique defini-tion of safety alarms; 1 ~ The physical location of the shift supervisor's office either adjacent to or within the control room complex.

Prior to the onsite 'review/audit, NRR will require a copy of the applicant's preliminary assessment and additional infor-mation which will be used in formulating the details of the onsite review/audit.

WNP-2 AMENDMENT NO. 17 July 1981 WNP-2 Position WNP-2 has undertaken an aggressive program to complete. a control room review program in accordance with this task. A report entitled "WNP-2 Preliminary Control Room Human Engineering Design Report," dated December 1981, provides a detailed description of the control room review efforts completed to date and future plans associated with this task.

B.l-42

WNP-2 AMENDMENT NO. 23 February 1982 I.D.2 PLANT SAFETY PARAMETER DISPLAY CONSOLE Position In accordance with Task Action Plan I.D.2, Plant Safety Parameter Display Console (NUREG-0660), each applicant and licensee shall install a safety parameter display system (SPDS) that will display to operating personnel a minimum set of parameters which define the safety status of the plant.

This can be attained through continuous indication of direct and derived variables as necessary to assess plant safety status (NUREG-0737).

Clarification These requirements for the SPDS are being developed in NUREG-0696, which is scheduled for issuance in November 1980.

WNP-2 Position FSAR 7.5.1.5, 7.5.1.6,,and 7.7.1.15 describe the SPDS and sup-porting technical data aquisition system to be implemented on WNP-2 in response to this issue. WNP-2 is working with the BWR Owners'roup to develop the emergency response information system (ERIS) as the BWR utility position responding to the concerns of Item I.D.2. The combination of these descriptions and the implementation of the ERIS concept adequately answers the concerns of Item I.D.2.

8. 1-43

WNP-2 AMENDMENT NO. 23 February l982 DELETED B.l-44

Insert following sentence on Page B.1-43:

The detailed design and extent to which the intent of the requirements provided in NUREG-0696 wiLL be provided in the next FSAR amendment.

W NL'll58UNKS'1 July 1981 NO ~ 1 7 I G ~ 1 PREOPERATIONAZ AND TOW-POWER TESTING Position (NUREG-0660)

The objective is to increase the capability of the shift crews to operate facilities in a safe and competent manner by assuring that training for plant changes and off-normal events is conducted. Near-term operating -license facilities will be required to develop and implement intensified training exer-cises during the low-power testing programs. This may involve the repetition of startup tests on different shifts for training purposes. Based on experiences from the near-term operating license facilities, requirements may be applied to other new facilities or incorporated into the plant drill requirement (Item I.A.2.5). Review comprehensiveness of test programs.

NRR will require new operating licensees to conduct a set of low-power tests to accomplish the requirement. The set of tests will be determined on a case-by-case basis for the first few plants. Then NRR will develop acceptance criteria for low-power test programs to provide "hands on" training for plant evaluation and off-normal events for each operating shift. It is not expected that all tests will be required to be conducted by each operating shift. Observation by one shift of training of another shift may be acceptable.

NRR will develop criteria in conjunction with initial near-term operating license reviews.

licensees will (1) define training plan prior to loading fuel, and (2) conduct training prior to full-power operation.

Clarification None WNP-2 Position The Supply System is committed to meet the intent of NUREG-0660 by performance of a special low power test subprogram which provides supplemental operator training in the areas of response to abnormal plant conditions and fami-liarity with critical systems. The special subprogram will amplify the well-established training value of the present Startup Test Program (STP) through (1) instruction on the con-tent, goals, and requirements of the existing program, (2) addition of selected special tests to the STP to demonstrate abnormal scenarios and use of'ritical systems and/or emergency operating procedures to control them, and (3) utili-B.1-45

.AMENDMENT NO. '17 July 1981 I

zation of the knowledge and experience gained during the STP in the training programs for future operators.

The overall Startup Test Program is'outlined in Chapter 14 while the conduct of operations is'iscussed in Chapter 13.

-During the preoperational and power ascension test phases', the operations personnel will be intimately involved in the per-formance of the various test procedures. With the impetus provided by the responsible test phase organization, the operations staff is charged with establishing the required

.plant/system conditions, initiating and controlling the desired test transient and returning the plant/system to its normal condition. The operations staff, provides the physi-cal ability to accomplish the,Startup Test Program. In this fashion, the completion of the Startup Test Program provides an unparalleled training opportunity for the operators.

The following, outlines those additional actions the Supply System will implement to augment the extensive training bene-fits inherent in the existing STP program:

I Development and Implementation of a Training Course on

~

the STP A. General Classroom Instruction (Prior to testing)

1) STP Overview a) Organization, Delineation of Responsibilities, Goals b)'dministrative and Emergency Procedures c) Preop and Power Ascension Test Schedule

2) Review Selected STP Specifics, for example;,

a) Pertinent Preop Test Purposes, Procedures, Anticipated Results b) Integrated System Cold Functional Tests c) Fuel Koading, Heatup, Power Ascension Test Purposes, Procedures, Anticipated Results d) Special Test Subprogram Test Purposes, Procedures, Anticipated Results

3) Review Expected Utilization of STP Data B~ 1-46

WNP-2 AMENDMENT NO ~ 17 July 1981 a) Documentation of Plant Safety b) Feedback/Confirmation of Anticipated Results B. Test Phase Instruction Performed by Test Director on a Shift Basis (during testing)

1) Review of the Immediate Test Schedule

2) Discussion of the Impending Tests: Procedures, Anticipated results, Precautions

3) Review/Disseminate Plant Response Data from Previous Shift(s)

C. Post-STP Completion Instruction Performed by Test Director (following testing)

1) Review of the Actual STP Results vs. Anticipated Results

2) Review Plant Design Changes/System Modifications Required I I. Development and Performance of a Special Test Subprogram Additional RCIC System Tests

1) RCIC Operation Following Loss of AC Power to the System

2) RCIC Operation to Prove DC Separation B. Integrated Reactor Vessel Level Instrumentation Functional Test C. Integrated Containment Pressure Instrumentation Functional .Test D. Simulated Loss of Control and Instrument Air Test E. Repetition of Some Normal STP Tests, for example:

=1) Feedwater Pump Trip/Recirc Runback Demonstration

2) Turbine Trip/Generator Load Rejection Within Bypass Valve Capacity

3) Pressure Regulator Setpoint Changes

ANENDNENT NO ~ 17 July . 1981

4) Recirculation Pump Trips

5) RHR Steam Condensing Mode Operation

6) Feedwater Level Setpoint Changes III. Utilization of the STP Data A. Refine the WNP-2 Simulator Response Models, as appropriate B. Incorporate a Major Plant Transient Response Section in Operator Training Program, as appropriate C. Update L'icense Program Training and Requalification Material, as appropriate.

It is anticipated. that every pertinent member of the opera-tions staff will obtain valuable knowledge and experience through participation in the WNP>>2 will receive appropriat'e classroom Startup Test Program. Each instruction; through judi-cious scheduling of tests, most will obtain personal exposure

.-to a variety of plant/system transient responses (or review of results thereof); and the training received will be con-tinually re-enforced through normal requalification program refinements. Future license candidates will also benefit from the training mateiial upgrades resulting from the STP experience.

With this program outline, the Supply System is meeting the

intent of NUREG-0660, Item I.G.1. Specific details of the training program, additional test procedures, and documen-tation methods will be devel'oped and made available for on-site NRC IaE, review prior to July 1982..

B. 1-48

nl1LalllJLllaLXJ Lt ~ I I July 1981 XX.B.1 REACTOR COOLANT SYSTEM VENTS Position Each applicant and licensee shall install reactor coolant system (RCS) and reactor vessel head high point vents remotely operated from- the control room. Although:the purpose of the system is to vent noncondensible gases from the RCS which may inhibit core cooling during natural circulation, the not lead to an unacceptable increase in the probability vents,'ust of a loss-of-coolant accident (LOCA) or a challenge to con-tainment integrity. Since these vents form a part of the reactor coolant pressure boundary, the design of the events shall conform to the requirements of Appendix A to 10 CFR Part 50, "General Design Criteria." The vent system shall be designed with sufficient redundancy that assures a low prob-ability of inadvertent or irreversible actuation.

Each licensee shall provide the following information con-cerning the design and operation of the high point vent system:

a. Submit a description of the design, location, size, and power supply for the vent system -along with results of analyses for loss-of-coolant accidents initiated by a break in the vent pipe.

The results of the analyses should demonstrate compliance with the acceptance criteria of- 10 CFR 50.'46.

b. Submit procedures and supporting 'analysis for operator use.of the vents that also include the information available to the operator for ini-tiating or terminating vent usage.

Clarification

a. General

1. The important safety function enhanced by this venting capability is core cooling. For events beyond the present design basis, this venting capability will substantially increase the plant's ability to deal with large quantities of noncondensible gas which could interfere with core cooling.

2. Procedures addressing the use of the reactor coolant system vents should define the con-ditions under which the vents should be used

WNP-2 AMENDMENT NO. 17 July 1981 as well as the conditions under which the vents should not be used. The procedures should be directed toward achieving a substantial increase in the plant being able to maintain core cooling without loss of con-tainment integrity for events beyond the design basis. The use of vents for accidents within the normal design basis must not result in a violation of the requirements of

'IO CPR 50.44 or 10 CPR 50.46.

The size of the reactor coolant vents is not a critical issue. The desired venting capa-bility can be achieved with vents in a fairly broad spectrum of sizes. The criteria for sizing a vent can be developed in several ways. One approach which may be considered is to specify a volume of noncondensible gas to be vented and in a specific venting time.

For containments particularly vulnerable to failure from large hydrogen releases over a short period of time, the necessity and desirability for contained venting outside the containment must be considered (e.g.,

into a decay gas collection and storage system).

,Where practical, the reactor coolant system vents should be kept smaller than the size corresponding to the definition of LOCA (10 CFR 50, Appendix A) . This will minimize the challenges to the emergency core cooling system (ECCS) since the inadvertent opening of a vent smaller than the LOCA definition would not require ECCS actuation, although result in leakage beyond technical speci-.

it may fication limits.. On PWRs, the use of new or existing lines whose smallest orifice is larger than the LOCA definition will require a valve in series valve that can be closed from the control room to terminate the LOCA

'that would result if an open vent valve could not be reclosed.

5. A positive indication of valve position should be provided in the control room.

6. The reactor coolant vent system shall'e operable from the control room.

B.2-2

W 8 P~J. AMENDMENT NO ~ 1 7 July 1981 7 ~ Since the reactor coolant system vent. will be part of the reactor coolant system pressure boundary, all requireme'nts for the reactor

'ressure boundary must be met, and, in addition, sufficient redundancy should be incorporated into the design to minimize the probability of an inadvertent actuation of the system. Administrative procedures> may be a viable option to meet the'single-failure criterion. For vents larger than the LOCA definition, an analysis is required to demonstrate compliance with 10 CFR 50.46.

8. The probability of a vent path failing to close, once opened, should be minimized; this is a new requirement. Each vent must have its power supplied from an emergency bus. A single failure within the power and control aspects of the reactor coolant vent system should not prevent isolation of the entire vent system when required. On'BWRs, block valves are not required in lines with that are used for venting.. safety'alves

9. Vent paths from the primary system to within containment should go to those areas that provide-good mixing with containment air.

10. The reactor coolant vent system (i.e., vent valves, block valves, position indication devices, cable terminations, and piping) shall be seismically .and environmentally qualified in accordance with ZEEE 344-1975 as supplemented by Regulatory Guide 1.100, 1.92 and SEP 3.92, 3.43, and 3.10. Environmental qualifications are in accordance with the May 23, 1980 Commission Order,and memorandum (CLZ-80-21).

11. Provisions to test for operability of the reactor coolant vent system should be part of the design. Testing should be performed in accordance with subsection IWV of Section XZ of the ASME Code for Category B valves.

12. Zt is important tnat the displays and controls added to the control room as a result of this requirement not increase the potential for operator error. A human-factor analysis should be performed taking into consideration:

B ~ 2-3

WNP-2 AMENDMENT NO-. 17 July 1981 (a) the use of this information by an opera-tor during both normal'nd abnormal plant conditions, (b) integration into emergency procedures, (c) integration into operator training, and (d) other alarms during emergency and need for prioritization of alarms.

b. BWR Design Considerations Since the BWR Owners'roup has suggested that the present BWR designs have an inherent capability to vent, a question relating to the capability of existing systems arises.

The ability of these systems to vent the RCS of noncondensible gas generated during an accident must be demonstrated. Because of differences among the head vent systems for BWRs, each licensee or applicarit should address the specific design features of this plant and compare them with the generic venting capability proposed by the BWR Owners'roup. In addition, the ability of these systems to meet the same requirements as the PWR vent system must be documented.

2 ~ In addition to RCS venting, each BWR licensee should address the ability to vent other systems, such as the isolation condenser which may be required to maintain adequate core cooling. If the production of a large amount of noncondensible gas would cause the loss of function of such a system, remote venting of that system is required. The qualifications of such a venting system should be the same as that required for PWR venting systems.

c. PWR Vent Design Considerations 1 ~ Each PWR licensee should provide a capability to vent the reactor vessel head. The reactor vessel head vent should be capable of venting noncondensible gas from the reactor vessel hot legs (to the elevation of the top of the outlet nozzle) and cold legs (through head jets and other leakage paths).

B.2-4

WNP-2 AMENDMENT NO ~ 17 July 1981

2. Additional venting capability is required for.

those portions o" each hot leg that cannot be vented through t'ie reactor vessel head vent:

or pressurizer. It is impractical to vent each of the many thousands of tubes in a U-tube steam genex'ator; however, the, staff believes that a procedure can be developed that assures sufficient liquid or steam can enter the U-tube region so that decay heat can be effectively removed from the,RCS.

Such operating procedures should incorporate this consideration.

3. Venting of the pressurizer is .required to assure its availability for system pressure and volume control. These are important considerations, especially during natural circulation.

WNP-2 Position Since the purpose of the Reactor Coolant System (RCS) vents is to vent noncondensible gases from the RCS which may inhibit core cooling during natural circulation, this requirement is not applicable to WNP>>2. The design of the WNP-2 Reactor Pressure Vessel (RPV) (as described in Chapter, 4), precludes noncondensible gases from inhibiting natural circulation cooling of the core. 'he gases which may be generated from the core would collect in the x'eactor dome above the water which, covers the core. Natural circulation through the core would continue unaffected by the noncondensible gases in the reactor vessel dome. 'ence, venting of the reaqtor. coolant system is not necessary to ensure continued natural cir-culation.

B ~ 2-5

WNP-2 AMENDMENT NO 17 July 1981 II".B.2 DESIGN REVIEW OF PZANT SH EKDING AND ENVIRONMENTAT QUALIFICATION OF EQUIPMENY FOR SPACES/SYSTEMS WHICH MAY BE USED IN POST-ACCID"."1T OPERATIONS Position With assumption of a post-accident release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalent of 50% of the core radioiodine; 100%

'the core noble gas inventory, and 1% of the core solids are of contained in the primary coolant), each licensee shall perform a radiation and shielding-design review of the spaces around systems that may, as a result of an accident< contain highly radioactive materials. The design review should identify the

'location of vital areas and equipment, such as the control room, radwaste control stations, erne gency power supplies, motor control centers, and instrument areas, in which person-nel occupancy may be unduly limited or safety equipment may be unduly degraded by the rad'ation fields during post-accident operations of these systems.

Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increa'sed permanent or temporary shielding, or, post-accident procedural controls. The design review shall determine which types of corrective actions are ne'eded for vital areas throughout the facility.

. Clarification The purpose of this item is to ensure that licensees'xamine their plants to determine what actions can be taken over the short term to reduce radiation levels and increase the capabi-lity of operators to control and mitigate the consequences of an accident. These actions should be taken pending conclu-

,sions resulting in the long term degraded core rulemaking, which may result in a need to consider additional sources.

Any area which will. or may require occupancy to permit an operator to aid. in the mitigation of or recovery from an acci-dent is designated as a vital area. For the purposes of this evaluation, vital areas and equipment are not necessarily. the same vital areas or equipment defined in 10 CFR 73.2 for security purposes. The security center is listed as an area to be considered as potentially vital, since access to this area may be necessary to take action to give access to other areas in the plant.

B.2-6

NNP-2 AMENDMENT NO 17 July 1981 The control room,. technical suppor" center (TSC),, sampling station and sample analysis area m st be included among those areas where access, is considered v:':al after an accident.

(See It'em III.A.1,.2 for discussion of the TSC and emergency operations facility.)'he evaluat: in to determine the necessary vital areas should also 'nclude, but not be limited to, consideration of the post-LOCA hydrogen control system, containment .isolation reset contro area, manual ECCS align-ment area (if any), motor contxol = nters, instrument panels, emergency power supplies, security center, and radwaste control panels. Dose rate determinations need not'e fox these areas if they are determined not to be vital.

As a minimum, necessaxy modifications must be sufficient, to provide fox vital system operation and for occupancy of the control room,'SC, sampling, station and sample analysis area.

In order to assure that personnel can perform necessary post-accident'perations in the vital areas, the following guidance is to be used by licensees to evaluate the adequacy of radiation protection to the operators:

a. Source Term:

The minimum radioactive source term should be equivalent to the source terms recommended- in Regulatory Guides 1.3, 1.4, 1.7 and Standard ReviewPlan 15.6.5 with appropxiate decay times based on plant design (i.e., you may assume the

, radioactive decay that occurs before fission pro-ducts can be transported to various systems).

1. Liquid-Containing systems: 100% of the core

., -equilibrium noble gas inventory, 50% of the

'core equilibrium halogen inventory, and 1% of all others are assumed to be mixed in the reactor coolant and liquids recirculated by residual heat removal (RHR), high pressure coolant injection (HPCI), and low pressure coolant injection (LPCI), or the equivalent of these systems. Xn determining the source term for recirculated, depressurized cooling water, you may assume that the, water contains no noble gases.

B ~ 2-7

WNP-'2 AMENDMENT NO. 17 July 1981

2. Gas-Containing Systems: 1QQS of the core equilibrium noble gas inventory and 25% of

. the core equilibrium halogen activity are assumed to be mixed in the containment atmosphere. For vapor-containing lines con-nected to the primary system (e.g., BWR steam lines), the concentration of radioactivity shall- be determined assuming the activity is contained in the vapor space in the primary coolant system.

b. Systems Containing the Source:

Systems assumed in your analysis to contain high levels of radioactivity in a post-accident situation should include, but not be limited to, i containment, residual heat removal system, .safety injection systems, chemical and volume control system (CVCS), containment spray recirculation system, sample lines, gaseous radwaste systems, and standby gas treatment systems (or equivalent of these systems); If any of these systems or others that could contain high levels of radioac-tivity were excluded, you should explain why. such systems were excluded. Radiation from leakage of systems located outside of containm'ent need not be considered for this analysis. Leakage measurement and reduction is treated under Item III.D.1.1; "Integrity of Systems Outside Containment Likely to Contain Radioactive Material for PWRs and BWRs." Liquid waste systems need not be included in this analysis.

Modifications'to liquid waste. systems will be considered after completion of, Item III.D.1.4, "Radwaste System Design Features To Aid in Accident Recovery and Decontamination."

c. Dose Rate Criteria:

The design dose rate for personnel in a vital area should be such that the guidelines of GDC 19 will not be exceeded during the course of 'the accident. GDC 19 requires that adequate radiation protection be provided such that the dose to personnel should not be in excess of 5 rem whole body, or its equivalent to any part of the body for the duration of the accident. When determining the dose to an operator, care must be B ~ 2-8

WNP-2 AMENDMENT NO. 17 July 1981 taken to determine the necessary occupancy times in a specific area. For example, areas requiring continuous occupancy will require much lower dose rates than areas where minimal occupancy is required. Therefore, allowable dose rates will be based upon expected occupancy, as well as the radioactive source, terms and shielding. However, in order to provide a general design objective, we are providing the following dose rate criteria with alternatives to be documented on a case-by-case basis. The recommended dose rates are average rates in the area. Local hot spots may exceed the dose rate guidelines. These doses are design objectives and are not to be used to limit access in the event of an accident.

Areas Requiring Continuous Occupancy: <15 mrem/hr (averaged over 30 days). These areas will require full-time occupancy during the course of the accident. The control room and onsite technical support center are areas where continuous occupancy will be required.

The dose rate for these areas is based on: the control room occupancy factors contained in SRP 6.4 ~

2. Areas Requiring Infrequent Access: These areas may require access on an irregular basis, not continuous coccupany. Shielding should be provided to allow access at a fre-quency and duration estimated by the licen-see. The plant radiochemical/chemical analysis laboratory, radwaste panel, motor control center, instrumentation locations, and reactor coolant and containment gas sample stations are examples of sites where occupany may be needed often, but not continuously.

d. Radiation Qualification of Safety-Related Equipment:

The review of safety-related equipment which may be unduly degraded by radiation during post-accident operation of this equipment relates to equipment inside and outside of the primary con-

.tainnment. Radiation source terms calculated to determine environmental qualification of safety-related equipment .consider the following:

B. 2-9

WNP>>2 AMENDMENT NO ~ 17 July 1981

1. LOCA events which completely,depx'essurize the pximary system should consider releases of the source term (1008 noble gases, 50% iodines, and 1% particulates) to the conntainment atmosphere.

2. LOCA events in which the primary system may not depressurize should consider the source term (100% nobles gases, 50% iodines, and 1%

particulates) to remain in the primmary coolant.

This method is used to determine the qualifica-tion doses for equipmment in close proximity to recirculating fluid systems inside and outside of containment. Non-LOCA events both inside and outside of containment should use 10% noble gases, 10% iodines, and OS pariticulates as a source term.

The following table summarizes these considerations:

Non-LOCA

.LOCA Source High-Energy Line Term (Noble Break Source Term Gas/Iodine/ (Noble Gas/Iodine/

Containment Particulate Particulate)

Outside (100/50/1) (10/10/0) in RCS in RCS Inside f (10/10/0)

(100/50/1 ) in RCS in containment ox'100/50/1) in RCS WNP-2 Position WNP-2 concurs with this task as presented in NVREGs 0578, 0660, and 0694. NUREG-0578 states that the review shall be done "with the assumption of a post-accident release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4 ..." NUREG-0660 states, NRR will require a radiaton and shielding design review of spaces around systems in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by radiation during operation following an accident resulting in a degraded core..." NUREG-0694 requests that a "radiation and B. 2-10

WNP-2 AMENDMENT NO ~ 30 June 1983 shielding design review that identifies the location of vital areas and equipment in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by radiation during operations following an accident resulting in a degraded core ..." be provided as a condition for issu-ance of a full power license. The NUREG-0660 clarification letter of September 5, 1980 expands on the post-accident/

degraded core scenario.

a WNP-2 performed the review as specified by the referenced NUREGs by using in-house personnel the architect/engineer and consultant. As a product of this review, WNP-2 produced an interim Shielding Evaluation Radiation Report (submitted to the NRC in January 1982) and a Final Shielding Evaluation Radiation Report (submitted in November 1982). These reports addressed all the issues needed to comply with the II.B.2 position except as followss WNP-2 takes exception to the portion of this task that specifies that a review of "safety-related equipment which may be degraded by radiation during post-accident operation be provided for a non-LOCA, High-Energy Line Break Source Term." The pipe break/missile analysis performed in 3.5 and 3.6 of the FSAR addresses non-mechanistic pipe breaks inside the outside pipe breaks do not lead mechanistically to a radiation containment'hese release due to fuel failures beyond those allowed in normal operation. Hence, the source term identified applied outside containment is entirely hypothetical and would be a new design basis beyond the scope of current regulations'E 2-11

WNP-2 AMENDMENT NO. 23 February 1982 II.B.3 POST-ACCIDENT SAMPLING CAPABILITY Position A design and operational review of the reactor coolant and containment atmosphere sampling line systems shall be per-formed to determine the capability of personnel to promptly obtain (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) a sample under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 18-3/4 rem to the whole body or extremities, respectively. Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products. If the review indicates that personnel could not promptly 'and safely obtain the samples, additional design features or shielding should be provided to meet the criteria.

A design and operational review of the radiological spectrum analysis facilities shall be performed to determine the capa-bility to promptly quantify (in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) certain radionuclides that are indicators of the degree of core damage. Such radionuclides are noble gases (which indicate cladding failure), iodines and cesiums (which indicate high fuel temperatures), and nonvolatile isotopes (which indicate fuel melting). The initial reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4 release. The review should also consider the effects of direct radiation from piping and components in the auxiliary building and possible contamination and direct radiation from airborne effluents. If the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or, equipment procurement shall be undertaken to meet the criteria.

In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions.

Procedures shall be provided to perform boron and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source term). Both analyses shall be capable of being completed promptly (i.e., the boron sample analysis within an hour and the chloride sample analy-sis within a shift).

Clarification The following items are clarifications of requirements iden-tified in NUREG-OS78, NUREG-0660, or the September 13 and October 30, 1979 clarification letters.

a. The licensee shall have the capability to promptly obtain reactor coolant samples and containment B.2-12

WNP-2 AMENDMENT NO ~ 17 July 1981 atmosphere samples. The combined time allotted for sampling and analysis should be 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or less from the time a decision is made to take a sample

b. The licensee shall establish an onsite radiologi-cal and chemical analysis capability to provide, within the 3-hour time frame established above, quantification of the following:

1. certain radionuclides in the reactor coolant and containment atmosphere that may be indi-cators of the degree of core damage (e.g.,

noble gases, iodines and cesiums, and non-volatile isotopes);

2. hydrogen levels in the containment atmosphere;

3. dissolved gases (e.g., Hp), chloride (time allotted for analysis subject to discussion below), and boron concentration of liquids.

4. alternatively, have inline monitoring capa-bilities to perform all or part of the above analyses.

c. Reactor coolant and containment atmosphere sampling during post-accident conditions shall not require an isolated auxiliary system (e.g the letdown system, reactor water cleanup system (RWCS)) to be placed in operation in order to use the sampling system.

'.d. Pressurized reactor coolant samples are not required if the licensee can quantify the amount of dissolved gases with unpressurized reactor coolant samples. The measurement of either total dissolved gases or H2 gas in reactor coolant samples is considered adequate.

Measuring the 02 concentration is recommendedg but is not mandatory.

e. The time for a chloride analysis to be performed is dependent upon two factors: if the plant's coolant water is seawater (1)or brackish water, and (2) if there is only a single barrier between primary containment systems and the cooling water. Under both of the above con>>

B ~ 2-13

WNP, Z AMENDMENT NO 17 July 1981 I

ditions the licensee shall provide for" a chloride analysis within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the sample bei'ng taken.'or all other cases, the licensee shall provide for the analysis=to be completed within 4 days. The chloride analysis does n'ot have to be done onsite.

'The design basis for plant equipment for reactor coolant and containment atmosphere sampling and analysis must assume that it is possible to obtain and a'nalyze a sample without radiation exposures to any individual exceeding the cri-teria of GDC 19 (Appendix A, 10 CFR Part 50)

(i.e., S rem whole body, 7S rem extremities).

(Note that the design and operational review cri-terion was changed -from the operational limits of 10 CFR Part 20-(NUREG-0578) to the GDC 19 cri-terion (October 30, 1979 letter from H. R. Denton to all licensees.))

g>> The analysis of pdimary coolant samples for boron

,is required for PWRs. (Note that Revision 2 of Regulatory Guide 1.97, when issued, will, likely specify the need for primary coolant. boron analy-sis capability at BWR plants.)

h. If inline monitoring is used for any sampling and analytical capability specified herein,.the J

licensee shall provide backup sampling through grab samples, and shall demonstrate the capabil-ity of analyzing the samples. Established planning for analysis at offsite facilities is .

acceptable. Equipment provided for backup s'ampling shall be capable, of providing- at least one sample per day for 7 days following onset of the accident,and 'at least one sample per week until the accident condition no longer exists.

The licensee's radiological and chemical 'sample analysis capability shyll include provisions to:

1. Identify and quan"ify the isotopes- of the nuclide categories discussed above to levels corresponding to the source terms given in Regulatory Guides 1.3 or 1.4 and 1.7. Where necessary and practicable, the ability to dilute samples to provide capability for measurement and reduction of personnel expo-sure should be provided. Sensitivity of onsite liquid sample analysis capability

WNP-2 AMENDMENT NO. 17 July 1981 should be such as to permit measurement of nuclide concentration in, the range from approximately 1 ACi/g to 10 Ci/g.

2. Restrict background levels of radiation in the radiological and chemical analysis faci-lity from sources such that the sample analy-sis will provide results with an acceptably small error (approximately a factor of 2).

This can be accomplished 'through the use of sufficient shielding around samples and out-side sources, and by the use of ventilation system design which will control the presence of airborne radioactivity.

j. Accuracy, range, and sensitivity shall be ade-quate to provide pertinent data to the operator in order to describe radiological and chemical status of the reactor coolant systems.

k. In the design of the post-accident sampling and analysis capability, consideration should be given to the following items:

I Provisions for purging sample lines", for reducinng plateout in sample lines, for mini-mizing sample loss or distortion, for pre-venting blockage of sample lines by loose material in the RCS or containment, for appropriate disposal of the samples, and for flow restrictions to limit reactor coolant loss from a rupture of the sample line. The post-accident reactor coolant and containment atmosphere samples, should be representative of the reactor coolant in the core area and the containment atmosphere following a transient or accident. The sample lines should be as short as possible to minimize the volume of fluid to be taken from containment. The residues of sample collection should be returned to containment or to a closed system.

2. The ventilation exhaust from the sampling station should be filtered with charcoal adsorbers and high-efficiency particulate air (HEPA) filters.

3 ~ Guidelines for analytical or instrumentation range are given in Table II.B.3-1.

B.2-15

WNP-2 AMENDMENT NO ~ 17 July'981 WNP-2 Position The presently designed WNP-2 sampl'ng system has been reviewed for adequacy with respect to post-accident sampling. A con-sultant has been contracted to review the system and make recommendations for .redesign.-

Based upon the Supply System's own investigations and the recommendations of the consultant, the WNP-2 architect/

engineer wi'll be directed to implement design changes as required. These changes will allow WNP-2 to adequately con-duct post-accident sampling.

The design details of the system will.be available for NRC review prior to April'982.

8 ~ 2-16

AMENDMENT NO 17 July 1981 II.B.4 TRAINING FOR'MITIGATING CORE DAMAGE Position (NVREG-0737)

Ticensees are required to develop a training program to teach the use of installed equipment and systems to control or miti-gate accidents in which the core is severly damaged. They must then implement the training program.

Clarification Shift technical advisors and operating personnel from the plant manager through the operations chain to the licensed operators shall receive all the training indicated in to H. R. Denton's March 28, 1980 letter.

Managers and technicians in the Instrumentation and Control (I&C), health physics, and chemistry departments shall receive training commensurate with their responsibilities.

WNP-2 Position A training program covering the intent of the above position will be developed prior July 1982.

Training of operating personnel responsible for monitoring= and controlling the reactor under degraded core conditions will be completed prior to operation of the reactor in a mode that could result in release of significant fission products as a result of core damage.

B 2-17

AMENDMENT NO ~ 17 July 1981 II.DE 1 'ELIEF AND SAFETY VALVE TEST. REQUIREMENTS Position A

Describe .a test, program and schedule for testing to qualify reactor, coolant system relief-"and safety valves und'er expected operating conditions for design basis transients and accidents.

This requirement shall be met before fuel loading. See.

Section =2.1.2 (Reference 4), and letters of. 'UREG-0578, September 27 (Reference 23) and November 9, 1979 (Reference 24).

Clarification Complete tests to qualify the reactor coolant system relief and safety 'valves under expected operating conditions for design basis trans'ients and accidents. k This requirement shall be met by Juiy 1, 1981. See

~ ~

NUREG-0578, Section 2.1.2 (Reference 4), and letters of

~

September 27 (Reference 23) and November 9, 1979 (Reference 24)

WNP-2 Position WNP-' is a partcipant in the BWR Owners'roup which is addressing this task. The final tes" program description and results'ill be submitted to NRR by October 1, 1981.

B ~ 2-18

e WNP-2 AMENDMENT NO. 17 July 1981 XI.D 3 RELXEF AND SAFETY VALVE POSITION XNDICATION Position Reactor coolant system rdlief and safety valves shall be pro-vided with a positive indication in the control room derived from 'a reliable valveposition detection device or a reliable indication of flow in the discharge pipe (NUREG-0737).

Clarification

a. The basic requirement is to provide the operator with unambiguous indication of'alve position (open or closed) so that appropriate operator actions can be taken.

b. The valve position should be indicated in the contxol room. An alarm should be provided in conjunction with this indication.

c. The valve position indication may be safety-grade. Xf the position indication is not safety-grade, a reliable single channel direct indication powered from a vital instrument bus may be provided if backup methods of determining position are available and are discussed in 'alve the emergency procedures as an aid to operator diagnosis and action.

d. The valve position indication should be seismi-cally qualified consistent 'with the component or system to which it is attached.

e. The position indication should be qualified for its appropriate environment (any transient or accident which would cause the relief or safety valve to lift).

WNP-2 Position WNP-2 concurs with the intent of the NRC staff position requiring direct indication of the position of, the main steam line safety/relief valves (SRV).

The Supply System will install acoustical monitors to measure the SRV opex'ation. The acoustical monitors wil'l provide qualitative flow and open/closed information for each safety/

relief valve.

B. 2-19

WNP-2 AMENDMENT NO ~ 17 July 1981 WNP-2 will be supplied with a single channel of acoustic instrumentation. The equipment will be seismically and environmentally qualified to Class 1 requirements.

Additional documentation will be available for NRC staff review ma~~ prior to November 12, 1981.

B.2-20

WNP-2 AMENDMENT NO. 17 July,. 1981 II.E.4.1 Dedicated Hydrogen Penetrations Position using external recombiners or purge systems for, post-r'lants accident combustible gas control of the containment atmosphere should provide containment penetration systems for external recombiner or purge systems that are dedicated to that service only, that meet the redundancy,and single- fail'ure- requirements of General Design Criteria 54 and 56 of Appendix A to 10 CFR 50, and that are sized to satisfy the flow requirements of the recombiner or, purge system.

The procedures for the use of combustible gas,control systems following an accident that results, in a degraded core and release of radioactivity to the conntainment must be reviewed and revised, .if necessary.

Clarification

a. An acceptable alternative to the dedicated penetration is a combined design that is single failure proof for containment. isolation purposes and single failure proof for operation of the

. recombiner or purge system.

b. The dedicated penetration or the combined single failure proof alternative shal'l,be sized such that the flow requirements for the use of. the recombiner or purge system are satisfied. The design shall be based on 10 CFR 50.44 requirements.

c. Components furnished .to satisfy this requirement shall be safety-grade.

d. Licensees that rely on purge systems as the pri-mary means for controlling combustible gases following a loss-of-coolant accident should be aware of the positions taken in SECY-80-399<

"-Proposed Interim Amendments to 10 CFR Part 50 Related to Hydrogen Control and Certain Degraded Core Considerations." This proposed rule, published in the Federal Resister on October 2, 1980, would require plants that do not, now have recombiners to have the capacity to install external recombiners by January.1, 1982.

(Installed internal recombiners are an acceptable alter'native to the above.)

B.2-21

WNP-2 AMENDMENT NO. 'l7 July 1981

e. Containment atmosphere dilution (CAD) systems are considered to be purge systems for,the purpose of implementing the requirements of this TMX Task Action item (NUREG-0737).

WNP-2 Position WNP-2 has permanently installed redundant single failure proof post-LOCA hydrogen recombiners as described in FSAR 6.2.5. Dedicated containment penetrations are provided that meet the requirements of General Design Criteria 54 and 56 of Appendix A to 10CFR50. (Reference FSAR 3.1.2.5.5, 3.1.2.5.7g 6.2.4.1.1, 6.2.4.3.2.2.3,1, and Figures 3.2-17 and 6.2-31g.)

The dedicated containment penetrations are sized to satisfy hyrogen recombiner flow requirements based on Regulatory Guide 1.7 (Revision 1, dated September "1976) -and 10CFR50;44 requirements.

No further WNP-2 action is required to comply with the intent 0 f this task.

B.2-22

WNP-2 'MENDMENT NO ~ 17 July 1981 II.E.4.2 Containment Isolation Dependability Position a ~ Containment isolation system designs shall comply with the recommendations of Standard Review Plan Section 6.2.4 (i.e., that there be diversity in the parameters sensed for the initiation of con-,

tainment isolation).

b. All plant personnel shall give careful con-sideration to the definition of essential and nonessential systems, identify each system deter-mined to be essential, identify each system to be nonessential, describe the basis

'etermined for..selection of each essential system, modify their containment isolation designs accordingly and report the results of the re<<evaluation to the NRC.

C~ All nonessential systems shall be automatically isolated by the containment isolation signal.

'd. The design of control systems for automatic con- ~

tainment isolation valves shall be such that resetting the isolation signal will not result in the automatic re-opening of containment isolation valves. Re-opening of containment isolation valves shall require deliberate operator action.

e. The containment setpoint pressure that initiates containment isolation for nonessential penetra-tions must be reduced to the minimum compatible

'with normal operating conditions.

Containment purge valves that do not satisfy the operability criteria set forth in Branch Technical Position=CSH 6-4 or the Staff Interim Position of October 23, 1979 must be sealed closed as defined in SRP 6.2.4, Item II.3.f during operational conditions 1, 2, 3, and 4.

Furthermore,'hese valves must be verified to be closed at least every 31 days.

g Containnment purge and vent isolation valves must close on a high radiation signal (NUREG-0737).

BE 2-23

NNP-2 AMENDMENT NO. 17 July 1981 Clarification

a. The reference to SRP 6.2.4 in position 1 is only to the. diversity requirements set forth in that document.

b. For post-accident situations, each nonessential penetration (except instrument lines) is required to have two isolation barriers in series that meet the requirements of General Design Criteria 54, 55, 56, and 57, as clarified by Standard Review Plan, Section 6.2.4. Xsolation must be performed automatically '(i.e., no credit can be given for operator action). Manual valves must be sealed closed, as defined by Standard Review Plan, Section 6.2.4, to qualify as an isolation barrier. Each automatic isolation valve in a nonessential penetration must receive the diverse isolation signals.

c~ Revision 2 to Regulatory Guide 1.141 will contain guidance on the classification of essential ver-sus nonessential systems and is due to be issued by June 1981. Requirements for operating plants to review their list of essential and nonessen-tial systems will be issued in conjunction with this guide including an appropriate time schedule for completion.

d. Administrative provisions to close all isolation valves manually before resetting the isolation signals is not an acceptable method of meeting position 4.

e. Ganged re-opening of containment isolation valves is not acceptable. Re-opening of isolation valves must be performed on a valve-by-'valve basis, or on a line-by-line basis, provided that electrical independence and other single>>fdilure criteria continue to be satisfied.

The containment pressure history during'ormal operation should be used, as a basis for arriving at an appropriate minimum pressure setpoint for initiating containment isolation. The pressure setpoint selected should be far enough above the maximum observed (or expected) pressure inside containment during normal operation so that inad-vertent containment isolation does not occur during normal operation from instrument drift or B.2-24

WNP-2 AMENDMENT NO. 17 July 1981 0

fluctuations due to the accuracy. of the -pressure sensor. A margin of 1 psi .above the maximum expected containment pressure should be adequate to accouunt for instrument error. Any greater than 1 psi will require detailed proposed'alues justification. Applicants for an operating license and operating,.plant licensees that have "

operated less than one year should use pressure history. data from similar plants that have operated more than one year, if possible, to.

at ainimum containment setpoint 'arrive pressure.

g. Sealed-closed purge isolation valves shall be under administrative control to assure that they cannot be inadvertently opened. Administrative control includes mechanical devices to. seal or lock the-.valve closed, or to prevent power from being supplied to the valve operator. Checking the valve position light in the control room is an'adequate method for verifying every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that the purge valves are closed.

WNP-2 Position For WNP-2, reactor water level (signals L, A, and V) and high drywell pressure (signal F) are used as diverse parameters indicative of a LOCA for initiation of'ontainment isolation.

Xn addition, high radiation measured in the Reactor Building ventilation exhaust plenum (signal 2) and in the main steam tunnel, (signal C), as well as various high flow and room tem-perature signals indicative of line breaks outside containment are used selectively. for system isolation, as indicated in.

Table ZZ.E.4.2-1. Tabel XZ.E.4.2-2 provides a description of isolation signals identified in Table ZZ.E.4.2.-1.

Systems penetrating containment were reviewed on a functional to determine whether they are, or may be required,to 'asis function following an accident to .provide a safety-related response to the accident. Systems penetrating containment which are not required to perform a safety function at any stage of the response to the accide'nt, between initiating event and cold shutdown, are considered "non-essential".

Systems penetrating containment which may be called upon to perform a safety'function, at any stage of the response to the accident, are considered "essential". Using these definitions (which differ from those used in NUREG-0578, but which are consistent with Appendix A to Branch Technical Position APCSB 3-1 and with standard usage), paths through containment are listed and categorized as essential or non-essential in Table B 2-25

WNP-2 AMENDMENT NO 17 July 1981 II.E.4.2-1 lines through containment for essential systems, which may be needed for long-term response to an accident, but not short-term (i.e., within 10 minutes), and would therefore be categorized "non-essential" per NUREG-0578, are indicated in Table II.E.4.2-1 by an asterisk (E*).

Table II.E.4.2-1 does not list systems for which all contain-ment isolation valves are check valves, relief valves, locked closed, or manual only, since the purpose of the review of essential and non-essential systems is, ultimately, to iden-tify any required logic changes or signal changes for automa-tic valves.

Pollowing identification of essential and non-essential systems, non-essential systems were reviewed to assess whether diverse parameters are used as signals to close the isolation valves.

Except 'as indicated below, containment isolation valves in non-essential systems were found to use diverse parameters as iso-lation signals.

Exceptions:

a. Main steam, main steam drain lines and reactor water sample (RRC) line isolate on low reactor water level (signal A). Xf low water level failed to initiate isolation valve closure, and fuel failure resulted from the ZOCAg Signal Cg high radiat'ion in the main steam tunnel, would close the isolation valves. In addition, signals indicative of a break in the main steam lines, as indicated in Table II.E.4.2-1, would initiate isolation valve closure.

b. The reactor water cleanup (RWCU) suction line valves isolate on low reactor water level (signal A). In addition, signals indicative of a break in the RWCU system would initiate isolation valve closure. High drywell pressure is not used as an isolation signal because it is desirable to operate the RWCU system under normal shutdown procedures, when high drywell pressure results from small steam leaks, small-break LOCAs, or failure of drywell coolers.

c. Normally closed containment isolation valves in non-essential systems do not, in some cases, use diverse paramete'rs as isolation signals, as indi-cated in Table II.E.4'.2-1. The requirement for diverse parameters as isolation signals is not B 2-26

WNP-2'MENDMENT.NO. 17 July .1981 considered to apply when the isolation valves are normally closed and under Administrative Control.

The exceptions listed above are considered acceptable. As a result of this review for diverse parameters, no design changes were found to be necessary for containment isolation.

\

An investigation reviewing the logic for the containment iso--

lation valves to v'erify that resetting the isolation signal would not result in the automatic re-opening of the contain-ment isolation valve has been completed. Valves RRC-V-19 and 20, EDR-V-19 and 20 and FDR-V-3 and 4 were. identified as being in violation of this requirement and have been corrected.

For addtional informaton, refer to FSAR Table 6.2-16, and FSAR Figures 6.2-31a through 31t.

WNP-2 is a participant in the'BWR Owners'roup. The position taken by the Owners'roup. and endorsed by WNP-2 with respect to containment setpoint pressure is as follows:

The containment isolation analytical setpoint pressure for Mark II containments is approximately 2 psig (drywell pressure). Under normal operating conditions, fluctuations in the atmospheric barometric pressure as well as heat inputs from such sources as pumps can result in containment pressure increases on the order of 1psi. Consequently, the isolation setpoint of 2 psig, provides a 1 psi margin above the maximum expected operating pressure.

't The 1 psi margin to isolation has proved to be a suitable value to minimize the possibility of spurious containment isolation. the same',time, it is such a low value (particularly 'in view of the small drywell volume of Mark II containments) that it provides a- very sen-sitive and positive means of detecting and protecting against breaks and leaks in the reactor coolant system. No change of the setpoint is necessary for these containment types.

The WNP-2 containment gurge valves s'atisfy the operability criteria set forth in Branch Technical Position CSB 6-4 and the Staff Interim Position of October 23, 1979. See,FSAR 6.2.1.1.8 and 6.2.4 for a description of the WNP-2 contain-ment, purge valves.

B.2-27

TABLE IlsE+4 ~ 2-I INBOARD 'OUTBOARD ISOLATION ISOLATION ,ISOLATION SIGNALS SYSTEM VALVE ~ VALVE CLASS ~ LOCA SYSTEM COMMENTS MAIN STEAM main steam lines MS-V-22AsB sCsD MS-V-28AsBsCsD CsGsDsP MS-V-67AsBsCsD

- MSIV-leakage control MS-V-22ATBTCTD MSLC-V-3ASBiCSD Leakage control for

- MS line drain MS-V-16 MS-V-19 CSGsDSP MSIVs hydraulic lines HY-V-17i 18 s pump 19s20sAsB seal water RRC-V-13A, B RRC-V-16A s B Valves must stay open to prevent

=

reactor coolant loss through seals.

HPCS

- to RPV HPCS-V-5 HPCS-V-4 E Essential safety system.

suppression HPCS-V-15 pool'uction test'line HPCS-.V-23 PsA minimum flow, line HPCS-V-12 LPCS to RPV LPCS-V-6 LPCS-V-5 Essential safety system.

suppression pool LPCS-V-1 suction test line ' LPCS-V-12 FsV minimum flow line LPCS-FCV-1 1

TABLE XI.E.4.2-1 (Continued)

INBOARD OUTBOARD ISOLATION ISOLATION ISOLATION SIGNALS SYSTEM 'VALVE VALVE CLASS LOCA SYSTEM COMMENTS SLC SLC-V-7 SLC-V-4Ai B .Outboard isolation valves are normally closed explosive valves. Must be available as backup to CRD system.,

- LPCI to RPV RHR-V-41AzBzCz RHR-V-42AzBzC E Safety system.

drywell spray RHR-V-16Az B E* Valves are normally RHR-V-17A z B closed and under administrative con-trol.

- wetwell spray RHR-V-27A,B PiV Can override isola-tion signals pro-vided LPCI in)ec-tion valve is closed.

- shutdown ~ling RHR-V-.50A)B RHR V 53AzB UzMzR Valves are closed return RHR-V-123A,B during operation at power and under administrative control.

shutdown cooling RHR-V-9 RHR-V-8 UzMzR Valves are closed suction during operation at power and under.

administrative control.

suppression pool RHR-V-4AzBzC E suction minimum flow line . RHR-PCV-64AzBzC E test line RHR-V-24A z B,C NE PzV

- heat exchange RHR-V-1 1Az B NE PzV Normaly closed, condition only open for hot standby.

- heat exchange RHR.V 73AzB - NE No isolation sig-vent nals. Valve normally closed, only open for hot standby

TABLE II.E.4.2-1 (Conti>>uad)

INBOARD OUTBOARD ISOLATION ISOLATION ISOLATION SIGNALS SYSTEM VALVE VALVE CLASSY LOCA SYSTEM COMMENTS RHR (Cont.)

CAC drains RHR-V-134A,B E*>> Normally closed, only -open when H2 recombiners are needed.-

condensate drain RHR-V-124AiB Normally closed, pots RHR-V-125AiB only open for hot standby.

head spray RCIC-V-66 RHR-V-23 UiMiR RCIC

- condensing mode RCIC-V-63 RCIC-V-64 KiX Normally closed

,steam supply 'alve only open for hot standby.

RCIC system necessary for core cooldown following isolation from turbine condenser and feedwater makeup.

turbine steam supply RCIC-V-63 RCIC-.V-8 E IiX pump minimum flow RCIC-V-1 9 E turbine exhaust RCIC-V-68 E

- turbine exhaust RCIC-V-110 E N vacuum breaker RCIC-V-113

- vacuum pump discharge RCIC-V-69 suppression pool RCIC-V-31 suction head spray RCIC-V-66 RCIC-V-13 CAC-V-2,4,6- Valves normally 8, 11, 13, 15'7. closed during operation.

CAC-PCV-1A i B Only opened if H2 2A B 3A B 4A B recombiners are needed.

TABLE II.E.4.2- ; nued)

INBOARD OUTBOARD ISOLATION ISOLATION ISOLATION SIGNALS SYSTEM VALVE VALVE CLASS LOCA SYSTEM COMMENTS CSP/CEP reactor building CSP-V-5~6~9 Valves normally to wetwell vacuum closed open auto-breakers matically to re-lieve negative pressure

- wetwell ventilation CSP-V-3,4 supply.

wetwell ventilation CEP-V-3A~B~ PgA exhaust 4A,B drywell ventilation CSP-V-l, 2 suppl y I

- drywe ll ventilation exhau st CEP-V-1AgB 2AiB FgA I

u$

I RCC tJ I

inlet RCC-V-5 RCC-V-104

- outlet RCC-V-40 RCC-V-21 FiA suppression pool PPC-V-153 PiA cleanup suction FPC-V-154

- suppression pool PPC-V-156 P,A cleanup return RWCU from reactor RWCU-V-1 RHCU-V-4 A JgEi YiW Valve does not close on high dry-well pressure sig-nal, to permit RWCU operation under small-break LOCA,, or small steam leak condition.

TABLE II.E.4.2-1 (Continued)

INBOARD

• OUTBOARD ISOLATION ISOLATION ISOLATION SIGNALS SYSTEM VALVE VALVE CLASS ~ LOCA SYSTEM COMMENTS RWCU (Cont.)

return to RFH RHCU-V-40 Connect to RFH system upstream of

'4 RFH isolation valves.

EDR EDR-V-'I 9 NE -'tA EDR-V-20 FDR FDR-V-3 F,A

'DR-V-4 CIA

-- to relief valve CIA-V-20 . Essential because accumulators system supports safety equipment.

- to ADS accumulators CIA-V-30Ai B insert Provides a desire-able reflood capability.

withdrawal ZaC

- RRC sample RRC-V-19 RRC-V-20 NE A

- TIP lines C51 J004 - ".." NE -- LiF Closed system with no direct interface with reactor con-tainment atmos-phere.

radiation monitors PI-VX-250'5I,253,256~

257i259 LEGEND! E. essential during short-teim response (within 10 minutes)

'E -- nonessential E* essential during long-term response (after 10 minutes)

WNP-2 AMENDMENT'O 1 7 July 1981 TABLE II.E.4.2-2 PRIMARY CONTAINMENT AND REACTOR VESSEL ISOLATION SIGNAL CODES Sicinal L* Reactor vessel low water level (Trip 3) - (A scram=

occurs at this level. also. This is the higher of the three low water -level signals.)

A* Reactor vessel low water level (Trip 2)

High radiation Main steam Line break Main steamline (steamline routing area high space temperature or steam line high steam

'f low )

High drywell pressure (core standby cooling systems are started.)

Zine break in cleanup system - high space temperature.

K* Line break in RCIC system line to turbine (high RCIC -pipe space temperature, high steam .flow, or low steam line pressure or high turbine exhaust pressure).

Line break in RHR shutdown piping (high suction flow )

I Zow main steam line pressure at inlet turbine (RUN mode only.)

High reactor vessel pressure.

High temperature at outlet of cleanup system non-regenerative heat exchanger.

Standby liquid control system actuated.

Reactor building ventilation exhaust plenum high radiation.

RM Remote manual switch located in main control room.

Low condenser vacuum.

BE 2-33

WNP-2 AMENDMENT NO. 1.7-July 1981 TABLE II.E.4.2-2 (Continued)

"K" plus RHR/RCIC equipment are high temperature.

High drywell'ressure and low reactor pressure.

RHR equipment area high temperature.

Reactor vessel low water level (Trip 1)

Reactor water cleanup system high differential flow.

These are the isolation functions of the primary containment and reactor vessel isolation system; other functions are given for information only.

B.2-34

WNP-2 AMENDMENT NO. 17 July 1981, II F ~ 1 ADDITIONAL ACCIDENT-MONXTORXNG XNSTRVMENTATION (NUREG-0737)

Position (Introduction to II.F.1.1 through XI.F.1.6)

Item IZ.F.1 of NUREG-0660 contains the following subparts;

a. Noble gas effluent radiological monitor:

b. Provisions for continuous sampling of plant effluents for post-accident releases of radioac-tive iodines and particulates and onsite labora-tory capabilities (this requirement was inadvertently omitted from NUREG-0660; see that follows, for position); 'I.F.1.2

c. Containment high-range radiation monitor;

d. Containment pressure monitor;

e. Containment water level monitor; and

f. Containment hydrogen concenration monitor.

NUREG-0578 piovided the basic requirements associated with items (a) through (c) above. Letters issued to all operating -nuclear power plants dated September 13, 1979 and October 30, 1979 provided clarification. of staff requirements associated with items (a) through (f) above. ZI.F.1.1 through XZ.F.1.6 present the NRC position on these matters.

Zt is important that the displays and controls added to the control room as a result of this requirement not increase the potential for operator error. A human-factor analysis should be performed taking into consideration:

a.'he use of this information by an operator during both normal and abnormal plant conditions,

b. integration, into emergency procedures,

c. integration'nto operator training, and

d. other alarms during emergency and need for prioritization of alarms.

B.2-35

WNP-2 AMENDMENT NO. 17

'uly 1981 II.F'.1.1 Noble Gas Effluent Monitor.'osition Noble gas effluent monitors shall be installed with an

.extended range designed to function during accident conditions as well as during normal operating'onditions.- Multiple moni-tors. are considered necessary to cover the ranges..of interest.

a. Noble gas effluent monitors w'ith an upper range.

capacity of 105 wCi/cc (Xe-133) are considered to be practical and should be installed in all operating plants.

b. Noble gas-effluent monitoring shall be provided for the total range of concentration extending from normal condition (as low as reasonably achievable (AZARA) concentrations to a maximum of mCi/cc (Xe-133) . Multiple monitors are con- '05 sidered to be necessary to cover the ranges of interest. The "range capacity of individual moni-tors should-overlap by a factor'f ten.

Clarification a 0 Licensees shall provide continuous monitoring of high-level, 'post-accident releases of radioactive noble gases from the plant. Gaseous shall meet the requirements specified in effluent'onitors the Table II.F.1-1. Typical plant effluent path-ways to be monitored are also given in the table.

b. The monitors shall be capable of functioning both during and following an accident. System designs shall accommodate a design-basis release and then be'apable of following decreasing concentrations of noble gases.

Ce Offline monitors are not required for the FNR secondary side main steam safety valve and dump valve discharge lines. For this application, externally mounted monitors viewing the main steam line upstream of the valves are acceptable with procedures to correct for the low energy gammas the external monitor's would not detect.

Isotopic identification is not,required.

B.2-36

WNP-2 AMENDMENT NO. 29 March 1983

d. Instrumentation ranges shall overlap to cover the entire range of effluents from normal (ALARA) through 'accident conditions.

The design description shall include the follow-ing information s F

1. System description, including:

(i), instrumentation to be used, including range or sensitivity, energy dependence or response, calibration frequency and technique, and vendor's model number, applicable; if (ii) monitoring locations (or points of sampling), including description of methods used to assure representative measurements and background correction; (iii) method location of instrument readout(s) and of recording, including descrip-tion of the method or procedure for transmitting or disseminating the information or data; (iv) assurance of the capability to obtain readings at least every 15 minutes during and following an accident; and, (v) the source of power to be used.

2. Description of procedures or calculational methods to be used for converting instrument readings or release rates per unit time, based on exhaust air flow and considering radionuclide spectrum distribution as a function of time after shutdown.

WNP-2 Position WNP-2 is in the process of having extended range noble gas effluent monitors installed. Instrument ranges will be 10 to 103 v Ci/cc for the radwast and turbine genera-tor building and 10 5 to 104 ~Ci/cc for the reactor building (Xe-133 equivalent). The power supplies will be from uninterruptible power.

B. 2-37

WNP-2 AMENDMENT NO. 29 March 1983 Each elevated release duct, turbine building exhaust, and radwaste building exhaust is monitored for radioactive ef fluent gases by off -line systems. In addition, the elevated release duct has an in-line monitoring system.

The off-line systems draw samples from each exhaust duct through isokinetic probes. The sample passes through partic-ulate and charcoal f ilters, a sample flow control system, and, a radioactive gas monitor and is returned to the original exhaust duct. The sample flow rate is automatically adjusted to compensate for e ffluent flow changes ~ The sys tern is equipped with local flow rate indication and remote flow rate trouble alarms to the control room. Each monitoring system has two separate detectors and instrument loops. (See Figures 11.5-5 and 11.5-6.) The ranges of these detectors are:

-6 -1

a. low range (10 ~Ci/cc to 10 .uCi/cc Xe-133)

b. medium range (10 2 ~Ci/cc to 103 ~Ci/cc Xe-133)

The det ctors are set in lead shielding to reduce the unwanted background. The low-range detector is a beta-sensi-tive scintillator. The medium-range detector is a beta scintillator which measures the beta radiation in the sample chamber .

The in-line monitor provides high-range detection of six decades up to 105 ~ Ci/cc Xe-133. This is a pa ir of ion chambers set into the elevated release duct. Each instrument Loop contains a detector (ion chamber) a power supply, a ratemeter, alarm modules, and a recorder. (See Figure 11.5-10. )

The low-range channel is equipped with a radioactive test source while the medium-range channel has a built-in electronic test circuit.

(Also, see 11.5.2. 2.1.5, 11.5.2. 2.1.6, 11.5.2. 2.1.7, 11.5. 2. 2. 3. 2, and Table 11. 5-1. )

B. 2-37a

WNP-2 AMENDMENT NO. 17 July 1981 TABLE II.F.1-1 HIGH-RANGE NOBLE GAS EFFLUENT MONITORS REQUIREMENT Capability to detect and measure con-centrations of noble gas fission products in plant gaseous effluents during and following an accident. All potential accident release paths shyll be monitored.

PURPOSE To provide the plant operator and emergency planning agencies with information on plant releases of noble gases during and following an accident.

DESIGN BASIS MAXIMUM RANGE Design range values may be expressed in Xe-133 equivalent values for monitors employing gamma radiation detectors or in microcuries per cubic centimeter of air at standard tem-perature and pressure (STP) for monitors employing beta radiation detector (Note: 1R/hr 9 ft ~ 6.7 Ci Xe-133 equiva-1 lent for point source) . Calibrations with a higher energy source are acceptable. The decay of radionuclide noble gases after an accident (i.e., the distribution of noble gases changes) should be taken into account.

10 >Ci/cc Undiluted containment exhaust gases (e.g.g PWR reactor building purge, PWR drywell purge through the standby gas treatment system).

- Undiluted PWR condenser air removal system exhaust.

10 ~Ci/cc - Diluted containment exhaust gases (e.g., >10:1 dilution, as with auxiliary building exhaust air).

BWR reactor building (secondary containment) exhaust air.

PWR secondary containment exhaust air.

10 WCi/cc Buildings with systems containing primary coolant or primary coolant offgases (e.g., PWR auxiliary buildings, BWR, turbine buildings).

- PWR steam safety valve discharge, atmospheric steam dump valve discharge.

B.2-38

WNP-2 AMENDMENT NO'7 July 1981 TABLE IZ.F.1-1 (Continued) 102 WCi/cc Other release points (e.g., radwaste buildings, fuel handling/storage buildings).

REDUNDANCY Not required; monitoring the final release point of several discharge inputs is acceptable P SPECI- (None) Sampling design criteria per ANSI FICATIONS N13. 1 POWER SUPPLY Vital instrument bus or dependable backup power supply, to normal ac.

Calibrate monitors using ftdetectors to CALIBRATION gamma Xe-133 equivalent (1 R/hr 9 1 6.7 Ci Xe-133 equivalent for point source).

Calibrate monitors using beta detectors to Sr-90 'or similar long-lived beta isotope of at least 0.2 MeV.

DISPLAY Continuous and recording as equivalent Xe-133 concentrations or wCi/cc of actual noble gases..

QUALIFI- The instruments shall provide sufficiently CATION accurate responses to perform the. intended function in the environment to which they, will be exposed during accidents.

DESIGN Offline monitoring is acceptable for all CONSIDERA- ranges of noble gas concentrations.

TIONS Znline (induct) sensors are acceptable for 10 Abaci/cc, off1'ine monitoring is recommended.

Upsteam filtration and particulates) radioactive iodines (prefiltering to remove is not required; however, design should consider all alternatives with respect to capability to monitor effluents following an accident.

For external mounted monitors (e.g., PWR main steam line), the thickness of the pipe should be taken into account in accounting for low-energy gamma radiation.

B.2-39

NNP-2 AMENDMENT NO ~ 17 July 1981 3

II P.1. 2

~ ~ ~ Sampling. and Analysis of Plant Effluents Position

~ ~

Because iodine gaseous effluent mdn'itors for the accident con-dition are not considered to be practical at this time, capa-bility for effluent monitoring of radioiodines for the accident condition shall be provided with sampling conducted by 'adsorption on charcoal or oth'er media, followed by onsite laboratory analysis.

-Clarification a ~ Licensees shill provide continuous sampling of plant gaseous. effluent for post-accident releases of radioactive iodines and particulates to meet the requirements of Table II.P.1-2. Licensees shall also provide onsite laboratory capabilities to analyze or measure. these samples. This requirement should not be construed to prohibit design and development of radioiodine and par-ticulate monitors to provide online sampling and analysis for the accident. condition. If gross gamma radiation measurement techniques are used, then provisions shall be made to minimize noble gas interference.

b. The shielding design basis is given in Table II.P.1-2. The sampling system design shall be such that plant personnel could remove samples, replace sampling media and transport the samples

. to the onsite analysis facility with radiation exposures that are'ot in excess of th'e criteria of GDC 19 of 5 rem whole-body exposure and 75 rem to the extremities. during the duration of the accident.

C ~ The design of the systems for the sampling of particulates and iodines should provide for sample nozzle entry velocities which are approxi-mately isokinetic (same volocity) with expected induct or instack air velocities. Por accident conditions, sampling may be complicated by, a reduction in stack or vent effluent velocities to below design levels, making it necessary to substantially reduce sampler intake flow rates to achieve the isokinetic condition. Reductions in air flow may well be. beyond the capability of available sampler flow controllers to maintain isokinetic conditions; therefore, the staff will B.2-40

WNP-2 AMENDMENT NO. 29 March 1983 accept flow control devices which have the capa-bility of maintaining isokinetic conditions with variations in stack or duct design flow velocity of + 20%. Further departure from the is'okinetic condition need not be considered in design.

Corrections for non-isokinetic sampling condi-tions, as provided in Appendix C of ANSI basis.

13.1-1969, may be considered on an ad hoc Effluent streams which may contain air with entrained water, e.g., air e jector discharge, shall have provisions to ensure that the absorber is not degraded while providing a representative sample, e.g., heaters.

I WNP-2 Position A shielded low flow particulate/iodine sampling system is being installed to monitor the post-accident effluent air from the reactor building ventilation exhaust. The sample system consists of a shielded filter holder, a running time meter, and a low volume positive displacement sample pump which draws a sample from the reactor building elevated release duct. The sample is drawn through the particulate and iodine filter assembly at a rate of about 0.1 cfm and then returned to the release duct. The pump is automatically started when the high-high level alarm on the associated noble gas monitor is activated. The pump will continue to run until manually reset. The sample filter holder has a quick-release allowing it to be removed and handled with remote tools reducing any potential personnel exposures To protect personnel, a 2-inch thick lead shield is posi-tioned around the filter holder. This will reduce the dose rate to S mR/hr at one foot from the filter under worst post-ulated conditions following a reactor accident based on Table II.F.1-2, Design Basis Shielding Envelope.

The shielded sample assembly will be used on the reactor building because of potential leakage from primary contain-ment into the reactor building following a reactor accident releasing fission products from the core. Particulate/iodine sampling of the other buildings'radwaste and turbine) exhausts will be handled by the normal effluent samplers where the post-accident release concentration is quite low.

B. 2-41

WMP-2 AMENDMENT HO. 29 March 1983 If there

release, were a.reactor accident w1th a core f1ss1on product the reactor building (secondary containment) immedi-ately isolates. It's atmosphere is maintained at a 0 '5" H20 vacuum by the standby gas treatment system (SGTS). The only potential airborne contamination that could reach the other builings is from the SGTS by-pass <leakage as listed in Table 6.1-16 which totals 0.74 scfh of which 0.35 scfh is into the radwaste building. Assuming an iodine concentration of 3.7 X 104VCi/cc (50% core inventory is released in the drywell atmosphere and 50% plates out) in the in-leakage air to the radwaste building, then the building exhaust (83,000 cfm) concentration will be 2.o X 10 3 pCi/cc ing volume dilution is ignored. The normal if the build-effluent sampler operates at 3 cfm; therefore, the charcoal cartridge 30 minute accumulation would be 6.7 mCi. This would result in a dose rate of 21 mR/hr at one foot from the cartridge.

Doubling the dose rate to account for particulates yields 42 mR/hr at one foot from the sample assembly. Because of the high exhaust flow rate (260,000 cfm) and less in-leakage (0.24 scfh) t'e turbine building exhaust is less concen-trated. Therefore, the radwaste and turbine building normal effluent sampling systems are considered adequate for post-accident sampling. (See 11.5.2.2.1.5 and Figure ll.5-5.)

B. 2-4la

WNP-2 WNP-2 AMENDMENT NO. 17 July 1981 TABLE II.F.1-2 SAMPLING AND ANALYSIS OR MEASUREMENT OF HIGH-RANGE RADIOIODINE PARTICULATE EFFLUENTS IN GASEOUS EFFLUENT STREAMS EQUIPMENT Capability to collect and analyze or measure representative samples of radioactive iodines and particulates in plant gaseous effluents during and following an accident. The capa-bility to sample and analyze not for radioiodine required for and particulate effluents is PWR secondary main steam safety valve and dump valve discharge lines.

PURPOSE To determine quantitative release of radioiodines and particulates for dose calcu-lation and assessment.

DESIGN BASIS 10 WCi/cc of gaseous radioiodine and par-SHIELDING ticulates, deposited on sampling media; 30 ENVELOPE minutes sampling time, average gamma energy (E) of 0.5 MeV.

SAMPLING MEDIA Iodine > 90% effective adsorption for all forms of gaseous iodine Particulats > 90% effective retention for 0.3 micron (g) diameter particles.

SAMPLING CONSIDERATIONS Representative sampling per ANSI N13.1-1969.

Entrained moisture in effluent stream should not degrade adsorber.

Continuous collection required whenever exhaust flow occurs.

Provisions for limiting occupational dose to personnel incorporated in sampling systems, in sample handling and transport, and in analysis of samples.

ANALYSIS Design of analytical facilities and preparation of ana-lytical procedurs shall consider the design basis sample.

B.2-42

WNP-2 AMENDMENT NO. 1 7 July 1981 TABZE II.P.1-2 (Continued)

Highly radioactive samples. may not be compatible with generally accepted analytical procedures;,in such cases,-

measurement of emissive gamma radiations and,'the use of shielding and distance'actors should be considered in design.

B.2-43

WNP-2 AMENDMENT NO.

17'uly 1981 II.P.1.3 Containment High-Range Radiation Monitor Position In containment radiation level monitors with a maximum range of 10 rad/hr shall be installed. A minimum of two such moni-tors that are physically separated shall be provided.

Monitors shall be developed and qualified to function in an accident environment.

Clarification

a. 'rovide.two radiation monitor systems in containment which are documented to meet the requirements of Table XI.P.1-3.

b. The specification of 108 rad/hr in the above position was based on a calculation of post-ac-cident containment radiation levels that included both particulate (beta) and photon (gamma) radiation. A radiation detector that responds to beta and gamma radiation cannot be qualified 'oth to post-LOCA (loss-of-coolant accident) contain-ment environments but gamma-sensitive instruments can be so qualified. -In order to follow the course of an accident, a containment monitor that measures only gamma radi'ation is adequate. The requirement was revised in'he October 30, 1979 letter to provide for a p)eton-only measurement with an uPPer range of 10 R/hre

c. The monitors shall be located in containment(s) in a manner as,to provide a reasonable assessment of area radiation conditions inside containment.

The monitors shall be widely separated so as to provide independent measurements and shall "view" a large fraction of the containment volume.

Monitors should not be placed in areas which are protected by massive shielding and should be reasonably accessible for replacement, main-tenance, or calibrati'on.'laicement high in a, reactor building dome is not recommended because of potential maintenance difficulties.

d. Por BWR Mark XII containments, two such monitoring systems should be inside both the primary con-tainment (drywell) and the secondary containment.

e. The monitors are required to respond to gamma photons with energies as low as 60 keV and to B.2-44

WNP-2 AMENDMENT NO. 17 July 1981 provide an essentially 'flat response for'amma energies'etween 100 keV and 3 MeV, as specified, in. Table II.F.1-3. Monitors that use thick shielding to increase the upper range will underestimate post-accident radiation levels in containment by several, orders of magnitude because of their insensitivity to low energy gam-mas and are not acceptable..

WNP-2'osition WNP-2 concurs with the intent of this task and is in the pro-cess of having high range radiation monitors installed'in pri-mary containment. The monitors will be redundant, seismically and environmentally qualified to Class I requirements and will operate in a design basis accideng environment. The monitors will have a range of 1 R/hr to .10'/hr. The complete design description will be provided prior to April 1982.

'E 2-45

WNP-2 AMENDMENT NO. 17

'uly 1981 TABLE IX ~ F.1-3 CONTAINMENT HIGH-RANGE RADIATION MONXTOR REQUIREMENT The capability to detect and measure the radiation level within the reactor contain-ment during and following an accident.

RANGE 1 rad/hr to 10 rads/hr (beta and gamma) or alternatively R/hr to 107 R/hr (gamma 1

only).

RESPONSE 60 keV to 3 MeV photons, with linear energy response + 20%) for photons of 0.1 MeV to 3 MeV. Instruments must be accurate enough to provide usable information.

REDUNDANT A minimum of two physically separated moni-tors (i.e., monitoring widely separated spa-ces within containment).

DESIGN AND Category 1 instruments as described in QUALIFICATXON Appendix A, except as listed below.

SPECIAL Xn situ calibration by electronic signal CALIBRATXON substitution is acceptable for all range decades above 10 R/hr. In situ calibration for at least one decade below 10 R/hr shall be by means of calibrated radiation source.

The original laboratory calibration is not an acceptable position due to the possible dif-ferences after in situ installation. For high-range calibration, no adequate sources exist, so an alternate was provided.

SPECIAL Calibrate and type-test representative speci-ENVIRONMENTAL mens of detectors at sufficient points to QUALIFICATIONS demonstrate linearity through all scales up to 106 R/hr. Prior to initial use, certify calibration of each detector for at least one point per decade of range between 1 R/hr and 103 R/hr.

B.2-46

AMENDMENT NO 17

.July 1981 II.P.1.4 . Containment Pressure Monitor Position A continuous indication of containment pressure shall be pro-.

vided in the control room of each operating reactor.

Measurement and indication capability shall include three times the design pressure of the containment for concrete, four times the design pressure for steel> and -5 psig for all containments.

Clarification

a. Design and qualification criteria are outlined in Appendix A.

b. Measurement and indication capability shall extend to 5 psia for subatmospheric containments.

c. Two or more instruments may be used to meet requirements. However, instruments that need to be switched from one scale to another scale to meet the range requirements are not acceptable.

d. Continuous display and recording of the contain-ment pressure over the specified range in the control room is required.

e. The accuracy and response time specifications of the pressure monitor shall be provided and justified to be adequate for their intended function.

WNP-2 Position WNP-2 concurs with the intent of. this task and is in the pro-cess of modifying the primary containment pressure instrumen-tation. Once modified the redundant instruments will have a range of 12 psia to 180 psig per Regulatory Guide 1.97, and, will be seismically and environmentally qualified to Class 1

,requirements. Continuous recording will be 'available in the control room. The complete design, description will. be pro-vided prior.to January 1982.

B.2-47

WNP-2 AMENDMENT NO. 17 July 1981 II.F.1.5 Containment Water Level Monitor Position A continuous indication of containment water level shall be provided in the control room for all plants. A narrow range instrument shall be provided for PWRs 'and cover the range from the bottom to the top of the containment sump. A wide range instrument shall also be provided foi PWRs and shall, cover the range from the bottom of, the containment to the elevation equivalent 'to a 600,000 gallon capacity. For BWRs, a wide range instrument shall be. provided and cover the range from

'he bottom to 5 feet above the normal water'level of the pool. "'suppression Clarification

a. -The containment wide-range water level indication channels shall meet the design and qualification criteria as outlined in Appendix A. The narrow-.

range channel shall meet the requirements of Regulatory Guide 1.89.

b. The: measurement capability of 600,000, gallons is based, on recent plant designs. For older plants with smaller water capacities, licensees may pro-pose deviations .from this requirement. based on .

the available water supply capability at their plant.

c. Narrow-range water level monitors are required for all'izes of sumps but are not required in

-those plants that do .not contain sumps inside the containment.

d. For BWR pressure-suppression containments, the emergency'core cooling system (ECCS) suction line inlets may be .used as a starting 'reference point for the narrow-range and 'wide-range water level monitors, instead of the bottom of the suppression pool.

e. The accuracy requirements of the water level monitors shall be provided and justified to be adequate for thei'r intended function.

WNP-2 Position WNP-2 concurs with the intent of this task and is in the pro-cess of having the suppression chamber water level instrumen-B.2-48

AMENDMENT NO. 17 July 1981 tation- modified. Once modified the instruments will be seismically and environmentally qualified to Class 1 require-ments, redundant, with readout in, the control room.

Instrument range will be from the center line of the HPCS suc-tion line to 5 ft. above normal water level. The complete design description will be provided prior to January 1982.

BE 2-49

WNP-2 AMENDMENT NO ~ 17

, July 1981

/ '

II.P;1.6 Containment Hydrogen Monitor 1 c

F, r'

Position 4 t

A continuous indication of hydrogen concentration in the con-tainment atmosphere shall be provided in the control room.

Measurement capability shall be provided; over the'ange of 0 to 10%. hydrogen concentration under both 'positive and 'negative-ambient pressure.

t Clarification

a. Design and qualification criteria are outlined in Appendix A.

b. . The continuous indication of hydrogen con-centration is not required during normal operation.

If an indication is not available at all times, continuous indication and recording shall be functioning within 30, minutes of the initiation of safety injection.

c. The. accuracy and placement of the hydrogen moni-tors shall be provided -and justified to be ade-( quate for their intended function.

P WNP-2 Position h

I WNP-2 concurs with the intent of this position. The existing monitors are. redundant providing continuous display and redun-dant recording in the control room. The instruments are seismically and environmentally qualified to Class 1 require-ments with a range of 0-30% hydrogen concentration. Complete

! design description, will be provided prior to January 1982.

B.2-50

WNP-'2 ANENDHENT NO ~ 17 July 1981 II.F.2 INSTRUHENTATZON FOR DETECTION OF INADEQUATE CORE COOLING Position Licensees shall provide a description of any additional instrumentation or controls (primary or backup) proposed for the, plant to supplement existing "instrumentation (including primary coolant saturation monitors) in order to provide an unambiguous, easy-to-interpret indication of inadequate core cooling (ICC). A description of the functional design requirements for the system shall also be included. A description of the procedures to be used with'the proposed equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided (NUREG-0737).

Clarification None WNP-2 Position a

( WNP-2 concurs with the intent of this task and is implementing a design study to supplement the existing reactor vessel water level instrumentation. The design study is considering installation of additional, redundant channels of water level indication. The indication will extend from below the active core to just below the vessel head flange. Per NUREG-0737, a complete design description will be forwarded for review to the NRC prior to January 2, 1982.

B.2-51

WNP-2 AMENDMENT NO ~ 17 July 1981 II.-K.1.5 Assurance. of Proper ESF Functioning Position Review all valve positions, positioning requirements, positive controls and related test and maintenance procedures 'to assure proper ESF functioning. See NRC Bulletins79-06A Item 8,79-06B Item 7, and'9-08 Item 6.

This requirement, shall be met, before fuel loading.

Clarification None WNP-2 Position Directives on valve positioning requirements, positive controls, and test 'and maintenance procedures associated with ESF systems are. being prepared. Motor-operated valves in safety systems will normally be maintained in a configuration such as to require the least number of valve automatic move-ments upon system actuation. System initiation logic is such that valves automatically move to the required position when required. The, position of vital manual ECCS valves is controlled .by the use, and documentation of locks on valve handwheels. Xn addition, numerous vital manual valves have position, status indicating lights in the WNP-2.Control Room.

WNP-2 will be equipped with ESF system status displays which continuously monitor the ESF systems and provide 'indication to operator of a. system bypass or inoperability introduced 'he during testing or maintenance which renders the system(s) unable to respond to an initiation signal. Typical parameters monitored inlcude:

a. Valve position

b. Power available to motor-operated valves

c. Initiation logic power available

d. Power'ources (including emergency diesels) availab'le

e. Breaker status Alarms are provided on a system level basis. Indication is provided on a component level and testing piocedures for ESF systems will basis.'urveillance include checks to ensure the system is returned.to standby status upon completion of testing.

V B.2-52

WNP 2 AMENDMENT NO 17 July 1981 When ESF equipment is removed from service for maintenance, WNP-2 procedures require documentation of removal and return to service-. Functional tests of equipment returned to service following maintenance are requiied, by these procedures to ensure operability.

B.2-53

NNP 2 AMENDMENT NO ~ 17 July 1981 t

II.K.1.10 Safety-Related

~ ~ System Operability Status Assurance Position Rev'iew.,and modify,= as required, procedures for removing safety-related systems from service and (restoring to service) to assure operability status is known. See Bulletins79-05A Item 10,79-06A Item '10, 79>>06B Item 9, and 79-08 Item 8.

This requirement shall be'et before fuel loading.

Clarification None WNP-2 Position

\

Refer to the responses for Items II.K.1.5 and I.C.6 for compliance with the intent of this position.

In addition, all safety-related systems removed from service (and restored) will be recorded in appropriate plant logs and on applicable equipment status boards.

~

~

BE 2-54

WNP 2 AMENDMENT NO;,17 Jul'y 1981 r

IZ.K. 1.22 Proper Functioning "of Heat Removal Systems Position Describe the actions, both automatic and manual, necessary for proper functioning of the auxiliary heat removal systems RCIC) that are used when the main feedwater.system is 'e.g.,

not operable. For'ny'anual action necessary, describe in summary form the procedure by which this;action i's taken -in a timely=sense. (ZE Bulletin 79-08).

Clarification None WNP-Position WNP-2 letter G02-,80-107 of May '23, 1980, responding to ZE

.Bulletin 79-08, provided the following:

The auxiliary heat removal systems provided to remove decay heat from= the reactor core and containment following loss of the feedwater systems are:

High Pressure Core Spray System (HPCS)

.Reactor Core Isolation Cooling (RCZC) System, Low Pressure Core Spray System (LPCS)

RHR System .LPCI Mode RHR System - Suppression Pool Spray Mode RHR System - Suppression Pool Cooling Mode Residual Heat Removal (RHR/Low Pressure Coolant Injection (LPCZ) System The description that follows details the operation of the systems needed to achieve initial core cooling followed by containment cooling and then followed by extended core cooling for long-term plant -shutdown, assuming the reactor is scrammed and isolated from the main condenser.

Initial, Core Cooling:

Following a loss of feedwatr and reactor scram, a low'eactor water level signal (level 2) will automatically initiate main steam line isolation valve closure. At the same time this signal will put the HPCS and RCIC Systems into the reactor coolant makeup injection mode. hese systems will continue to inject water into the vessel until a high water level signal (level 8) automatically t ips RCZC and closes the HPCS injection valve. The HPCS pump -emains running on minimum flow bypass.=

B.2-55

WNP-2 AMENDMENT NO. 17 July 1981 Following a high reactor water level 8 trip, the HPCS injec-tion valve will automaticaly reopen when 'reactor water level decreases to low water level 2. The RCIC System must be manually reset by the operator in the control room. before it will automatically re-initiate after a high water level 8 trip.

The HPCS and RCIC Systems have redundant supplies of water.

Normally they take suction from the condensate storage tank (CST). The HPCS and RCIC System suctions will automatically transfer from the CST to the suppression pool if the CST ater is depleted or the suppression pool water level increases to a high level.

The operator can manually initiate the HPCS and RCIC Systems from the control room before the level 2 automatic initiation level is reached. The operator has the option of manual control after automatic initiation. The operator can verify.

that these systems are delivering water to the reactor vessel by:

a. Verifying reactor water level increases when systems initiate.

b. Verifying systems flow using flow indicators in the control room.

c. Verifying system flow is to the reactor by checking control room position indication of motor-operated valves. This assures no diversion of system flow to other than the reactor.

Therefore, the HPCS and RCIC can maintain reactor water level at full reactor pressure and until pressure decreases to where low pressure systems such as the Low Pressure Core Spray (LPCS) or Low Pressure Coolant Injection (LPCI) can maintain water level.

Containment Cooling:

After reactor scram and isolation and establishment of satis-factory core cooling, the operator would start containment cooling. This mode of operation removes heat resulting from safety/relief valve (SRV) discharge to the suppression pool.

This would be accomplished by placing the Residual Heat Removal (RHR) System in the containment/suppression pool cooling mode, or the suppression pool spray mode, i.e., RHR suction from and discharge to the suppression pool.

B 2-56

WNP-2 AMENDMENT NO.'7 July 1981 The Operator could verify proper operation of the RHR system containment cooling function from the control room 'by:

~

a. Verifying RHR and Service Water (SW) .system flow, usinng system control room flow indicators.

b. Verifying correct. RHR and SW system flow paths using control room position indication of motor-operated valves.

c. On branch lines that-could divert =flow from the required flow paths, closing the motor-operated valves and noting the effect on RHR and SW flow rate.

When the reactor has been depressurized, the RHR system can be placed in the long-term shutdown cooling mode. The operator manually terminates the containment coling mode of one of the RHR loops and places the loop in the shutdown cooling mode as follows:

a., Trip the RHR pump to be used'or shutdown cooling,

b. Close .associated motor-operated valve in the suppression pool suction and LPCZ discharge line to the vessel, c.'pen shutdown cooling suction valves'from valves to the reactor vessel, and and'ischarge

d. Restart -the RHR Pump.

In this operating mode, the RHR system, can cool the reactor to cold shutdown. Proper operation and flow paths in this mode can be verified by methods similar to= those described for the containment cooling mode.

• In conclusion, the WNP-2 plant design is fully adequate to meet the intent of the requirements of auxiliary heat removal when= the main feedwater system is inoperable.

B.2-57

WNP-2 AMENDMENT NO. 17 July 1981 II.K.1.23 Reactor Vessel Level Instrumentation Position Describe all uses and types of ves'sel level indication'for both automatic and manual initiation of safety systems.

Describe other redundant instrumentation which the operator might have to give the same information regarding plant sta-tus. Instruct operators to utilize other available infor-mation to initiate safety systems (IE Bulletin 79-08).

Clarification None WNP-2 Position NEDO-24708 describes the multiple water level instrumentation provided in the BWR control room for the operator. An outline of the specific indication for WNP-2 is provided in the following paragraphs, which fully meets the intent of the plant requirements.and the NRC requirements.

Reactor vessel water level in the WNP-2 BWR is continuously monitored by eleven (11) indicators or recorders for normal,

.transient and accident conditions.. In general, those monitors used to provide manual safety equipment initiation are arranged in a redundant array with two 'instruments, one in each of two independent- electrical divisions. Thus, adequate information is provided to the operator for man'ual initiation of safety actions and for assurance of the vessel water level at all times.

Those sensors used to provide automatic safety equipment ini-tiation are arranged in a four quadrant vessel tap con-figuration with the four sensors divided electrically between two divisions.

In addition, the operating procedurs will reflect the requirements for the operators to also rely upon the infor-mation provided by other plant parameter indications relating to vessel level.

The range of reactor vessel- water level from below the top of the active fuel area up to the top of the vessel is covered by a combination of narrow and wide-range-instruments. Level is indicated and/or recorded in the control room..

BE 2-58

WNP-2 AMENDMENT NO 17 July 1981 A separate set (to that described above) of narrow-range and wide-range level instrumentation on separate condensing cham-bers provides reactor l~vel control via the reactor feedwater system. This set also indicates or records in the control room (three narrow-range level indicators and one wide-range level recorder).

The safety-related systems or functions served by safety-related reactor water level instrumentation are:

Reactor Core Xsolation Coolant System (RCXC)

High Pressure Core Spray System (HPCS).

Low Pressure Core Spray System (LPCS)

Residual Heat Removal/Low Pressure Coolant Inje ction (RHR/LPCI)

Automatic Depressurization System (ADS)

Nuclear Steam Supply Shutoff System (NSSSS)

Reactor Protection System (RPS)

Standby Gas Treatment System (SGTS)

Emergency Power System Secondary Containment Isolation Main Control Room and Critical Switchgear HVAC

.Standby Service Water System Containment Xnstrument Air System Trip of Non-essential Loads The modifications resulting from TMZ Task II.F.2 will change the number of level detectors but will not change the intent of the response to the task.. Upon completion of the- modifica-tions for NUREG Task ZI.F.2 this response will be updated to reflect the changes due to XI.F.2.'-

Low reactor vessel water level is used in the initiation logic of all systems listed above. Zn addition, the RCZC and HPCS systems shutdown on high reactor vessel water level. HPCS will automatically restart reached. (See response to if TMZ low reactor level is again Item IX.K.1.22 for further discussion.,) In the case of RCIC, manual resetting is required if high reactor vessel water level is reached.

BE 2-59

NNP-2 AMENDMENT NO ~ 17 July 1981 IZ.K.3.3 Failure of PORV or Safety Valve to Close Position Assure that any failure of a PORV or safety valve to close will be reported to the NRC promptly. All challenges to the PORVs or safety valves should be documented in the annual report.

This requirement shall be met before issuance of a full power license (NUREG-0694).

Clarification None WNP-2 Position Administrative procedures will be prepared to meet the intent of the above and implemented at WNP-2 prior to July 1982; The procedures will be available onsite for NRC Z&E review.

B.2-60

WNP-2 AMENDMENT NO ~ 17 July 1981 II.K.3. 13 Separation of HPCX and RCIC System Initiation Levels

a. Analysis

b. Modi fy Position Currently, the reactor core isolat'on cooling (RCIC) system and the high pressure coolant injection (HPCI) system both initiate on the same low water level signal and both isolate on the same high water level signal. The HPCX system will restart on low water level but the RCIC system will not. The RCIC system is a low-flow system when compared to the HPCX system. The initiation levels of the HPCI and RCXC systems should be separated so that the RCIC system initiates at a higher water level than the HPCI system. Further,'he RCXC system, initiation logic should be modified so that the RCIC system will restart on low water level. These changes have the potential to reduce the number of challenges to the HPCI system and could result in less stress on the vessel from cold water injection. Analyses should be performed to evaluate these changes. The analyses should be submitted to the NRC staff and changes should be implemented analyses (NUREG-0737).

if justified by the

a. Documentation provided results of evaluation and proposed modifications (if necessary) to staff by October 1, 1980. Provide sufficient supporting analysis to demonstrate that the systems, as modified,'ould not degrade proper system func-tions.

b. Modifications shall April be completed (if necessary) by 1, 1981.

See letter September 5, 1980, Enclosure 2, pg. 7 (Reference 33).

Clarification None WNP-2 Position At WNP-2'he HPCS and RCIC are both initiated a low-water level 2 (477.5 inches above vessel zero).

As a generic item, the possible separation of initiation levels for RCIC and HPCS was studied by GE for the BWR The results of that study were provided to the Owners'roup.

WNP-2 AMENDMENT NO 17 July 1981 Commission in a GE letter of October 'l, 1980 from R. A.

Buchholz to D. G. Eisenhut. The conclusions of, that study are endorsed by WNP-2, specifically, that the proposed separation of -RCIC and HPCS. initiation is "unnecessary.* The basis is that for rapid level changes associated with accident scenarios and severe transients their initiation would be essentially'simultaneous in that possible separation distan-ces could not preclude HPCS challenges; likewise, for slow level, changes due to small leaks or slow transients, adequate time e'xists for manual initiation of'CIC by the reactor operator, prior to HPCS auto-initiation. Justification of.

this basis is that over the lifetime of a unit, the expected occurrence of 'slow level decreases does not warrant installa-tion of equipment changes to decrease the number of challenges made to the HPCS. Manual operator response to maintain water level is consistent with abnormal operations demands.

GE and the BWR Owners'roup have submitted an analysis of the RCIC system for automatic reset following a high water level trip. The Owners'roup 'has recommended automatic closure steam supply valve on high vessel water level with the of'he RCIC turbine trip valve remaining open. This leaves the system in a standby mode capable of restarting at low water level. Very little modification required .to effect this change.

of existing circuitry is WNP-2 endorses the and will make the necessary'quipment modifica-Owners'roup'position tions. Changed circuit and logic diagrams for these changes will be 'available for review by November 12, 1981.

8.2-62

AMENDMENT NO. 17 .

July 1981 II.K.3.15 Modify Break-Detection logic to Prevent Spurious Isolation of High Pr'essure Coolant Injection and

. Reactor Core Isolation Cooling Position R

The high pressure coolant injection (HPCI) and'reactor'ore isolation cooling (RCIC) systems use, differential pressure sensors on elbow taps 'in the steam lines to their turbine drives to detect and isolate pipe breaks in the systems. The pipe break detection circuitry has resulted in spurious isola-tion of the HPCI and RCIC systems due to the pressure spike which accompanies startup of the, systems. The pipe break ,

dete'ction circuitry should be modified so that'ressure spikes resulting from HPCI and RCIC system initiation will not cause inadvertent. system isolation (NUREG-0737).

Clarification None WNP-2 Position WNP-2 does not have a steam-driven HPCI system, but instead has a motor-driven HPCS system for which- this modification does not apply. However, WNP-2 concurs with the intent of this position for the RCIC and will modify the RCIC pipe break detection circuitry to add a time delayed inhibit:to the isolation signals.: This minor change will eliminate the potential for isolation of the RCIC system due to the pressure spike caused by system startup.

A description of, the modification. as installed will be pro-vided prior to, november 12, 1981.

BE 2-63

WNP-2 AMENDMENT NO.'7 July 1981.

. II.K;3.16 Reduction of Challenges and Failures of Relief Valves - Feasibility Study and System-Modification Position The record of relief valve failures to close for all boiling water reactors (BWRs) in the past 3 years of plant operation is approximately 30 in 73 reactor-years (0.41 failures per reactor-year). This has demonstrated that the failure of a relief valve to close would be the most likely cause of a small-break loss-of-coolant accident (LOCA). The high failure rate is the result of a high relief valve challenge rate'nd a relatively high failure rate per challenge (0.16 failures per challenge). Typically, five valves are challenged .,in each event. This results in an equivalent failure rate per challenge of 0.03. The challenge and failure rates can be reduced in the following ways:

a. Additional anticipatory scram on loss of feedwater,

b. Revised relief valve actuation setpoints,

c. Increased emergency core cooling (ECC) flow,

d. Lower operating pressuresg

e. Earlier initiation of ECC systems, f.'eat removal through emergency condensers,

g. Offset valve setpoints, to open fewer valves per challenge,

h. Installation of additional relief valves with a block or isolation valve feature to eliminate, opening of the safety/relief valves (SRVs), coh-sistent with the ASME Code,

i. main Increasing the high steam line flow setpoint for steam line isolation valve (MSIV) closure,

j. Lowering the pressure setpoint for'SIV closure,

k. Reducing the testing frequency of the MSIVs,

l. More stringent valve leakage criteria, and

m. Early removal of leaking valves.

B.2-64

WNP-2 AMENDMENT NO ~ 17 July 1981 An investigatin of the feasibility and contra-indications of reducing challenges to the relief valves by use of the afore-mentioned methods should be conducted. Other methods should also be included in the feasibility study. Those changes which are shown to reduce relief valve challenges without compromising the performance of the relief valves or other systems should be implemented. Challenges to the relief valves should be reduced substantially (by an order of magnitude).

Clarification Failure of the power-operated relief valve (PORV) to reclose during the TMI-2 accident resulted in damage to the reactor.

core. As a consequence, relief valves in all plants, including BWRs, are being examined with a view toward their possible role in a small<<break LOCA.

The safety/relief valves (SRV) are dual-function pilot-operated relief valves that use a spring-actuated pilot for the safety function and an external air-diaphragm-actuated pilot for the relief function.

The operating history of the SRV has been poor. A new design, is used in some plants but the operational history is too brief to evaluate the effectiveness of the new design. Another way of improving the performance of the valves is to reduce the number of challenges to the valves. This may be done by the methods described above or by other means. The feasibility and contra-indications of reducing the number of challenges to the valves by the various methods should be studied. These changes which are shown to decrease the number of challenges without compromising the performance of the valves or other systems should be implemented.

The failure of an SRV to reclose will be the most probable cause of a small-break LOCA. Based on the above guidance and

,clarification, results of a detailed evaluation should be sub-mitted to the staff. The licensee shall document the proposed system changes for staff approval before implementation.

WNP-2 Position WNP-2 is a p'articipant in the BWR Owners'roup and endorses the position prepared by General Electric and the In summary this position shows that a significant Owners'roup.

reduction in relief valve challenges and failures is attained over industry history by the BWR five (5) design and installa-tion of Crosby safety/relief valves instead of the three-stage Target Rack valves. WNP-2, as a BWR five (5) design with B.2-65

WNP-2 AMENDMENT"NO 1 7 July 1981 Crosby relief valves, realizes approximately a factor of 12 reduction over industry history. Additionally, the position shows that further design modification will not substantially change the challenge and/or failure rate for the WNP-2, BWR five (5) Crosby SRV design.

This position was presented by GE to the NRC on behal'f of the Owners'roup in letter from D. B. Waters to D. G. Eisenhut, dated March 31, 1981, "BWR Owners'roup Evaluation of NUREG-0737 Requirements Items II.K.3.16 and II.K.3..18."

BE 2-66

WNP-2 AMENDMENT NO. 17 July 1981

.II.K.3.17 Report on Outages of Emergency Core Cooling Systems Licensee Report and Proposed Technical Specification Changes Position Several components of the emexgency core cooling (ECC) systems are permitted by'echnical specifications to have sub'stantial outage times (e.g., 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for one diesel-generator; 14 days for the HPCI system). In addition, there'are no cumulative outage time limitations for ECC systems. Licensees should submit a report detailing outage dates and lengths of outage for all ECC systems for the last 5 years of operation. The report should also include the causes of the outages (i.e.,

controller failure, spurious isolation) (NUREG-0737).

Clarification The present technical specifications contain limits on allowable outage times for ECC systems and components.

However, there are no cumulative outage time limitations on these same systems. "

It is possible that ECC equipment could meet present technical specification requirements but have a high unavailability because of frequent outages within the allowable technical'pecifications.

The licensees should submit a report detailing outage dates and length of outages for all ECC systems for the last 5 yeaxs of operation, including causes of the outages. This report will provide the staff with the quantification of historical unreliability due to test and maintenance outages, which will be used to determine if a need exists for cumulative outage requirements in the technical specificatio'ns.

Based on the above guidance and clarification,-a detailed report should be submitted. The report should contain (1) outage dates and duration of outages; (2) cause of the outage; (3) ECC systems or components involved in the outage'; and (4) corxective action taken. Test and maintenance outages should be included in the above listings which are to cover the last 5 years of operation. The licensee should. propose changes to improve the availability of ECC equipment, if needed.

Applicant fox an operating license shall establish a plan to meet these requirements.

WNP-2 Position The position statement by NRC re"uired a report 'to be sub-mitted by operating plants to identify actual ECCS outage B.2-67

WNP-2 AMENDMENT NO. 17 July 1981 experience for, the last five years of operation. The report requirement does not apply to WNP-2 since no operating experience is available. The Tech Specs for WNP-2 will be prepared and submitted- for NRC review and approval based on currently acceptable ECCS outage times, by December '30, 1981.

B.2-68

WNP-2, AMENDMENT NO. 1,7 July 1981 II.K.3.18 Modification of Automatic Depressurization System Logic-Feasibility for. Increased Diversity for Some Event Sequences

\

Position

'I The automatic depressurization system (ADS) actuation logic ..

should be modified to eliminate the need for manual actuation

'to assure adequate core, cooling. A feasibility and risk assessment study is required to'etermine the optimum approach. One possible scheme that should be considered is ADS actuation on low -reactor vessel water level provided no high pressure coolant inj'ection (HPCI) or high pressure coolant system (HPCS) flow exists and a low pressure emergency core cooling (ECC) system is running. This logic would complement, not replace, the existing',ADS actuation logic.

Clarification 1

None WNP-2 Position WNP-2 is a member of the BWR Owners', Group which has just recently completed the"required feasibility and risk assessment study. The results of th's study have, been transmitted in a letter from GE to NRC, D. B. Waters to D. G.

Eisenhut, dated March 31,1981. In thi's study, five ADS logic options, including retaining the current design, were con-sidered. WNP-2 has evaluated these options and has conclude'd that the current ADS, logic design, with implementation of the symptom-oriented emergency=procedure guidelines (EPGs), is adequate. As pointed out in the Owners'roup study, the tran-sients that are of concern are slow developing events, and with the incorporation of the EPGs the operator (under worst conditions) has 30 to 40 minutes in which to act. WNP-2 believes this provides the operator with sufficient time to evaluate the situation (using the EPGs) and take the necessary action.

BE 2-69

WNP-2 AMENDMENT NO. 17 July 198,1 II.K.3.21 Restart of Core Spray and Low Pressure Coolant Injection Systems Position

't The core spray and low pressure coolant injection (LPCI) system flow, may be stopped by the operator; These systems will not restart automatically.on loss of water level initiation signal is still if present. The core spray and IPCI an system logic should be modified so that these systems will restart, if required, to assure adequate this design modification affects core cooling.

Because several core cooling modes under accident conditions,'- preliminary design should be submitted for staff review and approval prior to making the actual'odif ication.

Clarification Modification of system design should be made in accordance with those, requirements set forth'n*Sections 4.12, 4.13, and

., 4.16 of IEEE Standard 279-1971 with regard to protective bypasses and completion of protective action once, func-'ion

-initiated.

WNP-2 Position WNP-2 as a participant in the BWR Owners'roup endorses the position presented in the letter dated,December"29, 1980 from D. B. Waters to the NRC (attention D. G. Eisenhut), subject:

"BWR Owners'roup Evaluation of NUREG-0737 Requirements".

The position presented in enclosure 2 to this letter conclu-des that the current system design is adequate and no design changes are required. WNP-2 concurs in this position.

B.2-70

i WNP-2 AMENDMENT NO. 17 July 1981 II.K.3.22 Automatic Switchover of Reactor Core Isolation Cooling System Suction-Verify Procedures and Modify Design Position The reactor core isolation cooling (RCXC) system takes suction from'he condensate storage tank with -manual switchover to the suppression pool when the condensate storage tank level is low. This switchover should be made automatically. Until the ~

automatic switchover is implemented, licensees should,verify that clear and cognizant procedures exist for the manual switchover of the RCIC system suction from the condensate storage tank to the suppression pool.

Clarification None WNP-2 Position WNP-2 committed to automatic RCIC suction transfer in response to WNP-2 docket Question 031.015. Per that commitment, WNP-2 is in the process of installing redundant, seismic and quality Class I level switches on the Seismic Class I RCIC suction piping. These switches will sense the loss of the condensate=

storage tank as a supply an'd switch the RCIC suction to the-suppression pool automatically. Complete design description will be provided. prior to,November 12, 1981.

B ~ 2-7 1

WNP-2 AMENDMENT NO. 17 July 1981 II.K.3.24 Confirm Adequacy of Space Cooling for High Pressure Coolant Injection and Reactor Core Isolation Cooling Systems Position Long-term operation of the reactor core isolation cooling

,(RCZC) and high pressure coolant in)ection (HPCI) systems may require space cooling to maintain the pump room temperatures within allowable limits. Licensees should verify the accept-ability of the consequences of a complete loss of alternating-current power. The RCZC and HPCZ systems should be designed to withstand a complete loss of offsite alternating-current power to their support systems, including coolers, for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Clarification None WNP-2 Position WNP-2 provides an emergency space cooling system to those equipment rooms containing the HPCS and RCIC pumps. Each emergency space cooling system, is powered from a Class 1E bus and is designed to withstand the effects of a safe shutdown earthquake. Loss of offsite power will not affect the abil-ity 'of the space coolers to operate. See FSAR 9.4.9 for a description of the capabilities of the Reactor Building Emergency Cooling system. The cooling water to the system is supplied from the standby service water system. Loss of off-site power will not affect the ability of the standby service water system to provide cooling water. See FSAR 9.2.7 and 6.2.2 for a description of the capabilities of the standby service water system.

In conclusion, loss of offsite alternating-currentpower will not cause adverse space cooling conditions for the RCZC and HPCS pump rooms.

No further WNP-2 action is required to comply with the intent of this task.

BE 2-72

WNP 2 AMENDMENT NO ~ 17 July 1981 Il.K.3. 25 Ef feet of Loss of Alternating-Current Power on Pump Seals Position The licensees should determine, on a plant-specific basis, by analysis or experiment, the consequences of a loss of cooling water to the reactor recirculation pump seal coolers. The pump seals should be designed to withstand a complete loss of alternating-current (ac) power for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Adequacy of the seal design, should be demonstrated.

, Clarification The intent of this position is to prevent excessive, loss of reactor coolant system (RCS) inven'tory following and antici-pated operational occurrence. Loss of ac power for this case is construed to be loss of offsite power. If seal failure is the consequence'f loss of cooling water to the reactor coolant pump (RCP) seal coolers for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, due to loss of offsite power, one acceptable solution would be to supply emergency power to the component cooling water pump. This topic is addressed for Babcock and Wilcox (B&W) reactors in Item II.K.2.16.

WNP-2 Position WNP-2 as a participant in the BWR Owners'roup endorses the position developed by General Electric for the Owners'roup-This position shows that a total failure of the recirculation pump seal cooling systems, followed by extreme degradation of the pump seals results in a primary coolant loss of less than 70 gallons per minute.* This loss is easily compensated for by normal water level controls and presents no hazard to the health and safety of the public. This position was presented to the NRC by General Electric.

• It should be noted that cooling can be provided to the seals via the seal purge system. This system is supplied water from the CRD pumps which are powered from emergency buses. Though the pumps are not part of automatic load sequence, they may be manually started and cooling w=ter supplied to the seals as needed from the condensate system (Condensate Storage Tank).

B.2-73

WNP-2 AMENDMENT NO 17 July 1981 EI.K.3.27 Provide Common Reference Kevel for Vessel Tevel Instrumentation Position Different reference points of the 'various reactor vessel water level instruments may cause operator confusion. Therefore<

all level instruments should be referenced to the same point.

Either the bottom of the vessel or the top of the active fuel are reasonable reference points (NUREG-0737).

Clarification None WNP-2 Position WNP-2 complies with this position. However, the common reference water level used is "zeroed" at the normal operating water level as shown in FSAR Figure 5.3-2. Justification for usi'ng the normal water level is as follows. When system upset conditions occur and vessel level varies, the operator's pri-

. mary focus is not how far above the top of the active fuel (TAF) the vessel level .is, but how far above or below "normal" it is. The majority of transients will not be severe enough to make core coverage the major concern+of the operator. To focus his. attention by.use.of a commo~ "zero" reference at TAF would detract from less severe system. upsets and may increase the probability of escalating the upset conditzon.

The water level references must allow the operator to accura-tely and quickly respond to both minor and severe system upsets.

Under various pressure and temperature conditions, the TAF indicated level will vary. Use of a "zero" reference on the fuel zone indicator is therefore not meaningful to the opera-tor during transient conditions without pressure/temperature compensation curves, and may even falsely imply. accuracy to the operator. Knowing how far TAF level actually is does not influence the operator's ction, only trend will influence his action. Thus use o a TAF reference "zero" is meaningless.

g, /o~

B.2-74

WNP-2 AMENDMENT NO. 23 February 1982 II.K.3.28 Verify Qualification of Accumulators on Automatic Depressurization System Valves Position Safety analysis reports claim that air or'itrogen accumula-tors for the automatic depressurization system (ADS) valves are provided with sufficient capacity to cycle the valves open five times at design pressures. GE has also stated that the emergency core cooling (ECC) systems are designed to withstand a hostile environment and still perform their func-tion for 100 days following an accident. Licensee should verify that the accumulators on the ADS valves meet thesecannot requirements, even considering normal leakage. If this be demonstrated, the licensee must show that the accumulator design is still acceptable.

Clarification The ADS valves, accumulators, and associated equipment and instrumentation must be capable of performing their functions during and following exposure to hostile environments and taking no credit for nonsafety-related equipment or instrumen-tation. Additionally, air (or nitrogen) leakage through valves must be accounted for in order to assure that enough inventory of compressed air is available to cycle the ADS valves.

WNP-2 Position The ADS valves accumulators and associated equipment and instrumentation for WNP-2 are designed to withstand a hostile 1 environment and perform their function for 100 days following an accident. The ADS nitrogen supply system was conser-

-vatively designed for 30 days operation following a LOCA (see FSAR 9.3.1.2). WNP-2 has modified the system so that nitrogen leakage from the system can be compensated for under any cir-cumstances through a remote connection which is accessible outside the secondary containment (outside the railway air lock). A portable nitrogen supply can then be connected to augment the existing nitrogen supply.

This supplementary, portable nitrogen supply consists of two nitrogen bottle connections (one for each supply header to the ADS valves) located in the corridor between the reactor building and the diesel generator building. A supplementary nitrogen bottle i.s manually valved in to its supply header when a local counter indicates the last nitrogen bottle in the railway air lock has gone on service or when header pressure (indicated in the control room) begins to decrease.

8.2-75

WNP-2 AMENDMENT NO. 30 June 1983 II'.3 '0 Revised Small-Break Loss-of-Coolant Accident Methods to Show Compliance with 10 CFR Part 50, Appendix K Position The analysis methods used by nuclear steam supply system (NSSS) vendors and/or fuel suppliers for small-break loss-of-coolant accident (LOCA) analysis for compliance with Appendix K to 10 CFR Part 50 should be revised, documented, and submitted for NRC approval. The revisions should account for comparisons. with experimental data, including data from the LOFT Test and Semiscale Test facilities (NUREG-0737) ~

Clarification i

As a result of the accident at TMI-2, the Bulletins and Orders Task Force was formed within the Of fice of Nuclear Reactor Regulator. This task force was charged, in part, to review, the analytical predictions of feedwater transients and small-break LOCAs for the purpose of assuring the continued safe operation of all operating reactors, including a deter-mination of acceptability of emergency guidelines for operators.

As a result of the task force reviews, a number of concerns were identified regarding the adequacy of certain features of small-break LOCA models, particularly the need to confirm specific model features (e.g., condensation heat transfer rates) against'pplicable experimental data'hese concerns, as they applied to each light-water reactor (INR) vendor's models, were documented in the task force reports for each INR vendor. In addition to the modeling concerns identified, the task force also concluded that, in light of the TMI-2 accident, additional systems verification of the s'mall-break LOCA model as required by II' of Appendix K to 10CFR50 was needed. This included providing predictions of Semiscale Test S-07-10B, LOFT Test (L3-1), and providing experimental verification of the various modes of single-phase and two-phase natural circulation predicted to occur in each vendor's reactor during small-break LOCAL Based on the cumulative staff requirements for additional small-break LOCA model verification, including both integral system and separate effects verification, the staff con-sidered model revision as the appropriate method for reflect-ing any potential upgrading of the analysis methods'he purpose of the verification was to provide the necessary assurance that the small-break LOCA models were acceptable to calculate the behavior and consequences of small primary B. 2-76

WNP-2 AMENDMENT NO. 30 June 1983 system breaks. The staf f believes that this assurance can alternatively be provided, as appropriate, by additional justification of the acceptability of present small-break LOCA models with regard to specific staff concerns and recent test data. Such justification could supplement or supersede the need for model revisions

• The specific staff concerns regarding small-break LOCA models are provided in the analysis sections of the BSO Task Force reports for each LWR vendor, (NUREG-0635, -0565, -0626,

-0611, and -0623). These concerns should be reviewed in total by each holder of an approved emergency core cooling system (ECCS) model and addressed in the evaluation as appro-priate.

The recent tests include the entire Semiscale small-break test series and IDFT Tests (L3-1) and L3-2). The staff believes that the present small-break LOCA models can be both qualitatively and quantitatively assessed against these tests. Other separatage effects test (e.g., ORNL core uncovery tests) and future tests, as appropriate, should also be factored into this assessments Based on the preceding information, a detailed outline of the proposed program to address this issue should be submitted.

In particular, this submittal should identify (1) which areas of the models, if any, the licensee intends to upgrade, (2) which areas the licensee intends to address by further just-ification of acceptability, (3) test data to be used as part of the overall verification/upgrade effort, and (4) the esti-mated schedule for performing the necessary work and submitting this information for staff review and approvals WNP-2 Position The General Electric Company final response to this concern was provided in GE letter MFN-132-81, R.H. Buchholz (GE) to D.G. Eisenhult "NUREG-0737, Item II K.3.30 Final Program Results ", dated June 26, 1981 'he ~

Supply System concurs with the results of the GE program that concludes no changes to the existing model are needed's an example, a model that presently does not properly account for horizontal countercurrent two-phase flow in the hot leg piping should either be revised to properly account for the phenomenon, or demostrated to produce a conserva-tive result for the entire spectrum of small breaks con-sidered.

B. 2-77

WNP-2 AMENDMENT NO. 30 June 1983 II. K. 3. 31 Plant-Speci fic Calculations to Show Compliance with 10 CFR Part 50.46 Position Plant-specific calculations using NRC-approved models for small-break loss-of-coolant acciden'ts (LOCAs) as described in Item II.K.3.30 to show compliance with 10CPR50.46 should be submitted for NRC approval by all licensees (NUREG-0737) ~

Clarification None WNP-2 Position The GE II. K. 3. 30 final response has been submitted, see response II.K.3.30. The plant specific analysis for WNP-2 has been completed and is reflected in a revision to 6 '.3.

B. 2-78

WNP-2 AMENDMENT NO ~ 17 July 1981 ZZ.K.3.44 Adequate Core ~

ooling for Transients with a Single Failure Position For anticipated transients combined with the worst single failure and assuming proper operator actions, licensees should demonstrate that the core remains covered or provide analysis to show that no significant fuel damage results from core uncovery. Transients which result in a stuck-open relief valve should be included in this category (NUREG>>0737).

Clarification None WNP-2 Position WNP-2 as a member of the BWR Owners'roup endorses the following position statement and analysis prepared by GE on behalf af the Owners'roup:

==

Introduction:==

This report has been prepared as the BWR Owners'roup generic-response to NUREG-0737 Task Item IZ.K.3.,44 which addresses the issue of adequate core cooling for transients with a single failure for those plants identified in Table ZI.K.3.44-4.

At the outset it should be noted that the conditions described in IX.K.3.44 (i.e., transients plus single failures) go beyond the current BWR design basis and that the item's reference to transients with multiple failures goes beyond the regulatory .

requirements as specified in Regulatory Guide 1.70, Revision 3.

The multiple failures specified involve consideration of a stuck-open relief valve (SORV) combined with the worst single failure. GE and the Owners'roup continues to support the current BWR design basis approach. This report is= intended to provide information to address Ztc.-.. ZI.K.3.44, but does not reflect our intention to change t)- current BWR design basis approach.

It is shown that, for the CE BWR/2 through BWR/6 plants, the core remains cove ed for any tran= nt with the worst single failure. This is achieved withouz any operator action to manually initiate emergency core cooling system (ECCS) or other inventory makeup systems. The worst transient with the worst single failure is shown to be the loss of feedwater (LOF) event with a failure of the high pressure ECCS or one isolation condenser (IC) loop, whichever is applicable.

BE 2-79

@or toe uvunQ3.ng" JUL evenc, sruaxeo wnxch lncluaea even more deg'raded conditions have been docuniented in Reference 1. 'The degraded conditions cover the failure of HPCS (or HPCI or FWCI or IC) and one SORV. Reference 1 shows that the core will remain covered and therefore, that no fuel failure would occurs Criteria,. Scope and Assumptions:

NUREG-0737 .Item II.K.3.44 requiies that the licensees demonstrate adequate core cooling to prevent the fuel from incurring significant damage for the anticipated transients combined with the worst single failure. In order to meet this requirement, eigher one of the following two criteria should be satisfied..

a. The reactor core remains covered -with water until stable conditions are achieved; or

b. No significant fuel damage results from core uncovery.

For BWR plants, this report will show that Criterion 1 is met.

The report makes the following assumptions:

a. A representative plant of each BWR product line, BWR/2 through BWR/6, is used to represent all of the plants of that product line.

b. The'anticipated transients as identified in NRC Regulatory Guide 1.70, Revision 3 were,.

considered.

c. The single failure is interpreted as an active failure.

'd. All'lant systems and components are assumed to functon normally, unless identified-as being failed.

Discussion:

Table II.K.3.44-1 lists all of the trans'ients which were con-sidered in this study. The event sequence of each transient was examined for each product line to determine the impact on core cooling. The following three factors were used to 'deter-mine the worst transient.and the worst single failure:

B.2-80

WNP-2 AMENDMENT 'NO. 1 7 July 1981 f

a. Reduction or'loss .of main feedwater or coolant makeup or'eat removal systems, especially high pressure systems, e.g., HPCX, FWCI HPCS, RCIC or, XC ~

4

b. Steam release paths causing'apid reactor coolant inventory loss, e.g , SRVs', turbine, or turbine bypass valves.

c. Power level, especially the timing of scram.

Based- on these considerations, a comparison was made among the transients in Table XX.K.3.44-1.

In Reference 2, the events of Table XI.K.'3.44-1 are-compared in detail for a typicla BWR/4 plant.= Xn particul'ar the impact on core cooling for each transient is evaluated by comparison to the analysis results for the ZOF event in the section titled "Applicability of Analyses." It is found that the ZOF event is the most severe transient from the core-cooling viewpoint due to its rapid depletion of reactor coolant inven-tory. This conclusion has generic applicability to all BWR product, lines covered by this study.

The same approach was also us'ed to select the single failures which would pose the greatest challenge to core cooling.

Among all of the possible failures considered (Table II.K.3.44-2 the following fail'ures are'dentified as the most important ones: I a., Failure of HPCX or HPCS or FWCX or one XC loop<

whichever is applicable.

b; Failure of RCXC.

c. One of the SRVs', which has opened as a result of the transient, fails to close.

Items a and b are the possible limiting failures because they represent loss of high pressure inventory makeup or heat removal systems which would be relied on following a loss of feedwater event. Item c is a possible limiting failure, because it results in the largest steam release rate from the vessel compared to other possible release paths (e.g., a stuck-open'turbine bypass valve)." No other failures iden-tified in Table XI.K.3.44-2 result in a direct challenge to core cooling capability.

Because of,the relatively low steam loss capacity through one SORV (Item c) compared to the makeup water capacity BE 2-81

WNP-2 AMENDMENT NO 17 July 198,1 of the highest capacity makeup water system, the failure of the highest capacity high pressure makeup system (Item a) would be worse than a stuck-open relief valve (Xtem c). For example, for a typical BWR/4, representative values of HPCI makeup and SRV flow are'8% and 6% of rated feedwater flow, respectively. Because of the higher makeup rate of HPCI/HPCS relative to RCXC (38 of rated fee'dwater flow), Item a would be worse than Item b. Table II.K.3.44-3 lists the worst com-bination of transient and single failure'for the GE BWR pro-

'duct lines covered by this study.

Even with the worst single failure in combination with the LOF event, the RCXC or- at least one XC loop will fuunction to pro-vide makeup and/or to remove decay heat while the vessel pressure remains, high. The design basis for the RCIC or the IC is such that they are capable of removing decay heat with the vessel being isolated. Analyses of the,LOF event with the worst single failure have been performed to support this conclusion. For example, for BWR/2 plants, such analyses are documented in Reference 1, Table 3.2.1.1.5-5. These analyses show that the isolation condenser heat removal capacity is greater than the decay heat generation rate and will lead to a safe and stable condition. Similar analysis have been per-formed for representative plants with the RCIC system. These analyses show that for the worst transient with the worst single failure, 'he minimum water level for different BWR pro-duct lines ranges from 6 feet to 11 feet above the top of the active fuel.

With even more degraded conditions, i.e., 'one SORV in addi>>

tion to the worst case transient with the worst single failure, reference plant .analyses in Reference 1, Tables 3.2.1.1.5-9 and 3.2.1.1.5-10 show that for the plants ana-lyzed the RCIC system can automatically provide sufficient inventory to keep the core covered even with a single failure plus a SORV. This capability is not a design basis for the RCIC system, and not all plants have been analyzed to

'emonstrate this capability.'f a plant should not have this capability, manual depressurization will avoid core uncovery for the case of LOF plus worst single failure plus SORV. It should be noted that manual depressurization is the proper operator action for all plants during loss of inventory con-ditions when the high pressure cooling system(s), are unable to restore and maintain RPV level. These proper operator actions are'allowed for in the NUREG-0737 requirement.

For plants without RCXC, manual depressurization will avoid core uncovery for the case of LOF plus .worst single failure plus SORV.

B.2 ~ -82

WNP-2 AMENDMENT NO.:17 July 1981

==

Conclusion:==

The anticipated transients in NRC Regulatory Guide 1.70, Revision 3 were reviewed for all BWR product lines BWR/2 through BWR/6 from a core, cooling viewpoint. The LOF event was identified to be the most limiting transient which would challenge core cooling. The BWR is designed so that the makeup or inventory maintenance systems or heat high'ressure removal systems (HPCI, HPCS, FWCI, RCIC or IC) are indepen-dently capable of maintaining the water level above the top of the active fuel given a loss of feedwater. The detailed ana-lyses show that even with the worst single failure in'om-bination with the LOF event, the core remains covered.

k Furthermore, =even with more degraded conditions involving one SORV in addition to the worst transient with the worst single failure, studies show that the core remains covered during the whole course of the transient either due to RCIC operation or due to manual depressurization.

It is. concluded that for anticipated transients combined with the worst single failure the core remains covered.

Additionally, it is'concluded that for severely degraded. tran-sticks open and 'an additional failure,it sients beyond the design basis where is assumed that, a SORV occurs the core remains covered with proper operator action.

BE 2-83

WNP-2 AMENDMENT NO ~ 17

'July,1981

'I Ref erences:

Section 3.2.1 (prepublication form), of "Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors," NEDO- 24708, March 31,'980

2. Section 3.2.2 (prepublication form) of. "Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors," NEDO-24708, June 30, 1980 3 ~ Section 3.5.2.1 (prepublication form) of "Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors," NEDO-24708, August 31, 1979 B. 2-84

'NP-2 AMENDMENT NO. '17.

July 1981 TABLE II.K 3, ~ 44-1 SVMMARY OF INITIATING TRANSIENTS

(

Reference:

NRC'egul'atory Guide 1.70, Revision 3)

Loss'f Feedwater Heating

2. Feedwater Controller Failure - Maximum Demand 3 ~ Pressure 'Regulator Failure - Open 4 ~ Inadvertent Safety/Relief Valve'pening

5. Inadvertent Residual Heat Removal (RHR) Shutdown Cooling Operation

'I

6. Pressure Re'gulator Failure Closed 7 ~ Generator Load Rejection

8. Turbine Trip

9. Main Steam Isolation Valve (MSZV) Closure

10. Loss of Condenser Vacuum Loss of Normal AC Power

12. Loss of Feedwater Flow

13. Failure of RHR Shutdown Cooling

'I 14 Recirculation Pump Trip

'5.

Recirculation Flow Control Failure Decreasinng Flow

16. Rod'Withdrawal Error V

17. Abnormal Startup of Idle Recirculation Pump

18. Recirculation Flow, Control Failure - Increasing Flow

19. Fuel Loading Error

20. Inadvertent Startup of High Pressure Core Spray (HPCS) or High Pressure Coolant Injection (HPCI) or Feedwater Coolant Injection (FWCZ) or Zsolation Condenser (ZC),

whichever is applicable.

B.2-85

AMENDMENT NO. 17 July 1981 TABLE ZZ ~ K. 3 ~ 44-2 LIST OF SINGLE FAILURES WHICH CAN POTENTIALLY DEGRADE THE COURSE OF A BWR TRANSIENT 1., One or all of the bypass valves fail to modulate 'open when required.

2. One of the bypass valves, which has opened as a result of-the transient, fails to close.,

3. Failure to trip the turbine or feedwater pumps on high water level.

II

4. One main steam isolation valve (MSIV) fails =to close when required;

5. One of the safety/relief valves fails to open when required.

6.. One of the safety/relief valves, which has opened as a result of the transient, fails to close.

7. Failure to trip one recirculation pump.

8. Failure to run back the recirculation pumps.

9. Failure of high pressure coolant injection (HPCI) or high

~

pressure core spray (HPCS) or feedwater coolant injection-(FWCZ) or one isolation condenser (ZC) loop, whichever is .

applicable.

10. Failure of reactor core isolation cooling (RCZC) or one IC loop, whichever is applicable.

11. Failure of one low pressure coolant injection (LPCI) loop or the low pressure core spray (LPCS) system.

12. Loss of one residual heat removal (RHR) system heat exchanger.

13. A single control rod stuck while the remainder of the contxol rods are moving.

14. Failure to achieve the rod block function (i.e., a single contxol rod will withdraw upon erxoneous withdrawal demand).

15. Loss of one initiating diesel generator event.

if loss of AC power was the B.2-86

WNP-2 AMENDMENT NO. 17 July 1981 TABLE II'.3 '4-3 THE WORST CASE OF- TRANSIENT WITH A SINGLE FAILURE FOR DIFFERENT, BWR PRODUCT LINES Product Line Transient with a Sin le Failure The Worst Case)

BWR/2 LOF + Failure of one IC Loop (Oyster Creek only)

LOF + Failure of FWCI (Nine Mile Point only)

BWR/3 LOF + Failure of FWCI (Millstone only)

LOF + Failure of HPCI (others)

BWR/4 LOF + Failure of HPCI BWR/5 LOF + Failure of HPCS BWR/6 LOF + Failure of HPCS B 2-87

WNP-2 AMENDMENT NO. 17 July, '1981-TABLE IX.K.3.44-4 PARTICIPATING UTILITIES NUREG-0737 This report applies to .the following plants, whose owners par-ticipated in the report's development.

Boston Edison Pilgrim 1 Caroline Power & Light Brunswick 1,& 2

.Commonwealth Edison LaSalle 1 & 2, Dresden 1-3, Quad Cities 1 & 2 Georg'ia Power- Hatch-1 & 2 Iowa Electric Light & Duane Arnold Power Jersey Central Power & Oyster Creek 1 Light Niagara Mohawk Power Nine. Mile Point 1 & 2 Nebraska Public Power Cooper District Northeast Utilities Millstone 1 Philadelph'ia Electric Peach Bottom 2 & 3; Limerick & 1 2 Power Authority of the Fi tzpatrick State of New'ork Tennessee Valley Authority Browns'erry 1-'3; Hartsville 1-4, Phipps Bend 1 & 2 Vermont Yankee Nuclear Vermont Yankee Power Detroit Edison Enrico Fermi 2 Mississippi Power & Grand Gulf 1 &

Light Pennsylvania Power & 1 & 2 2'usquehanna Light BE 2-88

WNP-2 AMENDMENT NO. 17 July 1981 TABLE II.K.3.44-4 (Continued)

Washington Public Power WNP-2 Supply System Cleveland Electric Perry 1 & 2 Illuminating Houston Lighting,& Power Aliens Creek Illinois Power Clinton Station 1 & 2 Public Service of Oklahoma Black Fox 1 & 2 Long Island Lighting Shoreham B.2-89

WNP-2 AMENDMENT NO. 17 July 1981 II.K.3.45 Evaluation of Depressurization with Other than Automatic Depressurization System Position Analyses to support depressurization modes other than full actuation of the ADS (e.g., early blowdown with one or two SRVs) should be provided. Slower depressurization would reduce the possibility of exceeding vessel integrity limits by rapid cooldown (NUREG-0737).

Clarification None WNP-2 Position WNP-2 as a member of the BWR Owners'roup endorses the following position statement and analysis prepared by GE on behalf of the Owners'roup.

The evaluation of alternate modes of depressurization other than full actuation of the Automatic Depressurization System (ADS) 'is made for those plants listed in Table II.K.3.45-5 with regard to the effect of such r'educed depressurization rates on core cooling and vessel integrity.

Depressurization by full ADS actuation constitutes a depressurization from about'050 psig to 180 psig in approxi-mately 3.3 minutes. Such an event, which is not expected to occur more than once in the lifetime of the plant, is well, within the design basis of the reactor pressure vessel. This conclusion is based on the analysis of 'several transients requiring depressurization via the ADS valves. Results of these analyses indicate that the total vessel fatigue usage is less than 1.0. Therefore, no change in the depressurization rate is necessary. However, to comply with the above request.

reduced depressurization rates were analyzed and compared with the full ADS actuation. The alternate modes considered cause vessel pressure to traverse the same pressure range in (1) depressurization case 1 (ranges from 6-10 minutes depending on size and ADS capacity), and (2) depressurization case 2 'lant (ranges from 15-20 minutes). The case 2 depressurization bounds the possible increase in depressurization time by pro-ducing an undesirably long core uncovered time. The case depressurization gives the results of an intermediate depressurization. These modes are achieved by opening a reduced number of relief valves. These blowdown rates are illustrated by Figure II.K.3.45-1.

B.2-90

NNP-2 AMENDMENT, NO 1 7 July 1981 Assumptions:

The major assumptions used for the core cooling analysis are:

a. No high pressure cooling systems are available.

b. All low pressure ECC systems are available.

c. Assumptions as stated in NED0-24708, Section 3.1..1.3, "Justification of Analysis Methods";

which includes the use of 1978 ANS Decay Heat (mean value).

Results:

a. Vessel Xntegrity The depressurization events considered are full ADS blowdown and blowdown over 10 and 20 minute

'ntervals. The reactor vessel stresses for these events are within the acceptance stress. limits defined by ASME Code Section III for emergency conditions (Level C). The core support struc-tures and other safety-related internal com-ponents are also within applicable emergency condition stress limits.

The ADS operating conditions which affect, fatigue usage of vessel or core support structures are not significantly different for fast and slow blowdown events. Specific calculations of fati-gue usage are not required for emergency con-ditions (Level C). However, available pressure vessel'atigue analyses show the usage per event to be <0.1 per full ADS event.

Xn summary, reactor vessel and core support structure integrity is assured for the blowdown rates considered if an ADS event should occur,.

and reduced rates of depressurization do not significantly decrea e fatigue usage.

b. Core Cooling Capability Examination of the roduced depressurization rates under consideration "ith respect to core cooling concerns shows that:

B.2-91

WNP-,2 AMENDMENT NO. '

1 7 July 1981

1. .Vessel depressurization for a case 2 'blowdown-

'(15-20 minutes) causes the core to be.uncov-.

ered*for a lengthy'.period of time even assuming system initiation at the earliest reasonable time.

2. Vessel depressurization for a case 1 blowdown (6-10 minutes), when actuated at the same level as the full ADS case, will result in less -vessel inventory at .the time of ECCS injection 'and can result in longer-.periods of, core uncovery.

B.2-91a

WNP-2 AMENDMENT NO ~ 17 July 1981

3. Vessel deoressurization for a case 1 blowdown (6-10 minutes) when actuated considerably earlier than at the ADS initiation setpoint can result in some improvement in core cooling. Hoewever, the operator is required to act more quickly in these cases (i.e.,

within '1-5 minutes after the accident) . This earlier depressurization also reduces the time available to start high pressure system

-injection and hence to avoid the need'or manual. depressurization. It also increases the frequency of depressurization.

The results of the calculations are presented in Tables II.K.3.45-1 through II.K.3.45-4. They show the total core uncovered time and remaining vessel inventory at the time, of low pressure ECCS injection. A discussion of these results follows below.

Discussion:

a The results are based upon calculations performed with the assumptions stated earlier using a representative BWR/3 and a BWR/6 to show consistency of results across the product lines.

The transients considered are an outside steamline break and a stuck-open relief valve. The ADS will depressurize. the vessel .

to the low pressure 'ECCS injection setpoint when no high pressure cooling systems are available. The depressurizations are initiated at different times based on the downcomer 'sed water level. The first initiation time considered is when the water level is at the top of the active fuel which is,con-,

sistent with the original design for most plants and thus is the basis for comparison. The second initiation time con-sidered is the downcomer water level of,34 fe'et from the bot-tom of the vessel which still provides the operator, with a reasonable time to attempt to start the high pressure systems.

The last initiation time conside'red is the high pressure make-up system setpoint (Level 2 for BWR/6 and Level 1 for BWR/3) plus 60 seconds which is the earliest time in which depressurizaton could be expected to occur.

core cooling criteria used in assessing -the impact of a

'he reduced depressurization rate are:

a. Inventory in the core and lower plenum at the of low pressure ECCS injection as predicted 'ime by the SAFE model (Reference 1)'.

B.2-92

WNP-2 AMENDMENT NO. 17 July 1981

b. The total time which the top of the active fuel (TAF) remains uncovered as predicted by the SAFE .model (Reference 1).

The first criterion demonstrates the increased mass loss due to boiloff for the longer blowdown, since mass loss. due to flashing will be independent of the depressurization rate. pro-viding the boundary pressure values are the same for all the rates. The second criterion is a measure of the resultant core temperature.

Table II.K.3.45-1 gives the results for a BWR/6 assuming an outside steamline break. As the length of depressuriization is increased the vessel inventory at the time the ECCS injec-tion decreases and the total core uncovered time increases.

Table XZ.K.3.45-1 further shows that the actuation times based on higher water levels (i.e., 34'nd Level 2 + 60 seconds) longer depressurizations exhibit the same trends.

Furthermore, for any particular depressurization rate, raising the actuation level increases the vessel inventory at ECCS injection and decreases the total core uncovered time.

However, this also decreases'he time the operator has available to try to get high pressure level control systems working in order to avoid the need to depressurize.

Table ZX.K.3.45-2 shows that these same results are exhibited

=for the case of a stuck-open relief valve. Table IX.K.3.45-3 shows the results for a BWR/3 as5uming an outside steamline break. Examination of the table shows the same trends as Table II.K.3.45-1, and therefore the results are applicable to all product lines. Table XX.K.3.45-4 shows that these general trends 'are independent of the models used by exhibiting the same trends "for a BWR/3 using standard Appendix K licensing assumptions.

==

Conclusion:==

The cases considered show that no appreciable improvement can be gained by a slower depressurization based on core cooling considerations. A significantly slower depressurization rate will result in increased core. uncovered time. A moderate decrease in the depressurizaton rate necessitates an earlier actuation time resulting in less time available for operator action to start high pressure ECC without significant benefit to vessel fatigue usage. This will also result in an increased frequency of ADS actuation.

Finally, it is of paramount importance to note that the ADS is not a normal core cooling system; it is a backup for high pressure cooling systems (feedwater, RCZC, HPCZ/HPCS). Zf ADS operation is ever required in a BWR< it will be because core B.2<<93

WNP-2 AMENDMENT NO ~ 17

.-'July 1981 cooling is threatened. Since a 'full ADS blowdown is well within the design basis of the reactor pressure vessel and ADS is properly designed to minimize the threat to .core cooling, no change in the depressurization rate is- necessary.

h 1

B.2-94

WNP-2 AMENDMENT NO. 17 July 1981

References:

1. NEDO-24708, "Additional Information Required for NRC Staff Generic Report on Boiling Hater Reactors", August 1979.

BE 2-95

WNP-2 AMENDMENT NO ~ 17 July 1981 TABLE II+K+3~ 45=1 RESULTS FOR BWR/6 OUTSIDE STEAMLZNE BREAK NO HIGH PRESURE SYSTEMS AVAILABLE LIQUID INVENTORY IN DEPRESSURIZATZON CORE CORE AND LOWER PLENUM DEPRESSURZZATZON INITIATION UNCOVERED AT LOW PRESSURE ECCS CASE TIME (SEC) TIME (SEC) - INJECTION (LBS)

FULL ADS 1086. 0 26 1 '03 x 105 CASE 1 1086.0 117 1.528 x 105 CASE 1 610.6, 10 1.779 x 105 34'evel 2** 78.3 No 1.993 x 105

+ 60 Sec. Uncovery Level 2 78 ~ 3 No 1.937 x 105

+ 60'Sec. Uncovery J

Level 2 78%3 390 1.755 x 105

+TOP OF ACTIVE FUEL

• • HIGH PRESSURE INITIATION SETPOZNT PLUS 60 SECONDS BE 2-96

WNP-2 AMENDMENT NO ~ 17 July '1981 TABLE IZ.K~3 '5-2 RESULTS FOR BWR/6 STUCKWPEN RELIEF VALVE NO HIGH PRESURE SYSTEMS AVAILABLE LIQUID INMMVORY IN DEPRESSURI ZATION CORE CORE AND LOWER PLENUM DEPRESSURI ZAT ION, INITIATION UNCOVERED AT TAN PRESSURE ECCS

. CASE LEllEL TTEE (SEC) TIME (SEC) INJECTION (LBS) 642.6 No 1 '36 x 105 Uncovery CASE 1 642.6 15 1 '87 x 105-CASE- 1 391. 8 No 1.889 x 105 Uncovery 34'evel CASE 2** 77.7 No 1.961 x 105

+ 60 Sec. Uncovery

• TOP OF ACTIVE FUEL

• • HIGH PRESSURE INITIATION SETPOZNT PLUS 60 SECONDS B.2-97

AMENDMENT NO+ 17 July 'l981 TABLE XI+K+3~ 45-3 RESULTS FOR BWR/3 OUTSIDE STEAMLINE BREAK NO HIGH PRESURE SYS'5!EMS AVAXLABLE LX{}UZD INVENTORY XN DEPRESSURXZATZON CORE CORE AND LOWER PLENUM DEPRESSURIZATION INXTIATZON UNCOVERED AT LOW PRESSURE ECCS CASE LEVEL TIME (SEC) TIME (SEC) INJECTXON (LBS)

FULL ADS TAF* 1527. 8 155 2.027 x 105 I

CASE CASE 1

1 FULL ADS 34'01.

Level 1**

1527 ~ 8 364 '

6 170 51 No

~

1 '75 x 105 2.291 x 105 2.446 x 105

+ 60 Sec. Uncovery CASE 1 Level 1 364 ' 10 ~ 2.394 x 105

+ 60 Sec.

• TOP OP ACTIVE FUEL

• • HIGH PRESSURE INITIATION SETPOZNT PLUS 60 SECONDS B 2-98

I WNP-2 AMENDMENT NO 17 July 1981 ~

TABLE ZIvKv3 ~ 45 4

'ESULTS FOR BWR/3 OUTSIDE STEAMLZNE BREAK ON APPENDIX K ASSUMPTIONS WITH NO HIGH PRESSURE SYSTEMS

LIQUID INVENTORY ZN DEPRESSURIZATZON CORE CORE AND LOWER PLENUM DEPRESSURIZATZON INITIATION UNCOVERED AT LCW PRESSURE ECCS CASE LEVEL TIME (SEC) TIME (SEC) INJECTION .(LBS)

FULL ADS 759.4 264 1.960 x 105 CASE 1 759 4 277 1 913 x 105 PULL ADS Level 1** 145.6 175 2.210 x 105

+ 60 Sec.

Level 1 145.6 191 2 165 x 105

. + 60 Sec.

"TOP OF ACTIVE FUEL

• +HIGH PRESSURE INITIATION SETPOZÃL PLUS 60 SECONDS B.2-99

AMENDMENT NO. 17 July 1981 TABLE ZX.K.3.45-5 NUTMEG-0737 This report applies to the following plants, whose owners par-ticipated in the report's development.

Boston Edison Pilgrim 1 Caroline Power & Zight Brunswick 1 & 2 Commonwealth Edison Z aSalle 1 & 2, Dresden 2 & 3, Quad Cities 1 & 2 Georgia Power Hatch 1 & 2 Xowa Electric Light & Duane Arnold Power Jersey Central Power & Oyster Creek 1 Light Niagara Mohawk Power Nine Mile Point 1 & 2 Nebraska Public Power Cooper District Northeast Utilities Millstone 1 Northern States Power Monticello Philadelphia Electric Peach Bottom 2 & 3; Zimerick 1 & 2 Power Authority of the Fitzpatrick State of New York Tennessee Valley Authority Browns Ferry 1-3; Hartsville 1-4, Phipps Bend 1 & 2 Vermont Yankee Nuclear Vermont Yankee Power Detroit Edison Enrico Fermi 2 Long Zsland Lighting Shoreham Mississippi Power & Grand Gulf 1 & 2 Light B.2-100

WNP-2 AMENDMENT NO 17 July 1981 TABL"- II.K-3.45-5 (Continued)

Pennsylvania Power & Susquehanna- 1 &, 2 Light Washington Public Power . WNP-2 Supply System Cleveland Electric Perry 1 & 2 Illuminating Houston Lighting & Power Aliens Creek Illinois Power Clinton Station 1 & 2 Public Service of Black Fox 1 & 2 Oklahoma B.2-101

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WNP-2 AMENDMENT NO'7 July 1981 II.K.3.46 Response to fist of Concerns from ACRS l"

Consultant (Michelson Concerns)

Position General Electric should provide a response to the Michelson concerns as they relate to BWRs. See NUREG-0660, Appendix C, Table c.3, Item 46 (Reference 1)and NUREG-0626, Section 4, Item A.17 (Reference 6c).

Clarification None WNP-2 Position GE, acting for the BWR Owners'roup, responded to these con-cerns in a letter, "Response to Questions Posed by Mr. C.

Michelson", R. H. Buchholz (GE) to D. 'F. Ross, dated February 21, 1980. Submittal of this letter completes the action required by this task.

B.2-103

WNP-2 AMENDMENT NO ~ 17 July 1981 IXI.A.1.1 Upgrade Emergency Preparedness Position Provide an emergency response plan in substantial compliance with NUREG-0654, "Criteria for preparation and Evaluation of Radiological Emergency'esponse Plans and Preparedness in Support of Nuclear Power Plants," (which may be modified after May 13, 1980 based on public comments) except that only a description of a completion schedule for the means for- pro-viding prompt notification to the population (Appendix 3), the staffing for emergencies in acdition to that already required (Table B.1), and an upgraded meteorological program (Appendix

2) need be provided. NRC will give substantial weight to FEMA findings on offsite plans in judging the adequacy against NUREG-0654. Perform an emergency response exercise to test the integrated capability and a major portion of the basic elements existing within emergency preparedness plans and organizations (NUREG-0694).

Clarification

~ ~

None WNP-2 Position The Washington Public Power Supply System WNP-1, -2, -4 Hanford Site Emergency Plan was submitted to the NRC in March 1981. The Emergncy Plan complies with the intent of 10 CFR 50..47, 10 CFR 50.54, and Appendix E. The Emergency Plan commplies with the intent of the essential elements of NUREG-0654 but does not compl.'y with Regulatory Guide 1.101 as this guide was withdrawn per Federal Register Notice, Tuesday, October 21,'980.

An exercise to demonstrate the capabilities of the Supply

'll t 9 1* P Operations Facility to test the integrated capability of the Emergency Preparedness plans and organizations manent facilities are not available.

if the per-B ~ 3-1

WNP-2 AMENDMENT NO. 30 June 1983 available in the TSC and EOP on a near real time basis prior to fuel loading with the capability for remote interrogation. In addition, these parameters are available as input to the Supply System emergency dose projection system for the Purpose of the prefuel load exercise.

A backup meteorological measurements system will be designated prior to fuel loading. The backup meteorological measurements system will be capa-ble of being 'accessed at the EOP, TSC, and control room, and will be remotely interrogable.

A Class A meteorological model and dose assess-ment capability (Reference Appendix 2, NUREG-0654) is available for use as part of the emergency dose piojection system. The system is capable of producing initial transport, diffu-sion, and dose estimates within the plume expo-sure EPZ within 15 minutes following the classification of an accident. An emergency dose calculation manual describing the meteoro-logical model and use of the system, including a methodology for projecting doses out to the plume exposure pathway .EPZ, is available to the NRC and copies will be placed in appropriate emergency response facilities prior to the prefuel 1'oad exercise.

An ongoing study is being conducted by the Supply System to evaluate the impact of season-al, diurnal, and terrain-induced flows on the meteorological model. The emergency dose pro-jection system display, which utilizes color-graphic video output of the site grid map and areas of projected radiological hazard, and/or extensive tabular data concerning projected doses and field sampling can be rapidly tele-copied to the NRC and other emergency response organizations'his is a desirable method for transmission of this data due to time limita-tions encountered in sending a digitized map and a voluminous amount of data via a 1200 baud data transmission links Current estimates of the time required to transmit the video display output run about five to ten minutes per display as opposed to about 20 seconds for the same data to be tele faxed.

B. 3-lb

WNP-2 AMENDMENT NO. 23 February 1982 The facsimile method would minimize operator cross training and computer compatibility problems associated with an on-line computer link and would be more reliable in an emergency situation. This method would also alleviate possible assessment delays caused by loading the .

dose assessment computer with multiple interroga-tions during an emergency when time may be critical.

All "real time" meteorological and source data which serves as input to the emergency dose projection system will be available via a conven-tional data link capability to permit independent verification of dose calculations. The selection of a more comprehensive meteorological model incorporating advanced temporal and spatial aspects (sometimes referred to as "Class B" meteorology) is being developed by the Supply System. Final model design and implementation scheduling await the issuance of guidelines by the NRC. The models used by all emergency response facilities will be evaluated for compatibility.

d. Emergency Staffing The operating crew for each eight-hour shift at the plants normally consists of a shift manager, control room supervisor, shift support supervisor, licensed reactor operators, equipment operators, shift technical advisor, health physics/chemistry technicians, maintenance per-sonnel, and security force personnel. The shift manager on duty has the immediate responsibility for the plant at all times, and has full authority and responsibility for recognizing and declaring emergencies. Each shift is provided with a complement of qualified individuals who have specialized training in the operation of the reactor and other plant systems, plant instrumen-tation, radiation safety, and maintenance.

The plant manager has full authority and respon-sibility for the overall supervision and admin-istration of the nuclear plant. Technical support is'rovided by a staff experienced in reactor physics, nuclear power plant systems technology and operation, and health physics and chemistry. The normal plant operating organiza-B.3-1c

WNP-2 AMENDMENT NO. 17 July 1981 III.A.1.2 Upgrade Emergency Support Facilities Position (NUREG-0660 Clarification letter dated September 5, 1980)

Each operating nuclear power plant shall maintain an onsite" technical support center (TSC) separate from and in close proximity to the control room, that 'has the capability to display and transmit plant status to those individuals who are knowledgeable of and responsible for engineering and manage-ment support of reactor. operations in the event of an acci-dent. The center shall be habitable to the same degree as the, control room for postulated accident conditions. The licensee shall revise his emergency plans as necessary to incorporate the role and location of the TSC. Records that pertain to the as-built conditions and layout of structures, systems, and components shall be readily available to personnel in the TSC.

An operational support center (OSC) shall be established separate from the control room and other emergency response facilities (EOFs) as a place that op rations support person-nel can assemble and report in an emergency situation to receive instructions from the operating staff.

Communications shall be provided between the OSC, TSC, EOF, and control room.

An Emergency Operations Facility (EOF) '(near-site) will be operated by the licensee for continued evaluation and coor-of all licensee activities related to an emergency 'ination having or potentially having environmental consequences. The EOF must meet habitatability requirements to ensure that per-sonnel can remain in the'facility throughout the entire course of the accident including evacuation of the surrounding area.

The facility will have sufficient space to accommmodate repre-sentatives from Federal, State, and Local governments as appropriate. In addition, the major State and Local response agencies may provide for data analysis jointly with the opera-tor at this location. The Emergen-y Operations Facility (EOF) will provide information needed by Federal, State and Local authorities for implementation of "ffsite emergency plans in addition to a centralized meeting location for key represen-tatives from the agencies.

Clarification None 8 ~ 3-2

WNP-2 AMENDMENT NO ~ 17 July 1981 WNP-2 Position WNP-2 concurs with the intent of this position and its clari-fication as described in NUREG-0696 and is implementing the following:

a 0 Technical Support. Center (TSC)

WNP-2 is establishing a habitable, onsite Technical Support Center. The TSC will be complete 'and functional prior to July 1982, or a temporary location will be established and equipped to fulfill the intent of the functional requi'rements of the TSC.

Plant status data and communication systems beween TSC, control room and EOF will be completed by July 1982 or temporary subsituations will be implemented to meet the intent of the function requirements. Plant records will be available in the TSC. Procedures will be written to include the accident assessment function that would be conducted in the TSC and control room.

Those plant and meterological parameters necessary to conduct the accident assessment and initial emergency response evaluation will be available in the TSC.

The TSC will be designed and constructed to meet the intent of the requirements addressed in NUREG-0578, TMI-2 Lessons Zearned Task Force Status Report and Short-Term Recommendations, July 1979, and NUREG-0696, Functional Criteria for Emergency Response Facilities.

Design descriptions for the TSC facility, includ-ing the plant status data and communication systems, will be provided by November 12, 1981..

b. Operational Support Center (OSC)

An onsite Operational Support Center will be established at WNP-2. Communication will be pro-vided between OSC, TSC, EOFp and the control room. Procedures will be prepared and imple-mented to establish tne responsibilities and management communications for personnel assigned gf .

to the OSC. The OSC will be located ~the- mi+WI~ a.

Building

+/e ~~c~,Q B 3-3

WNP-2 AMENDMENT NO ~ 17 July 1981

c. Emergency Operations Facility (EOF)

The Washington Public Power Supply System's Hanford Site will have a common plant support facility containing EOF space, instrumentation and displays to support the Hanford projects, WNP-2, -1 and -4. The siting of this facility +8 zs.

described in the Supply System's letter of December 12, 1980 from G. D. Bouchey to B.

Grimes. The facility will have adequate. protec-tion factors to support, habitability throughout the course of an accident.

Design descriptions for. the EOF will be provided by November 12, 1981.

d. OSC/TSC/EOF Emergency Plan The WNP-2 Emergency Plan addresses the OSC/TSC/EOF approaches defined above.

B.3-4

WNP-2 AMENDMENT NO ~ 17, July 1981 III.D.1.1 Integrity of Systems Outside Containment Likely to Contain Radioactive Material for Pressurized Water Reactors and Boiling Water Reactors Position -

(NUREG-0737)

Applicants shall implement a program to red'uce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels. This program shall include the following:

a. Immediate Eeak Reduction

1. Implement all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment.

2. Measure actual leakage rates with system in operation and report them to the NRC.

b. Continuing Leak Reduction Establish and implement a program of preven-tive maintenance to reduce leakage to as-low-as-practical levels. This program shall include periodic integrated leak tests at intervals not to exceed each refueling cycle.

Clarification

.'pplicants shall provide a summary description, together with initial leak-test results, of the'ir program to reduce leakage from systems outside conta.inment that would or could contain primary coolant or other highly radioactive fluids or gases during or following a serious transient or accident.

a. Systems that should be leak tested are as follows (any other plant system which has similar func-tions or postaccident characteristics even though not specified herein, should be included)':

Residual heat removal (RHR)

Containment spray recirculation High pressure injection recirculation Containment and primary coolant sampling Reactor core isolation cooling

WNP-2 AMENDMENT NO ~ 17 July .1981 Makeup and letdown (PWRs only)

,Waste gas, (includes headers and cover gas system outside -of containment in addition to decay or storage system)

Include a list of systems containing radioactive materials which are excluded from .program .and provide justification for exclusion.

b.' Testing of gaseous systems should inclu'de helium leak detection or equivalent testing methods.

~ Should consider program to reduce leakage poten-tial release paths due to design and operator deficiencies as discussed in our letter to all operating nuclear power plants regarding North Anna and related incidents, dated October 17, 1979.

WNP-2 Position WNP-2 has performed a systems design review 'and will develop a surveillance/preventive maintenance program to limit to as-low-as-practical, leakage from systems outside containment which .could transport highly radioactive 'fluids during a.

serious transient or accident.

a ~ A systems review, for potential leakage paths outside containment was conducted to determine those systems which penetrate containment and could contain highly radioactive fluids in the case of a serious transient or reactor accident.,

'hree unisolated leak paths were found which

.could potentially transport highly radioactive fluids. Auto-isolation will be installed on

\

these potential leak paths in support o'f existing manual, remote isolat'on capabilities and to maintain radiological consequences to as-low-as-reasonably achievable.

b. A leakage surveillance and preventive will be developed for those systems maintenance'rogram*

within secondary containment which could be used to transport highly r. Qioactive fluids in the case of a serious reactor transient or accident.

This program includes the following features:

1. Designation of systems included within the leakage surveillance and preventive main-tenance program.

B.3-6

0 WNP-2 AMENDMENT NO ~ 17.

July 1981

2. A system listing identifying the components to be inspected, method of ispection or measurement and frequency or surveillance.

3. Routine operator inspections of visually accessible portions of designated systems at normal operating conditions or test mode.

4. Detailed leakage inspection and measurement

. for designated systems during initial test program and thereafter.

5., An aggressive preventive maintenance program with high priority assigned to leakage-related work or designated systems.

6. A review cycle for leakage-related work requests to evaluate posible modifications to keep leakage as-low-as-reasonably achievable.

/

• This program is to be initiated prior to July 1982, exception being taken to those systems which cannot be tested until startup due to required plant conditions. Program documen-tation will be availble onsite for NRC ZSE review approxi-mately by April 1982.

B. 3-7

WNP-2 AMENDMENT NO. 17 July 1981 III.D.3.3 Improved Inplant Iodine Instrumentation Under Accident Condi" ions Position (NUREG-0737)

a. Each licensee shall provide equipment and asso-ciated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident.

b. Each applicant for a fuel-loading license'o be issued prior to January 1, 1981 shall provide the equipment, training, and procedures necessary to accurately determine the presence of airborne radioiodine in areas within the plant where plant personnel may be present during an accident.

Clarification Effective moitoring of increasing iodine levels in the buildings- under accident conditions must include the use of portable instruments using sample media that= w'ill collect iodine selectively over xenon (e.g., silver ziolite) for the following reasons:

a. The physical size of the auxiliary and/or fuel handling building precludes locating stationary monitoring instrumentation at all areas where airborne iodine concentration data might be.

required.

b. Unanticipated isolated "hot spots" may occur in locations where no stationary monitoring instru-mentation is located.

c. Unexpectedly high background radiation levels near stationary monitoring instrumentation after an accident may interfere with filter radiation readings.

d. The time required to retrieve samples. after an accident may result in high. personnel exposures if .these filters are located in high-dose-rate areas.

After January 1, 1981, each applicant and licensee shall have the capability to remove the sampling cartridge to a low background, low contamination area for further analysis.

Normally, counting rooms in auxiliary buildings will not have B.3-8

WNP-2 AMENDMENT NO 17 July 1981 sufficiently low backgrounds for such analyses following an accident. In the low background area, the sample should first be purged of any entrapped noble gases using nitrogen The licensee shall have gas or clean air free of noble gases.

the capability to measure accurately the iodine concentrations present on these samples under accident conditions. There should be sufficient samplers to sample all vital areas.

For applicants with fuel loading dates prior to January 1, 1981, provide by fuel loading (until Januax'y 1, 1981) the

~ capability to accurately detect the presence of iodine in the region of interest following an accident. This can be accomplished by using a portable or cart-mounted iodine sampler with attached single-channel analyzer (SCA). The SCA window should be calibrated to the 365 KeV of iodine-131 using the SCA. This will give an initial conservative estimate of

,presence of iodine and can be used to determine if'espiratory protection is required. Caxe must be taken to asure that the counting system is not saturated as a result of too much acti-vity collected on the sampling cartridge.

WNP-2 Position WNP-2 intends to meet the intent of the position defined in this item. To summarize the intent of the program being implemented, six (6) continuou's air monitoring systems are provided for air sampling'lant areas where personnel may be present during accident conditions. In addition, ten (10) low volume air sampling systems will be strategically located throughout the, plant in frequently occupied areas to con-tinuously draw air samples for subsequent analysis.

Grab samples will be obtained using High Volume Air Samplers, both AC and DC powered operation.

During accident conditions activated charcoal cartridges will be used for radioiodine analysis iri con)unction with a Ge(Li)

Gamma Spectroscopy System located 'n a low background, low contamination area such as the Radiochemistry Iab in the Near-Site Facility. Prior to analysis- cartridges will be purged in a fume hood using plant service air or bottled nitrogen which is stored on site.

Station procedures are provided for obtaining and evaluating both routine and non-routine air ~-mples. In addition to ini-tial training provided for Health ."hysics/Chemistry personnel, periodic drills are conducted in accordance with the WNP-2 Emergncy Plan Section 17.

B.3-9

NNP-2 AMENDMENT NO. 17 July 1981 III.D.3.'4. Control Room Habitability Requirements Position p

'n'accordance with Task Action Plan Item XII,D.3.4 and control room habitability,'icensees shall ass'ure that control room .

operators will be adequately protected against the- effects of accidental release of toxic and radioactive'ases and that the nuclear power plant can .be safely operated or shut down,'under design basis accident conditions (Criterion 19,. "Control of Appendix A, ."General Design Criteria 'oom,"

to '10 CFR Part for;Nuclear*'ower'Plant,"

50).'larification

a. All licensees must make a submittal to the NRC regardless of whether or not they met the cri-teria of the referenced Standard Review Plans (SRP) sections. The new clarification specifies that licensees that meet the criteria of the SRPs should provide the basis for their conclu-sion that SRP 6.4 requirements are met.

Licensees. may establish this basis by referencing past submittals to the NRC and/or providing new or additional information to supplement past submittals.

b. All. licensees w'ith control rooms that meet the.

criteria of the following, sections of the.

Standard Review Plan:

2.2.1-2.2.2 Identification of Potential Hazards in Site Vicinity 2 2.3 Evaluation of Potential Accidents 6.4 Habitability Systems shall report their findings regarding the speci-fic SRP sections as explained below. The following documments should be used for guidance:

1. Regulatory Guide 1.78, "Assumptions. for Evaluating the Habitability of Regulatory Power Plant Cont el Room During a Postulated Hazardous Chemicai Release";

2. Regulatory Guide '..95, "Protection of Nuclear Power Plant Control Room Operators Against an Accident Chlorine Release"; andi B.3-10

WNP-2 AMENDMENT NO. 17 July 1981

3. K. G. Murphy and K. M. Campe, "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criterion 19,"

13th AEC Air Cleaning Conference, August 1974.

Licensees shall submit the results of their fin-dings as well as the basis for those findings by January 1, 1981. In providing the basis for the habitability finding, licensees may reference their past submittals. Licensees should, however, ensure that these submittals reflect the current facility design and that the information requested in Attachment 1 is provided.

c. All licensees with control rooms that do not meet the criteria of the above-listed references, Standard Review Plans, Regulatory Guides, and other References.

These licensees shall'erform the necessary evaluations and identify appropriate modifications.

Each licensee submittal shall include the results of the ana-lyses of control room concentrations from postulated acciden-tal release of toxic gases and control room operator radiation exposures from airborne radioactive material and direct radiation resulting from design basis accidents. The toxic gas accident analysis should be performed for all potential hazardous chemical releases occurring either on the site or within 5 miles of the plant site boundary. Regulatory Guide 1.78 lists the chemicals most commonly encountered in the evaluation of control room habitability but is not all inclu-sive.

The design basis accident (DBA) radiation source term should be for the loss-of-coolant accident (LOCA) containment leakage and engineered safety feature (ESF) leakage contribution out-side containment as described in Appendix A and B of Standard Review Plan Chapter 15.6.5. In addition, boiling water reac-tor (BWR) facility evaluations should add any leakage from the main steam isolation valves (MSIV) (i.e., valve stem leakage, valve seat leakage, main steam isolation valve leakage control system release) to the containment leakage and ESF leakage following a LOCA. This should not be construed as altering the staff recommendations in Section D of Regulatory Guide

1. 96 (Revision 2) regarding MSIV leakage control systems.

Other DBAs should be reviewed to determine whether they might constitute a more severe control room hazard than the LOCA.

B.3-11

WNP-2 AMENDMENT NO.

17'uly 19&1 In addition to the accident analysis results, which should either identify the possible need for control room modifica-tions, or provide. assurance that the habitability systems will operate under all postulated conditions to p'ermit the control room operators to remain in the control room to take appropriate actions required by General Design Criterion 19, the licensee should submit sufficient information needed for an independent evaluatin of the adequacy of the habitability systems. Attachment 1 lists the information'hat should be provided along with the licensee's evaluation.

WNP-2 Position The WNP-2 control room meets the intent of the .requirements of the subject SRPs. The control room design description is being reviewed by NRC as part of the WNP-2 FSAR (FSAR 6.4}.

Information required to permit an independent review is provied in the FSAR.

B 3-12

f