ML20246E640

Safety Evaluation Report Related to the Operation of Comanche Peak Steam Electric Station,Units 1 and 2.Docket Nos. 50-445 and 50-446.(Texas Utilities Generating Corp)

ML20246E640
Person / Time
Site:	Comanche Peak
Issue date:	04/30/1989
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0797, NUREG-0797-S21, NUREG-797, NUREG-797-S21, NUDOCS 8905110322
Download: ML20246E640 (223)

Text

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l NUREG-0797 Supplement No. 21 Safety Evaluation Report related.to the operation of '

Comanche Peak Steam Electric Station, Units 1 and:2

. Docket. Nos. 50-445 and 50-446 Texas Utilities. Electric Company, et. al.

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation April 1989-ga"%,

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I AVAILABILITY NOTICE >

l Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:

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1. The NRC Public Document Room, 2120 L Street, NW, Lower Level, Washington, DC 20555

2. The Superintendent'of Documents, U.S. Government Printing Office, P.O. Box 37082,

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Washington, DC 20013-7082

3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publica-tions, it is not intended to be exhaustive.

Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investi-gation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence.

The following documents in the NUREG series are available for purchase from the GPO Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proceed-ings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regula-tions in the Code of Federal Regulations, and Nuclear Regulatory Commission issuances.

Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.

Documents available from public and special technical libraries include all open literature items, such as books, journal and periodical articles, and transactions. Federa' Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries.

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Single copies of NRC draft reports are available free, to the extent of supply, upon written request to the Office of Information Resources Management, Distribution Section, U.S.

Nuclear Regulatory Commission, Washington, DC 20555.

Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, 7920 Norfolk Avonue, Bethesda, Maryland, and are available there for reference use by the public. Codes and standards are usually copy-righted and may be purchased from the or.'ginating organization or, if they are American National Standards, from the American National Standards institute,1430 Broadway, New York, NY 10018.

NUREG-0797 Supplement No. 21 Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Units 1 and 2 Docket Nos. 50-445 and 50-446 Texas Utilities Electric Company, et. al.

i U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation April 1989

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l ABSTRACT  !

Supplement 21 to the Safety Evaluation Report related to the operation of the ,

Comanche Peak Steam Electric Station (CPSES), Units 1 and 2 (NUREG-0797), has 1 l been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Corrnission (NRC). The facility is located in Somervell County,  ;

Texas, approximately 40 miles southwest of Fort Worth, Texas. This supplement reports the status of certain issues that had not been resolved when the Safety Evaluation Report and Supplements 1, 2, 3, 4, 6, and 12 to that report were published. This supplement also lists the new issues that have been identifico since Supplement 12 was issued and includes the evaluations for licensing items resolved in this interin period.

Supplement 5 has not been issued. Supplements 7, 8, 9, 10, and 11 were limited I to the staff evaluation of allegations investigated by the NRC Technical Review Team. Supplement 13 presented the staff's evaluation of the Comanche Peak Response Team (CPRT) Program Plan, which was formulated by the applicant to resolve various construction and design issues raised by sources external to TU Electric. Supplements 14 through 19 presented the staff's evaluation of the CPSES Corrective Action Program for large- and small-bore piping and pipe supports (Supplement 14) cable trays and cable tray hangers (Supplement 15);

conduit supports (Supplement 16); mechanical, civil / structural, electrical, instrumentation and controls, and systems portions of the heating, ventilation, and air conditioning (HVAC) system workscopes (Supplement 17); HVAC structural design (Supplement 18); and equipment qualification (Supplement 19). Supple-ment 20 presented the staff's evaluation of the Comanche Peak Response Team implementation of the CPRT Program Plar and the issue-specific action plans, as well as the CPRT's investigations to determine the adequacy of various types of programs and hardware at CPSES.

Items identified in Supplements 7, 8, 9, 10, 11, and 13 through 20 are not included in this supplement, except to the extent that they affect the applicant's Final Safety Analysis Report.

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l TABLE CF CONTENTS Page ABSTRACT ............................................................ iii 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT .................. 1-1 1 1.1 Introduction .............................................. 1-1 1.4 Identi fication of Agents and Contractors . . . . . . . . . . . . . . . . . . 1-4 1.5 Summa ry of Pri ncipal Revi ew Matters . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.6 Modifications to this Facility During the Course of the Staff Review .......................................... 1-4 1.7 Outstanding Issues ........................................ 1-4 1.8 C o n f i rma to ry I s s u e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.9 Licensing Conditions ...................................... 1-6 3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS ........ 3-1 3.6 Protection Against Effects Associated With the Postulated Rupture of Piping .............................. 3-1 3.6.1 Inside Containment .............................. 3-1 3.6.1.1 Reactor Coolant System Main Loop Piping ................... ........ 3-1 3.6.1.2 Systems Other Than RCS Main Loop ....... 3-2 3.6.2 Outside Containment ............................. 3-3 3.6.2.5 Elimination of Arbitrary Intermediate Pipe Breaks ............................ 3-3 3.7 Seise.ic Design ............................................ 3-4 3.7.1 Seismic Input ................................... 3-4 3.9 Mechanical Systems and Components ......................... 3-6 3.9.1 Special Topics for Mechanical Components ........ 3-6 3.10 Seismic and Dynamic Qualification of Seismic Category I Nechanical and Electrical Equipment ............ 3-7 3.10.1 Seismic and Dynamic Qualification ............... 3-7 3.11 Environmental Qualification of Electrical Equipment important to Safety and Safety-Related Mechanical Equipment ................................................. 3-7 Comanche Peak SSER 21 v

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s TABLE OF CONTENTS (continued) l Page- J 3.11.4.1 Electrical Equipment Important to Safety ................................ 3-7 3.11.4.2 Safety-Related Mechanical Equipment ... 3-8 5

REACTOR COOLANT SYSTEM ......................................... 5-1 5.4 Component and Subsystem Design ............................ 5-1 5.4.2 Steam Generators ................................ 5-1 5.4.2.2-.. Steam Generators Inservice Inspection .. 5-1 6 ENGINEERED SAFETY FFATURES ..................................... 6-1 6.1 Engineered Safety Features Materials ...................... 6-1 6.1 '. 2 Organic Materials ............................... 6-1

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6.3 Errergency Core Cool i ng Sys tem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.3.5 Performance Evaluation'.......................... 6-1 7 INSTRUMENTATION AND CONTROLS ................................... 7-1 7.2 R e a c to r T ri p Sys tem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1 ~'

.7.2.6 Generic Implications of ATWS Events at Salem Nuclear Power Plant (Generic Letter 83-28)................................... 7-1 7.2.7 ATWS Mitigation System Actuation Circuitry (AMSAC)......................................... 7-9 9' AUXILIARY SYSTEMS .............................................. 9-1

9. 5 . 0 t he r Au xi l i a ry Sys tems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 9.5.1 Fire Protection ...... .......................... 9-1 9.5.1.1 Fire Protection Program Requireunts ... 9-1 9.5.1.2 Administrative Cor.trols ................ 9-2 9.5.1.3 Fire Brigade and Fire Brigade Trainin 9-3 9.5.1.4 General Plant Guidelines ............g...

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Page 9.5.1.5 Fire Detection and Suppression ......... 9-11 9.5.1.6 Fire Protection of Specific Areas ...... 9-13 9.5.1.7 Summary of Deviations From Appendix A to BTP APCSB9.5-1 and Appendix R to 10 CFR Part 50 ...................... 9-14 9.5.1.8 Conclusions ............................ 9-14 10 STEAM AND POWER-CONVERSION SYSTEM ............................. 10-1

'10.4'Other Features ........................................... 10-1 10.4.7 Condensate and Feedwater System ................ 10-1 11 RADI0ACTI VE WASTE MANAG EMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-1 11.2 System Description and Eval uation . . . . . . . . . . . . . . . . . . . . . . . 11-1 11.2.3 Solid Radioactive Waste Treatment System ....... 11-1 14 INITIAL TEST PROGRAM .......................................... 14-1 15 ACCIDENT ANALYSIS .............................................. 15-1 15.3 Infrequent Transients and Postulated Accidents ........... 15-1 15.3.8 Loss-of-Coolant Accident ....................... 15-1 22 THI-2 REQUIREMENTS ............................................ 22-1 22.2 Discussion of Requirements ............................... 22-1 II.D.1 Performance Testing of Boiling Water Reactor and Pressurized Water Reactor Relief and Safety Valves .................................. 22-1 II.E.1.1 Auxiliary Feedwater System (AFWS)

Reliability Evaluation ......................... 22-2 II.F.2 Instrumentation for Detection of Inadequate Core Cooling ................................... 22-2 II.K.3.31 Plant-Specific Calculations to Show Compliance With 10 CFR 50.46 ................... 22-2 i l

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APPENDIX A CONTINUATION OF CHRON0 LOGICAL LISTING OF CORRESPONDENCE APPEhDIX B BIBLIOGRAPHY APPENDIX D LIST OF PRINCIPAL CONTRIBUTORS APPENDIX E ERRATA TO CDMANCHE PEAK SAFETY EVALUATION REPORT AND SUPPLEMENTS APPENDIX L THE EFFECTS OF PAINT AND INSULATION DEBRIS ON THE <

PERFORMANCE OF POST-ACCIDENT FLUID SYSTEMS AT COMANCHE PEAK STEAM ELECTRIC STATION, UNITS 1 AND 2 APPENDIX M NRC STAFF EVALUATION AND RESOLUTION OF TECHNICAL CONCERNS AND ALLEGATIONS REGARDING PROTECTIVE C0ATINGS INSIDE OF THE REACTOR CONTAINMENT BUILDING AT COMANCHE PEAK STEAM ELECTRIC STATION, UNIT 1 APPENDIX Y REVIEW 0F LICENSEE AND APPLICANT RESPONSES TO NRC GENERIC LETTER 83-28, ITEM 1.2 - POST-TRIP REVIEW:

DATA AND INFORMATION CAPABILITIES APPENDIX W CONFORMANCE TO GENERIC LETTER 83-28, ITEM 2.1 (PART 1) - EQUIPMENT CLASSIFICATION (RTS COMPONENTS)

APPENDIX X CONFORMANCE TO GENERIC LETTER 83-28. ITEM 2.1 (PART 2) -

REACTOR TRIP SYSTEM VENDOR INTERFACE APPENDIX Y CONFORMANCE TO GENERIC LETTER 83-28, ITEMS 3.1.3 AND 3.2.3 - POST-MAINTENANCE TESTING 0F REACTOR TRIP SYSTEM COMPONENTS AND OTHER SAFETY-RELATED COMP 0NENTS APPENDIX Z CONFORMANCE TO GENERIC LETTER 83-28, ITEM 4.5.2 -

REACTOR TRIP SYSTEM RELIABILITY APPENDIX AA TECHNICAL EVALUATION REPORT - TMI ACTION - NUREG-0737 (II.D.1)

Comanche Peak SSER 21 viii

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1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1 Introduction The Nuclear Regulatory Commission (NRC) Safety Evaluation Report (SER), NUREG-0797, on the application of the Texas Utilities Generating Company (TUGC0) (the applicant) for a license to operate the Comanche Peak Steam Electric Station (CPSES), Units 1 and 2, was issued in July 1981. On January 16, 1987, TUGC0 informed the NRC that it had adopted a new corporate signature and would be known as TV Electric (Texas Utilities Electric Company). Since the issuance .

of the SER, the following supplements have been issued:

. Supplement I was issued in October 1981. It described the resolution of a large portion of the outstanding and confirmatory issues identified in the SER.

. Supplement 2 was issued in January 1982. It included the report of the Advisory Comittee on Reactor Safeguards (ACRS) to the NRC Chairman by letter dated November 17. 1982, which was appended as Appendix F.

Applicant and staff responses to comments by the ACRS were also included.

. Supplement 3 was issued in March 1983. It addressed outstanding and confirmatory issues resolved since Sunplement 2 was issued. The staff's evaluation of the applicant's emergency plans was also described.

. Supplement 4 was issued in November 1983. It included the staff's evaluation report on design modifications made to the Westinghouse model D4 and D5 steam generators installed at Comanche Peak.

. Supplement 5 has been cancelled. It was to have been limited exclusively to the CYGNA Independent Assessment Program. The issues from the CYGNA Independent Assessment Program have been addressed in the applicant's Corrective Action Program. The staff's evaluations of the CYGNA issues are provided in the respective SSERs for each Corrective Action Plan design work scope. Therefore, the planned supplement was never issued.

. Supplement 6 was issued in November 1984. It addressed outstanding and confirmatory issues resolved since Supplement 4 was issued. Noteworthy in this supplement was a partial exemption to General Design Criterion (GDC) 4 of Appendix A to Part 50 of Title 10 of the Code of Federal l Regulations (10 CFR Part 50) deleting the requirement for installing jet i I

impingement shields for the Unit 1 primary coolant loop piping at postulated break locations. l

. Supplement 7 was issued in February 1985. It was limited exclusively to the staff's evaluation of allegations investigated by the Technical Review Team (TRT) pertaining to plant electrical / instrumentation systems and testing programs.

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. Supplement 8 was issued in March 1985. It was limited exclusively to the staff's evaTuation of allegations investigated by the TRT pertaining to the plant's civil / structural and other miscellaneous construction and plant-readiness testing items, i

. Supplement 9 was issued in April 1985. It was limited exclusively to the '

staff's evaluation of coating requirements inside containment and allegations of coating deficiencies investigated by the TRT.

. Supplement 10 was issued in May 1985. It was limited exclusively to the staff's evaluation of allegations investigated by the TRT pertaining to the mechanical and piping areas.

. Supplement 11 was issued in June 1985. It was limited exclusively to the staff's evaluation of allegations investigated by the TP.T pertaining to quality assurance / quality control (QA/QC) practices in the design and construction of Comanche Peak.

. Supplement 12 was issued in October 1985. It updated the SER further by providing the results of the staff's review of information submitted by the applicant by letter and in Final Safety Analysis Report (FSAR) amend-ments addressing several of the issues and license conditions listed in Sections 1.7,1.8, and 1.9 of the SER that were unresolved at the time Supplement 6 was issued. Supplement 12 also identified several new issues that had been identified since Supplement 6 was issued and that were unresolved.

. Supplement 13 was issued in May 1986. It presented the staff's evaluation of the Comanche Peak Response Team (CPRT) Program Plan, which was formulatea by the applicant to resolve various design and construction issues raised by the Atomic Safety and Licensing Board, allegers, the Citizens Association for Sound Energy (CASE), and NRC inspections, as well as those raised by Cygna Energy Services during its independent design assessment.

. Supplement 14 was issued in March 1988. It presented the staff's evaluation of the applicant's Corrective Action Program related to large-

-and small-bore piping and pipe supports.

. Supplements 15 through 19 were issued in November 1988. They presented the staff's evaluation of the Corrective Action Program as related to cable trays and cable tray hangers (Supplement 15); conduit supports (Supplement 16); the mechanical, civil / structural, electrical, instrumentation and controls, and systers portiens of the heating, ventilation, and air conditioning system (HVAC) workscopes (Supplement 17); HVAC structural oesign (Supplement 18); and equipment qualification (Supplement 19).

. Supplement 20 was issued in November 1988. It presented the staff's evaluation of the Comanche Peak Response Team (CPRT) implementation of the CPRT Program Plan and the issue-specific action plans, as well as the CPRT's investigations to determine the adequacy of various types of '

programs and hardware at CPSES.

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The purpose of this supplement is to update the SER by presenting the results of the staff's review of information that the applicant submitted by letter and in FSAR amendments. The staff review addresses several of the issues and license conditions listed in Sections 1.7,1.8, and 1.9 of the SER that were unresolved at the time Supplement 12 was issued.

1 Each section or appendix of this Supplemental Safety Evaluation Report (SSER) is designated and titled so that it corresponds to the section or appendix of the SER that has been affected by the staff's additional evaluations and, except where specifically noted, does not replace the corresponding SER section or appendix. A new section (7.2.7), which was not part of the original SER, has been added. Appendix A is a continuation of the chronology of correspon-dence between the NRC and the applicant that updates the correspondence listed in the SER and Supplements I through 12. Appendix 0 includes references other than NRC documents and correspondence cited in this supplement.* Appendix D contains a list of principal contributors to this supplement. Appendix E contains a list of errata identified in preceding supplements. Appendices L and H contain the results of the staff evaluation of the applicant's pre- and post- operation coatings testing and surveillance program and the corrective actions taken with regard to the backfit test program, traceability, coatings procedures, and the Coatings Exempt Log. Appendices V through AA provide reports prepared by support contractors on review of applicant responses to several Generic Letter 83-28 action items. No changes were made to SER Appendices C, F, G, H, I, J, K, N, 0, P, Q, R, S, T, or U in this supplement.

Management and coordination of all the outstanding regulatory actions for Comanche Peak are under the overall direction of Mr. Christopher I. Grimes, the NRC Comanche Peak Project Division Director. Mr. Grimes may be contacted by calling (301) 492-3299 or by writing to the following address:

Mr. Christopher I. Grimes Comanche Peak Project Division Office of Nuclear Reactor Regulation Mail Stop 7H-17 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Copies of this supplement are available for public inspection at (1) the NRC's Public Document Room at 2120 L Street, N.W., Washington, D.C. 20555; (2) the Local Public Document Room, located at the Somervell County Public Library on The Square, P.O. Box 1417, Glen Rose, Texas 76043; and (3) the mini Local Public Document Room at the University of Texas at Arlington Library, 701 South Cooper, P.O. Box 19447, Arlington, TX 76019.

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• Availability of all material cited is described on the inside front cover of this document.

Comanche Peak SSER 21 1-3

1.4 Identification of Agents and Contractors The Texas Utilities Electric Company (TU Electric), formerly Texas Utilities Generating Company (TUGCO), is the agent with overall responsibility for design, construction, and operation of the Comanche Peak facility. TU Electric is a subsidiary of Texas Utilities Company. In February 1988, TV Electric entered into an agreement with the Texas Municipal Power Agency (TMPA), one of the minority owners, to purchase TMPA's ownership interest in the Comanche Peak J facility. In December 1988, TU Electric completed its purchase of the ownership interest of Brazos Electric Power Cooperative, Inc.

Although Gibbs & Hill, Inc. was retained as the original architect-engineer responsible for the design and engineering of Comanche Peak, TU Electric has gradually assumed more of that responsibility. TU Electric currently has responsibility for design and engineering of Comanche Peak, but it has contracted portions to engineering services contractors, such as Stone and Webster Engineering Corporation, Ebasco, Inc., and Impell Corporation, all of which work under TU Electric-approved quality assurance programs.

1.5 Summary of Principal Review Matters In addition to the seven principal review matters summarized in the SER, the staff is evaluating the following:

The implementation of the applicant's ccrrective action programs for review and reinspection of the CPSES design and construction in response to issues raised by the staff, the Atomic Safety and Licensing Board, and other sources. The successful completion of these programs, initiated in the fall of 1984, is pivotal to the licensing of CPSES.

1.6 Modifications to the Facility During the Course of the Staff Review As a result of the applicant's corrective action programs (see Section 1.5),

there have been a number of changes to the facility design. The applicant includes details corcerning these changes in FSAR amendments, which the staff reviews. The acceptability of specific design changes is covered in appropri-ate subsections of this report; other changes will be included in a future supplement.

1.7 Outstanding Issues Section 1.7 of the SER, as supplemented, identified a total of 41 outstanding issues at the time SSER 12 was issued. Those issues which were resolved in previous SSERs are no longer listed in this section.

The following outstanding issues from SSER 12 are resolved in this supplement:

(1) Leak before break - partial exemption from 10 CFR Part 50, Appendix A, GDC 4 (Section 3.6.1.1) i (2) Elimination of arbitrary intermediate pipe breaks (Section 3.6.2)

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(3) Containment sump performance (Section 6.1.2, Appendix L, and Appendix M)

(4) Additional storage capacity needed for the solid waste system before licensing Unit 2 (Section 11.2.3)

Outstanding issues that are not fully resolved are listed below. Resolution of these issues will be addressed by the staff in a future supplement to the SER.

(1) Generic implication of anticipated transient without scram (ATWS) events, Generic Letter 83-28 (Section 7.2.6)

(2) Application of leak-before-break methodology to reactor coolant system (RCS) branch lines (Section 3.6.1.2)

(3) Seismic and dynamic qualification of mechanical and electrical equipment (Section3.10)

(4) Environmental qualification of electrical and mechanical equipment in a harsh environment (Section 3.11)

(5) Environmental impact of main steamline break outside containment (Section3.11.6)

(6) Accident monitoring design - comparison to Regulatory Guide 1.97, Revision 2 (Section 7.5.2)

(7) Generic implication of anticipated transient without scram (ATWS) events, Generic Letter 83-28 (Section 7.2.6)

(8) Plant-specific design of ATWS mitigation system actuation circuitry (AMSAC)(Section7.2.7)

(9) Control room design (Section 22, I.D.1)

(10) Safet parameter display system design and parameter analysis (Section 22 I.D.2 1.8 Confirmatory Issues Section 1.8 of the SER, as supplemented, identified a total of 27 confirmatory issues at the time SSER 12 was issued. Those issues which were resolved in previous SSERs are no longer listed in this section.

The following confirmatory issues from SSER 12 are resolved in this supplement:

(1) Performance testing of boiling water reactor and pressurized water reactor i

relief and safety valves for Unit 1 (Section 22.2, II.D.1) i- (2) Recommendation GL-3 verification by test of the capability of the turbine-driven auxiliary feedwater (AFW) pump to operate for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> '

without ac power (Section 22.2, II.E.1.1) l Comanche Peak SSER 21 1-5 L - -_ _ --_ ___ _ --

(3) Plant-specific calculations to show compliance with 10 CFR 50.46 l (Section 22.2, II.K.3.31) i

. Confirmatory issues that are currently outstanding are listed below. Resolution of these issues will be addressed in a future supplement to the SEP..

(1) Inservice inspection program for compliance with 10 CFR 50.55a(g)

(Sections 5.2.4.1 and 6.6.1) 3 (2) Fracture toughness properties of Unit 2 reactor vessel materials (Sec-tions 5.3.1.2, 5.3.1.3, 5.3.2, and 5.3.3)

(3) Performance of natural circulation test per Branch Technical Position  ;

RSB 5-1 requirements (Section 5.4.3)

(4) Airlock testing requirements (Section 6.2.5.1) 1 (5) Change main steam isolation valve bypass valve operators to manual '

(Section10.3.1) i 1

(6) Qualifications of revised organization (FSAR Amendment 55); qualifications of persons assigned to unfilled positions (Section 13.1.1)

(7) Containment isolation dependability - purge valve operability analysis for loss-of-coolant (LOCA)-related dynamic loads (Section 22, II.E.4.2)

(8) Automatic trip of reactor coolant pumps during LOCA - Generic Letter 85-12 (Section 22, II.K.3.5)

(9) Performance of boiling water reactor and pressurized water reactor relief and safety valves for Unit 2 (Section 2.2.2, II.D.1) 1.9 License Conditions In Section 1.9 of SSER 12, the staff identified a total of 19 proposed license conditions. Those license r ditions which were resolved in previous SSERs are-  :

no longer listed in this sect 1on.

The following proposed license condition from SSER 12 was resolved in this supplement:

(1) Implementation of instrumentation for the detection of inadeouate core cooling (Section 22.2, II.F.2)

License conditions that are currently unoer consideration are listed below.

(1) The licensee must continue to control mineral exploration within the exclusion area; i.e., at distances beyond 2,250 feet from safety-related ctructures per General Design Criterion (GDC) 4, 10 CFR Part 50, Appendix A (Sections 2.1.2, 2.2, and 2.3).

(2) Ultrasonic testing (UT) of the bore and keyways of the low-pressure turbine discs must be performed at the first refueling cutage, and a UT Comanche Peak SSER 21 1-6

inspection must be performed every subsequent refueling outage, or about every 1-1/2 years. Should the UT techniques applied permit detection of i stress corrosion cracks in the bore / keyway region with depths of about 1 0.1 inch and should no cracks be detected at the first inspection, the licensee will be permitted an extension of the turbine disc inspectiun interval to about every 3 years, or every other refueling outage (Sections 3.5.1.3 and 10.2.2).

(3) The licensee shall implement and maintain in effect all provisions of the approved fire protection Analysis Report (as amended) andprogram, as described as approved in the SER in theitsFinal Safety and suppleneents, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior apprcval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire (Section 9.5.1).

(4) The licensee is required to submit for NRC review and approval the final evaluation and recommendations from the Transamerica Delaval, Inc. (TDI),

Diesel Generator Owners Group Program applicable to Comanche Peak and to perform those actions in response to that program for the train A and train 8 diesel generators. This must be accomplished before Unit I restart after the first refueling outage. In addition, those items / actions committed to by the applicant by letter dated August 15, 1984, and other items / actions identified by the staff in Section 9.5.9.4 of SSER 6 must be completed before fuel loading and/or before exceeding 5% of thermal rated power as specified therein (Section 9.5.9).

(5) The licensee shall have a licensed senior reactor operator (SRO) on each shift who has had at least 6 months of hot operating experience on a same-type plant, including at least 6 weeks at power levels greater than 20% of full power, and who has had startup and shutdown experience. For those shifts where such an individual is not available on the plant staff, an advisor shall be provided who has had at least 4 years of power plant experience, including 2 years of nuclear plant experience, and who has had at least 1 year of experience as a licensed SR0 at a similar type of facility. Use of advisors who are licensed only at the reactor operator (RO) level will be evaluated on a case-by-case basis. These advisors shall be retained until the experience levels identified in the first sentence above have been achieved. The NRC shall be notified at least 30 days before the date the licensee proposes to release the advisors 1 from further service (Section 13.1.2).

(6) The licensee shall maintain in effect and fully implement all provisions of the Commission-approved physical security plan, guard training and qualification plan, and safeguards contingency plan, including amendments made, pursuent to the authority of 10 CFR 50.54(p). The approved plans that contain safeguards information, as specified in 10 CFR 73.21, are collectively entitled as follows: " Comanche Peak Steam Electric Station Physical Security Plan," Revision 1 (dated August 29,1978), Revision 2 (dated September 24,1981), Revision 3 (dated Octcber 8, 1982), and Revision 4 (dated May 26,1983), as supplemented by changes to Revision 4, i

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dated January 23, 1984; " Comanche Peak Steam Electric Station Security Training and Qualification Plan," Revision 1 (dated July 1, 1981) and Revision 2 (dated December 6, 1982); and the " Comanche Peak Steam Electric Station Safeguards Contingency Plan," Revision 1 (dated August 18,1981) and Revision 2 (dated February 14,1983)(Section13.6).

(7) Performance of the following preoperational tasks and/or pretests may be deferred until after fuel load, but must be satisfactorily completed before initial criticality (Section 14):

(a) main steam isolation valve (MSIV) testing to demonstrate MSIV 5-second stroke times, MSIV auto-test timing to verify limit switch setting, and MSIV operability 30 minutes after occurrences of loss of instrument air (b) repeat of the ICP-PT-45-06 containment cooling systems to demonstrate that the system will meet Technical Specification limits in critical areas (c) repeat of the ICP-PT-37-09 check valve and hot functional safety injection in accordance with Technical Specification surveillance task for check valves (d) repeat of the ICP-PT-37-03 turbiree-driven auxiliary feedwater steam supply line check valve and drain pot level control valve (e) repeat of the ICP-PT-55-09 reactor coolant pumps test and ICP-PT-49-02 seal water and letdown flow performance test (f) repeat of the ICP-PT-55-11 thermal expansion preoperational test pertaining to. snubbers, springs, and supports (g) retest of the control room ventilation system (8) The licensee is rec,uired to provide for staff acceptance a report on the

! results of the Westinghouse Owners Group study of the steam generator tube rupture event, including plant-specific application of these results to Comanche Peak, before startup of the second operating cycle (Section 15.4.4).

(9) Completion of the description cf the program for conducting the plant function and task analysis per TMI Action Plan Item I.C.1 is to be per-formed, and the results are to be submitted for staff acceptance before startup of the second operating cycle of Unit 1 (Section 22, I.C.1).

Comanche Peak SSER 21 1-8

3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS 3.6 Protection Against Effects Associated With the Postulated Rupture of Piping .

l 3.6.1 Inside Containment 3.6.1.1 Reactor Coolant System Main Loop Piping In a letter dated August 28, 1984, the staff granted the applicant a partial exemption from General Design Criterion (GDC) 4 (10 CFR Part 50) for CPSES Unit 1. The partial exemption eliminated the requirement to install jet impingement shields at eight locations in the primary piping loop of the reactor coolant system (RCS) for Unit 1. On the basis of its evaluation of the analyses for CPSES, contained in Westinghouse Topical Report UCAP-10527, the staff determined that the applicant had presented an acceptable technical justification for not installing protective devices to mitigate the dynamic effects associated with a postulated pipe break in the RCS piping of CPSES Unit 1.

In granting the exemption, the staff had determined that the advanced fracture mechanics techniques used by the applicant provided assurance that leakage  ;

from a flaw in the primary system piping would be detected before such a flaw reached a size that could lead to unstable crack growth. For this reason, further protection provided by jet impingement shields against the dynamic effect of jet impingement resulting from the discharge from a double-ended guillotine break in the primary piping was found to be unnecessary. Although the analysis in the Westinghouse report encompassed relief from the need to install pipe break protective devices (i.e., pipe whip restraints, reactor cavity noncrushable insulation, and jet impingement shields) in Units I and 2, the applicant's request for exemption pertained solely to the installation of jet impingement shields associated with eight pipe break locations per loop in CPSES Unit 1. The partial exemption was incorporated into Amendment 8 of the Comanche Peak Steam Electric Station Unit 1 Construction Permit (CPPR-126) by letter dated October 26, 1984.

In letters dated harch 12 and July 15, 1965, the applicant requested a partial exemption from GDC 4 for CPSFS Unit 2. The Westinghouse report provided the bases for obviating the reqd rement for protection cf structures, systems, and components from all dynamic effects associated with the eight RCS pipe breaks per loop in Units 1 and 2. Furthermore, the applicant stated that granting the exemption would not affect the emergency core cooling system design bases, the containment and compartment design bases, the equipment qualification bases, the engineered safety features systems response, and the design of RCS heavy component supports.

Comanche Peak SSER 21 3-1

The Commission approved the final rule on this matter, which stated that applicants for operating licenses seeking to modify design features to take advantage of the rule are required to reflect the revised design in an amend-ment to the FSAR. On April 11, 1986, the final rule (limited-scope GDC 4 rule change) was published in the Federal Register (51 FR 12502). By letter dated June 20, 1986, the staff informed the applicant of7he final rule regarding the relief sought by the applicant for Unit 2. By letter dated November 10, 1986, the applicant transmitted advance ccpies of FSAR revisions that incorporated design changes related to the implementation of GDC 4 as modified by the final rule of April 11, 1986. The FSAR revisions were subsequently incorporated into the CPSES FSAR by Amendment 61 by letters dated December 19, 1986.

Amendment 61 to the CPSES FSAR reflects a modified design basis for CPSES Units 1 and 2. The dynamic effects associated with eight postulated primary loop breaks per loop have been removed from the design basis of CPSES Units 1 and 2.

However, the dynamic effects associated with the postulated rupture of the primary coolant loop branch line nozzles are still considered in the design basis. Accordingly, the jet impingement shields will not be installed in either Unit 1 or Unit 2. The applicant is evaluating the optimum time to remove the primary coolant whip restraints and reactor cavity noncrushable insulation from Units 1 and 2. Their removal will be documented in a future FSAR amendment. It should be noted that, although the main RCS loop breaks are not part of the modified design basis, they are still included in the load and stress evaluation for the RCS heavy component supports. Any future changes to the CPSES Unit 1 or Unit 2 RCS heavy component supports based on the modified design basis will require prior staff approval.

Conclusion The technical justification for the exemption previously granted by the staff l in the letter dated August 28, 1984, provided the basis for eliminating from the design basis the dynamic effects associated with the postulated primary coolant loop pipe breaks. These dynamic effects include pipe whip and jet impingement. Since the technical basis has been reviewed and approved previously by the staff in granting the August 28, 1984, exemption for Unit 1, the staff considers that the justification associated with that exemption is also adequate to demonstrate compliance with revised GDC 4. Therefore, the modified design basis dccumented in Amendment 61 of the CPSES FSAR is accept-able for Units 1 and 2. Furthermore, the staff concludes that for CPSES, the jet impingement shields, reactor cavity noncrushable insulation, and pipe whip restraints associated with primary coolant loop pipe breaks are unnecessary for compliance with revised GDC 4. This outstanding issue for CPSES Units I and 2 is, therefore, closed.

3.6.1.2 Systems Other Than RCS Main Loop lhe applicatier. of alternative pipe break criteria to high-energy piping systems inside containment, excluding the reactor coolant system primary loop, of CPSES Units 1 ano 2 is addressed in Section 3.6.2.5.

Comanche Peak SSER 21 3-2

The applicant requested on April 15, 1988, the elimination of the evaluation of the dynumic effects for certain main coolant loop branch line breaks using the advanced fracture mechanics techniques (leak-before-break analyses) allowed by the broad-scope GDC 4 rule change published on October 27, 1987 (52 FR 41288).

The lines included in this request are the pressurizer surge, the residual heat removal suction hot leg, and the accumulator injection lines. This '

request is under review. Therefore, this issue will be classified as an outstanding issue.

3.6.2 Outside Containment 3.6.2.5 Elimination of Arbitrary Intermediate Pipe Breaks In a letter dated May 2,1985, the applicant requested staff approval to eliminate from design consideration those pipe breaks cenerally referred to as

" arbitrary intermediate breaks." Arbitrary intermediate breaks (AIBs) are defined as those break locations that, on the basis of pipe stress analysis results, are below the stress and fatigue limits specified in Branch Technical Position (BTP) NEB 3-1 (NUREG-0800), but are selected to provice a minimum of two postulated breaks betwcen the terminal ends of c piping system. The applicant specifically requested NRC approval of the application cf alternative pipe break criteria to high-energy piping systems both inside and outside containment, excluding the reactor coolant system primary loop, of CPSES Units 1 and 2 as follows:

(1) Elimination from the structural design basis of AIBs in all high-energy piping systems identified in the letter of May 2,1985, Attachment B-2, Table 1. Breaks will centinue to be postulated at intermediate locations where the stress criteria or fatigue criteria (applicable to Class 1 piping only) of BTP MEB 3-1 are exceeded.

(2) Exclusion of the dynamic effects (pipe whip, jet impingement, and compart-ment pressurization loads) associated with AIBs from the CPSES desion basis.

(3) Elimination of the requirement for pipe whip restraints and jet shields associated with previously postulated AIBs.

The applicant responded to staff requests for aoditional information on the elimination of AIBs on July 29 and October 21, 1986. The latter response indicated that additional information on welded piping attachments and analyses  ;

of potential fatigue effects at welded attachments within five nominal pipe diameters of the AIB to be eliminated would be provided for staff review.

Generic Letter 87-11, " Relaxation in Arbitrary Intenrediate Pipe Rupture l Requirements," was issued on June 19, 1987. The generic letter revises BTP MEB 3-1 by eliminating requirements to consider all dynamic and environmental effects resulting from arbitrary intermediate pipe ruptures. However, a new and related provision, Section B.I.c.(5), in revised BTP MEB 3-1 specifies that safety-related equipment be environmentally qualified in accordance with the Standard Review Plan (HUREG-0800), Section 3.11. BTP MEB 3-1 now specifies that required pipe ruptures and leakage cracks (whichever control) must be included in the design bases for environmental qualification of electrical and mechanical equipment both inside and outside containment. In a letter dated

)

Comanche Peak SSER 21 33 1

July 15, 1987, the applicant advised the staff that, after considering the staff position in Generic Letter 87-11, it anticipated that the generic letter eliminated the need to respond to the staff's request for additional information dated September 11, 1986, regarding welded piping attachments.

In the request for staff approval of its application of the alternative pipe break criteria to high-energy piping systems described above, the applicant .

provided assurance that elimination of AIBs would not affect the environmental  !

analysis for equipment qualification. In addition, the applicant's July 29, 1986, response to the staff's request for information ensures that elimination of AIBs will not reduce the environmental qualification of safety-related equip-ment. On the basis of its evaluation and the staff positions of Generic Letter 87-11, the staff finds that the applicant's alternative pipe break criteria for high-energy piping are acceptable. However, the staff requires that the appli-cant not allow any relaxation in the environmental qualification of safety-related mechanical and electrical equipment as specified in Generic Letter 87-11 unless specific approval of the deviation is granted. This outstanding issue for CPSES Units 1 and 2 is, therefore, closed.

Conclusion On the basis of its evaluation of the applicant's submittals and the staff positions in Generic Letter 87-11, the staff has determined that the appli-cant's request to eliminate AIBs from the design considerations for CPSES Units 1 and 2 is acceptable'and can be implemented without undue risk to public health and safety, provided that, unless otherwise permitted by the staff, the applicant maintains the environmental qualification of safety-related mechanical ano electrical equipment.

3.7 Seismic Design 3.7.1 Seismic Input In a letter dated January 25, 1985, the applicant requested staff approval of the use of American Society of Mechanical Engineers, Boiler and Pressure Vessel Code (ASME Code), Code Case N-397, which contains an alternative method of seismic analysis with shifting of spectral peaks, and Code Case N-411, which concerns frequency-dependent damping values. In a letter dated June 5, 1985, the staff asked the applicant to clarify further the use of the subject Code cases. In a letter dated September 25, 1985, the staff indicated its approval of the use of the subject Code cases for broad application to seismic analysis of piping systems by the spectrum analysis method. However, specific cpproval of the subject Code cases for CpSES was conditioned on the receipt of a satis-factory response from the applicant to the staff's letter of June 5, 1985, requiring implementation of several ccnditions. The applicant responded to the NRC letters by letter dated November 18, 1985.

The staff has reviewed the applicant's letter dated November 18, 1985, regarding implementation of six criteria. The applicant has agreed to comply with the subject criteria. The staff's evaluotion of the applicant's commitments is as follows:

)

I Comanche Peak SSER 21 3-4

(1) The applicant has committed to identify in the FSAR all pipe stress packages in which the two Code cases are used. This is acceptable. The applicant intends to use the two Code cases for both the reanalysis and l requalification effort for Unit 1 and the as-built reconciliation for Unit 2. The staff approves the use of the two Code cases for the reanalysis effort and requires that any new analysis be made the analysis of record.

(2) The applicant has committed to identify the pipe stress packages in accordance with Item (1) above and to designate whether the Code cases were used for new analyses, reconciliation, or support optimization. This is acceptable. However, as indicated in Item (1) above, any stress report affected by the use of the two Code cases must be revised as the report of record.

(3) The applicant has committed to use the two Code cases only for piping systems reanalyzed by response spectrun methods, as specified in Condition (2) of Regulatory Guide 1.84, Revision 24 (June 1986), " Design and Fabrication Code Case Acceptability ASME Section III, Division 1,"

for Code Case N-411. This is acceptable.

(4) The applicant has committed to check all predicted maximum displacements affected by the use of the two Code cases for adequate clearance from adjacent structures, components, and equipment (including mounted equip-ment). This is acceptable. When the use of the Code cases results in removal of supports, the applicant committed to evaluate the piping system stability appropriately. This is acceptable. However, the staff requires that the adequacy of the remaining supports in the affected system be properly evaluated and the stability of the affected piping system be verified by experienced piping design engineers, familiar with pipe support hardware, through a physical walkdown of the affected piping system.

This verification should include a determination that the support conditiens assumed in the analytical model are consistent with the physical charac-teristics of the actual hardware used.

(5) The applicant has committed to revise the piping system design specifica-tions and required design documents to reficct the use of these Code cases.

This is acceptable.

(6) The applicant intends to use the Code cases in a manner consistent with the guidance in Welding Research Council (WRC) Bulletin 300, " Technical Position on Industry Practice." The applicant does not intend to follow each and every recommendation of WRC Bulletin 300 as though these recom-mendations are set design rules for its licensing requirements. The staff .

agrees that WRC Bulletin 300 provides recommendations rather than absolute (

requirements. The intent of the bulletin is to address areas in the l' design process that have been of concern in the past and to motivate the user to develop (or to use existing) in-house procedures for better control i of the piping design process, including interfaces with other discipliries.

l The staff understands that the applicant and its contractors have proce-dures in place that address the piping design process, including interface control, whether or not the subject Code cases are being used.

1 l

Comanche Peak SSER 21 3-5 L--_._____-_-______-__-__-__

l 1

I i

I l

1 Conclusion The staff fincs that the applicant's response to the staff's letters conditioning the use of AShE Code cases for CPSES is acceptable. The staff, therefore, approves the use of the Code Cases N-397 and 411, as indicated above.

3.9 Mechanical Systems and Components 3.9.1 Special Topics for Mechanical Components In a letter dated September 12, 1986, the applicant requested staff approval to use ASME Code Case N-253-2, " Construction of Class 2 and 3 Components for Elevated Temperature Service,Section III, Division 1. This Code case was developed to provide rules for Class 2 or Class 3 components when metal temperatures exceed those for which allowable stress values are given by Section III, Division 1. The staff understands that this Code case will be used for the recualification of those diesel generator exhaust components at CPSES that are covered under Class 3 rules.

The staff has completed its evaluation of the applicant's request dated October 26, 1987 to use a yield strength value of 42 ksi for selected pipe supports fabricated from A500, Grade B (cold-formed) structural tube shapes.

The staff understands that, generally, the pipe supports at Comanche Peak Steam Electric Station, Units 1 and 2 fabricated from A500 Grade B material were design-validated to a yield strength of 36 ksi (as specified in Revisions 10 through 14 to ASME Code Case N-71). The use of a 36-ksi yield strength at 100'F for A500, Grade B tube steel (as specified in Revisions 10 through 14 to the Code case) provided more stringent requirements. The staff believed these requirements to be justified until the currently available test data established that there was no significant reduction of the yield strength of the cold-formed material in the heat-affected zone of weldments. As a result of the recent test data, the yield strength was increased to 46 ksi in Revision 15 to ASME Ccde Case N-71.

By letter dated January 21, 1985, the applicant rcquested authorization to use ASME Code Case N-378 for application in the examination of bottom-mounted instrumentation seal tubes for CPSES Units I and 2.

i Conclusion The staff has reviewed the applicant's request to use ASME Code Case N-253-2 in the requalification of those diesel generator exhaust components at CPSES that are covered under Class 3 rules. The staff finds the use of this Code case to be acceptable provided that (1) the rules of the Code case shall govern the design and materials selection and (2) the stamping and data report shall indicate the Code case number and the applicable revision.

Although at this time the staff has not yet finalized its review of the accepta-bility of using an increased yield strength of 46 ksi for A500, Grade B tube shapes (as specified in Revision 15 to ASME Code Case N-71), the staff finds I that a 42-ksi yield strength at 100*F (as specifieo in Revision 9 to the Code l case and previously approved by the staff) can be used without undue risk to the public health and safety and is, therefore, accepta'ie.D 1

Comanche Peak SSER 21 3-6

Code Case N-378, " Examination of Piping Support Material,Section III, Division 1, Class 1," is an acceptable Code case and has been approved by the i

t Commission for general use. (It was endorsed by Revision 23 of Regulatory i Guide 1.88, issueo in September 1985.) Therefore, use of Code Case N-378 for the examination of bottom-mounted instrumentation seal tables at CPSES Units 1 and 2 is approved as written, with no additional requirements other than those specified in the Code case.

3.10 Seismic anc Dynamic Qualification of Seismic Category 1 Mechanical e.rd Electrical Equipment 3.10.1 Seismic and Dynamic Qualification In Supplement 12, the staff proposed the following license condition: 3 i

The applicant is required to complete the final closecut inspection of the installation of seismic and dynari.ic qualified equipment with all deficiencies corrected. Staff acceptance is subject to overall acceptance of the adequacy of the quality assurance program instituted i at Comanche Peak by the NRC Technical Review Team. The applicant shall complete qualification, including full documentation, before exceeding ,

5% power of operation for the differential pressure indicating switches '

(ESE-40) and the auxiliary rack (ESE-XX).

When this license condition was proposed, the staff anticipated issuance of the Unit 1 cperating license within a short period of time. Subsequent licensing delays ensued while the applicant formulated and implemented corrective actions to address numerous fscility-wide design and quality of construction issues that were identifiec. These corrective actions included major design validation efforts and a number of plant rrcdifications, as well as development of a specific workscope fer equipment qualification. The staff's evaluation of the latter is documented in SSER 19. SSER 19 ccncludes that the corrective action workscope provides a comprehensive program for resolving concerns identified by the CPRT and the NRC, including issues identified in the Comanche Peak SER and its supplements. It further concludes that the workscope implementation will ensure that the seismic and dynamic qualification of equip-ment at CPSES satisfies the validated plant design and applicable requirements of 10 CFR Part 50.

The staff ar.ticipates verifying the adequacy of implementation of the seismic and dynamic electrical and mechanical equipment qualification pregram at CPSES prior to Unit 1 fuel loading, and documenting its fircings in a future supplement. Therefore, the above license conoition is no longer applicable.

This matter will be reclassified as an outstanding issue.

3.11 Environmental Ovalification of Electrical Equipment Important to Safety and Safety-Related Mechanical Equipment 3.11.4 Qualification of Equipment l

3.11.4.1 Electrical Equipment important to Safety '

Comanche Peak SSER 21 3-7 4

3.11.4.1.2 Electrical Equipment Inportant to Safety Equipment Requiring Additionel Information and/or Correction In Supplement 12, the stoff proposed the following license condition:

All electrical equipment within the scope of 10 CFR 50.49 must in that regulation be perenvironmentally) subsections (9 ,qualified (h), (1), by (j),the anddeadlir4 (k), as appropriate.

When this l kar.se condition was proposed, the staff anticipated Unit I fuel loading withio a short period of time. Subsequent licensing delays ensuca while the applicant formulated and implemented corrective actions to address numercus facility-wide design and quality of construction issues that were identified. These corrective actions included major design revalidation efforts and a number of plant modifications, as well as development of a specific workscope for equipment qualification. The staff's evaluation of the latter is documented in SSER 19. SSER 19 concludes that the corrective action workscope prevides e comprehensive program for resolving concerns identifico by the CPRT and the NRC, including issues identified in the Comanche Peak SER and its supplements. It further concludes that the workscope implementation.

will ensure that the environmental qualification of electrical equipment at CPSES satisfies the validcted plant design and applicable requirements of 10 CFR Part 50.

The staff anticipates verifying the adequacy of implementation of the environmental qualification progran for electrical equipment at CPSES prior to Unit 1 fuel loading, ano documenting its findings in a future supplement.

Therefore, the above license condition is no longer applicable. This matter .

will be reclassified as an outstanding issue. l 3.11.4.2 Safety-Related Mechanical Eauipment In Supplement 6,* the staff proposed the following license condition:

Before power ascending above 5% of full power, the applicant must complete the safety-related mechanical equipment qualification program for all safety-related mechanical equipment located in a harsh environment and establish qualification or submit a justification for interim operation for each equipment item.

When this license condition was propesed, the staff anticipated issuance of the Unit 1 operating license within a short period of time. Subsequent licensing delays ensued while the applicant formulated and implemented corrective actions to address numerous fccility Nice design and cuality of construction issues that were identified. There corrective actions included

• License Condition (14) of SSER 6 was split in ESER 12 into two separate license conditicns. SSER 12 shows License Condition (14) as shown in Section 3.11.4.1,2 above. License Ccndition (19), pertaining to safety-related mechanical equipment in a harsh environment, was inadvertently omitted in SSER 12, but is added by an errata in Appendix E of this supplement.

Comanche Peak SSER 21 3-8

major design validation efforts crd a number of plant modifications, as well as development of a specific workscope for equipment qualification. The staff's evaluation of the latter is documented in SSER 19. SSER 19 concludes that the corrective action worksenpe provides a comprehensive program for resolving concerns identified by the CPRT and the NRC, including issues identified in the Comanche Peak SER and its supplen+r.ts. It further concludes that the workscope implementation will ensure that the environmental qualifi-cation of safety-related mechanical equipment in a harsh environment at CPSES satisfies the validated plant design and applicable requirements of 10 CFR Part 50.

The staff anticipates verifying the adequacy of implementation of the environ-niental qualification program for safety-related mechanical equipment in a harsh environment at CPSES prior to Unit i fuel loading, and documenting its findings in a future supplement. Therefore, the above license condition is no longer applicable. The matter will be reclassified as an outstanding issue.

Comanche Peak SSER 21 3-9

h ic 5 REACTOR COOLANT SYSTEM 5.4 Component and Subsystem Design 5.4.2 Steam Generators 5.4.2.2 Steam Generators Inservice Inspection In Supplement 4 (Appendix H) of.the SER, the staff requested additional infomation on the frequency and operating conditions at which accelerometer readings will be made at Comanche Peak and on the schedule for updating these data to the NRC. Submittal of this information was considered a confirmatory -

issue to be resolved in a future supplement to the SER.

In a letter dated March 28, 1988, the applicant requested staff approval to .

eliminate the vibration monitoring program fcr Comanche Peak, Units 1 and 2, based on the operating experience of Model D-4/D-5 steam generators.

The staff concurred with the applicant's'reouest to remove the vibration

. instrumentation and eliminate the vibration monitoring program in a letter dated June 1, 1988. The confirmatory issue described above is, therefore, considered closed.

Comanche Peak SSER 21 5-1

i i

6 ENGINEERED SAFETY FEATURES 6.1 Engineered Safety Features Materials 6.1.2 Organic Naterials In Appendix L of SER Supplement 9, the staff evaluated the protective coatings area at CPSES and concluded that the applicant's proposal to amend the FSAR to eliminate the comitment that coatings inside the containment buildings be qualified was acceptable. Appendix M of Supplement 9 contains the NRC

, Technical Review Team (TRT) assessment of allegations and concerns related to protective coatings. The TRT's assessment revealed specific deficiencies in the protective coatings area and resulted in recommendations for corrective actions. These recomended actions were listed in Appendix M of Supplement 9.

The corrective actions, which were related to the backfit test program, traceability, coatings proccdures, and the Coatings Exempt Log, were modified on the basis of the staff's conclusions in Appendix L of Supplement 9.

On the basis of its evaluation of the protective coatings area and the IRT's assessment of allegations and concerns related to this area, the staff recommended in Supplement 9 that the applicant (1) document the status of some protective coating systems inside the containment buildings and (2) implement a surveillance program for protective coating maintenance.

The applicant provided the requested information in letters dated June 7, 1985, November 18, 1985, and December 16, 1986. The staff's evaluation of these submittals is discussed in Appendices L and M of this supplement.

Conclusion On the basis of its evaluation of the applicant's submittals of June 7, 1985, November 18, 1985, and December 16, 1986, and as discussed in Appendices L and M of this supplement, the staff concludes that the applicant's proposed surveillance program for protective coating systems inside the Unit I and 2 containment builoings, the documentation of the status of the existing coatings work, and the actions taken by the applicant on the protective coating systems meet the guidelines in Appendices L and M of Supplement 9 and are, therefore, ,

acceptable. The staff also concludes that the applicant need not perform the in situ temperature and pressure testing for coating adhesion previously recomended in Supplement 9. Therefore, the outstanding issue regarding containment sump performance for CPSES Units 1 and 2 is considered closed.

6.3 Emergency Core Cooling System 6.3.5 Performance Evaluation In a letter dated June 2,1986 Westinghouse notified the NRC of the need for some additions to and corrections of the emergency core cooling system (ECCS)

Comanche Peak SSER 21 6-1

evaluation models that contain the WREFLO6D and the BART codes. The problems with these codes were discussed at a neeting in Bethesda, Maryland, on June 23, 1986. On October 22, 1986, the staff issued Generic Letter 86-16, " Westinghouse ECCS Evaluation Models," which stated the staff's position on the use of Westinghouse ECCS models.

The generic letter stated that, for those plants that were analy7ed with the 1978 and 1981 versiens of the Westinghouse ECCS evaluation model, the change to i the WREFLOOD code would result in a 6 F to 12"F increase in peak cladding j temperature (PCT). Westinghouse informed the NRC that the increase would not {

cause the PCT in current analyses to exceed 2200 F. Therefore, the staff {

determined that a new ECCS reanalysis is not required for plants using the 1978 and 1981 versions of the models. Additionally, Westinghouse does not plan to modify the 1978 and 1981 ECCS evaluation models or use them for future ECCS analyses.

By letter dated May 18, 1987, the applicant provided a response to Generic Letter 86-16, maintaining that the generic letter does not require an ECCS reanalysis for CPSES Units 1 and 2.

As stated in Generic Letter 86-16, the staff has accepted Westinghouse's assessment that the errors in the 1978 and 1981 ECCS evaluation models using WREFLOOD each result in a 6'F to 12"F underprediction of PCT and that there is ample margin for all facilities between the maximum PCT given in 10 CFR Part 50.46 and the calculated PCT. The applicant affirmed that this is the situation for CPSES Units 1 and 2, which have calculated PCTs of 2010.7 F and 1808'F, respectively. These PCTs are well below the 10 CFR Part 50.46 PCT limit of 2200 F; therefore, the staff concludes that reanalyses are not required for Units 1 and 2.

Comanche Peak SSER 21 6-2

I 7 INSTRUMENTATION AND CONTROLS 7.2 Reactor Trip System 7.2.6 Generic Implications of ATWS Events at Salem Nuclear Power Plant (GenericLetter83-28)

On July 8, 1983, the NRC staff issued Generic Letter 83-28, which indicated actions to be taken by licensees and applicants based on the generic implica-tions of the anticipated transient without scram (ATWS) events at the Salem Nuclear Power Plant that occurred on February 22 and 25,1983. The staff's evaluation of the responses submitted by the applicant for CPSES Units 1 and 2 to Action Items 1.2, 2.1 (Parts 1 and 2), 3.1.1, 3.1.2, 3.1.3, 3.2.1, 3.2.2, 3.2.3, 4.1, 4.5.1, and 4.5.2 of Generic Letter 83-28 is provided below.

(1) Post-Trip Review Data and Information Capability, Action Item 1.2 The staff developed the following review guidelines after its initial evaluation of the various utility responses to Action Item 1.2 of Generic Letter 83-28.

These guidelines incorporate the best features of these submittals and in effect. represent a " good-practices" approach to post-trip review. The staff t

has reviewed the applicant's response to Item 1.2 against these guidelines.

(a) The equipment that provides the digital sequence of events (SOE) record and the analog time-history records of an unscheduled shutdown should provide a reliable source of the necessary information to be used in the post-trip review. Each plant variable that is necessary to determine the cause and progression of the events following a plant trip should be monitored by at least one recorder (such as an SOE recorder or a plant process computer) for digital perameters and bystrip charts, a plant process computer, or analog recorder for analog (time-history) variables.

Performance characteristics guidelines for SOE and time-history recorderr are as follows.

. Each SOE recorder should be capable of detecting and recording the sequence of events with sufficient time discrimination capability to ensure that the time responses associated with each monitored safety-related system can be ascertained and that a determination can be made as to whether the time response is within acceptable limits based on FSAR Chapter 15 accident analyses. The recommended guideline for the SCE time discrimination is approximately 100 milli-teconds. If current SCE recorders do not have this time discrimina-tion capability, the applicant should show that the current time discrimination capability is sufficient for adequate reconstruction of the course of the reactor trip and post-trip events. As a minimum, this should include the ability to adequately reconstruct the transient and accident scenarios presented in FSAR Chapter 15.

Comanche Peak SSER 21 7-1

. Each analog time-history data recorder should have a sample interval small enough so that the' incident can be accurately reconstructed following a reactor trip. As a minimum, the applicant should be able to reconstruct the course of the transient and accident sequences  !

evaluated in the accident analysis of FSAR Chapter 15. The recom-mended guideline for the sample interval is 10 seconds. If the time-history equipment does not meet this guideline, the applicant should show that the time-histor.y capability is sufficient to accurately reconstruct the transient and accident sequences presented in FSAR Chapter 15. To support the post-trip analysis of the cause of the trip and the proper functioning of involved safety-related equipment, each analog time-history data recorder should be capable of updating and retaining information from approximately 5 minutes before the trip until at least 10 minutes after the trip.

. All equipment used to record SOE and time-history information should be powered from a reliable, noninterruptible power source. The power source used need not be Class 1E.

(b) The SCE and time-history recording equipment should monitor sufficient digital and analog parameters, respectively, to ensure that the course of the reactor trip and post-trip events can be reconstructed. The parameters  ;

monitored should provide suff'cient information to determine the root ]

cause of the unscheduled shutdown, the progression of the reactor trip, i and the response of the plant parameters and protection and safety systems ,

to the unscheduled shutdowns. Specifically, all input parameters asso-  !

ciated with reactor trips and safety injection and other safety-related i systems, as well as output parameters sufficient to record the proper func-  !

tioning of these systems, should be recorded for use in the post-trip j review. The parameters deemed necessary, as a minimum, to perform a  ;

post-trip review that would determine if the plant remained within its l safety-limit design envelope are presented in Table 7.1. These parameters l were selected on the basis of staff engineering judgement following a com-  !

piete evaluation of utility submittals. If the applicant's SOE and time- 5 history recorders do not monitor all of the parameters suggested in this >

table, the applicant should show that the existing set of monitored parameters is sufficient to establish that the plant remained within the design envelope for the accident conditions analyzed in FSAR Chapter 15.

l (c) The information gathered by the S0E and time-history recorders should be stored in a manner that will allow for data retrieval and analysis. The data may be retained in either hardcopy (e.g., computer printout or strip chart record) or in an accessible memory (e.g., magnetic disc or tape).

This information should be presented in a readable and meaningful format, taking into consideration good human-factors practices such as those outlined in NUREG-0700, " Guidelines for Control Room Design Reviews,"

September 1981.

(d) Retention of data from all unscheduled shutdowns provides a valuable reference source for the determination of the acceptability of the plant vital parameter and equipment response to subsequent unscheduled shutdowns. Information gathered during the post-trip review is to be retained for the life of the plant for post-trip review comparisons of subsequent events.

Comanche Peak SSER 21 7-2

Table 7.1 ~ Pressurized water reactor parameter list Sequence- Time of-Events History recorder. recorder Parameter / signal x'(1) Reactor trip x (1) Safety injection x Containment isolation x(1) Turbine trip x Control-rod position x(1) x Neutron flux, power x x ' Containment pressure (2) Containment radiation

-x Containment sump level x(1) x Primary system pressure x x Primary system temperature x Pressurizer level x Peactor coolant pump status x(1) x Primary system flow (3) Safety injection; flow; pump / valve status Main steam isolation valve position

~

x x x' Steam generator pressure x x Steam generator level x x Feedwater flow i

x x Steam flow h (3) Auxiliary feedwater system; flow; pump / valve status l x ACandDCsystemstatus(busvoltage) x Diesel generator status (start /stop, on/off) x Power-operated relief valve position (1) Trip parameters.

(2) Parameter may' be monitored by either a sequence-of-events (SOE) recorder ,

or a time-history recorder.  !

(3) Acceptable recorder options are (a) system flow recorded on an SOE recorder, (b) system flow recorded on a time-history recorder, or (c) equipment )

status recorded on an SOE recorder.

l Science Applications International Corporation provided support to the staff in this evaluation. The contractor's technical evaluation report is included as Appendix V to this supplement.

The applicant provided information regarding its post-trip review program data and information capabilities for CPSES by letters dated November 3, 1983, April 30,1964, and June 2, 1986. The staff has evaluated the applicant's i submittals against the review guidelines described above. The staff discussed deviations from the guidelines with the applicant's representatives by telephone on March 13, 1986. A brief description of the applicant's responses and the staff's evaluation of the responses against the review guidelines l follows.

Con.anche Peak SSER 21 7-3

(a) The applicant has described the performance characteristics of the equipment used to record the SOE and time-history data needed for post-trip review. On the basis of its review of the applicant's submittals and a telephone conversation with the applicant, the staff finds that the SOE recorder and time-history recorder characteristics conform to the guidelines in Item (a) above and are, therefore, acceptable.

(b) The applicant has established and identified the parameters to be monitored and recorded for post-trip review. On the basis of its review, the staff finds that the parameters selected by the applicant will include all but two of those identifica in Table 7.1. Feedwater flow and steam flow are not CPSES trip parameters; although these parameters are not monitored on the SOE recorder, they are monitored on time-history recorders. The staff finds this acceptable. Consequently, the staff finds that the applicant's selection of parameters meets the intent of the guidelinesinItem(b)aboveandis,therefore, acceptable.

(c) The applicant has described the means for the storage and retrieval of the information gathered by the SOE and time-history recorders and for the presentation of this information for post-trip review and analysis. On the basis of its review, the staff finds that this information will be presented in a readable and meaningful fonnat and that the storage, retrieval, and presentation conform to the guidelines in Item (c) above and are, therefore, acceptable.

(d) The applicant's submittal of November 3,1983, indicates that the data and information used during post-trip reviews will be retained in an accessi-ble manner for the life of the plant. On the basis of this information, the staff finds that the applicant's program for data retention conforms to the guidelines in Item (d) above end is, therefore, acceptable.

On the basis of its review of the applicant's submittals and a telephone conversation with the applicant, the staff concludes that the applicant's post-trip review data and information capabilities are acceptable.

(2) Equipment Classification and Vendor Interface Reactor Trip System Components (Ecuipment Qualification), Action Item 2.1 (Part 1)

Action Item 2.1 (Part 1) requires that licensees and applicants confirm that all components whose functioning is required to trip the reactor are identified as safety related in documents, procedures, and information-handling systems used in the plant to control safety-related activities, including maintenance, work orders, and parts replacement.

The Idaho National Engineering Laboratory, EG&G Idaho, Inc., provided support to the staff in this evaluation. The contractor's report is included as Appendix W to this supplement.

In letters dated November 3 and November 21, 1983, the applicant stated that all components that are required to perform the reactor trip function had been reviewed to verify that they were classified as safety-related equipment. The Comanche Peak SSER 21 7-4

l l

applicant further stated that future maintenance work orders and orders for l replacement parts will be required by procedurec to be evaluated for proper safety classification before they are approved.

On the basis of its review of the applicant's submittals, the staff finds that the applicant has confirmed that a program exists for identifying, classifying, and treating components that are required for performance of the reactor trip function as safety related. This program meets Action Item 2.1 (Part 1) of Generic Letter 83-28 and is, therefore, ecceptable.

Reactor Trip System Components (Vendor Interface), Action Item 2.1 (Part 2) )

I Action Item 2.1 (Part 2) requires that licensees and applicants confinn that an interface has been established with the nucicar steam supply system (NSSS) vendor or with the vendors of each of the components of the reactor trip system that  ;

includes (a) periodic communication between the licensee / applicant and the NSSS vendor or the vendors of each of the components of a system and (b) a system of positive feedback that confirms receipt by the licensee / applicant of transmittals of vendor technical information.

The Idaho National Engineering Laboratory, EG&G Idaho, Inc., provided support to the staff in this evaluation. The contractor's report is included as Appendix X to this supplement.

In a letter dated November 21, 1983, the applit. ant stated that Westinghouse is the NSSS vendor for CPSES Units 1 and 2 and that the reactor trip system (RTS) is included as part of the Westinghouse interface program established for this plant. The letter also confirmed that this interface program includes both periodic communication between Westinghouse and the applicant and positive feedback from the applicant in the form of signed receipts for technical information transmitted by Westinghouse.

On the basis of its review of the applicant's submittal, the staff finds that the applicant has confirmed that a vendor interface program exists with the NSSS verdor for components that are required for performance of the reactor trip function. This program meets Action Item 2.1 (Part 2) of Generic Letter 83-28 and is, therefore, acceptable.

(3) Post-Maintenance Testing i Reactor Trip System Components, Action Items 3.1.1 and 3.1.2 Action Items 3.1.1 and 3.1.2 require that licensees and applicants submit a statement indicating that they have reviewed plant test and maintenance procedures and technical specifications to ensure that post-maintenance 1 operability testing of safety-related components in the reactor trip system is l required. The statement is expected to verify that vendor-recommended test '

guidance has been reviewed, evaluated, and, where appropriate, included in the test and maintenance procedures or the technical specifications.  ;

The applicant proviued a response to Action items 3.1.1 and 3.1.2 in a letter dated November 3,1983, which advised the hRC staff that testing and maintenance procedures were reviewed. The response also confirmed that post-maintenance Con.anche Peak SSER 21 7-5

testing is adequately defined in the instrument and control procedures, as well as in the station administrative procedures controlling the authorization and performance of maintenance. The staff noted that the applicant's response indicated that post-maintenance testing requirements were being defined during the prework review of maintenance action rcouests (MARS). In February 1986, the MAR system was replaced with a work request / work order system; however, the. staff verified by review of the new administrative procedure that post-maintenance testing is adequately addressed. The applicent stated in the response that CPSES procedures contain the current testing guidance for the reactor trip system with one exception: testing to verify the state of the P-4 interlock had not yet been incorporated into plant procedures, because the hardware changes had only recently been installed. The applicant committed in

'its response to revise the procedures before fuel load.

On the basis of its review of the applicant's submittal, the staff concludes I that the applicant's commitment meets the intent of Action Items 3.1.1 and 3.1.2 of Generic Letter 83-28 ard is, therefore, acceptable.

Reactor Trip System Components, Action Item 3.1.3 The requirements for Action Items 3.1.3 and 3.2.3 are identical except that Item 3.1.3 applies these requirements to the reactor trip system components and Item 3.2.3 applies them to all other safety-related components. Because of this similarity, the responses to both items were evaluated together.

Action Items 3.1.3 and 3.2.3 require that licensees and applicants identify, if applicable, any post-maintenance test requirements in existing technical specifications that can be deacnstrated to degrade rather than enhance safety.

Appropriate changes to these test requirements, with supporting justification, are to be submitted for staff approval.

The Idaho National Engineering Laboratory, EG&G Idaho Inc., provided support to the staff in this evaluation. The contractor's report is included as Appendix Y to this supplement.

In submittals dated November 3, 1983, June 7, 1985, and August 19, 1985, the applicant stated that there were no post-maintenance testing requirements in the technical specifications for either the reactor trip system or other safety-related components that degraded safety.

On the basis of the applicant's statement that no post-maintenance test requirements were found in the technical specifications that degraded safety, the staff finds the applicant's responses for Action Items 3.1.3 and 3.2.3 of Generic Letter 83-28 to be acceptable.

All Other Safety-Related Components, Action Items 3.2.1 and 3.2.2 l

Action Items 3.2.1 and 3.2.2 require that licensees and applicants submit a statement indicating that they have reviewed plant test and maintenance proce-dures and technical specifications to ensure that post-maintenance operability

{ testing of all safety-related components is required. The statement is expected to verify that vendor-recommended test guidance has been reviewed, evaluated, and, where appropriate, included in the test and maintenance procedures or the technical specifications.

Comanche Peak SSER 21 7-6

The applicant's response of November 3, 1983, to Action Items 3.2.1 and 3.2.2 advised the NRC staff that existing plant procedures require a review of each maintenance action to determine post-maintenance testing and to ensure that the I operability of each piece of affected equipment is properly demonstrated. The applicant contaitted to issuing a list of specific testing requirements by compo-rent tag number for the responsible organizations to use as a reference. The applicant stated that CPSES test and maintenance procedures were still being developed for other safety-related components, but those procedures that were prepared had been thoroughly reviewed for testing requirements and technical accuracy, and consideration was given to vendor and engineering recommendations.

On the basis of its review of the applicant's submittal, the staff concludes that the applicant's commitment meets the intent of Action Items 3.2.1 and 3.2.2 of Generic Letter 83-26 and is, therefore, acceptable.

All Other Safety-Related Components, Action Item 2.2.3 See Action Item 3.1.3 above.

(4) Reactor Trip System Reliability Vendor-Related Modifications, Action Item 4.1, and Preventive Maintenance and Surveillance Program for Reactcr Trip Breakers, Action Item 4.2 Generic Letter 83-28 addresses intermediate-term actions to be taken by licensees and applicants aimed at ensuring that a comprehensive program of preventive maintenance and surveillance testing is implemented for the reactor trip breakers (RTBs) in pressurized water reactors. In particular, Action Item 4.1 of the generic letter requires licensees and applicants to verify that all vendor-recontr. ended RTB modifications have been implemented or that a written evaluation of the technical reasons for not implementing a modification exists.

Action Item 4.2 recuires them to submit a description of their preventive maintenance and surveillance program to ensure reliable RTB operation. The program is to include the following:

Item 4.2.1: A planned program of periodic maintenance, including lubrication, housekeeping, and other items recommended by the equipment supplier. .

l Item 4.2.2: Trending of parameters affecting operation and measured during testing to forecast degradation of operation.

In letters dated November 3, 1963, and June 7, 1985, the applicant provided responses to Action Items 4.1, 4.2.1, and 4.2.2 of Generic Letter 83-28.

In regard to Action Item 4.1, the applicant stated in its June 7, 1985, response that the undervoltage trip attachment (UVTA) replacement of reactor trip and bypass breakers, as delineated in the Westinghouse letter of April 21, 1983, has been implemented. The installation and testing were completed on June 15, 1984, with the full involvement of Westinghouse representatives.

On the basis of its evaluation of the applicant's submittal, the staff finds the applicant's position on Item 4.1 to be acceptable.

Comanche Peak SSER 21 7-7

_ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ . __ 1

Items 4.2.1 and 4.2.2 are open pending receipt of additional information from the appliccnt.

System Functional Testine, Action Item 4.5.1 Action Item 4.5.1 requires that licensees and applicants submit a statement committing to independent, on-line functional testing of the diverse trip fea-tures. In a letter dated November 3,1983, the applicant conraitted to on-line functional testing of the reactor trip system, including independent testing of the short trip coil in accordance with station surveillance procedures.

On the basis of its review of the applicant's sut,mittal, the staff concludes that the applicant's commitment meets the intent of Action Item 4.5.1 of Generic Letter 83-28 and is, therefore, acceptabic.

System Functional Testino, Action Item 4.5.2 Action Item 4.5.2 requires that licensees and applicants with plants not currently designed to permit periodic on-line functional testing of the reactor trip system, including independent testing of the diverse trip features of the reactor trip breakers, justify not making modifications to permit such testing.

The Idaho National Engineering Laboratory, EG8G Idaho Inc., provided support to the staff in its evaluation. The contractor's report is included as Appendix Z to this supplement.

In a letter dated November 3, 1985, the applicant stated that CPSES Units 1 and 2 will be designed to allow on-line testing of the reactor trip system and that this on-line testing will include independent testing of the undervoltage and shunt trip attachments of the reactor trip breakers.

On the basis of its evaluation of the applicant's submittal, the staff finds that CPSES Units 1 and 2 will be designed te permit on-line functional testing of the reactor trip system, including independent testing of the diverse trip features of the reactor trip breakers. Thus, the applicant meets Action Item 4.5.2 of Generic Letter 83-28.

Conclusion The status cf the actions required by Generic Letter 83-28 is as follows:

Action Item Status 1.1 Closed in SSER 12 1.2 Closed in this supplement 2.1 Closed in this supplement l 2.2.1 Under staff review 2.2.2 Under staff review 3.1.1 Closed in this supplement 3.1.2 Closed in this supplement 3.1.3 Closed in this supplement 3.2.1 Closed in this supplement Comanche Peak SSER 21 7-8

l Action item Status 3.2.2 Closed in this supplement 3.2.3 Closed in this supplement 4.1 Closed in this supplement 4.2.1 Input required from applicant 4.2.2 Input required from applicant 4.2.3 Input required from applicant 4.2.4 Input required from applicant 4.3 Open pending the resolution of the technical specifications 4.4 Not applicable j 4.5.1 Closed in this supplement 4.5.2 Closed in this supplement 4.5.3 Under staff review The resolution of the action items that are not closed (2.2.1, 2.2.2, 4.2.1, 4.2.2,.4.2.3, 4.2.4, 4.3, and 4.5.3) is classified as an outstanding issue.  ;

7.2.7 ATWS Mitigation System Actuation Circuitry (AMSAC)

On July 26, 1984, the Code of Federal Regulations was amended to include 10 CFR Part 50.62, " Requirements for Reduction of Risk Trom Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants" (known as the "ATWS rule"). An ATWS is an expected operational transient (such as loss of feedwater,. loss of condenser vacuum, or loss of offsite power) that

'is accompanied by a failure of the reactor trip system (RTS) to shut down the reactor. The ATWS rule requires specific improvements in the design and operation of commercial nuclear power facilities to reduce the likelihood of failure to shut down the reactor following anticipated transients and to mitigate the consequences of an ATWS event.

1 On behalf of the Westinghouse Owners Group, westinghouse submitted Topical Report WCAP-10858, "AMSAC Generic Design Package," for staff review on July 25, 1985. This document describes proposed generic AMSAC designs for compliance with 10 CFR Part 50.62.

The staff has reviewed WCAP-10858 and has concluded that the generic designs in WCAP-10858 adequately meet 10 CFR 50.62 and follow staff review guidelines.

Although WCAP-10858 presents a generic design, many details and interf. aces are of a plant-specific natizre. The staff will review the implementation of plant-specific designs tc, evaluate compliance with the ATWS rule and provide its evaluations in a future supplement to the SER, as necessary. Therefore, this issue will be classified as an outstanding issue, j

(

i Comanche Peak SSER 21 7-9

9 AUXILIARY SYSTEMS 9.5 Other Auxiliary Systems 9.5.1 Fire Protection A site audit (NRC Inspection Report 84-44) of the Comanche Peak fire protection ,

program was conducted from October 24, 1984, through November 2, 1984. This audit, which included personnel from both the Office of Nuclear Reactor Regulation (NRR) and Region IV, resulted in a number of open items.

The staff issued Supplement 12 to the SER (SSER 12) in October 1985. This supplement contained details on a review of the applicant's fire protection program through Amendment 55 of the FSAR. On October 9, 1987, the applicant submitted an extensive amendment to the FSAR, just prior to an audit that was conducted from October 19 through October 23, 1987. This amendment included revisions to the fire safe shutdown analyses, the fire hazards analysis, and organizational and design modifications as identified in Section 9.5.1 of the FSAR.

The applicant submitted Revision 1 to the Fire Protection Report by letter dated April 29, 1988. This document, which provides the basis of the CPSES fire protection program, has been incorporated into the FSAR by reference.

Revision 1 to the Fire Protection Report includes information and analyses that address many of the issues raised during the October 1987 site audit.

The staff performed a followup site visit to review the resolution of many audit issues from May 2 through 6, 1988.

The applicant compared the CPSES fire protection program with the guidelines contained in Appendix A to Branch Technical Position (BTP) APCSB 9.5-1 (NUREG-0800), as well as with Sections G J, and 0 of Appendix R to 10 CFR Part 50. Previously, the staff used the guidelines of BTP CMEB 9.5-1 to evaluate the adequacy of the fire protection program. However, the staff subsequently concluded that the applicable fire protection criteria for CPSES included Appendix A to BTP APCSB 9.5-1; Sections G, J, L, and 0 of Appendix R to 10 CFR Part 50 (applied as a guideline); and the guidance issued in Generic Letters 81-12 and 86-10.

The staff has reviewed and evaluated the CPSES fire protection program as I described in the applicant's FSAR through Amendment 71 and performed an additional audit and a site visit (October 19-23, 1987, and May 2-6, 1988, respectively) as documented below.

9.5.1.1 Fire Protection Program Requirements Fire Protection Program In SSER 12, the staff stated that the CPSES fire protection program met the guidelines of BTP CMEB 9.5-1 and was, therefore, acceptable. During the Comanche Peak SSER 21 9-1 l

Octcber 24 through November 2. 1984 (NRC Inspection Report 84-44) audit, the l staff found that the applicant's progran did not specifically assign the responsibility for fire brigade training and training records. In addition, 4 the program did not indicate that a cuality assurance (QA) program had been established for the fire protection program.

During an October 19-23, 1987 site audit, the applicant presented Procedure FIR-101, " Fire Protection Program," which had been revised to iddress these staff concerns. The staff four.o that the revisions to Procedure FIR-101 adequately addressed the assignment of responsibility for fire brigade training and records maintenance and clearly established that a QA program for fire protection would be provided. This issue was closed in NRC Inspection Report 50-445/87-22.

Generic Letter 86-10 " Implementation of Fire Protection Requirements," dated April 24, 1986, states that inclusion of the fire protection program in the FSAR is a prerequisite for the licensing of all applications now under review.

By letter dated April 29, 1988, the applicant submitted its revised Fire Protection Report. This report includes the limiting conditions for operation, action statements, and surveillance requirements for the CPSES fire protection program. The staff reviewed the procedures included in the Fire Protection Report and found that they provide a level of protection equivalent to the fire protection sections of the Westinghouse Standard Technical Specifications (STS) and are, therefore, acceptable.

Fire Hazards Analysis The staff concluded in SSER 12 that the CPSES fire hazards analysis (FHA) met the guidelines of BTP CMEB 9.5-1. The applicant has since revised the FHA and included it in the Fire Protection Report dated September 22, 1987. Revisions to the Fire Protection Report submitted to the NRC cn April 29, 1988, reflected changes to the plant design described in the Fire Safe Shutdown Analyses Report. As a result of these revisions, an additional deviation related to the residual heat removal (RHR) isolation valves was identified, as was a new deviation related to the main steam isolation valves. In addition, a number of changes were made to items that had previously been considered deviations.

Those changes that may have affected previous evaluations are discussed in the applicable sections of this evaluation. The evaluation of new deviations that were identified is discussed in the " Safe Shutdown Capability" section of Section 9.5.1.4.

9.5.1.2 Administrative Controls The staff concluded in SSER 12 that the administrative controls ider.tified by the applicant met the guidelines of BTP CMEB 9.5-1. Durino the staff audit conducted in 1984, four items were identified for which procedures were inadequate. These items are as follows:

(1) failure to designate who is responsible for obtaining a fire permit for controlling ignition sources (2) failure to delete a temporary instruction for protection of the new fuel area, even though the permanent procedure was in place Comanche Peak SSER 21 9-2

(3) discrepancies between the proposed technical specifications and the fire protection surveillance procedures (4) failure to include a fire pump performance curve in the preoperational test procedure During the site audit of October 19-E3, 1987, the applicant demonstrated that all of these discrepancies had been adequately addressed. These items are now closed, as stated in NRC Inspection Report 60-445/87-22.

9.5.1.3 Fire Brigade and Fire Brigade Training In SSER 12, the staff steted that the tire brigade and training progran met the guidelines of BTP CMED 9.5-1. During the staff audit conducted in 1984, the staff noted that the fire brigade composition was defined in several conflicting ways in plant procedures. Ir. addition, the staff found that the applicant's fire protection training procedure did not adequately accress tracking the continuing ouelification of fire bricade members. During the Octcber 19-23, 1987 audit, the applicant demonstrated that these issues have been resolved.

Therefore, the staff considers these items closed, as stated in hRC Inspection Report 50-445/87-22.

9.5.1.4 General Plant Guidelines Building Design (1) Enclosure of Fire Areas Section D.1.(j) of Appendix A of BTP APCSB 9.5-1 states that floors, walls, er.d ceilings (including penetration seals, fire doors, and dampers) enclosing separate fire areas should have a minimum fire rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

The staff stated in SSER 12 that all fire-rated assemblies at CPSES are (1) tested for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> in accordance with American Society for Testing and Materials (ASTN) Standard E-119, (2) designed in accordance with 3-hour-rated fire barrier designs obtained from the Fire Resistance Director.y published by Underwriters Laboratories (UL), or (3) constructed of a minimum of 8-inch-thick reinforced concrete in accordance with the " Uniform Building Code" (International Conference of Building Code Officials) for a minimum fire-resistance rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The staff concluded that the fire-rated walls and floor / ceiling assemblies are provided in accordance with the guidelines of BTP CMEB 9.5-1, Section C.5.a. and are, therefore. acceptable.

During the audit of October 19-23, 1987, the staff identified several barriers ,

separating redundant trains of safe shutdown equipment that were not 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated. Specifically, the staff found unrated steel hatches in fire-area boundaries. The applicant presented an analysis demonstrating that it was r.ot

( probable that a fire would prcpagate through the hatch and affect rtdundant trains of safe shutdown equipment because of the following:

(1 la combustible loading on either side of the hatches (2 automatic suppression on at least onc side of the hatch (3 1-hour-rated, fire-resistive coating on both sides of the hatch Comanche Peak SSER 21 9-3 t _ _ _ - - _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

The staff reviewed the analysis and found it to be acceptable. However, the staff noted that the unrated hatches, which were a deviation from Section D.1.(j) of Appendix A to BTP APCSB 9.5-1, were not identified in the FSAR. In Amendment 71 to the FSAR, the applicant identified the unrated steel hatches as a deviation.

(2) Enclosure of Stairwells, Elevators, and Chutes Section D.4.(f) of Appendix A to BTP APCSB 9.5-1 states, " Stairwells, elevators, and chutes should be enclosed in mansonry towers with a minimum fire rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.'- In the FSAR, the applicant stated that stairwells providing access and egress routes to areas containing safe shutdown equipment had 2-hour-rated barriers. Because of the negligible combustible loading inside stairwells and because the walls do not separate safe shutdown equipment, the steff finds the use of 2-hour-rated assemblies in stairwell boundaries to be an acceptable

' deviation from Section D.4.(f) of Appendix A to BTP APCSB 9.5-1.

During the audit of October 19-23, 1987 audit, the staff noted that a number of stairuell walls were not 2-hour-rated construction. The applicant presented an evaluation that had been performed to determine the rating of fire-area and stairwell boundaries. This evaluation was used to identify those boundaries that were not built specifically to the specifications of an independent testing organization. Where specific installation criteria were not adhered to, the evaluation was used to determine if the criteria for items such as wall thickness and material type were met or exceeded.

As a result of this evaluation, the applicant found that the installation specifications for six stairwell walls could not be directly related to an installation criterion established by a recognized approval agency. The applicant issued Deficiency Report C-87-414, which requires field-verification of the design of these walls. If the walls are not in compliance with appli-cable criteria, the walls will be brought into ccmpliance with the necessary modifications. Since the applicant has initiated the appropriate corrective action, this issue is considered closed.

(*s) Protection of Cable Tray Supports Generic Letter 86-10 states, " Cable tray supports should be protected, regard-less of whether there is a sprinkler system. However, they need not be protected if...an analysis is perfomed which takes into account the fire loading and automatic suppression available in the area and which demonstrates that the unprotected supports will not fail and cause a loss of cable-tray fire barrier required for the postulated fire."

During a site visit of May 2-6, 1988, the applicant presented an analysis to the staff that demonstrated the adequacy of the unprotected cable tray supports at CPSES. The analysis took into consideration in situ ano transient combustibles and fire-suppression systems in the areas of concern. On the basis of this analysis, the applicant instituted administrative controls to limit transient rcmbustibles so that cable tray supports would not be exposed to temperatures resulting from a fire that would cause the steel to fail. The staff reviewed Comanche Peak SSER 21 9-4

the analysis and found that it was conservative and provided reason 6ble assurance that unprotected cable tray supports would not fail under fire l conditions. Therefore, the unprotected cable tray supports that were addressed in the applicant's evaluation are found to be accept 6ble. 1 (4) Sealing or Closure of Penetrations in Fire Barriers Appendix A to BTP APCSB 9.5-1 Section D.1.(j), states, " Penetrations in fire barriers, including conduits and piping, should be sealed or closed to provide a fire-resistive rating at least equal to that of the fire barrier itself. j Docr openings should be protected with equivalent-rated doors, frames, and j h6rdware that have been tested and approved by a nationally recc9nized 3 '

l abora tory. "

During the audit of October 19-23, 1987, the staff was concerned that the methed used to seal conduits 4 inches in oiameter and smaller was not in accordance with rcted configurations and had not been identified as a deviation trom staff guidance. The applicant stated that conduits that had either suppression on both sides or detection on both sides would be sealed only on one side. Conduits that had either no detection or no suppression on at least one side would be sealed at the first opening on both sides. The staff was concerned that this plan would allow for detection only on both sides of the barrier, with no suppression en either side of the barrier. The applicant agreed to revise its position. If detection and suppression do not exist on both sides of the barrier, the applicart will seal conduits 4 inches and smaller at the barrier or will install (at the first opening) smoke and gas seals on both sides of the barrier. The staff found this approach to be acceptable.

The applicant identified six bus duct penetrations that did not have seals in  !

accordance with a tested configuration. The applicant performed an analysis '

that demonstrated that the seals provided are similar to the tested configuration and will provide the necessary protection. The staff reviewed the applicant's analysis during the site visit of May 2-6, 1988, and found it to be acceptable.

During f!RC inspection 50-445/05-16; 50-446/85-13, the inspector was concerned that documentation for certain BISCO seals used at the plant might not be adequate to justify the ratings fcr the seals. Specifically, American Nuclear Insurers (ANI) had identified one type of BISCO stal being used at CPSES as the same type of seal that had failed a fire test curing testing of a sample of BISCO seals. During the audit of October 19-23, 1987, the staff reviewed documentation provided by the applicant that demonstrated that all of the BISCO seals being installed at CPSES have passed fire tests. 1his inspection item is now considered closed, as stateo in f!RC Inspection Report 50-445/87-22.

(5) Modification of Fire Doors During the audit of October 19-23, 1987, the staff observed that a number of modifications, primarily for security hardware, had been made to fire doors.

Although the fire doors and frar.;es contained labels demonstrating compliance with UL testing criteria, the staff was concerned that the modifications would degrade the performance of the doors under fire conditions. The applicant Comanche Peak SSER 21 9-5

i presented documentation from UL that described how security modifications could be made without jeopardizing the rating of the doors; however, these Suidelines may not have been implemented during modification of the plant fire doors.

The applicant committed to evaluate all fire doors now installed to determine if the modifications comply with the guidance provided by UL. !! ben compliarce cannot be established, the applicant committed to either bringing the door into compliance or replacing it with cre that conforms to the guidelines. The cpplicant also committed to ensuring that all future modifications conform to established UL guidance. The staff finds these ccmmitment to be acceptable.

In Amendment 71 to the FSAR, the applicant identified several doors that are not mounted in accordance with UL criteria. The doors described in the FSAR dmendment are mounted to steel angles Coated With Thermo-Lag fireproofing.

These doors, which are a deviation from Section D.1.(j) of Appendix A to BTP APCSD 9.S-1, were inspectea curing the site visit of May 2-6, 1988. The staff finds that these doors could be expected to perform adequately under fire conditions. Therefore, the deviation from Section D.1.(j) of Appendix A to LTP APCSB 9.5-1 is acceptable.

(6) Prior Deviation for Heating, Ventilation, and Air. Conditioning (HVAC)

Penetration Dampers SSER 12 addressed a number of deviations dealing with heating, ventilation, and air-conditioning (HVAC) penetrations of fire-rated barriers. Because of demonstrated difficulties in the operation of these dampers under air-flow conditions, the applicant instituted a program to completely change-out the dampers, with the exception of those in stairwells. A deviation is associated with the stairwell dampers, because they cannot be mounteo completely inside the barrier because of interference with the tornado pressure-relief dampers.

The fire dampers, which protrude approximately 2 inches, are covered with a 1-hour-rated, fire-resistive material. Combustible leading on both sides of the stairwell dampers is low. TFere is reasceable assurance that these dampers would prevent the propagation of fire from one side of the barrier to the other, since the dampers are essentially in the barrier and would function normally. Thernfore, the staff finc's this deviation to be acceptable.

Ali of the replacement dampers are being installea in accordance with tested configurations and are certified to close under the air-flow conditions encountered within the plent. Therefore, with the exception of the deviations associated with the stairwells, the damper deviations described in SSER 12 are no longer necessary.

Safe Shutdown Capability (1) , Separation of Redundant Pressurizer Transformers During the 1984 audit, the staff found that the redundant presseriter trans-formers located in the safeguards building were not in compliance with the separation criteria of Section III.G.2 of Appendix R to 10 CFR Part 50. During the audit of October 19-23, 1987, the applicart stated that, on the basis of Fire Separation Calculation 152, Pevision 3, and Westinghouse's thermal Comanche Peak SSER 21 9-6

hydraulic analysis (WCAP-11331), the pressurizer transformers are no longer required to achieve safe plant shutdown. Therefore, the staff considers this item to be closed.

(2) Separation of Redundant Residual }l eat Removal (RHR) Inlet Isolation Valves By letter dated October 2, 1987, the applicant requested an additional deviation from Section III.G.2.d of N.pendix R for the RHR inlet isolation valves, because the redundant valves are within the same fire area and are not protected by automatic suppression. The valves in one set of redundant valves are located within 20 feet of each other. Valves 1-8701A and 1-87018 are located in the corridor outside the steam generator compartment (fire zone 1018); valves 1-8072A and 1-80728 are located within the steam generator compartment (fire zone 101C). The valves in the corridor are separated by approximately 40 feet.

Intervening combustibles consist of three cable trays that do not run directly between the valves. The valves inside the compartment are separated by approximately 6 feet; howcver, a partial-height concrete wall extends from just below the valve bonnet up several elevations. Thermistor strip-heat detection is provioed in both of the fire zones containing these valves.

Combustible loading inside the containment is 34,200 Btu /ftr. The loading consists mainly of reactor coolant pump lubrication oil. All four pumps have oil collection systems.

The staff was concerned that a fire inside the containment could spread between redundant RhR inlet isolation valves and affect the ability to safely shut down the plant. The combustible-loading inside the containment is low; therefore, if a fire were to occur, it would develop slowly and dissipate its heat to the i large air volume. In addition, detection is provided in both of the zones that contain the redundant valves. The detection system, which alarms in the control room, would alert the operators to a fire in the area of the valves and ensure plant fire-brigade response. Since access to the containment is restricted ouring plant operation, it is unlikely that transient combustibles or ignition sources would be introduced into the area. It is not probabic, therefore, that a fire could occur inside the containment that would disable the redundant valves in both sets of RHR inlet isolation valves. On the basis of this evaluation, the staff finds the deviation from Section III.G.2.d of Appendix R to be acceptable.

(3) Separation of Main Steam Isolation Valves (MSVs)

In letters datea April 29 and May 20, 1988, the applicant requestec a deviation from Appendix R to 10 CFR Part 50, Section III.G.2.b, regarding the lack of adequate separation between the main steam isolation valves (MSIVs) and control solenoids. MSIVs 1-HV233A, 1-HV2334A, 1-HV3440A, and 1-HV2336A are located in separate compartments at the 873-foot, 6-inch elevation in fire zone 178. A common walkway on the 880-foot, 6-inch elevation provides recess to all four compartnecnts. The compartments are separated by partial-height concrete '

barriers that extend out to the open walkway. The MSIVs are separated by a minimum horizontal distance of 15 feet around the partial barriers. l The MSIVs are equipped with stored energy actuators. They are held open by hydraulic pressure, and fail closed on loss of hydraulic fluid. Hydraulic pressure is released to close the valves by energizing either one of the two Comanche Peak SSER 21 9-7 i

_ ___ _ ___ _ _ _ _ _ - - )

redundant solenoid valves associated with each MSIV. The redundant solenoids for each of the MSIVs are located in each of the respective MSIV compartments.

The solenoids within each compartment are separated by approximately 18 inches. In the event of loss of offsite power, all four MSIVs must be operable.

A fire in this area does not require these MSIVs to be closed, except in the case of a simultaneous loss of offsite power.

The staff was concerned that a fire in the area where the MSIVs are located could damage the redundant solenoids on one or more of the MSIVs and prevent them from operating. However, area-wide detection and suppression are provided in the area that contains the MSIVs. The MSIVs are separated by a minimum of 15 feet, and the hydraulic fluid associated with each valve is the only combustible in the area. The hydraulic fluid is contained in a steel reservoir, and the piping is not subject to leakage. Because of the presence cf sprinklers anc the low combustible loading in the area, it is unlikely that a fire would propagate between valves.

The staff was also concerned that a fire in the area of a single valve might damage the redundant solenoids associated with that valve; thus preventing that valve from being closed. The applicant has committed to providing a means of manually operating one of the solenoids to allow the MSIV to clcse. Since a fire in this area would not be reason to immediately close the MSIVs, the auto-matic suppression and responding plant fire brigade could extinguish a fire affecting an MSIV, and plant shutdown would not be jeopardized. If a fire damaged both of the solenoids associated with the MSIV, control would still be possible using the manual closure method.

On the basis of information provided by the applicant in the deviation request and of the evaluation above, lack of separation between the MSIVs and redundant solenoids is found to be an acceptable ocviation from Section III.G.2.b of Appendix R to 10 CFR Part 50.

(4) Adjacent Hanholes Providing Access to Safe-Shutdown Cables During the audit of October 19-23, 1987, the staff observed that there are two adjacent manholes that provide access to service water pump power and control cables. At the time of the audit, both manhole covers were removed for maintenance purposes. The staff was concerned that a flannable-liquids spill and subsequent fire at a time when both covers were removed could jeopardize the redundant trains of safe-shutdown cables. The concern was heightened when the staff observed that the manholes were approximately 40 feet from the unlo6cing area for emergency diesel fuel oil and could be directly adjacent to the path that tanker trucks would travel to the unloading station.

The manhole covers are of substantial, steel construction ard, when in place, provide an environmentally tight cover. The applicant had performed an evaluation to demonstrate that the manhoie covers would provide the equivalency of a 3-hour barrier; however, the evaluation did not address the flammable-liquids issue. During the aucit, the applicant conmitted to administrative controi of the manhole covers to ensure that only one cover would be removed at any time during plant operation. The applicant also presented a procedure change requiring the operations department to ensure that the manhole covers Comanche Peak SSER 21 9-8

i are in place during diesel fuel unloading operations. The staff finds this resolution to be a satisfactory method of ensuring the integrity of bcth trains of service water pump cables.

(5) . Deviation on Lack of One-!!our Separation Between Various Equipment In SSER 12, the staff approved a deviation from Section III.G.2 of Appendix R to 10 CFR Part 50 for a lack of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> separation between redundant service water pumps. In a letter dated October 2,1987, the applicant stated that this deviatior. request should be expanded to include the service water recirculation valves, branch circuits, exhaust fans, and branch circuit motor control centers (MCCs). The previous deviation was granted on the basis of (1) negligible combustible loadir.g (2) the presence of early warning smoke detection, and (3) area-wide automatic suppression. This area was reviewed during the audit of October 19-23, 1987. The staff determined that the previous conclusions for granting the deviation remain valid and that expanding the deviation to include the additional equipment will not adversely affect plant safety. Therefore, the lack of 1-hour separation between the aforementioned components is an acceptable deviation from Section III.G.2 of Appendix R to 10 CFR Part 50.

(6) Manual Actions Within the Same Fire Area During the audit of October 19-23, 1987, the staff raised a concern regarding the necessity for plant operators to perform certain manual actions within the same fire area as that containing the postulated fire. During the May 2-6, 1988, site visit, the applicant presented the st6tf with evaluations detailing each of the manual actions in question, the fire protection features of the fire area in which the manual action was to be taken, and justification for acceptability of the situation. These evaluations were reviewed in detail by the staff, and each manual action in question was walked down in the field.

As a result of this review, two additieral plant modifications were deemed necessary. These modificaticos are:

(1) Fire Area CA (Containment) - Control and power cables associated with J valve 8112 (scal return) must be separated from cable interactions that could produce a spurious safety injection signal. This would preclude the need for marual actions to restore seal return in the fire area if a fire caused the loss cf this return. During the May 1988 site visit, the applicant modified its evaluation to identify this separation criterion. The applicant also stated that these sets 1 of cable will be separated in accordance with plant Design Document DBD-ME-020, " Fire Safe Shutdown Analysis" which establishes separation criteria for redundant safe shutdown devices and cables inside containment. l (2) Fire Area AA-5 (Auxiliary Building-South) - Manual action to realign I the component cooling water (CCW) valves may be required tc Legin withir.

30 minutes et a fire occurring in the same fire area. Although this is a fairly short time ir, relation to fire control concerns, the fire area is a large area that covers raultiple flocrs of the auxiliary building.

The cables that could cause the spurious operation ir, question are some distance from the valves and on another elevation. The staff reviewed the fire protection features of the fire area and founo that they provide Comanche Peak SSER 21 9-9 1

1

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i i

I reasonable assurance that a fire that might cause spurious salve operation I would not propagate within the ared and prevent the operators from gaining access to the valves. Because of the large volume of the fire area and the multiple elevations, the staff also determined that smoke and gases generated from the fire would not present a habitability problem for the 3 operators performing the manual action. The concern raised by the staff  !

was that the valves in question were approximately 15 feet above ground and l would be difficult to reach in the short amount of time required. Because of this, the applicant, by letter dated May 20, 1988, committed to making the necessary plant modifications to facilitate quick access to the valves.

7 With the acdifications identified above, the manual acticns within the same fire areas were found to be acceptable. Staff acceptance is based on the fact that there is sufficient distance to perform manual operations from the closest cable that could cause the spurious operations and that the fire protection ,

features of the fire areas and the timeframes of the manual operations required '

in relation to anticipated fire development and control are adequate.

Lighting Section III.J of Appendix R to 10 CFR Part 50 requires that all areas needed for operation of safe shutdown equipment, including access ano egress routes thereto, be provided with emergency lighting units with at least an 8-hour-battery power supply.

SSER 12 stated, " Emergency lighting will be installed in all areas of the plant that may have to be manned for safe shutdown operations and at access and egress routes to and from all areas." The staff found the emergency lighting to be acceptable on this basis. During the 1984 audit, however, the staff found that e number of lights were misaligned and that some areas requiring safe shutdown operations did not have emergency lights. This issue was identified as an open item.

During the audit of October 19-23, 1987, the applicant presented procedures that were designed to ensure the proper alignment of emergency lights. The staff observed that a number of lights were misaligned; however, the applicant stated that because of the present construction phase of the plant, it was difficult to maintain the lights in alignment. The applicant also presented a procedure for identifying locations that require emergency lights. Although the areas that were identified in the 1984 audit had been provided with lights, new areas requiring lighting had been identified as a result of changes in the safe shutdown analyses. On the basis of the applicant's commitment that all areas requiring emergency lights would have lights installed before plant operation and the staff's evaluation of the applicant's plan to meet this  !

commitment, the staff finds the emergency lighting to be acceptable.

Communication At the time of the 1984 audit, plant procedures identifiec the Gaitronics page system as the method used to contact fire-brigade and other emergency response personnel. The audit team was concerned that a control room fire could disable the page system, leaving no emergency communications system. During the audit Comanche Peak SSER 21 9-10

l of October 19-23, 1987, the applicant provided details of a recently installed radio system that would provide communications independent of the control room.

Based on the installation of the new radio system, the staff considered the concerns raised during the 1984 audit to be resolved, and the item was closed in NRC Inspection Report 50-445/07-22.

During a review of the radio system, the staff noted that the system could be disabled by a fire in certain areas of the plant and that a fire in these areas could require manual operator actions in the field. If the radio system were disabled by fire, the only method of communication between the operator and the control room would be the plant paging system. Some of the required manual operations in these areas involve regulating valve flow. The staff was concerned that, because of the lack of proximity between the valves end the pagers, the paging system might not provide adequate communication during this l type of operation. The applicant simulated these manual operations, using the

! pagers as the method of communication between the control room ano the operator in the field. The applicant demonstrated that, even if the pagers nearest the valves were inoperabic, the paging system would provide an adequate means of communication auring manual operations, such as regulating flows, should the radio system be dis 6 bled.

, 9.5.1.5 Fire Detection and Suppression Fire Detection Section E.1 of Appendix A to BTP APCSB 9.5-1 gives the minimum requirements for fire detection systems. Detection systems should comply with National Fire Protection Association (NFPA) 72D, " Standard for the Installation, Maintenance, and Use of Proprietary Protective Signaling Systems."

NFPA 72D requires fire alarm control panels to be listed or approved for their intended purpose. During the 1984 audit, the staff observec that the fire alarm panels used in the plant were neither listed nor approved in accordance with NFPA 720. This was an open item. The applicant addressed this item by sending an alarm panel that was originally to be used for training purposes to Factory Nutual (FM) for testing. At FM, a series of tests identical to those used to approve commercial systems was performed en the alarm panel. During the audit of October 19-23, 1987, the applicant supplied the audit team with an FM report documenting approval of the plant fire alarm panels. The staff reviewed the FM report and found it to be acceptable; therefore, this open item was closed and documented in NRC Inspection Report 50-445/82-22.

NFPA 720 indicates that detector placement should be in accordance with NFPA 72E, " Automatic Fire Detectors," which provides guidance on the location and spacing of detectors. During the audit of October 19-23, 1987, the staff was concerned that early warning smoke detectors might not be located in accordance with NFPA 72E. The applicant presented an evalection in which detector place-ment in each plant area was reviewed for compliance with NFPA 72E. During the site visit on May 2-6, 1988, the staff reviewed this evaluation and performed a walkdown inspection for a number of the plant areas evaluated.

Comanche Peak SSER 21 9-11

The staff found the applicant's evaluation to be acceptable; therefore, the staff finds that the spacing of detectors in those areas identified by the applicant where detector spacing does not comply with NFPA 7EE is an acceptable deviation from Appendix A to BTP APCSB 9.5-1, Section E.1.

Fire Protection Water Supply System As a result of problems with microbiologically induced corrosion (f4IC) in the l fire-water pipf ng, the applicant is replacing the current lake fire-water i supply with dedicated fire-water tanks. This modification includes the addition of redundant 504,000-gallon storage tanks and three, 50-percent-capacity fire pumps (2000 gpm at 160 psi). Two of the fire pumps will be diesel driven; the third will be electric. The staff reviewed the new design during the audit of October 19-23, 1987, and found that it complied with the guidance outlined in Section E.2 (" Fire Protection Water Supply Systems")

of Appendix A to BTP APCSB 9.5-1.

Sprinkler and Standpipe Systems Section E.3.(c) of Appendix A to BTP APCSB 9.5-1 states, " Automatic sprinkler systems should, as a minimum, confern to the requirements of appropriate standards such as NFPA 13, ' Standard for the Installation cf Sprinkler Systems'."

During the 1984 audit, the staff observed that a number of sprinkier systems in the plant did not conform to the requirements of NFPA 13. Specifically, sprinkler spacing exceeded the maximum requirements for distance from the ceiling. As a result of this open item, the applicant reviewed the installed sprinkler system against the requirements of NFPA 13. The applicant identified a number of areas where sprinkler installation was in conflict with the code.

These areas were then addressed by a major retrofit program to bring the sprinkler systems into compliance with NFPA 13. During the audit of October 19-23, 1987, the staff evaluated the sprinkler installations for com-pliance with NFPA 13. The staff fourd that all of the areas evaluated were in compliance with NFPA 13 and with Section E.3.(c) of Appf.ndix A to BTP APCSB 9.5-1. Therefore, the open item from the 1984 audit was closed.

In a reference to partial area sprinkler coverage, Generic Letter 86-10 states that "... detection and suppression providing less than full-area coverage may be adequate to comply with Appendix R. Where full-area suppression and detection is not installed, licensees must perform an evaluation to assess the adecuacy of partial suppression and detection in these areas."

During the site visit on May 2-6, 1988, the staff reviewed the applicant's evaluation of the adequacy of suppression in those areas of the plant with less than full-area coverage. The evaluation documented areas within the plant that were not provided with full-area suppression and included cetailed reviews when these areas contained safe shutdown equipment. The staff performed a walkdown inspection during the site visit and founti that the conclusions of the applicant's evaluation were velid. Therefore, the staff fincs those areas with less than full-area suppression (as documented by Calculation 0210-063-0004, " Partial Sprinkler Coverage Evaluation") to be acceptable and in compliance with the guidance provided in Generic Letter 66-10.

Comanche Peak SSER 21 0-12

During the May 2-6, 1988, site visit, the applicant presented a calculation identifying serveral areas in the plant where hose coverage was not in compliance with NFPA 14, because the hose proviaed at the stations could not reach all treas. In order to addrcss this concern, the applicant presented pre-fire strategies for the areas of concern. The staff reviewed these strategies, which instruct the responding fire brigade to bring adcitional icngths of hose to ensure that all areas can be reached, and found them te be acceptable. 1herefore, the staff finds the lack of total hose coverage in certain plant areas to be an acccptable deviction from NFPA 14.

IE Information Notice 83-41 discusses cases in which inadvertent actuations of fire suppression systers had adversely affected the operability of safety-related equipment. During the October 19-23, 1987 audit, the staff was, concerned about whether the applicant had adequately addressed this issue. The applicant presented an evaluation indicating that safety-related equipment had been inspected to ensure that the placement of fire suppression systems would not affect safety system operability if the. fire protection systems were to operate. The staff reviewed the evaluation and determined that it adequately addresses the issue of fire protection systems adversely affecting the operation of safety-r elatea systems.

llalon Suppression Systems Section E.4 of Appendix A to BTP APCSB 9.5-1 states, "The use of Halon fire extinguishing egents shoulc, as a minimum, comply with the requirements of NFPA 12A and 12B, 'Halogenated Fire Extinguishing Agent Systems - Halon 1301 and Halon 1211'."

During the October 19-23, 1987, audit, the staff was concerned that the halon system in the cable spreading room night not be in compli6nce with NFPA 12A.

The applicant it.dicated that a review of the system against the requirements of NFPA 12A had not been performed. During the May 2-6, 1988 site vuit, the applicant indicated that a full-discharge test would be perfomed on the systems, as would a complete code-compliance review. The applicant also indicated that if discrepancies to the code were found, appropriate design changes would be made to ensure ccde cnmpliance. The staff found this approach to be acceptable.

9.5.1.6 Fire Protection of Specific Areas Safety-Rolated Pumps The applicent has provided an early warning detection system that alarms locally and in the control rcom for all roots housing sufety-related pumps, with the following exceptions:

. Rcora 161 - waste evaporatcr feed pump

. Room 170 - recycle evaporator feed pump

. Rcom 172 - recycle evapcrctor feed pump The pumps listed above are classified as Safety Class 3 ccmponents. Safety Class 3 applies to those components that are not classified as Safety Clers 1 or 2 and whose failure hculd result in release to the environment of Comanche Peak SSER 21 9-13

radioactive gases ncrmally required to be held for decay. This classification also applies to components that are necessary to provide or support (1) a safety system function, (2) control of outside containment airborne radioactivity released in an accident, or (3) removal of decay heat from spent fuel. The three pumps listed above are not required for safe shutdown of the plant.

The applicant evaluated the three safety-related pump rooms and determined that, because of the substantial reinforced concrete construction of the rooms, the insignificant amount of in situ combustibles, and adequate protection in Room 174 (which houses these pumps), a fire in any one of these rooms would not endanger other safety-related equipment required for safe plant shutdown. In addition, because of the substantial construction of the mechanical components in these rooms, the structural integrity of the pumps would not be compromised by a fire in the respective pump rooms.

During the site visit of May 2-6, 1988, the staff discussed these conclusions with the applicant and inspected the specific rooms of concern. On the basis of this review, the staff agrees with the applicant's evaluation. Therefore, the staff finds the lack of detection in the three rooms containing safety-relateo pumps to be acceptable.

9.5.1.7 Summary of Deviations From Appendix A to BTP APCSB 9.5-1 and Appendix R to 10 CFR Part 50 SSER 12 provides details on 16 deviations from Appendix A to BTP APCSB 9.5-1 and Appendix P, to 10 CFR Part 50. On the basis of the evaluations in this report, the staff concludes that the 11 additional deviations listed below are also acceptable.

(1 unrated steel hatches in the fire-area boundaries (2 2-hour-rated stairwell boundaries (3 fire doors mounted to steel engle and not in conformance with Underwriters Laboratories guidelines ,

(4) dampers outside barriers I (5) untested penetration seal configurations in bus ducts (6) untested penetration seal configurations in flexible conduit (7) lack of separation and suppression for the redundant residual heat removal inlet isolation valves (8) lack of separation tetween MSIVs and MSIV solenoids (9) lack of 1-hour separation between service water isolation valves, scrvice water recirculation valves, branch circuits, exhaust fans, and branch circuit motor control centcrs (10)failuretocomplywithNFPA72Eincertainplantareas (11) failure to comply with NFPA 14 in certain plant areas 9.5.1.8 Conclusions On the basis of its evaluations, audit, and site visit, as specifically noted in previcus paragraphs, the staff concludes that the CPSES fire protection progr6m, as outlined in the applicant's FSAR through Amendment 71, provides a level of fire safety that is in conformance with, or equivalent to, staff guidance in Appendix A to BTP APCSB 9.5-1 and Appendix P. to 10 CFR Part 50.

In addition, the staff concludes that the following condition related to fire protection shall be placed in the operating license:

Comanche Peak SSER 21 9-14

The licensee shc11 implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report (as amended) and as approved in the SER and its supplements, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior apprert.1 of the Comission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

As noted previously, some areas of the fire protection program were found to be acceptable on the basis of actions to be performed by the applicant in the future. These areas are summarized below:

(1) Stairwell Boundaries

, The applicant could not demonstrate that six stairwell boundaries, which should be qualified as 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> rated, were cualified as such. The applicant will verify the construction of these walls and upgrade the walls to a 2-hour rating if a current rating of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> cannot be documented.

(2) Conduit Seals The applicant comitted to sealing conduits 4 inches and smaller in diameter at the barrier or on both sides of a barrier at the first opening, if detection and suppress <on do not exist on both sides of the barrier. The staff found the use of untested seals in flexible conduit to be acceptable. The staff also found the use of an untested penetration seal configuration in bus ducts to be acceptable.

(3) Fire Doors The staff founo that UL guidelines that describe hew security modifications that could be made to fire doors without jeopardizing the rating of the doors may not have been implemented. The applicant comitteo to a review of ell fire doors now installed to determine i if the modifications comply with the guidance provided by UL. When '

compliance cannot be established, the applicant comitted to either bringing the door into compliance or replacing the door.

(4) Manholes Subject to Flamable-Liquids Fire The staff found that redundant service water cables in manholes were subject to a flamable-liquids fire when both marholes are open. The applicant committed to providing administrative procedures that prohibit both manhole covers from being removed at the same time.

(5) Emergency Lighting The staff reviewed emergency lighting in areas containing safe shutdown equipment and the access and egress routes thereto. On the basis of modifications to the safe shutdown analyses, adcitional areas requiring CGnanche Peak SSER 21 9-15

emergency lighting were identified. The applicant has committed to having emergency lighting installed and properly aligned in all areas of the plant that may have to be staffed for safe shutdown operations and at access and cgress routes to these areas.

(6) Fire Water Supply A'new fire water supply and additional fire pumps are being installed at the plant. The staff reviewed the oesign for compliance with staff guidelines and found it to be acceptable.

(7) Halon Suppression System The applicant could not demonstrate that the installation of the Halon suppression system is in accordance with the requirements of NFPA 12A.

The applicant will perform a ccde-compliarce study on the installed system and will make t'ne necessary redifications to comply with the code if deviations are identified.

Comanche Peak SSER 21 9-16

l-I 10 STEAM AND POWER-CONVERSION SYSTEM 10.4 Other Features.

10.4.7 Condensate ana Feedwater Systems

! Inspection Report 50-446/84-36, Open Item 8436-02, references Section 10.4.7 of the SER and states, "The licensee has not demonstrated, during the main fcedwater acceptance test, or has he prepared, in the initial startup program, to demonstrate the ability to recover from a low-level water transient without causing an unacceptable feedwater/ steam generator water hammer."

The applicant's response to Q210.6,* incorporated in Amendment 60 to-the FSAR, stated that the applicant has committed to perform a preoperational test to demonstrate that no unacceptable feedwater/ steam generator water hammer would occur following a limiting transient. The applicant also stated that the staff agreed that a single test would be sufficient to demonstrate water hamer adequacy for the entire feedwater/ auxiliary feedwater/ steam generator systems.

-This test would be recovery from a low-level transient using the auxiliary feedwater pumps and with the steam generator level below the auxiliary feedwater nozzle at the start of the transient to ensure a limiting transient.

It was the staff's intent in the SER that the applicant perform a single test to demonstrate the plant's ability to recover from low steam generator level fol-lowing loss of main feedwater and initiation of auxiliary feedwater. It is the staff's understanding that the applicant has performed this test. Furthermore, it is the staff's understanding that the epplicant does not plan to perform any test on the main feedwater system to demonstrate the plant's ability to recover from a low-level water transient without causing an unacceptable feed-water / steam generator water hammer because of the configuration of the steam generator (D4 model - Unit 1; D5 model - Unit 2). It is not the staff's intention to require that the applicant perform such a test.

• Q210.6 states, " Provide the preoperational test report demonstrating the adequacy of the feedwater configuration to reduce or eliminate water hammer as stated in Section 10.4.7 of the CPSES SER dated July 1981."

Comanche Peak SSER 21 10-1

11 RADI0 ACTIVE WASTE MANAGEMENT 11.2 System Description and Evaluation 11.2.3 Solid Radioactive Waste Treatment System The storage of solid radioactive waste is discussed in Section 11.4.2.7 of the CPSES FSAR. The amount of solid radioactive waste storage required is dependent upon the availability of waste disposal sites while the plant is operating. If

, waste disposal sites are not available, more storage may be required to continue l operation. The available storage is considered adequate to license CPSES Units 1 and 2. Design changes to provide additionel storage, if needed to continue i operation, will be evaluated when submitted by the licensee. The outstanding issue of storage capacity for the solid radwaste system for CPSES Unit 2 is, therefore, considered closed.

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l Comanche Peak SSER 21 11-1

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14 INITIAL TEST PROGRAM See Section 10.4.7 for the staff's evaluation of the preoperational test to demonstrate the ability to recover from a low-level water transient without causing an unacceptable.feedwater/ steam generator water hammer.

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i Comanche Peak SSER 21 14-1 l

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15- ACCIDENT ANALYSIS 15.3 Infrequent Transients and Postulated Accidents l-15.3.8 Loss-of-CoolantAccident(LOCA) l Small-Break LOCA See Section 22.2, Item II.K.3.31, for the staff's evaluation of the applicant's small-break LOCA analyses.

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. Comanche' Peak SSER 21 15-1 w______- _ _ _ _ _ _ _ _ _ _ - - _ _ . - _ .

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l 22 TMI-2 REQUIREMENTS 22.2 Discussion of Requirements 11.D.1 Performance Testing of Boiling Water Reactor and Pressurized Water Reactor Relief and Safety Valves The staff concluded in SSER 1 that, on the basis of a review of the information submitted by the applicant, the general approach used in responding to this item was acceptable and provided adequate assurance that the requirements for perfomance testing of the relief, safety, and block valves and the associated piping would be satisfied. The staff indicated that it would continue its '

review to confirm that the Electric Power Research Iristitute/ Pressurized Water Reactor (EPRI/PWR) generic test program is acceptable for the Comanche Peak plant-specific design. ,

The results of the EPRI/PWR generic test program were transmitted to the NRC on September 30, 1982. In January 1983, EPRI issued a valve selection /justifi-cation report for the PhR safety and relief valve test program, as well as a test condition justification report. At the same time, Westinghouse issued a report on the inlet fluid conditions for pressurizer safety and relief valves in Westinghouse-designed plants. Preliminary information on the plant-specific evaluation of the adequacy of the overpressure protection system for CPSES Unit 1 was submitted by the applicant on July 8, 1981, and on March 31, 1982.

On the basis of its review of this preliminary information and Amendment 40 to the FSAR, the staff requested additional information on July 5, 1985. The applicant responded to this request on June 17, 1986. On March 24, 1987, the staff transmitted a second request for information to which the applicant responded on April 15, 1987.

The Idaho National Engineering Laboratory, EG&G Idaho, Inc., provided support to the staff in its evaluation. The contractor's report is included as Appendix AA to this supplement.

In its review of the EPRI valve test program, the staff observed that Crosby 6M6 pressurizer safety valves experienced valve chattering upon closure during tests where water was passed; however, their performance was stable imediately after valve maintenance. As a result, the staff technical evaluation report recomended that the applicant develop a formal procedure for inspection of the safety valves after each activation. On July 6, 1987, the staff requested that the applicant consider performing an inspection of the Crosby 6M6 valve internals each time water passed the valves as a result of an operational transient. By letter dated February 12, 1588, the applicant committed to performing an inspection of Crosby 6M6 pressurizer safety valve internals as soon as possible after each activation caused by an operational transient, but before power ascension.

Comanche Peak SSER 21 22-1

On the basis of its review of the information and the commitment submitted by the applicant, the staff has concluded that the applicant has provided an acceptable response to the requirements of NUREG-0737, Item II.D.1 for Comanche Peak Unit 1. Therefore, this confirmatory issue is closed for Unit 1.

Pending submittal by the applicant of the Unit 2 pressurizer safety and relief line thermal hydraulic analysis for staff review, this remains a confirmatory issue for Unit 2.

II.E.1.1 Auxiliary Feedwater System (AFWS) Reliability Evaluation 4 Recommendation GL-3 states, "At least one AFWS pump and its associated flow path )

and essential instrumentation shculd automatically initiate AFWS flow and be capable of being operated independently of any ac power for at least two hours."

THI Action Plan Item II.E.1.1 requires verification by test that the turbine-driven auxiliary feed pump can operate a minimum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> without ac power.

This issue was designated as a confirmatory issue in SSER 1.

During the performance of Procedure ICP-PT-37-03, Revision 0, " Auxiliary Feedwater Turbine Drive Pump " the pump was run for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with no ac power as part of the 48-hour run. During this 2-hour run, there was no ventilation to the room, ard the ventilation dampers were shut. The temperature and humidity require-ments for the 2-hour portion of the test are 104 F and 70 percent, respectively.

The final temperature and humidity were 107.7 F and 100 percent humidity, respectively. Gibbs & Hill analyzed the deviation and concluded in GTN-65740 i that the test was acceptable because the conditions were more severe than the test criteria, and the intent of the test was not lessened. On the basis of NRC Inspection Report 85-06, dated September 12s 1985, and its review of this information, the staff concludes that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> test meets the requirements of Item II.E.1.1. This confirmatory issue is, therefore, closed.

II.F.2 Instrumentation for Detection of Inadequate Core Cooling In SSER 12, the staff imposed the following license condition:

The applicant is required to complete implementation of instrumentation for the detection of inadequate core cooling (in conformance with TMI Action Plan Item II.F.2) before Unit 1 fuel loading, except for the Heated Junction Thermocouple (HJTC) system, which is to be implemented before startup following the first refueling outage. In addition, the applicant must submit a letter report for staff review 90 days following completion of preoperational testing of the reactor vessel level instrumentation system (RVLIS) at the beginning of the second operating cycle. See Section 22 of SSER 6.

The design for the HJTC has been completed. Installation of the HJTC probes and preoperational system tests are scheduled to be performed before Unit 1 fuel load. The staff will verify completion of both of these items during an onsite inspection. The capability of the RVLIS to detect the approach to inadequate core cooling will be demonstrated during preoperational testing.

This license condition is, therefore, resolved.

Comanche Peak SSER 21 22-2

II.K.3.31 Plant-Specific Calculations To Show Compliance With 10 CFR 50.46 TMI Action Plan Item II.K.3.31 of NUREG-0737 outlines the Commission require-ments for the industry to demonstrate that its small-break LOCA methods comply with Appendix K to 10 CFR Part 50. The techni.:al issues to be addressed were listed in NUREG-0611, including comparison with semiscale experimental test results. In response to Item II.K.3.30, the Westinghouse Owners Group (WOG) elected to reference the NOTRUMP code as the new, licensing small-break LOCA model. The NOTRUMP code and methodology are described in Westinghouse Topical Reports WCAP-10079 and WCAP-10054. The staff reviewed and approved NOTRUMP as the new licensing tool for calculating the small-break LOCA response for Westinghouse plant designs. The staff further concluded that the WOG had met the requirements of Item II.K.3.30. j i

Referencing the new computer code did not imply deficiencies in the WFLASH code  !

(which was previously used for small-break LOCA analysis) in complying with i Appendix K to 10 CFR Part 50. The decision to use NOTRUMP was based on the industry's desire to perform licensing evaluations with a computer program specifically designed to calculate small-break LOCAs with greater phenome-nological accuracy than is possible by using WFLASH.

Item II.K.3.31 of NUREG-0737 requires that each licensee or applicant submit a new small-break LOCA analysis using the model approved under Item II.K.3.30.

HRC Generic Letter 83-35, " Clarification of TMI Action Plan Item II.K.3.31,"

allowed licensees and applicants to comply on a generic basis by demonstrating that the WFLASH analyses are conservative when compared with analyses performed using NOTRUMP.

In response to this guidance, the WOG submitted Westinghouse Topical Report WCAP-11145, which contains generic comparisons to WFLASH analyses for various plant types. These include comparisons for four-loop plants of the Comanche Peak design. If plant-specific analyses are perfonned for Comanche Peak using NOTRUMP, lower peak cladding temperatures (PCT) should be expected than if the generic NOTRUMP analysis was perfornied (about 537*F lower than the 1790'F PCT l currently calculated with the WFLASH small-break LOCA evaluation model).

Although the calculated peak temperatures are significantly lower for the NOTRUMP analyses than for the WFLASH analyses, the 4-inch break remains the limiting break size.

The staff has reviewed KCAP-11145 and accepted it as a licensing basis for small-break LOCA analysis. The applicant has referenced WCAP-11145 (which consists of the results of calculations using approved methodology) instead of submitting a plant-specific analysis and has met the criteria in Generic Letter 83-35. The staff, therefore, concludes that the CPSES Units 1 and 2 FSAR analyses of small-break LOCAs have been demonstrated to be conservative in comparison with the NOTRUPP evaluation model analyses. This meets the requirements of Item II.K.3.31 and 10 CFR 50.46 for CPSES and closes out this confirmatory issue.

Comanche Peak SSER 21 22-3

APPENDIX A CONTINUATION OF CHRON0 LOGICAL LISTING 0F CORRESPONDENCE L This appendix continues the chronological listing of routine licensing correspondence between the NRC staff and the applicant since issuance of Supplement 12.

September 12, 1985 Letter from. applicant transmitting change to Revision 4 of Physical Security Plan.

September 12, 1985 Letter from applicant transmitting scenario manual for 1985 emergency preparedness exercise.

September 19, 1985 Letter from applicant concerning plans to meet requirements of Generic Letter 84-16.

September 19, 1985 Lettor from applicant transmitting a request for exemption from schedular requirements for full participation in emergency preparedness

! exercise.

September 19, 1985 Letter from applicant concerning qualift-cation and training of operators.

September 19, 1985 Letter from applicant concerning qualifica-tion and training of operators.

September 20, 1985 Letter from applicant concerning response to Generic Letter 85-12 and TMI Item II.K.3.5.

September 25, 1985 Letter to applicant concerning FSAR Amendment 55 and protection of control room operators against a chlorine release accident.

September 25, 1985 Letter to applicant concerning use of ASME l' Code Cases N-397 and N-411.

September 27, 1985 Letter to applicant concerning operator licensing exams (Generic Letter 85-18). .

i September 27, 1985 Letter to applicant concerning reportir.g requirements on primary coolant iodine spikes (Generic Letter 85-19).

Comanche Peak SSER 21 1 Appendix A

September 30, 1985. Letter from' applicant transmitting deferred MSIV preoperational test status report, v September 30, 1985 Letter to applicant transmitting request for additioral. information regarding FSAR Amendments 54 and 55.

September 30, 1985- Letter to applicant transmitting Revision 2 to staff evaluation of.CPRT Program Plan.

0ctober 3, 1985. Letter from applicant transmitting Revision 0 to CPP-022.

' October 3, 1985 Letter to applicant transmitting FEMA report verifying adequacy of State and local government emergency preparedness plans.

i October 3, 1965 Letter from' applicant concerning supplementary evaluation of visual weld inspection techniques.

~ October 14, 1985 Letter from applicant concerning results of its evaluation of design review.

' October 14, 1985 Letter from applicant concerning implementa-tion of ATWS mitigation system actuation circuitry per Generic Letter 85-06.

October 15, 1985 Letter from applicant transmitting table providing revised description of Amendment 56 to FSAR.

October 24,-1985 Letter from applicant transmitting revised Emergency Plan procedures, including Revision 3 to EPP-309, Revision 5 to EPP-310, and.

Revision 5 :to EPP-101.

' October 24 1985 Letter from applicant transmitting list of training and drills conducted for offsite agencies over the past 2 years, and concern-ing exemption request from emergency exercises.

October 24, 1985 Letter to applicant transmitting Supple-nient 12 to the SER.

October 25, 1985 Letter from applicant transmitting revised objectives and guidelines for-1985 emergency preparedness exercise.

October 28, 1985 Letter from applicant concerning corporate organizational changes.

I p

Comanche Peak SSER 21 2 Appendix A

i October 30,-1985 Letter from applicant concerning estimates for future operator licensing examinations l in response to Generic Letter 85-18.

November 7, 1985 Letter from applicant concerning corporate policy statement being developed regarding conduct of control room operations and management responsibilities per IEN 85-53.

November 18, 1985 Letter from applicant concerning imple-mentation of NRC conditions for use of ASME Code Cases N-397 and N-411.

November 18, 1965 Letter from applicant concerning results of review of coatings nonconformance reports.

November 22, 1985 Letter from applicant.concerning CPRT program plan.

November 26, 1985 Letter from applicant concerning status of items to be completed before fuel load and/or before exceeding 5% themal-rated power.

December.3, 1985 Letter from applicant concerning request for additional infomation regarding FSAR Amend-ments 54 and 55.

December 3, 1985 Letter to applicant concerning potential for loss of post-LOCA recirculation capability due to insulation debris blockage (Generic Letter 85-22).

December 5, 1985 Letter from applicant transmitting various sections of pump and valve specifications.

December 5, 1985 Letter to applicant concerning changes to service and circulating water systems since issuance of Final Environmental Statement (FES)inSeptember1981.

December 5, 1985 Letter to applicant concerning interim guidance on the emergency planning standard specified in 10 CFR 50.47(b)(12) regarding Comanche Peak.

December 5, 1985 Letter to applicant concerning Texas Utilities Generating Company's responses to Generic Letter 83-28 (Items 1.2, 4.1, 4.2.1, and 4.2.2).

December 10, 1985 Letter to applicant transmitting request for additional information regarding alternate pipe break criteria.

Comanche Peak SSER 21 3 Appendix A

December 16,;1985 ' Letter from applicant transmitting Revision 4 to ODA-204, Revision 0 to EOS-0, and others.

December'16, 1985 . Letter from applicant concerning implement-ation of accident monitoring system and i

proposed schedule for neutron flux monitoring instrumentation.

December 16, 1985 Letter from applicant transmitting Revision B t:

o LOCA and seismic analysis.

December 16, 1985 Letter from applicant concerning control'of corbicule and microbiologically induced Corrosion.

i December 16, 1985 Letter from applicant concerning use of emergency response guidelines during containmentLsump circulation.

December 16, 1985 Letter from applicant.concerning fracture-toughness properties of Unit 2 reactor vessel materials.

50ecember 17, 1985 Letter from applicant concerning use of ASME Code Case N-378 for examination of bottom-mounted seal tables.

December 20, 1985 Letter from applicant transmitting table describing FSAR Amendment 57.

December 30, 1985- -Letter to applicant transmitting proposed license. condition regarding security contingency guard training.

January 15; 1986 Letter to applicant transmitting a copy of the environmental assessment for the GDC $,

request for exemption for Unit 2.

- Janua ry 20, 1986 Letter from applicant transmitting justifi-cation for the requests.for exemption from 10 CFR Part 50 Appendix J containment airlock leakage testing requirements.

January 23, 1986 Letter to applicant concerning technical resolution of Generic Issue B-19, " Thermal-HydraulicStability"(GenericLetter86-02).

January 24, 1966 Letter from applicant transmitting Endorse-ments 16 and 17 to Nuclear Energy Liability l Insurance Association (NELIA) Policy NF-274. 1 Comanche Peak SSER 21 4 Appendix- A

- _ - _ -_ _ _ _ A

January 27, 1986 Letter from applicant transmitting Revision 3 to CPRT Program Plan and Issue-Specific Action Plans.

January 31, 1986 Letter from applicant transmitting Revision 1 to Appendix D, "CPRT Sampling Approach Applications and Guidelines."

February 4,1986 Letter from applicant concerning extension of Construction Permit CPPR-126.

February 4, 1986 Letter to applicant transmitting request for additional information regarding extension of CPPR-126.

February 5, 1986 Letter to applicant concerning extension of latest construction completion date.

February 5, 1986 Letter from applicant transmitting Revision 1 of Appendix E, CPRT procedure for classification, evaluation of design and construction dis-crepancies.

February 7, 1986 Letter from applicant concerning Comanche Peak task force review and evaluation of allegations of intimidation and harassment.

February 7, 1986 Letter from applicant transmitting response to remaining SER open items for SPDS.

February 7, 1986 Letter from applicant transmitting Appendix E of the CPRT Program Plan.

February 10, 1986 Letter to applicant concerning applications for license amendments (Generic Letter 86-03).

February 10, 1986 Letter to applicant transmitting order extending latest construction completion date for CPPR-126. l February 13, 1986 Letter to applicant transmitting a policy I statement regarding engineering expertise on shift (Generic Letter 86-04).

February 14, 1986 Letter from applicant concerning modifica-tions to water treatment for nonradwaste systems after issuance of FES in September 1981.

February 17, 1986 Letter from applicant transmitting response to Section IV of Generic Letter 85-12.

Comanche Peak SSER 21 5 Appendix A

February 20, 1986 Letter from applicant concerning FEMA approval  ;

of offsite emergency plans regarding guard requirements.

p

February 21, 1986 Letter from applicant transmitting Revision 1 to inservice testing program.

February 27, l9'86- Letter from applicant concerning investigation of Ebasco Nastran cracking element.

February 27, 1986 Letter to appifcant concerning request for change to CPSES. Technical Specifications-concerning replacement of MSIV bypass valve  ;

l actuators.

February 28, 1986 . Letter from applicant concerning withdrawal of request for approval of paragraphs and tables from Subsection NF of ASME Code.

February 28, 1986 Letter from applicant transmitting revised CPRT Issue-Specific Action Plan regarding testing.

February 28, 1986 Letter from applicant concerning recommended dynamic system test configuration.to address slenderness ratio limits for tension members in cable tray supports.

February 28, 1986 Letter to applicant transmitting request for additional information regarding occupational radiation exposure.

February 26, 1986- Letter from applicant concerning role of overview quality team.

March 3, 1986 Letter to applicant concerning exemption request from full-participation emergency exercise.

March 6, 1986 Letter to applicant concerning emergency communications capabilities.

March 7.-1986- Letter to. applicant transmitting Promatec Part 21 report concerning defective material in fire stops.

March 13, 1986 Letter to applicant transmitting approval of request to use ASME Code Cases N-397 and N-411.

March 14, 1986 Letter from applicant transmitting Ebasco response to questions dircussed at NRC audit of cable tray design. '

Comanche Peak SSER 21 6 Appendix A

L I

i March 14, 1986 Letter from applicant concerning criterie for train C ccnduits 2-inches and less in dia-neter (Regulatory Guide 1.29 issue).

March 19, 1986 Letter from applicant concerning meeting to discuss use'of shift advisors.

March 20,-1986 Letter from applicant concerning status of S&W piping and pipe support qualification program.

tiarch 20, 1980 Letter to applicant concerning San Lnufre Unit I loss-of-power and water-hammer event (GenericLetter86-07).-

March 23, 1986 Letter to applicant concerning availability of Supplement 4 to NUREG-0933, "Prioritization of Generic Safety Issues" (Ceneric Letter 86-08).

March '24, 1966 Letter to applicant concerning ectivities authorized by CPPR-126 and CPPR-127.

March 25,1986 Letter to applicant transmitting comments regarding CPRT Program Plan.

March 26, 1986 .

Letter from applicant transmitting revisions to Energency Plan Procedures EPP-104 and EPP-118.

tiarch 28, 1986 Letter to applicant concerning allegation management system.

March 31, 1986 Letter to applicant concerning technical resolution of Generic Issue B-59 regarding h-1 loop operation in BWRs and PWRs (Generic Letter 86-09).

April 2,1986 Letter from applicant transmitting endorsement 18 to NELIA Policy NF-274. ..I April 3,1986 Letter from applicant concerning implementation of procedures per Office of Inspection and Enforcercent Bulletin (IEB) 79-14.

i April 4. 1986 Letter from applicant transmitting.CPRT report. j l

April 11, 1986 Letter from applicant transmitting Ebasco Volume I, Book 9. "Multimooe Response Multiplier- i Studies." i i

April 11, IS66 Letter from applicant concerning perched water I evaluation resolving concerns with water leakage into Category I structures, i

-Comanche. Peak SSER 21 .7 Appendix A

I n

L ,

April '17, 1986 Letter from applicant concerning proposed civil penalty. i f> Aprf1'18, 1986 Letter from applicant concerning on-shift  ;

experience.

April 21, 1986 Letter to applicant transmitting copies of j draft HUREG CR-4405 concerning PRAs. j April 24 1986 Letter to applicant concerning implementation of fire protection requirements (Generic Letter 86-10). ,

April 28, 1986

' Letter to applicant transmitting request for additional information regarding CPRT results reports.

.May 1, 1986 Letter from applicant transmitting Revision 2 to Appendix II, " Generic Maintenance Matrix

. and Justifications," of TDI Diesel _ Generator Owners Group report.

May 2, 1986 Letter from applicant concerning changes in management.

'May 2, 1986 Letter from applicant concerning CPRT results reports.

-May 2, 1986 Letter from applicant concerning CPRT results report for Issue-Specific Action Plans I.A.4, I.B.3 II48, III.D, and VII.B.2.

-May 2, 1986- Letter to applicant concerning its response to Ceneric Letter 85-12 regarding implementa-tion of THI Action Item II.K.3.5.

May 9, 1986 Letter to applicant transmitting request for CPRT checklists to continue NRC review within 30 days.

, .May 13, 1986 Letter from applicant transniitting Revision 4 to CPRT Issue-Specific Action Plans III.A.1 and III.A.9 and Revision 2 to VII.B.I.

'May 14, 1986 Letter from applicant transmitting public version of revised Emergency Plan procedures.

,May 14, 1986..

Letter from applicant concerning slenderness ratio limitations for cable trays.

May 15, 1986 Letter to applicant transmitting investigative report regarding document control allegations.

Comanche Peak SSER 21 8 Appendix A l

I - _ _ - _ - _ _ -

i i

l l May 15, 1986 Letter to applicant transmitting request for edditional information regarding CPRT Issue-Specific Action Plan VII.b.2, " Valve Dis- ,

assembly."

May 19, 1986 Letter to applicant transmitting request that it respond to Regulatory Guide 9.3.

May 19, 1986 Letter from applicant concerning response to Reg. Guide 9.3.

May 23, 1986 Letter from applicant concerning results report for CPRT Issue-Specific Action Plan VII.b.2.

May 28, 1986 Letter from applicant concerning schedule for full-participation emergency preparedness exercise. l May 30, 1986 Letter from applicant concerning response to Generic Letter 86-04.

June 2, 1986 Letter from applicant concerning additional information for response to Generic '

Letter 83-28.

June 3, 1986 Letter to applicant transmitting listed items in support of invoice for operating license review costs.

June 6, 1986 Letter from applicant transmitting Senior Review Team results report including Revision I to CPRT Issue-Specific Action Plan I.A.3 and Revision 1 to VII.A.4.

June 6, 1986 Letter from applicant transmitting Gibbs &

Hill, Inc., analysis of cable splice termina-tions located in raceways in response to Regulatory Guide 1.76, Revision 1.

June 9, 1986 Letter to applicant transmitting Supplement 13 to the SER.

June 9, 1986 Letter to applicant concerning establishment of detailed audit program master schedule for overview cuality team reviews.

June 9, 1986 Letter from applicant concerning procedures that will contain approved CPRT checklist.

June 11, 1986 Letter to applicant transmitting investigative i report regcrding regenerated training record.

Comanche Peak SSER 21 9 Appendix A

.' June 11, 1986 Letter to applicantLacknowledging receipt of response to IEB 85-03 and granting 2-month f .- . extension to finalize response to Item 1(E).

June'13, 1986 Letter from applicant concerning response to

. NUREG-0737, TMI Action' Item II.D.1.

June 13, 1966 Letter to applicant transmitting request for information regarding status of CPRT Results Report I.A.4.

, June 17, 1986 Letter from applicant transmitting description of Amendment 59 to the FSAR.

June.19, 1986 Letter to applicant transmitting request for

additional information regarding CPRT Program Plan.

June 20, 1986 Letter to applicant transmitting Commission approval of. change to GDC 4 and effect on applicant's exemption request.

June 20',-1986 Letter to applicant con'cerning continuation-p of periodic. meetings on CPRT activities.

June 30. 1986- Letter from applicant transmitting a table providing the description of Amendment 58 to the FSAR.

July 1, 1986 Letter from applicant transmitting Amendment 58 to the FSAR.

July 9, 1986 Letter from applicant transmitting S&W technical issues report to. assist in the evaluation of pipe stress analysis.

July 11.-1986 Letter from applicant transmitting procedure manuals regarding CPRT checklists.

July-11, 1986 Letter from applicant concerning extension for response to Regulatory Guide 9.3.

July 14, 1986 Letter from applicant concerning response to IEB 85-03.

July 15, 1986 Letter to applicant transmitting request for additional information regarding CPRT Results Report VII.B.2.

. July 16, 1986 Letter to applicant transmitting SER on Item 2.1 of Generic Letter 83-28.

Comar,che Pe'ak SSER 21 10 Appendix A

I L

. July 23, 1986 Letter from applicant concerning status of I CPRT Issue-Specific Action Plan I.A.4.

July 23, 1986 Letter from applicant transmitting Revision 2 to Appendix E of the CPRT Program Plan. l July 23, 1986 Letter from applicant transmitting revised overview of the CPRT activities.

July 25, 1986- Letter from applicant transmitting Revision 2 to TDI. Diesel Generator Owners Group design review and quality revalidation report.

e July 25, 1986 ' Letter from applicant concerning CPRT. Issue-Specific Action Plan VII.B.2 results report.

July 28,.1986 Letter from applicant transmitting Revision 2 to CPPP-7. .

' July 2'8, 1986 Letter from applicant concerning implemen-tation procedures for S&W's resolution of generic technical issues related to pipe

. stress analysis and pipe support design.

July 29, 1986~ Letter from applicant concerning elimination of arbitrary intermediate pipe breaks.

July 31, 1986 Letter to applicant concerning review of three issue-specific action plans pertaining to hot functional testing, receipt and storage of purchased material, and onsite fabrication.

August 1, 1986 Letter from applicant concerning response to IEB 85-01.

August 6, 1986 Letter from applicant transmitting pages 23-34 of Revision 6 to Emergency Plan Procedure EPP-309.

August 6, 1986 Letter to applicant concerning CPRT results discussed in an August 6, 1986 conference call.

August 11, 1986 Letter from applicant transmitting CPRT Issue-Specific Action Plan III.A.S.

August 11, 1986 Letter from applicant transmitting CPRT results reports, including Revision 2 to Issue-Specific Action Plan report I.A.4 and Revision 1 to Issue-Specific Action Plan report I.A.S.

Comanche Peak SSER 21 11 Appendix A

i y

August 15, 1986 Letter.to applicant transmitting SER regarding TDI. Diesel Generator Owners . Group findings and recommendations from program addressing operational and regulatory issues.

i L

-August 19, 1986 Letter from applicant transmitting Revision 0 to 1 CPRT Results Report III.A.4.

August 20, 1986.' Letter to applicant concerning operator licensing examinations. (Generic Letter 86-14).

August 26, 1986 Letter from applicant concerning approval request for use of ASME Code Case N-253-4.

September 2, 1986 Letter from applicant transmitting response to Regulatory Guide 9.3.

September 8, 1986 Letter from a'pplicant transmitting (1) corrected information on Brazos Electric Power Cooperative,.Inc.: ' Antitrust Operating License Review and (2) responses'to Regulatory Guide 9.3.

Letter to applicant transmitting request for September-11, 1986 additional information regarding elimination of arbitrary intermediate pipe breaks.

September 12, 1986 Letter from applicant amending earlier approval request for use of ASME Code Case N-253-4 to Case N-253-2.

September'12, 1986 Letter from applicant transmitting Revision 1 to CPRT Results Report I.D.3.

September 12, 1986 Letter from applican- transmitting revised FSAR pages.

September 16, 1986 Letter from applicant (Spiegel & McDiarmid) transmitting antitrust review responses for Brazos Electric Power Cooperative,'Inc.

September-17, 1986 Letter from applicant transmitting evaluation and resolution of generic technical issues for cable tray hangers.

September 19, 1986 Letter from applicant transmitting S&W Comanche Peak project procedures.

September 22, 1986 Letter to applicant concerning compliance j with 10 CFR 50.49 (Generic Letter 86-15). ]

September 24, 1986 Letter from applicant transmitting additional information regarding Generic Letter 85-12.

1 i

' Comanche Peak SSER 21 12 Appendix A l

.- u j

i 4 SeptAmber 26,.1986 Letter from applicent transmitting CPRT

. results report, Revision 1, to Issue-Specific Action Plan I.D.2.

0ctober 3, 1986 Letter from applicant transmitting schedule for future FSAR. amendments resulting from analysis-to justify cable splices in raceways.

October.7, 1986 Letter from applicant concerning proposed FSAR amendment incorporating changes to S&W piping.

and pipe support requalification effort. 3 i

October 10, 1986 Letter from applicant transmitting revised ,

procedure nianuals for CPRT checklists. <

October 10, 1986 Letter from applicant transmitting FSAR-revisions reflecting modifications to design

- basis in response to modified GDC 4.

October 17, 1986 Letter from applicant. transmitting Revision I to CPRT Issue-Specific Action Plan VII.A.9.

October 21, 1986 Letter from applicant concerning elimination of arbitrary intermediate pipe breaks.

October 22, 1986- Letter to applicant concerning Westinghouse ECCS evaluation models (Generic Letter 86-16).

October 30, 1986 Letter to applicant transmitting SER regarding lesser separation between Class 1E and certain non-Class 1E circuits.

October 31, 1986 Letter from applicant transmitting TUGCO, Texas Municipal Power Agency, and Brazos Electric Cooperative Inc. annual reports.

November 4, 1986 Letter from applicant concerning evaluation and resolution of CYGNA train A and B issues as applicable to train C conduits.

November 4, 1986 Letter from applicant transmitting Revision 1 to evaluation and resolution of generic technical issues for conduits and conduit supports.

j November. 4, 1986 Letter to applicant concerning acceptability  !

of proposed revision to the FSAR design criteria for piping and pipe supports.

November 10, 1986 Letter from applicant concerning implemen-tation of revised GDC 4.

Comanche Peak SSER 21 13 Appendix A

"1 1

' November 12, 1986 Letter from applicant concerning response-to  !

IEB 86-03.

November 14, 1986' Letter from applicant transmitting piping and pipe support requalification program for Unit I large-bore piping.

. November 19, 1986 Letter from applicant transmitting Volumes I and II of CPRT results reports.

' November 21. 1986 Letter to applicant transmitting Federal Regist_er notice of receipt'of antitrust information per Regulatory Guide 9.3.

November 21, 1986 Letter to applicant acknowledging receipt of updated antitrust information.

November 24, 1986- Letter from applicant transmitting FSAR-revision regarding heat treatment and shot-peening on Unit I steam generator tubes.

November 25, 1986 Letter from applicant transmitting detailed description of FSAR Amendment 60.

November 25, 1986- Letter from applicant transmitting Form 8-K.

- . November 28, 1986 Letter to applicant concerning certification that systems are abating or controlling atmosphere pollutants, contaminants, or water pollution to a greater extent.

December 2, 1986 Letter from applicant transmitting CPRT responses regarding welding.

December 8, 1986 Letter from applicant concerning habitability of control room during postulated chlorine release.

' December 8, 1986 Letter from applicant concerning use of manual operators on MSIV bypass valves.

December 8, 1986 Letter from applicant concerning control room pressurization.

December 10, 1966 Letter from applicant transmitting revised Emergency Plan procedures.

December 16, 1986 Letter from applicant concerning coating

. performance program.

- December 18, 1986 Letter from applicant transmitting S&W small-bore piping and pipe supports generic issues report.

Comanche Peak SSER 21 14 Appendix A

i December 19,.1986 Letter from applicant transmitting Revision 1 to CPRT Issue-Specific Action Plans V.A, I.B.2, I.B.1, and VII.A.7.

December 19, 1986 Letter from applicant transmitting cable separation test program..

December 19,.1986- Letter from applicant transmitting'FSAR

! Amendment 61.

December. 19,-1986' Letter from applicant concerning revisions contained in FSAR Amendment 61.

December 22, 1986- _

Letter from' applicant concerning antitrust update and Regulatory Guide 9.3.

December 22, 1986 Letter from applicant. transmitting design bases consolidation program plan.

December 23, 1986 . Letter from applicant transmitting electrical generic issues report resolving specific design bases issues.

' December 23, 1986' Letter from applicant transmitting Revision 0 to evaluation _and resolution of generic issues for HVAC systems.

. December 29, 1986 Letter to applicant transmitting evaluation supporting its response to Generic Letter 83-28. Item 2.1.

. December 31, 1986 Letter to applicant transmitting request for information regarding list of discrete program components for each corrective action plan organization.

December 31, 1986 Letter from applicant concerning security training and qualification program.

December 31, 1986 Letter from applicant concerning followup 1 report regarding allegations.

' January 5, 1987 Letter.from applicant transmitting Revision 0 to generic issues technical report regarding TERA equipment qualification.

January 8,.1987' Letter.to applicant concerning public availa-bility of NRC operator licensing examination question bank (Generic Letter 87-01).

January 13, 1987 Letter to applicant transmitting acceptance of CPRT Program Plan overview quality team and audit program.

I Comanche Peak SSER 21 15 Appendix A

l January 16, 1987 Letter from applicant transmitting Revision 0 to instrumentation and controls generic issues report.

January 16, 1987 Letter from applicant transmitting Revision 1 to CPRT Results Reports I.A.1,.I.B.4, II.C V.D.

and VII.A.3.

January 16, 1987 Letter from applicant transmitting additional i information concerning presentation on opera-

.bility of electrical' equipment'to ACRS.

' January 16, 1987 Letter from. applicant concerning adoption of new corporate signature.

January'22, 1987 Letter from applicant transmitting Endorse- ,

ment 19 to NELIA Policy NF-274. l l

' January 26, 1987 Letter from applicant transmitting revisions J to Comanche Peak Steam Electric Station i manuals.

January 27, 1987 Letter from applicant concerning updated approach to resolving cable. tray hanger issues.

February 3, 1987 Letter to applicant transmitting request for additional information regarding BISCO electrical penetration seals and Namco switches.

February 9, 1987 Letter from applicant concerning BISCO l electrical penetration seals and penetration  ;

switches per IEB 79-28. l J

February 13, 1987 Letter from applicant concerning NUREG-1216.

February 13, 1987 Letter from applicant transmitting WCAP-11174 regarding heat treatment and shot-peening of l Unit I steam generator tubes.

February 18, 1987 Letter from applicant concerning mechanical generic issues report.  ;

l Letter to applicant transmitting request for February 18, 1987 additional information regarding piping spool piece size and location of spool piece within chemical and volume control system.

February 19, 1987 Letter to applicant concerning verification of I seismic adequacy of mechanical and electrical equipment.

Comanche Peak SSER 21 16 Appendix A  ;

i

LFebruary 27. 1987 Letter from applicant transmitting Revision 0 to CPRT Results Report III.B and Revision 1 to Report VII.B.1.

February 27, 1987 Letter to applicant concerning seismic adequacy of equipment under provisions of Unresolved Safety.IssueA-46(GenericLetter87-03).

March 3, 1987 Letter to applicant transmitting evaluation of its response to Generic Letter 83-28, Item 4.5.2.

March 6, 1987 Letter to applicant concerning temporary s

exemption from provisions of FBI criminal history rule for temporary workers (Generic Letter 87-04).

March 11, 1987 Letter to applicant concerning masonry wall supports and note b in FSAR Table 17A-1.

March 13, 1987 Letter from applicant concerning definition of " mild environment" equipment category.

March 13,.1987 Letter to applicant concerning periodic veri-fication of leak-tight integrity of pressure isolation valves (Generic Letter 87-06).

March 19, 1987 Letter to applicant concerning final rule-making for revisions to operator licensing (GenericLetter87-07).

March 19, 1987 Letter to applicant transmitting acceptability of Revision 4 to Guard Training and Qualification Plan.

March 24, 1987 Letter to applicant concerning status of protective coating system and surveillance programs.

March 24, 1987 Letter to applicant transmitting request for adoitional information regarding HUREG-0737, TMI Action Item II.D.I.

March 25, 1987 Letter to applicant transmitting summary of confirmations needed to close out Generic Letter 83-28 Item 2.2.1.

March 26, 1987 Letter from applicant transmitting' Revision 3 to S&W Procedure CPPP-7.

March 27, 1987 Letter from applicant concerning schedule to eliminate arbitrary intermediate pipe breaks.

Comanche Peak SSER 21 17 Appendix A

L i i

i March 31, 1987 Letter to applicant transmitting sumary of March 10, 1987 meeting with its represen-tative regarding security organization brief- )

ing.

April 1, 1987- Letter from applicant transmitting Revision 5 to Physical. Security Plan.

April 1, 1987 Letter from applicant transmitting detailed description of FSAR Amendment 62.

April 6, 1987 Letter to applicant transmitting Region IV security inspector response to question regarding miscellaneous amendments to 10 CFR 73.55.

April 8, 1987 Letter from applicant concerning removal of hot cell.

April 13, 1987 Letter from applicant concerning antitrust matters.

April 13, 1987 Letter from applicant transmitting revised documents addressing generic technical issues for cable tray hangers.

April 14, 1987 Letter from applicant transmitting. documents regarding generic technical issues for cable tray hangers.

April 15, 1987 Letter from applicant transmitting conenents on Supplement 12 to the SER, Section 9.5.1.

April 15, 1987 Letter from applicant concerning response to NUREG-0737. TMI Action Item II.D.1.

April 24, 1987 Letter from' applicant concerning eddy current inspectability of heat-treated steam generator tubes.

April 27, 1987 Letter from applicant transmitting Revision I to CPRT Issue-Specific Action Plans 1.A.2, III.A.5, and VI.A.

April 29, 1987 Letter from applicant concerning response to Generic Letter 85-06.

May 5, 1987 Letter from applicant concerning progre::s of prudency audit.

May 9, 1987 Letter to applicant concerning applicability of Unit 1 Significant Hazards Consideration to Unit 2.

Comanche Peak SSER 21 18 Appendix A

i May 11, 1987 Letter from applicant concerning response to ,

IEB 86-02.  !

May 11, 1987 Letter to applicant concerning implementation of 10 CFR 73.55 (Generic Letter 87-08).

May 12, 1987 Letter to applicant transmitting request for additional information regarding CPRT Program Plan identifying activities required for licensing.

May 18, 1987 Letter from applicant concerning response to Generic Letter 86-16.

May 19, 1987 Letter from appifcant transmitting Revision 2 to generic issues report regarding conduits and conduit supports.

May 19, 1987 Letter from applicant transmitting Texas Municipal Power Agency and Brazos Electric Power Cooperative, Inc., annual reports.

June 3, 1987 Letter to applicant concerning protection of emergency preparedness exercise from disclosure to participants.

June 4, 1987 Letter from applicant concerning schedule for full-participation emergency field exercise.

June 5, 1987 Letter from applicant transmitting Revision 3 to the Security Training and Qualification Plan.

June 8, 1987 Letter from applicant transmitting Revision 1 to CPRT Results Reports VII.A.1, VII.A.2, and VII.B.4.

June 8, 1987 Letter to applicant transmitting safety evaluation regarding utility design changes and implementation of revised GDC 4 as described in FSAR Amendment 61.

June 8, 1987 Letter from applicant concerning investigation of administrative controls on fuel building fire protection program.

June 10, 1987 Letter from applicant transmitting documents concerning as-built documentation for piping and pipe supports.

June 12, 1987 Letter to applicant concerning implementation of 10 CFR 73.57 (Generic Letter 87-10).

Comanche Peak SSER 21 19 Appendix A

' June 15, 1987 Letter. from applicant transmitting FSAR Amendment 63.

June 15, 1987 . Letter from applicant transmitting descrip- I tion of administrative corrections made in FSAR Amendment 13.

i.

-June 19,'1987 Letter to applicant concerning relaxation in  !

arbitrary intermediate pipe rupture require-ments(GenericLetter87-11).

June 23,'1987' Letter to' applicant transmitting a request for.additioral information regarding evalua-tion of Revision 1 to CPRT Issue-Specific Action  ;

Plans.I.D.2, VII.A.4, and VII.A.8? .

l

. June 25, 1987 Letter from applicant transmitting Revision 4 to  !

Volumes I and II of CPRT Program Plan. j June 26, 1987 Letter from applicant concerning station I blackout issue.

June'30, 1987  ; Letter from applicant transmitting revised FSAR Table 3.2-1.

JulyL8, 1987. Letter from applicant transmitting Revision I to CPRT Issue-Specific Action Plan III.A.I.

r July.9. 1987. Letter.to applicant concerning loss of-residual heat removal while reactor coolant system is partially filled (Generic o Letter 87-12). I 1

' July.10, 1987. Letter from applicant transmitting Endorse-ments'20'and 21 to NELIA Policy NF-274.

' July.15, 1987 Letter from applicant concerning Generic Letter 87-11 and further work in area of i arbitrary intennediate pipe rupture require- '

uents.

July 17, 1987 Letter to applicant transmitting oversize reference chart on corrective action programs.

July 21,'1987 Letter frem applicant concerning FSAR Table 17A-1.  ;

July.22,.1987 Letter to applicant requesting additional information on groundwater withdrawals (Unit 2).

i Comanche Peak SSER 21 20 Appendix A

._____-_L

b July'27, 1987 Letter to applicant transmitting request for c1hrification on items listed regarding

' responses to Regulatory Guide 9.3.

July 31, 1967 Letter from applicant transmitting revisions to the CPRT checklists.

August 3, 1987 Letter from applicant transmitting description of FSAR Amendment 64.

August 3, 1987- Letter from applicent transmitting Revision 1 to large-bore pipe stress and pipe support E

generic issues report.

August 4,1987 Letter from applicant transmitting table of comments regarding fire protection program.

August 4, 1987 Letter to applicant concerning operator licensing examinations (Generic Letter 87-14).

August 14, 1967 Letter from applicant transmitting Revision 1 to CPRT Issue-Specific Action Plans I.D.2 and VII.A.4.

August 14, 1987 Letter to applicant transmitting. Standard Technical Specifications for development of Unit 1 plant-specific Technical Specifi-cations.

August.20, 1987 Letter from applicant concerning inter-relationship between CPRT and corrective

-action programs.

August 25, 1987 Letter from applicant transmitting Revision 1 to CPRT Results Report III.C.

August 25, 1987 Letter to applicant transmitting request for additional information regarding groundwater withdrawal rates.

August 28, 1987 Letter from applicant concerning corrective ,

action program. )

Aug'ust 28, 1987 Letter from applicant transmitting initial markup of Westinghouse Standard Technical Specifications for Unit 1.

September 1, 1987 Letter to applicant concerning proper use of health physics network emergency communi-cations system.

September 1, 1987 Letter to applicant concerning schedule for team inspection to ensure compliance with BTP CHEB 8.5-1.

1

- Comanche Peak SSER 21 21 Appendix A

i

' September 2, 1987- Letter to applicant transmitting request for additional infomation regarding response-to Generic Letter 83-28. Item 4.2.

-September 8, 1987 Letter from applicant transmitting response {

to NRC Eulletin 87-01.  !

September 8, 1987 Letter from applicant concerning antitrust matters.

September 8, 1987 Letter from applicant transmitting executive sunenary of the post-construction hardware validation program.

. September 8,1987 Letter from applicant transmitting Revision 0 to Quality Assurance Procedure NQA 3.07-1.01.

September 9, 1987 Letter from applicant concerning groundwater withdrawal rates from 1979 to present.

September 11, 1987 Letter from applicant transmitting Emergency Plan revisions including Revision 2 to EPP-118 and EPP-202.

Septeniber 16, 1987 Letter to applicant transmitting request for edditional information regarding CPRT results report for Issue-Specific Action Plan V.A.

September 21, 1987 Letter to applicant concerning September 4, 1987, meeting discussing Region IV preoper-ational inspection experience in security and how experience might be applied to plant effort.

September 23, 1987 Letter from applicant transmitting calculations and position papers concerning trapeze support post members, including Revision 0 to IM-P-011.

September 23, 1987 Letter from applicant transmitting post-instruction hardware validstion program attribute matrix.

September 29, 1987 Letter from applicant transmitting Revision 1 to CPRT Results Reports II.E, V.A. and IX.

October 2, 1987 Letter from applicant transmitting Revision 4 to the Plant Security Training and Qualification Plan.

October 2, 1987 Letter from applicant transmitting Revision 0 to the Fire Protection Report.

' Comanche Peak SSER 21 22 Appendix A

October 2,.1987 Letter from applicant transmitting revised i environmental qualification program addressing j mild environments.

October 2, 1987 Letter from applicant concerning revised 1 deviations from Branch Technical Position '

APCSB 9.5-1, Appendix A, and 10 CFR 50, Appendix R.

October 2, 1987 Letter from applicant transmitting revised procedure manuals describing CFRT checklists.

October 2, 1987 Letter from applicant transmitting additional fire protection deviation from 10 CFR 50, Appendix R, Section III.G.2.

October 2, 1987 Letter to applicant concerning proposed date ,

of emergency exercise.  !

October 5, 1987 Letter from applicant transmitting CPRT Results Report VI.B.

October 9, 1987 Letter from applicant transmitting plant-specific requirements as designated in WCAP-10858.

October 13, 1987 Letter from applicant concerning changes in management.

October 13. 1987 Letter to applicant transmitting safety evaluation regarding request to eliminate arbitrary intermediate pipe breaks in high energy piping system.

October 23, 1987 Letter to applicant requesting additional information regarding extension of CPPR-127 (Unit 2).

October 26, 1987 Letter to applicant transmitting safety evaluation regarding TMI Action Item II.K.3.31.

October 26, 1987 Letter from applicant concerning adoption of Revision 15 to ASME Code Case N-71.

October 27, 1987 Letter from applicant transmitting corrections to FSAR changes regarding environmental qualifications program.

October 30, 1987 Letter from applicant concerning marked-up proposed Unit 1 Technical Specifications.

Comanche Peak SSER 21 23 Appendix A

October 30, 1987 Letter from applicant transmitting additional information regarding Revision 3 to CPRT Results Report II.C.

November 2, 1987 Letter from applicant transmitting Revision 0 to status report on resolution of large-bore piping and pipe support generic issues.

Novembe- 2, 1987 Letter from applicant transmitting Revision 0 to status ' report on resolution of small-bore piping and pipe support generic issue.

November 4, 1987: Letter to applicant concerning policy state-mentondeferredplants(GenericLetter87-15).

November 5, 1987 Letter from applicant transmitting information regarding plant thermal-hydraulic analysis of fire shutdown scenario.

November 6, 1987 Letter from applicant transmitting additional information regarding Generic Letter 83-28.

- November 6, 1987 Letter from applicant transmitting project status report regarding cable tray and cable tray hangers.

- November 6, 1987 Letter to applicant concerning list of generic issues.

November 9, 1987 Letter from applicant transmitting CPRT Results Reports I.C. II.A. and II.D. which were approved by a special review team.

4 November 9. 1987 Letter to applicant concerning extensive development and application criteria for cable tray hangers.

November 11, 1987 Letter from applicant transmitting Revision 0 to corrective action program project status report regarding conduit supports train C,

? inches in diameter.and less.

November 12, 1987 Letter to applicant transmitting NUREG-1262 regarding implementation of 10 CFR 55 in regard to operator licenses (Generic Letter 87-16).

November 18, 1987 Letter from applicant transmitting Revision 0 to status report regarding conduit supports trains A and B and train C, larger than 2 inches in diameter.

November 20, 1987 Letter from applicant transmitting descrip-tion of FSAR Amendment 65.

. Comanche Peak SSER 21 24 Appendix A

l November 20, 1987 Letter from applicant concerning inspection of coated weld samples.

l- November 25, 1987 Letter from applicant concerning plant program descriptions section of the corrective action program.

I December 3, 1987 Letter to applicant transmitting marked-up ,

first draft Unit 1 Technical Specifications. i December 3, 1987 Letter from applicant transmitting additional information regarding groundwater wells (Unit 2).

December 4, 1987 Letter to applicant concerning schedule for SALP evaluation.

December 7, 1987 Letter from applicant transmitting CPRT Results Reports V.B and VIII.

December 15, 1987 Letter from applicant transmitting table of interpretations regarding codes, standards, and regulatory guides.

December 18, 1987 Letter from applicant concerning clarification of interrelationships between CPRT and corrective action programs.

December 30, 1987 Letter from applicant transmitting additional infonnation regarding CPRT Results Report V.A.

December 31, 1987 Letter from applicant transmitting answers to 14 questiens from the ASLB regarding CPRT Results Report IX.

December 31, 1987 Letter from applicant transmitting Special Review Team Report VII.C.

December 31, 1987 Letter from applicant transmitting Revision 0 to CPRT collective evaluation report and Revision I to Volume IV-A through Volume IV-E.

January 6,1988 Letter to applicant transmitting first draft linit 1 Technical Specifications and descrip-tion of SER items.

January 8, 1988 Letter from applicant transmitting Revision 0 to corrective action program status report regarding equipment qualification.

January 11, 1966 Letter from applicant concerning compliance with NRCB 87-02.

Comanche Peak SSER 21 25 Appendix A

i

-o l 1

January 15, 1988 Letter from applicant transmitting Revision 0 to corrective action program status report regarding generic electrical issues.

L .. Ja nua ry 15, 1988 Letter from applicant transmitting description of FSAR Amendment 66..

January 15, 1988 Letter to applicant concerning use of-increased yield strength for A500 grade B tube shapes.

January 18, 1988 Letter from applicant transmitting CPRT results reports regarding Issue-Specific Action Plans I.D,1 and VII.B.3.

January 18, 1988 Letter from applicant concerning design of future modifications to avoid concentrating essential equipment in superpipe areas.

January 19. 1988 Letter from applicant concerning availability of Revision 1 to emergency dose assessment model program document.

January 20, 1988 Letter from applicant transmitting answers to 14 questions from the ASLB.-

January 20, 1988 Letter to applicant concerning integrated safe assessment program (Generic Letter 88-02 .

January 21, 1988 Letter to applicant transmitting request for a response to all open issues identified in the technical evaluation of concerns and allegations for various electrical and-mechanical areas.

January 22, 1988 Letter from applicant concerning revised response to NUREG-0737 TMI Action Item II.F.2.

i.

January 22, 1988 Letter to applicant transmitting evaluation accepting Revision 4 to CPRT Program Plan and ,

corrective action program.

January 25, 1988 Letter from applicant transmitting Revision 0 to l

project status report regarding mechanical system interaction.

January 29, 1986 Letter from applicant transmitting Endorse-ment 23 to NELIA Policy NF-274.

February 1,1988 Letter from applicant transmitting Revision 0 to project status report regarding instrumentation and controls generic issues.

Comanche Peak SSER 21 26 Appendix A

1 February 4, 1988 Letter from applicant concerning responses to HVAC open items M-1 through M-14.

1 February 5, 1988 Letter from applicant transmitting description of changes for FSAR Amendment 68.

February 5, 1988 Letter from applicant transmitting description of FSAR Amendment 67.

February 5, 1988 Letter from applicant transmitting detailed response to NRCB 87-02.

February 8, 1988 Letter from applicant transmitting Revision 0 to status report regarding civil / structural generic issues.

February 11, 1988 Letter from applicant transmitting notarized response to NRC Bulletin 87-02.

February 12, 1988 Letter from applicant concerning commitment to perform inspection of Crosby 6M6 pressurizer safety valves after each actuation caused by an operational transient.

February 16, 1988 Letter from applicant transmitting revised Emergency Plan procedures.

February 16, 1988 Letter from applicant concerning ATWS mitigation system actuating circuitry.

February 17, 1988 Letter to applicant concerning resolution of Generic Safety Issue 93, " Steam Binding of Auxiliary Feedwater Pumps" (Generic Letter 88-03).

February 18, 1988 Letter from applicant concerning conserva-tiveness of attenuation factors now being used for performing jet impingement analysis.

February 18, 1988 Letter from applicant transmitting Revision 0 to project status report regarding HVAC generic technical issues.

February 19, 1988 Letter from applicant transmitting Form 8-K. l February 23, 1988 Letter to applicant concerning response to Generic Letter 86-16 regarding Westinghouse l ECCS evaluation models. l February 24, 1988 Letter from applicant transmitting project  ;

status report regarding instrumentation and 1 controls.

Comanche Peak SSER 21 27 Appendix A l I'

february 29, 1988. Letter from applicant transmitting final submittal of chronologies of confirmed issues per ASLB 860808.

February 29, 1988 Letter 1from applicant transmitting CPRT Results Report VII.A.9 and collective significance report.

March 2, 1988 Letter to applicant transmitting second draft of Unit 1 Technical Specifications.

March 3, 1988 Letter to applicant concerning acceptability of using ASME Code Case N-253-2 provided rules of code case govern design and materials selection.

March 3, 1988 Letter to applicant transmitting request for additional information regarding response to IEB 85-03.

March 10,1988 Letter to applicant transmitting request for edditional information regarding radiation protection organization.

March 11,1988 Letter from applicant concerning preliminary results of jet impingement analysis.

March 15, 1988 Letter from applicant concerning steam generator tube rupture.

March 16, 1988 Letter from applicant concerning design deficiencies identified in corrective action program.

March 17, 1988 Letter to applicant concerning boric acid corrosicn of carbon steel reactor pressure bounoary components (Generic Letter 88-05).

March 18 1988 Letter from applicant transmitting additional chronologies of confirmed issues per ASLB 860808.

March 18, 1988 Letter from applicant concerning collective evaluatior, report and completed external source issues matrix.

March 18, 1988 Letter from applicant transmitting additional information regarding compliance with Revision 1 of NiSAC generic design package.

March 18, 1988 Letter from applicant transmitting detailed I description of FSAR Amendment 69.

Comanche Peak SSER 21 28 Appendix A

March 18, 1988 Letter from applicant concerning schedule for full-participation emergency exercise.

March 22, 1988 Letter to applicant concerning removal of organization charts from Technical Specifi-cations (GenericLetter88-06).

March 23, 1988 Letter from applicant transmitting response to NRCB 88-02.

March 24,.1988 Letter from applicant transmitting errata to small- and large-bore piping and pipe support status reports.

March 24, 1988 Letter to applicant transmitting Supple-ment 14 to the SER.

March 25, 1988 Letter from applicant concerning safety issues management system.

March 25, 1988 Letter from applicant concerning schedule for submitting Unit 1 proof and review Technical Specifications.

March 27, 1988 Letter from applicant transmitting comments on draf t Regulatory Guide Task HF 601-4 and value/ impact statement.

March 28, 1988 Letter from applicant concerning removal of tube vibration instrumentation installed in steam generators and approval requested for elimination of vibration monitoring program.

March 29, 1988 Letter from applicant transmitting responses to certain consnents and questions from NRC arising from December 9, 1987, public meeting.

March 31, 1988 Letter from applicant concerning availability of documents in Supplement 14 to the SER.

April 6, 1986 Letter from applicant transmitting errata to equipment qualification and electrical project status reports.

April 6, 1988 Letter from applicant transmitting INPO report dealing with operational aspects of facility.

April 7, 1988 Letter to applicant transmitting modified enforcement policy regarding 10 CFR 50.49 (GenericLetter88-07).

April 8,1988 Letter from applicant transmitting response to NRCB 88-01.

Comanche Peak SSER 21 29 Appendix A

s iApril8,-1988 Letter from applic6nt transmitting additional information regarding IEB 85-03.

. April 11,'1988 Letter from applicant transmitting additional information regarding response to Generic Letter 83-28.

April 14,1988 Letter from applicant transmitting additional information regarding applicability of corrective action program.

April 15, 1988 Letter from-applicant transmitting whipjet >

' program report for review and approval to allow elimination of pipe breaks and associated dynamic effects for listed systems.

1 April 15,1988 Letter from applicant transmitting final )

revisions and instruction sheets for manuals describing the use of plant response team checklists.

.]

April 15, 1988.

Letter to applicant transmitting request for responses to questions regarding cable pulling operations and cable splices.

April 21, 1988 Letter from applicant transmitting response to questions arising from December 9, 1987, public meeting.

April 22 -1988 Letter from applicant transmitting response to IEB 84-03.

April 22, 1988 Letter from applicant transmitting detailed description of FSAR Amendment 70.

April 29, 1988 Letter from applicant concerning fire protection SER and additional comments regarding Supplement 12 to the SER.

' April 29,1988 Letter from applicant transmitting changes to the FSAR regarding fire protection.

April 29, 1988 Letter from applicant transmitting errata to project status report regarding mechanical supplements A (system interaction), and B (fire protection).

May 2, 1988 Letter from applicant transmitting changes to FSAR Sections 17.1 and 17.2.

May 3, 1988 Letter to applicant concerning mail sent or delivered to NRR (Generic Letter 88-08).

Comanche Peak SSER 21 30 Appendix A

I i

l May-4, 1988 Letter from applicant transmitting changes to j FSAR Section 3.68 incorporating ANSI /ANS i Standard 58.2.

May.10, 1988 Letter from applicant transmitting Endorse-ment 22 to NELIA policy NF-274.

hay.'13, 1988. Letter from applicant concerning Radiation j Protection Manager position. ,

1 May 16, 1988 Letter from applicant transmitting updated S&W corrective action program design validation package indexes for various procurement status reports.

May 17, 1988 Letter from applicant concerning response to l Generic Letter 88-03.  !

May 17, 1988 Letter'to applicant concerning active review of FFAR amendments.

May 20, 1988 Letter from applicant transmitting errata to project status report for instrumentation and controls and HVAC.-

May 20, 1988 Letter from applicant concerning manual actions to be taken in the event of a fire in the plent.

May 20, 1988 Letter from applicant concerning response to Generic Letter 83-28. Item 2.2.1.

May 20, 1988 Letter from applicant transmitting additional infomation regarding cable pulling operations and cable splices.

May 26, 1988 Letter from applicant transmitt ng errata to i

1 mechanical and civil / structural project status reports.

May 27, 1988 Letter from applicant transmitting detailed description of FSAR Amendment 71.

June 1, 1988 Letter from applicant concerning FSAR changes i through Supplement 12 in response to SER supplements.

1 June 1, 1988 Letter to applicant concerning staff position that inspections and corrective actions required by IEB 88-01 are to be completed by fuel load.

' June 1, 1986 Letter to applicant transmitting concurrence with removal of tube vibration instrumentation installed in Unit 1 steam generator.

Comanche Peak SSER 21 31 Appendix A 1

! l

1

' June 6, 1988. Letter from applicant transmitting Endorse-ment 24 to NELIA Policy NF-274.

. June 6, 1988 Letter from applicant transmitting additional information regarding response to IE8 78-04.

June 6, 1988 Letter from applicant transmitting infoma-tion supporting extension of Unit 2 construc-tion schedule.

June 9, 1988 Letter from applicant transmitting summary of programmatic enhancements relevant to NRC review and validation of design and hardware installation.  !

June 15, 1988 Letter from applicant concerning change in engineering contractor responsibilities for cable tray hangers.

June 17, 1988 Letter.from applicant transmitting ANC0 Engineers, Inc. report regarding ccnduit tests.

June 17, 1988 Letter from applicant transmitting assessment of RCS Cold hydrostatic test report.

June 20, 1988 Letter from applicant transmitting Westinghouse  !

Owners Group report identifying midloop j concerns per Generic Letter 87-12. i f

June 22, 1988 Letter from applicant transmitting TU Electric,  ;

Braros Electric Power Cooperative, Inc., and '

Texas Municipal Power Agency annual reports.

June 23, 1988 Letter from applicant concerning changes to second draft Unit 1 Technical Specifications. l June 24, 1988 Letter from applicant concerning response to Generic Letter 88-05.

I June 24, 1986 Letter from applicant concerning clarification of "significant safety-related" as used in CPRT collective significance reports.

l July 1, 1988 Letter from applicant transmitting descrip-tion of FSAR Amendment 72.

July 1, 1988 Letter to applicant concerning purchase of GSA-approved security containers (Generic Letter 88-10).

July 1, 1988 Letter from applicant transmitting joint stipulation and joint motion for dismissal and exhibits for filing in proceeding.

Comanche Peak SSER 21 32 Appendix A

--_.-_______---___j

4 ,

4' July 8, 1988 Letter from applicant concerning response to i t NRCB 88-04.

V July 8, 1988 Letter from applicant concerning application of leak-before-break methodology to branch lines.

July 11, 1988 Letter from applicant concerning response to J NRCB 88-03. j j

July 12, 1988 Letter to applicant concerning NRC position

^

on radiation embrittlement of reactor vessel materials and impact on plant operations (Generic Letter 88-11).

l July 15, 1988 Letter from applicant transmitting changes l to FSAR regarding normal maximum and minimum offsite power grid system voltages.

July 22, 1988 Letter from applicant transmitting responses to Supplements I and 2 of NRCB 87-02.

July 29, 1988 Letter to applicant concerning application of Technical Specifications improvements to Technical Specifications under development.

August 2, 1988 Letter from applicant concerning engineering contractor's and utility's engineering respon-sibilities and organizations.

August 2, 1988 Letter to applicant concerning removal of fire protection system technical specifications from Technical Specifications (Generic Letter 88-12).

August 2, 1988 Letter to applicant concerning qualifications of the Radiation Protection Manager.

August 5, 1988 Letter from applicant transmitting FSAR Amendment 73.

August 5, 1988 Letter from applicant concerning update to proposed Unit 1 Technical Specifications sections 6.5.2.2 and 6.5.2.6.

August.5, 1988 Letter from applicant concerning unreviewed generic changes to proposed Unit 1 Technical Specifications.

August 8, 1988 Letter from applicant transmitting endorse-ments 1, 2, and 3 to NELIA Certificate NW-167 and MAELU certificate MW-190.

Comanche Peak SSER 21 33 Appendix A

\

August 8, 1988 Letter from applicant transmitting proposed final draft of the Offsite Dose Calculation Manual.

' August 8, 1988 Letter to applicant transmitting Supple-ment 16 to the SER.

August 8, 1988 Letter to applicant concerning instrument air supply system problems affecting safety-related equipment (Generic Letter 88-14).

August 8, 1988 Letter to applicant concerning operator licensing examinations (Generic Letter 88-13).

August 10, 1988 Letter to applicant transmitting Amendments 9 ar.d 8 to CPPR *.73 and CPPR-127, respectively.

August 12, 1988 Letter to applicant transmitting Supple-ment 15 to the SER.

August 17, 1988 Letter to applicant transmitting summary of August 8, 1988, meeting regarding performance i of plant personnel on recent and past NRC-administered operator licensing examinations.

August 19, 1988 Letter from applicant transmitting plant l meteorological tower parameters for January through April 1989.  ;

August 22, 1988 Letter from applicant transmitting Revision 10 to the Emergency Plan.

August 22, 1988 Letter from applicant transmitting Revision 5 to i the Security Training and Qualification Plan.

August 22, 1988 Letter from applicant transmitting Revision 6 to the Plant Physical Security Plan.

August 23, 1988 Letter from applicant transmitting Revision 1 to ER-ME-01 report concerning cold system hydro-static test. c August 26, 1988 Letter from applicant transmitting Endorse-ment 4 to NELIA Certificate NW-167 and Endorse-ment 5 to MAELU Certificate MW-190. <

l August 29, 1988 Letter from applicant transmitting facility isolator test procedure for core cooling monitors in response to Generic Letter 82-33.

September 6, 1988 Letter from applicant concerning corporate management change.

I Comanche Peak SSER 21 34 Appendix A i

1 1 p

September 7, 1988 Letter from applicant concerning qualifications of the Radiation Protection Manager candidate.

September 7, 1988 Letter from applicant concerning pressurizer

' surge line stratification leak before break.

September 12, 1988 Letter to applicant'concerning electric power systems and inadequate control over the design process (Generic Letter 88-15).

September 14, 1988~ Letter from applicant concerning completion status of primary portion of S&W engineering function-evaluation.

September 15, 1968 Letter from applicant transmitting Supple-ment 3 to human factors detailed control room design review.

September 16, 1988 Letter from applicant transmitting current version of response team findings and related project commitments data base used to coordinate various documents.

September 23, 1988 Letter from applicant transmitting Revision 0 to evaluation regarding service water coating removal program.

September-28, 1988 Letter from applicant concerning additional examinations during system functional testing heatup.

September 28, 1988 Letter from applicant transmitting plant primary meteorological tower parameters for.

April through June 1988.

October 4, 1988 Letter to applicant concerning removal of arameter limits from Technical cycle-specificp(GenericLetter88-16).

Specifications October 7, 1988 Letter from applicant transmitting final draft 1 process control program.  !

October 7, 1988 Letter from applicant transmitting comments regarding draft Environmental Protection Plan.

October 7, 1986 Letter to applicant acknowledging receipt of proposed changes to Unit 1 Technical Specifi-cation ~6.5.2.

October 12, 1988' Letter from applicant transmitting proprietary WCAP-11987 and nonproprietary WCAP-11988 regarding pressurizer surge line stratifica-tion.

Comanche Peak' SSER 21 35 Appendix A l

q

{ .

October 13,~1988' Letter to applicant concerning inspection initiative to evaluate maintenance perfom-  ;

ance.

October.17,,1988 Letter to applicant concerning loss of decay.

heat removal;(Generic Letter 88-17).

October- 17, 1988 Letter to applicant transmitting notice of -

environmental assessment extending construc-tioncompletiondate-(Unit 2). l

. October 19, 1988 Letter from applicant transmitting detailed description of.FSAR Amendment 74. j 1

October 19,.1988 Letter to-applicant transmitting notice of-environmental assessment and finding of no significant impact.

f October 20, 1988 Letter to applicant transmitting request. for -

additional information regarding natural circulation cooldown testing..

October 20, 1988 Letter.to applicant concerning plant record storage on optical disks.(Generic Letter 88-18).

October 21, 1988 Letter from applicant ccncerning response to NRCB 88-08.

October 21,1 1988 Letter from applicant transmitting changes to FSAR regarding impact of main steam

'line break.

l-October 21, 1988- Letter' from applicant. transmitting coments regarding'significant changes in license

~

L l activity that warrant antitrust review.

.0ctober 21, 1988 Letter to applicant tonsmitting draft SALP reports for the period September 1987 through August 1988.

L . October 24,'1988 Letter from applicant transmitting changes I to FSAR regarding post-LOCA-environment equipment qualification.

~

l October' 28, 1988 Letter to applicant concerning use of deadly force by licensees -(Generic Letter 88-19).

October 31, 1988 Letter from applicant concerning response to NRCB 88-08.

Comanche Peak SSER 21 36 Appendix A

=

October 31, 1988 Letter from applicant concerning changes to ,

information contained in PSAR that will be ')

specifically identified in correspondence i to NRC.

-October 31,-1988 Letter from applicant cancerning imple-- i mentation schedule for ATWS mitigation system

.f actuation circuitry installation. 1 November 1, 1988 Letter from applicant concerning completion of environmental qualification verification 1 program for Unit 2. i l

November 2, 1988' Letter from applicant concerning response to l Generic Letter 85-02.

November'7, 1988 Letter to applicant transmitting additional lj infonnation regarding use of emergency <

notification system and health physics  !

network in responding to emergency events.

November 14, 1988 Letter from applicant transmitting information and renewal application for permit to discharge waste water.

November 16, 1988 Letter from applicant transmitting proposed changes to Unit 1 second draft Technical.Speci-fications as part of ongoing improvement effort..

November 16, 1988 Letter to applicant transmitting request for.

additional information in order to complete review of- FSAR Amendments 55 through 73.

November 17, 1988 Letter from applicant transmitting Revision 0 to CPSES Technical Specification improvement  ;

program. l November 17, 1988 Letter from applicant concerning results of study on Technical Specifications relocation.

November 18,1988 Letter to applicant transmitting order extending latest construction completion

! date for CPPR-127 (Unit 2). 1 November 18, 1988 Letter to applicant concerning status of p S&W engineering functional evaluation and I acceptability of proposed integration of remaining efforts into plant activities.

November 18, 1988 Letter to applicant transmitting order extending latest construction completion l.

date for CPPR-126 (Unit 1).

i Comanche Peak SSER 21 37 Appendix A J

h November 22, 1966 Letter to applicant transmitting marked-up

-Unit 1 proof and review version of Technical Specifications.

November 23, 1988 Letter from applicant transmitting comments on draft Environmental Protection Plan.

-November 23, 1988 Letter to applicant concerning individual plant examinations to determine severe acci-dent vulnerabilities (Generic Letter 88-20).

November 28, 1988 Letter from applicant' transmitting Revision 6 to the Physical Security Plan and Revision 5 to the Training and Qualification Plan.

November 28, 1988- Letter from applicant concerning response-to Generic Letter 88-11, " Radiation of Reactor Vessel Materials."

November 28, 1906 Letter from applicant concurring with assessment of September 1987 SALP report.

November 30, 1988 Letter from applicant concerning completion f of constituents in response to NRCB 88-04. "

December 5, 1988 Letter to applicant transmitting request for additional information concerning steam generator tube rupture analysis to resolve proposed License Condition 18 of SSER 12.

December 6, 1988' Letter to applicant transmitting Supple-ment 18 to the SER.

December 6,'1988 Letter to applicant transmitting Supple-nent 20 to the SER.

December 8, 1988 Letter to applicant transmitting Amendments 10 and 9 to CPPR-126 (Unit 1), and CPPR 1.27 (Unit 2),respectively.

December 9, 1988 Letter from applicant transmitting additional information regarding FSAR review.

December 9, 1988 Letter to applicant transmitting Supplement 19 l to the SER.

December 9, 1988 Letter to applicant transmitting final SALP ,

reports for September 1987. l December 15, 1988 Letter to applicant transmitting marked-up Unit 1 proof and review version of Technical Specifications. 1 i

i Comanche Peak SSER 21 38 Appendix A  ;

i

_______-____-_______ - -_ _ _ D

LDecember 16,.1988 Letter from applicant transmitting additional information requested regarding natural circulation cooldown testing for facility and a marked-up draft revision to FSAR Appendix SA.

. December 16, 1988 Letter from applicant concerning technical requirements to improve Unit 1~ service water reliability during Unit 2 construction and I testing.

December 16 -1988 Letter from applicant transmitting response

-to SER items concerning Technical Specifica-tions.

December 16, 1988 Letter from applicant transmitting additional information regarding FSAR review.

December 20,:1988 Letter from applicant concerning stecm generator tube rupture analysis.

December 20, 1988 Letter from applicant transmitting. tabulation of plant primary meteorological tower parameters for July through September 1988.

December 21, 1988 Letter from applicant concerning delay of its response to Generic Letter 88-17, " Loss of Decay Heat Removal Capability."

December 21,_1988 . Letter from applicant transmitting additional information requested regarding FSAR Sections 6.5, 9.1'.3, 9.2.2, 9.4.1, 9.5.4, 10.3, and 10.4.9.

December 21,'1988 Letter from applicant concerning Technical Specification Improvement Program change of name to Technical Requirements Manual and describing updated administrative control y process.

December 28, 1988 Letter from applicant _ concerning maximum radioactivity in floor drain tank for accident analysis and waste gas processing.

December 30, 1988 Letter from applicant transmitting proposed FSAR changes regarding radiation embrittlement of reactor vessel materials per Generic Letter 88-11. ,

December 30, 1988 Letter from applicant transmitting nonpro-prietary WCAP-11966 and proprietary Revision 1 to WCAP-11312 regarding reactor trip breaker maintenance and surveillance.

LComanche Peak SSER 21 39 Appendix A

3p j

-The following correspondence was found to have'been omitted from earlier 1 supplements and should be inserted in its proper order.

SSER'12-September 18, 1984 Letter to applicant concerning Comanche Peak review and items identified by the NRC Technical Review Team.

October 8, 1984 Letter from applicant concerning program plan and. issue-specific action plans in response to the request for additional

~

information by the NRC Technical Review Team, h

October 11, 1984 Letter to applicant concerning staff I evaluation findings pertaining to contain-ment isolation items.

November 2, 1984 Letter from applicant concerning GAP and the NRC Technical Review Team.

November 12 -1984- Letter from applicant concerning CPRT Program Plan, October 8, 1984.

' November 21,.1964 Letter from applicant transmitting Revision 1 to the CPRT Program Plan.

December 14, 1984 Letter to applicant concerning resolution of Confirmatory Issue 18 in SSER 6.

l

' January 8, 1985 Letter to applicant concerning formation of two' panels of NRC senior staff managenent to address Comanche Peak hearing issues on intimidation.

January 8, 1985 Letter to applicant concerning Comanche Peak review. [

' January 11, 1985 Letter to applicant deferring final deter-mination of order imposing civil penalty, a

January 15, 1985 Letter from applicant transmitting list and corrective actions regarding concern that

! craft welding supervisors lacked sufficient understanding of procedure requirements. ,

. January 24, 1985 Letter to applicant transmitting comments on L issue-specific action plan for consideration in next-revision.

L L'

Comanche Peak SSER 21 40 Appendix A I # . - _ _ _ _ _ _ _

wf-

,_ g ,

l January 24, 1985.

. Letter from applicant transmitting'the ;lq annual. financial. reports. '

January 25, 1985 - Letter'to applicant acknowledging r.ceipt e of-letter updatingLinitial schedule for' completion of NRC emergency appraisal Appendix A items.

Febr'uary 8, 1985 Letter to applicant concerning cost for operating license review.

February. 12, 1985- Letter to applic' ant.concerning request for information on employee survey.

~

March 21,1985 Letter to applicant concerning request for access to the Comanche Peak plant site.

Letter to applicant transmitting FEMA's, final-April 4, 1985 radiological emergency preparedness exercise report.

' April 15. 1985 Letter from applicant-concerning piping and supports.

April 23, 1985 Letter from applicant concerning CPRT Program Plan and issue-specific action plans.

May_18,.1985 Letter from applicant concerning schedule for submitting additioral information in response to Generic Letter.83-28.

. hay 23, 1985 Letter to applicant concerning 10 CFR 20.408 termination reports (Generic Letter.

85-08).

June 28, 1985 Letter from applicant concerning CPRT. Program Plan and self-initiated actions.

July 2, 1985 Letter from applicant transmitting the annual financial reports for 1984.

August 1, 1985 Letter to applicant concerning CPRT Program Plan. I August 6, 1985 Letter to applicant concerning consnercial storage at power reactor sites of low-level ,

radwaste not j Letter 85-14) generated by utility (Generic

. j August 9, 1985 Letter to applicant transmitting Revision 2 to the NRC staff's evaluation of the CPRT Program q Plan. ] i j

Comanche Peak SSER 21 41 Appendix A I l

' August 9. 1985 Letter to the applicant concerning staff's evaluation of the CPRT Program Plan.

August 16, 198'5 Letter from applicant concerning CPRT quality aspects.

August 30, 1985 Letter from applicant concerning increased i allowable limits for pipe support design organizations.

J August 30, 1985 Letter from applicant transmitting objectives l and guidelines for emergency preparedness 1 exercise. ]

J SSER 6 August 29.-1983 Letter from aoplicant transmitting _ Amendment 43 to the FSAR.

September 1, 1983 Letter from applicant concerning supplemental response to NRC Generic Letter 83-10c, "TMI Action Plan II.K.3.5 - Automatic Trip of Reactor Coolant Pumps."

January 15, 1984 Letter from applicant concerning equipment environmental qualification justification for interim operation.

January 10, 1984 Letter from applicant transmitting a table describing Amendment 50 to the FSAR.

August- 16, 1984 Letter from applicant transmitting the seventh biweekly update on the. status of Comanche Peak fuel loading.

August 20, 1984 Letter from applicant concerning changes to proposed Unit 1 Technical Specifications.

August 23, 1984 Letter from applicant concerning request for partial exemption from 10 CFR 50, '

Appendix J.

August 27, 1984- Letter from applicant transmitting Amend-ment 52 to the FSAR.

' August 27, 1984 Letter from applicant requesting additional information on the preoperational integrated leak rate test performed for Comanche Peak.

August 29, 1984 Letter from applicant concerning preservice inspection plan update.

-Comanche Peak SSER 21 42 Appendix A

l a

L IAugust 31 -1984 Letter to applicant' requesting additional information concerning the design of cable tray supports.

~ September 6, 1984 Letter from applicant transmitting the f eighth biweekly update on the status of important schedule-related issues.

.'SSER 4' June 15, 1983 Letter from applicant transmitting 1982 snnual financial reports.

July 15, 1983 Letter from applicant transmitting Revision 4 to the Physical Security Plan. 1

, j 1 " July 19, 1983 Letter from applicant transmitting an 1 affidavit for Amendment 41.to the FSAR. ]

'SSER 2 January 5, 1982- Letter from applicant concerning' exhaust damper for ESF filter train in the safe-

, guards building.

January 5, 1982 Letter from applicant transmitting an affidavit for Amendment 29 to the FSAR.

SSER 1 June 3, 1981 Letter from applicant concerning the conduct of operations and personnel staffing at Comanche' Peak.

June 8, 1981 Letter from applicant concerning review of emergency procedures and training procedures (in response to' Generic Letter 81-04).

July 6, 1981 Letter to applicant concerning submittal-of INP0 evaluation reports to NRC in-response to Generic Letter 81-23A.

July 14, 1981 Letter from applicant concerning proposed secondary water chemistry monitoring program (Revision 1).

October 5, 1981 Letter from applicant transmitting an affidavit for Amendment 10 to the FSAR.

Comanche Peak SSER 21 43 Appendix A

)

L APPENDIX B BIBLIOGRAPHY

1. American Society for Testing and Materials, Standard E-119 " Fire Test of Building Construction and Materials."

i

2. American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Code Case N-71, " Additional Materials for Subsection

.NF, Class 1, 2, 3, and MC Component Supports Fabricated by Welding,"

Section III, Division 1.

3. . Code Case N-2S3-2, " Construction of Class 2 and 3 Components for ITevated Temperature. service,"Section III, Division 1.

4. , Code' Case H-378, " Examination of Piping Support Material,"

Ye'ction III, Division 1.

S. . Code Case N-397, " Rules to Spectral Broadening Proceduret of N-1226.3 T57 Class 1 '2, and 3 Piping,"Section III, DivSion 1.

6. . Code' Case N-411, " Alternative Damping Values for Seismic Response Spectra Analysis of Class 1, 2, and 3 Piping,"Section III, Division 1."

7. Electric Power Research Institute IEPRI), " Safety and Relief Valve Test Report," September 1982.

8. , " Pressurized Water Reactor Safety and Relief Valve Program Test fonHition d Justification Report," June 1982.

9. Gibbs & Hill, Inc., GTN-65740, "TUGC0 Test Deficiency Report 1017,"

May 19, 1983.

10. International Conference of Building Code Officials, "Unifonn Building Code."

[

11. National Fire Protection Association (NFPA) 12A/128, "Halongenated Fire Extinguishing Agent Systems - Halon 1301 and Halon 1211."

12. , NFPA 13. " Standard for the Installation of Sprinkler Systems."

13. , NFPA 14 " Standard for Stanapipe anc hose Systems."

,- 14. , NFPA 72D, " Standard for the Installation, Maintenance, and Use of Proprietary Protective Signaling Systems."

L l

Comanche Peak SSER 21 1 Appendix B l

15. , NFPA 72E, " Automatic Fire Detectors."

16. Rahe, E.P. (Westinghouse) letter to R.C. DeYoung (NRC), "Part 21 and Deficiency Reports re Misoperation of 05-416 Reactor Trip Ur.dervoltage Devices," April 21, 1983.

17. Rahe, E.P. (Westinghouse) letter to H. Thompson (NRC), " Addendum to BART-A1: Computer Code for Best Estimate Analysis of Reflood Transients (Special Report: Thimble Modeling in Westinghouse ECCS Evaluation Model),"

dated June 2, 1986.

18. Underwriters Laboratories, Inc., Fire Resistance Directory, Northbrook, IL, 1 January 1983.

19. Welding Research Council, Bulletin 300 " Technical Position on Industry Practice," December 1984.

20. Westinghouse Electric Corporation, Topical Report WCAP-10054, " Westinghouse Small Break ECCS Evaluation hodel Using NOTRUMP Code," December 1984.

21. . Topical Report WCAP-10079, "NOTRUMP: A Modal Transfer Small Dreak l and General Network Code," November 1982.

22. , Topical Report, WCAP-10527, " Technical Bases for Eliminatir,g Large Loop Pipe Rupture as the Structural Design Basis for Comanche Peak, Units 1 and 2," April 1984 (Westinghouse Proprietary Class 2).

23. , Topical Report WCAP-10858, "AMSAC Generic Design Package," July T95.

24. , Topical Report WCAP-11145, " Westinghouse Small Break LOCA ECCS TvaTuation Model Generic Study with the NOTRUMP Code," May 1986.

25.

Xnal..

a ysis Topical of Fire Report WCAP-11331 Safe Shutdown " Comanche Scenario," Peak October SES Thermal / Hydraulic 1986.

f Comanche Peak SSER 21 2 Appendix B

APPENDIX D LIST OF PRINCIPAL CONTRIBUTORS Contributor Organization F. Ashe Office of Nuclear Reactor Regulation Comanche Peak Project Division B. Belke Office of Nuclear Material Safety and Safeguards Division of High Level Waste Management C. Brown Idaho National Engineering Laboratory (EG&G)

Office of Nuclear Reactor Regulation b H. Garg TVA Projects Division R. Hogan Office of Nuclear Reactor Regulation Emergency Preparedness Branch D. Kelly Nuclear Regulation Comission Region IV J. Kramer Office of Nuclear Reactor Regulation Human factors Assessment Branch D. Lasher Office of Nuclear Reactor Regulation Instrumentation & Control Systems Branch J. Lyons Office of Nuclear Reactor Regulation Comanche Peak Project Division M. Malloy Office of Nuclear Reactor Regulation Comanche Peak Project Division P. McKee Office of Nuclear Reactor Regulation Comanche Peak Project Division J. Moore Nuclear Regulatory Comission Office of the General Counsel A. Singh Nuclear Hegulatory Comission Region IV W. Smith Nuclear Regulatory Comission Region IV Comanche Peak SSER 21 1 Appendix D !

l I

j

Contributor Organization L D. Terao Office of Nuclear Reactor Regulation l TVA Projects Division A. Toalston Office of Nuclear Reactor Regulation Electrical Systems Branch l

i J. Wilson Office of Nuclear Reactor Regulation Comanche Peak Project Division J. Wing Office of Nuclear Reactor Regulat'on Chemical Engineering Branch.

l M. Young Nuclear Regulatory Commission Office of the General Counsel i

l l

Comanche Peak SSER 21 2 Appendix D

APPENDIX E j ERRATA TO COMANCHE PEAK SAFETY EVALUATION REPORT AND SUPPLEMENTS The following errata to the Comanche Peak SER.were not included in Appendix E

!.. of any previous supplements.

Page 9-24, line 36- Change " Category I" to _" Category II".

Page 15-9, lines 20 and 21 Change " described in Section 15.5 of this report." to " described in Section 15.4.2 of this report."

The following errata are applicable to SER Supplements 1 through 12:

Supplement 1 i Page 4-3, line 1 Change " October 2,1981" to " October 5,1981".

Supplement 6 Page 1-5, lines 19, through~21 Change " Deferred; applies to two unit operation" to " Resolved in SSER 1".

Page 1-10, line 12 Change "10 CFR 50.56 to "10 CFR 50.46".

Page_1-15, line 21 Add "3.11.4.2 and" before "3.11.5" and change "Section" to " Sections".

Page 3-5, line 36 Change "1983." to "1982."

Page 3-9, line 33 Change "1983." to "1982."

Page 5-1, line 18 Change "therfy" to "thereby".

Page 14-1, line 20 Change " Coding" to " Cooling".

-Page.16-2, line 15 Delete.

Page 16-2, line 16 Delete.

Page 16-2, line 17 Delete.

Page 16-2, line 18 Delete.

Page 16-2, line 19 Change "(32)" to "(28)".

I Comanche Peak SSER 21 1 Appendix E

Supplement 12 Page V, lines 28 and 29 Change title of Section 3.11 to " Environmental Qualification of Electrical Equipment Important to Safety and Safety-Related Mechanical Equipment."

Page 3-19, line 16 Change title of Section 3.11 to " Environmental Qualification of Electrical Equipment Important to Safety and Safety-Related Mechanical Equipment."

Page 1-3, lines 32 and 33 Change " Deferred in SER; resolved in SSER 12" to " Deferred in SER; preservice inspection and inservice inspection.

Changed to Confirmatory Issue (16) in SSER 4; resolved inservice testing for Unit 1 in SSER 12".

Page 1-4, line 7 Add ", License Condition (14) in.SSER 6 changed to License Conditions (14) and (19) in SSER 12".

Page 1-5, lines 23 through 26 Change " Applies to two-unit operation; awaiting information from applicant" to

" Resolved in SSER 1".

Page 1-10, line 8 Change "SSER 1" to "SSER 6."

-Page 1-12, lines 17 through 21 Change "Added in SSER 4; resolved preservice inspection in SSER 12; awaiting inservice inspection submittal" to "Added in SSER 4; resolved preservice inspection for Unit 1 in SSER 12; awaiting inservice inspection submittal".

Page 1-12, line 18 Change "6.6.2" to 6.6.1".

Page 1-15, line 31 Change "SSER 12." to "SSER 6."

Page 1-16, line 19 Add.[ License Condition] "(19) Before ascending above 5% of thermal-rated power, the applicant must complete the safety-related mechanical equipment qualification program for all safety-related mechanical equipment located in a harsh environment and establish qualification or submit a justification for interim operation for each equipment item for the first cycle of operation (SERSection3.11).* See Sections 3.11, 4.2, 3.11.5, and 1.9, License Condition (14),ofSSER6."  ;

Comanche Peak SSER 21 2 Appendix E

_-_______-__________________________a

s.

Page 3-19. line 26 Change " Confirmatory

" Confirmatory Issue (18)". Issue (14)" to

-n Page 5-1, line 37 Change " Outstanding Issue (7)" to

" Confirmatory Issue (16)".

Page 5-1, line 43 Change " Outstanding Issue (7)" to

( " Confirmatory Issue (16)".

Page 6-2, line 37 Change " Outstanding

" Confirmatory (Issue (7)" to Issue' 16)".

L i

i i-l l

l Comanche Peak SSER 21 3 Appendix E i-

)

.o APPENDIX L' THE~ EFFECTS OF PAINT AND INSULATION DESRIS ON THE PERFORMANCE OF POST-ACCIDENT FLUID SYSTEMS AT

' COMANCHE PEAK STEAM ELECTRIC STATION,. UNITS 1 AND 2 On-the basis of its evaluation in Appendix L of SER Supplement 9, the' staff requested that the applicant propose a pre- and post-operational coatings

' testing and surveillance program for CPSES Units 1 and 2. The staff also

provided specific guidelines for developing such a program.' By letters dated June 7 and November 18, 1985 and December 16, 1986, the applicant provided the requested information for staff review.

The applicant has developed a surveillance program for testing, inspection, and documentation of the protective coating systems inside the containment buildings of Units 1 and 2. The pro inspection personnel; (2) gram includes (1) qualification and training of: inspec tional methods for each inspection and test, inspection equipment, the fre-quency of testing and inspection, the acceptance criteria for each inspection and test, and recordkeeping to document each inspection and test; (3) verifica-tion of storage and handling of protective coatings; (4) calibration of mea-suring and test equipment; (5) reporting, disposition, and tracking of coating degradation and deficiencies; (6) completion, issuance, and control of docu-mentation; and (7) maintenance and control of the Coatings Exempt Log for Unit 1.

The applicant provided the methods and criteria for operational surveillance of the coatings inside the containment buildings of Units'1 and 2. Before plant operation and at each respective refueling outage, a surveillance of the protective coatings will .be conducted to identify and report any current or incipient coating degradation or failure. Coating repairs will be performed in accordance with the as-low-as-reasonably-echievable (ALARA) guidelines for radiation exposure.

The surveillance program addresses the selection of painted areas for inspection that have high radiation exposure, that may not have full quality assurance or quality control verification (as inoicated in the Coatings Exempt Log), or that .

are adjacent to the containment sump areas.

The applicant's surveillance program does not include in situ temperature and pressure testing for. coating adhesion, as recommended in Supplement 9 of the

- SER. .The applicant justified this omission on the grounds that (1) adequate assurance of the acceptability of the coatings' initial conditions will be provided through the comprehensive backfit test program and final walkdown inspections; (2) inservice conditions will be verified through the coating surveillance and testing program during each refueling outage; and (3) the l

Comanche Peak SSER 21 1 Appendix L

recomended testing is destructive to the coatings and, in view of other actions prescribed in the surveillance program, is considered (by the applicant) an unwarranted hazard to testing personnel.

The staff has determined that, with the applicant's commitment to conduct periodic tests, inspections, and curveillarice of the protective coating systems, the previously recommended in situ temperature and pressure testing would not provide significant additional information on coating conditions.

Accordingly, in situ testing for coating adhesion need not be performed. The staff finds that the proposed surveillance program meets all of the other guidelines for a preoperational and periodic coatings testing and surveillance program that appear in Appendix L of Supplement 9 and that the program is acceptable. This outstanding issue is, therefore, closed.

Comanche Peak SSER 21 2 Appendix L

t l

APPENDIX M NRC STAFF EVALUATION AND RESOLUTION OF TECHNICAL CONCERNS AND ALLEGATIONS REGARDING PROTECTIVE C0ATINGS INSIDE THE REACTOR CONTAINMENT BUILDING AT COMANCHE PEAK STEAM ELECTRIC STATION, UNIT 1 Appendix M of SER Supplement.9 contains the results of the NRC Technical Review Team's (TRT's) assessment of allegations and concerns in the protective coatings area. The TRT's assessment resulted in recommendations for corrective actions to be taken by the applicant. These corrective actions, which were related to the backfit. test program, traceability, coatings procedures. and the Coatings Exempt Log, were modified on the basis of the staff's conclusions in Appendix L of Supplement 9. .The staff.recuested that the applicant submit information regarding these corrective actions and documenting the status of protective coating systems inside the containment building.

The applicant provided the requested information for staff review in letters dated June 7 and November 18, 1985, and December 16, 1986. The results of.the staff's evaluation of these submittals are provided below.

(1) Backfit Test Program The applicant applied the Elcometer calibration correction to each Elco-meter reading that was obtainea during the period of improper calibra-tion. An evaluation of the adhesion test results by the applicant showed, with a 95 percent confidence limit, that as much as 20 percent of the coated surfaces on miscellaneous steel items inside the containment build-ing of Unit 1 could have failed to meet the minimum test criterion. Accord-ingly, 36,000 square feet of coating on miscellaneous steel surfaces have been added to the Coatings Exempt Log for Unit 1. The staff has detemined that this action meets the guidelines for the backfit test program in Appendix M of Supplement 9 and is, therefore, acceptable.

(2) Traceability The applicant provided a listing of all nonconformance reports on protec-tive coating systems inside the containment buildings with technical justi-fications for the use-as-is disposition of the discrepant coating materials.

The technical justifications were reviewed by the Comanche Peak Coating Engineering Manager and an independent, third-party consultant and found to support the acceptability of the batches of the discrepant coating materials listed in the nonconfomance reports. The staff determined that this provision meets the guidelines for traceability in Appendix M of Supplement 9 and is, therefore, acceptable.

Comanche Peak SSER 21 1 Appendix M

k (3) Coatinas Procedure After a review of the technical requirements for containment building coat-ing work, which resulted in the rewriting of all the procedures and instruc-tions pertaining to coating application, inspection, testing, and documen-tation, the applicant provided a nonsafety-related containment coatings program. The program includes criteria to achieve quality coating material and workmanship;.namely, the use of coating materials that meet the design-basis-accident.(DBA) conditions, compliance with the technical requirements of paint application specifications, quality verification of coating work, and traceability of coating quality verification documentation. The staff l determined th6t this provision meets the guidelines for coatings procedures l in Appendix M of Supplement 9 and'is, therefore, acceptable. '

(4) Coatings Exempt Log i

The applicant has updated the estimates of the exempted coating areas inside l the containment building of Unit 1. In so doing, the applicant included in '

the Coatings Exempt Log for Unit 1 (1) the total coating surface area on miscellaneous steel items that failed the adhesitn test, (2) the coating areas that have unsatisfactory dry film thicknesses. (3) the coating areas that are not qualified under DBA conditions and (4) the coating work that was applied in inaccessible or limited-access areas where all specified requirements were not met.. The staff has determined that this action meets the guidelines for the Coatings Exempt Log in Appendix M of Supplement 9 and is, therefore, acceptable. The outstanding issue regarding containment sump performance is considered closed.

I i

i Comanche Peak SSER 21 2 Appendix h

s APPENDIX V REVIEW 0F LICENSEE AND APPLICANT RESPONSES TO NRC GENERIC LETTER 83-28 ITEM 1.2 - POST-TRIP REVIEW: DATA AND INFORMATION CAPABILITIES l

l Comanche Peak SSER 21 Appendix V

h SAIC-85/1523-3 REVIEW OF LICENSEE AND APPLICANT RESPONSES TO NRC GENERIC LETTER 83-28 (Required Actions Based on Generic Implications df Salem ATWS Events), Item 1.2*

" POST-TRIP REVIEW: DATA AND INFORMATION CAPABILITIES" FOR COMANCHE PEAK STEAM ELECTRIC STATION, UNITS 1 AND 2 (50-445, 50-446) i Technical Evaluation Report Prepared by Science Applications International Corporation 1710 Goodridge Drive McLean, Virginia 22102 Prepared for U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Contract No. NRC-03-82-096 l-i 1

1

- Comanche Peak SSER 21 Appendix V

i FOREWORD This report contains the technical evaluation of the Comanche Peak Steam Electric Station, Units 1 and 2 response to Generic Letter 83-28 (Required Actions Based on Generic Implications of Salem ATWS Events), Item 1.2 " Post Trip Review: Data and Information Capabilities."

For the purposes of this evaluation, the review criteria, presented in part 2 of this report, were divided into five separate categories. These are:

1. The parameters monitored by the sequence of events and the time history recorders,

2. The performance characteristics of the sequence of events recorders,

3. The performance characteristics of the time history recorders,

4. The data output format, and

5. The long-term data retention capability for post-trip review material.

All available responses to Generic Letter 83-28 were evaluated. The plant for which this report is applicable was found to have adequately responded to, and met, categories 2, 4 and 5.

The report describes the specific methods used to determine the cate-gorization of the responses to Generic Letter 83-28. Since this evaluation report was intended to apply to more than one nuclear power plant specifics regarding how each plant met (or failed to meet) the review criteria are not presented. Instead, the evaluation presents a categorization of the responses according to which categories of reveiw criteria are satisfied and which are not. The evaluations are based on specific criteria (Section 2) derived from the requirements as stated in the generic letter.

Comanche Peak SSER 21 Appendix V

TABLE OF CONTENTS

.Section p g, Introduction. . . . . . . . . . . . . . . . . . . . . . .

, i

1. Background. . . . . . . . . . . . . . . . . . . . . . . . . . 2

2. Review Criteria . . . ... . . . . . . . . . ,,. . . . . ,_,_, 3 l

u

3. Evniustion. . . . . . . . . . . . . . . . . . . . . . . , , .

g 4.- Conclusion. . . . .-. . . . . . . . . . . . . . . . . . . . , _

9

5. R e fe re n c e s . . . . . . . . . . . . . . . . . . . . , , , , , ,. 10

.6.

Summry Of Evaluation of GL 8 3-28 Iten 1. 2 . . . . . . . . . . . . . . . . . . . 11 I

1 i

Comanche Peak SSER 21 Appendix V

_ _ _ = _ _ _ _ . _ _ -

INTRODUCTION l

l SAIC has reviewed the utility's response to Generic Letter 83-28, item 1.2 " Post-Trip Review: Data and Information Capability." The response (see references) contained sufficient information to determine that the data and I

information capabilities at this plant are acceptable in the following areas.  !

e The sequence-of-events recorder (s) performance charac-teristics.

e The output format of the recorded data, o The long-term data retention, record keeping, capa-bility.

However, the data and information capabilities, as described in the submittal, either fail to meet the review criteria or provide insu.fficient information to allow determination of the adequacy of the data and information capabilities in the following areas.

e The parameters monitored by both the sequence-of-events and time history recorders.

e The time history recorder (s) performance characteris-tics.

l f

Comanche Peak SSER 21 1 Appendix V

1. Background On February 25, 1984, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant failed to open upon an automatic reactor trip signal from the reactor protection system. This incident occu.rred during the plant startup and the reactor was tripped manually by the operator about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breakers has been determined to be related to the sticking of the under voltage trip attachment. Prior to this incident; on February 22, 1983; at Unit 1 of the Salem Nuclear Power Plant an automatic trip signal was generated based on steam generator low-low level during plant startup.

In this case the reactor was tripped manually by the operator almost coinci-dentally with the automatic trip. At that time, because the utility did not have a requirement for the systematic evaluation of the reactor trip, no investigation was performed to determine whether the reactor was tripped automatically as expected er manually. The utilities' written procedures required only that the cause of the trip be determi'ned and identified the responsible personnel that could authorize a restat t if the cause of the trip is known. Following the second trip which clearly indicated the problem w'ith the trip break ~ers, the question was raised on whether the circuit breakers had functioned properly curing the earlier incident. The most useful source of information in this case, namely the sequence of events printout which would have indicated whether the reactor was tripped automatically or manually during the February 22 incident, was not retained after the incident. Thus, no judgment on the proper functioning of the trip system during the earlier incident could be made.

Following these incidents; on February 28, 1983; the NRC Executive Director for Operations (EDO), directed the staff to investigate and report on the generic implications of these occurrences at Unit 1 of the Salem j

Nuclear fower Plant. The results of the staf f's inquiry into the generic '

implications of the Salem Unit incidents is reported in NUREG-1000. " Generic Implications of ATWS Events at the Salem Nuclear Power Plant." Based on the results of this study, a set of required actions were developed and included in Generic Letter 83-28 which was issued on July 8,1983 and sent to all licensees of operating reactors, applicants for operating license, and construction permit holders. The required actions in this generic letter consist of four categories. These are: (1) Post-Trip Review, (2) Equipment Comanche Peak SSER 21 2 Appendix V

a.

Classification and Vender Interface, (3) Post Maintenance Testing, and (4)

' Reactor Trip System Reliability linprovements.

The first required action of the generic letter, Post-Trip Review, is the subject of this TER and consists of action item 1.1 " Program Description and Procedure" and action item 1.2 " Data and Information Capability." In

'the next section the review criteria used to assess the adequacy of the utilities' responses to the requirements of action item 1.2 will be discussed.

2. Review Criteria The intent of' the Post Trip Review requirements of Generic Letter 83-28 is to ensure that the licensee has adequate procedures and data and l information sources to understand the cause(s) and progression of a reactor trip. This understanding should go beyond a simple identification of the ,

l course of the event. It should include the capability to determine the root cause of the reactor trip and to determine whether safety limits have been exceeded'and if so to what extent. Sufficient information about the reactor trip event should be available so that a decision on the acceptability of a l

reactor restart can be made.

The following are the review criteria developed for the requirements of Generic Letter 83-28, action item 1.2: j The equipment that provides the digital sequence of events (SOE) record -l and the analog time history records of an unscheduled shutdown should pro-vide a reliable source of the necessary information to be used in the post j trip review. Each plant variable which is necessary to determine the cause(s) and progression of the event (s) following a plant trip should be monitored by at least one recorder [such as a sequence-of-events recorder or a plant process computer for digital parameters; and strip ~ charts, a plant process computer or analog recorder for analog (time history) variables).

Each device used'to record an analog or digital plant ~ variable should be l described in sufficient detail so that a determination can be made as to whether the following performance characteristics are met:

Comanche Peak SSER 21 3 Appendix V

e Each sequence-of-events recorder should be capable of detecting and recording. the sequence of events with a sufficient time j discrimination capability to ensure that the time responses asso- I ciated w'ith each monitored safety-related system can be ascer-tained.. and that a determination can b6 made as.to whether the time response is within acceptable limits based.on FSAR Chapter 15 Accident Analyses. The recommended guideline for the SOE time

' discrimination is approximately 100 msec. If current SOE recorders do not have this time discrimination capability the licensee or applicant should show that the current time discrimi-

, nation capability is sufficient for an adequate reconstruction of the course of.the reactor trip. As a minimum this should include  ;

the ability to adequately reconstruct the accident scenarios pre-sented in Chapter 15. of the plant FSAR.

o- Each analog time ~ history data recorder should have a sample inter-

' val small enough so that the incident can be accurately reconstructed following a reactor trip. As a minimum, _the licensee or applicant should be able to reconstruct the course of the accident sequences evaluated in the accident analysis of the plant FSAR (Chapter 15). The recommended r,uideline for the sample interval is 10 sec. If the time history equipment does not meet this guideline, the licensee or applicant should show that the current time history capability is sufficient to accurately recon-struct the accident sequences presented in Chapter 15 of the FSAR.

2 e To support the post trip analysis of the cause of the trip and the

{ '

proper functioning of involved safety related equipment, each e analog time history data recorder should be capable of updating and retainfng information from approximately five ninutes prior to the trip until at least ten minutes after the trip.

e .The. information gathered by the sequence-of-events and time history data collectors should be stored in a manner that will allow for retrieval and analysis. The data may be retained in either hardcopy (computer printout, strip chart output, etc.) or in an accessible memory (magnetic disc or tape). This information should be presented in a readable and meaningful format, taking, Comanche Peak SSER 21 4 Appendix V

into consideration good human factors practices (such as those outlined in NUREG-0700).

e All equipment used to record sequence of events and time history information should be powered,from a reliable and non-interruptible power source. The power source used need not be safety related.

The sequence of events and time history recording equipment should monitor sufficient digital and analog parameters, respectively, to assure that the course of the reactor trip can be reconstructed. The parameters monitored should provide sufficient information to determine the root cause of the reactor trip, the progression of the reactor trip, and the response of the plant parameters and systems to the reactor trip. Specifically, all input parameters as.sociated with re6ctor trips, safety injections and other safety-related systems as well as output parameters sufficient to record the proper functioning of these systems should be recorded for us'e in the post trip review. The parameters deemed necessary, as a minimum, to perform a post-trip review (one tha.t would determine if the plant remained within its design envelope) are presented on Tables 1.2-1 and 1.2-2. If the appli-cants' or licensees' SOE recorders and time history recorders do not monitor a'l of the parameters suggested in these tables the applicant or licensee should show that the existing set of monitored parameters are sufficient to establish that the plant remained within the design envelope for the appro-priate accident conditions; such as those analyzed in Chapter 15 of the plant Safety Analysis Report.

Information gathered during the post trip review is required input for future post trip rev,iews. Data from all unscheduled shutdowns provides a valuable reference source for the determination of the acceptability of the plant vital parameter and equipment response to future unscheduled shut-downs. It is therefore necessary that information gathered during all post trip reviews be maintained in an accessible manner for the life of the plant.

l Comanche Peak SSER 21 5 Appendix V

Table 1.2-1. PWR Parameter List SOE Time History Recorder Recorder Parameter / Signal x R'eactor Trip (1) x Safety injection x Containment Isolation (1) x Turbine Trip x Control Rod Position (1) x x Neutron Flux, Power x x Containment Pressure  !

(2) Containment Radiation x Containment Sump Level

.(1) x x Primary System Pressure (1) x x Primary System Temperature (1)'x Pressurizer Level (1) x Reactor Coolant Pump Status (1) x x Primary System Flow (3) Safety Inj.; Flow, Pump / Valve Status x MSIV Position x x Steam Generator Pressure (1) x x Steam Generator Level (1) x x Feedwater Flow (1) x x Steam Flow (3) Auxiliary Feedwater System; Flow.

Pump /Value Status x AC and DC System Status (Bus Voltage) x Diesel Generator Status (Start /Stop, On/Off) x PORY Position (1): Trip parameters j (2): Parameter may be monitored by either an SOE or time history recorder.

(3): Acceptable recorder options are: (a) system flow recorded on an SOE l

recorder, (b) system flow recorded on a time history recorder, or (c) equipment status recorded on an SOE recorder.

Comanche Peak SSER 21 6 Appendix V

t-f This page has not been included because it applies to boiling-water reactors only.

I Comanche Peak SSER 21 7 Appendix V

3. Evaluation The parameters identified in part 2 of this report as a part of the review criteria are those deemed necessary to perform an adequate post-trip review. The recording of these parameters on equipment that meets the guidelines of the review criteria will result in a source of information that can be used to determine the cause of the reactor trip and the plant j response to the trip, including the responses of important plant systems.

The parameters identified in this submittal as being recorded by the sequence of events and time history recorders do not correspond to the parameters specified in part 2 of this report.

The review criteria require that the equipment being used to record the sequence of events and time history data required for a post-trip review meet certain performance characteristics. These characteristics are intended to ensure that, if the proper parameters are recorded, the record-ing equipment will provide an adequate source of information for an effec-tive post-trip review. The information provided in this submittal does not indicate that the time history equipment used would meet the intent of the performance criteria outlined in part 2 of this report. Information supplied in the submittal does indicate that the s'0E equipment meets the performance criteria specified in part 2 of this report.

The data and information recorded for use in the post-trip review should be output in a format that allows for ease of identification and use of the data to meet the review criterion that calls for information in a readable and meaningful format. The information contained in this submittal indicates that this criterion is met.

The data and information used during a post-trip review should be retained as part of the plant files. This information could prove useful during future post-trip reviews. Therefore, one criterion is that infor-mation used during a post-trip review be maintained in an accessible manner for the life of the plant. The information contained within this submittal indicates that this criterion will be met.

I Comanche Peak SSER 21 8 Appendix V

Y

4. Conclusion

.The information supplied _in response to Generic Letter 83-28 indicates that the. current post-trip review data and information capabilities are .;

adequate in the' following areas:

1. The recorded data is output in a readable and meaningful format.

1

2. The information'reerded for the post-trip review is maintained in an accessible manner for the life of the plant.

l 3. The sequence of events recorders meet the minimum performance characteristics.

l The information supplied in response to Generic Letter 83-28 does not ,

Indicate that the post-trip review data and information capabilities are adequate in the following areas.

1. Based upon the informat. ion contained in the' submittal, all of the parameters specified in part 2 of this report that should be-recorded for use in a post-trip review are not recorded.

2. Time history recorders, as described in the submittal, do not meet the minimum performance characteristics.

It'is possible that the current data and information capabilities at this nuclear power pl a nt are adequate to meet the intent of these review I criteria, but were not completely described. Under these circumstances, the licensee should provide an updated, more complete, description to show in more detail the data and information' capabilities at this nuclear power pl a nt. If the information provided accurately represents all current data and information capabilities, then the licensee should tither snow that the data and information capabilities meet the intent of the criteria in part 2 of this report, or detail future modifications that would enable the licensee to meet the intent of the evaluation criteria. J Comanche Peak SSER 21 9 Appendix V

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...--_,---.__-___,__w___,_,_,,__,,_____u -

REFERENCES NRC Generic Letter 83-28. " Letter to all licensees of operating reactors, applicants for operating license, and holders of construction permits regarding Required Actions Based on Generic Implications of Salem ATWS Events." July 8, 1983.

NUREG-1000 Generic Implications of ATWS Events at the Salem Nuclear 1 Power Plant, April 1983. j Letter from' B.R.' Clements, Texas Utilities Generating Company, to D.G.

Eisenhut, NRC, dated November 3,1983 Accession Number 8311100183 in i response to Generic Letter 83-28 of July 8,1983, with attachment. '

I CPSES response to NRC Generic Letter 83-28 dated November 7, 1983.

Letter from D.R. Clements, Texas Utilities Generating Company, to D.G.

Eisenhut, NRC, dated April 30, 1984, Accession Number 8405080179 providing information in response to Generic letter 83-28.

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l Comanche Peak SSER 21 10 Appendix V

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Comanche peak i

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1. Parameters recorded: Unsatisfactory See. attached table for discrepancies.

2. SOE recorders performance characteristics: Satisfactory Dedicated computer: (Westinghouse P2500) time discrimination between events: approximately 4ms, powered by non-interruptible power source.

3. Time history recorders performance characteristics: Unsatisfactory ,

Dedicated computer: Sampling rate is 10 sec except for 2 variables which are scanned at 25 sec, powered by non-interruptible source, the time history duration is from 2 min before to 3 min after the trip.

4. Data output format: Satisfactory SOE data: time, parameter descriptor, change of state are among the output

' Analog data: ' time, parameter descriptor, and parameter value are among the output

5. Data retention capability: Satisfactory l

Pertinent data will be stored for the life of the plant.

Comanche Peak SSER 21 11 Appendix V

1 Desirable PWR Parameters for Post-Trip Review (circled parameters are not recorded)

SOE Time History Recorder Recorder Parameter / Signal x Reactor Trip (1)x Safety Injection

.@ Containment Isolation (1)x Turbine Trip

@ Control Rod Position (1) x x Neutron Flux, Power x x Containment Pressure

@ Containment Radiation

@ Containment Sump Le'.el (1) x x Primary System Pressure (1) x x Primary System Temperature (1)x Pressurizer Level (1) x Reactor Coolant Pump Status (1) x x Primary System Flow

@ Safety Inj.; Flow, Pump /Yalve Status

@ MS!V Position x x Steam Generator Pressure (1) x x Steam Generator Level (1)@ x Feedwater Flow (1)@ x Steam Flow

@ Auxiliary Feedwater System; flow.

Pump /Value Status

@ AC and DC System Status (Bus Voltage) x Ditsel Generator Status (Start /Stop, On/Off)

@ PORV Position (1): Trip parameters (2): Parameter may be monitored by either an SOE or time history recorder.

(3): Acceptable recorder options are: (a) system flow recorded on an SOE recorder (b) system flow recorded on a time history recorder, or (c) equipment status recorded on an SOE recorder.

Comanche Peak SSER 21 12 Appendix V

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lif:1 M. APPENDIX W a

s; sj . CONFORMANCE TO GENERIC LETTER 83-28

ITEM 2.1 (PART 1) - EQUIPMENT CLASSIFICATION g (RTS COMPONENTS) fi b

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i Comanche Peak SSER 21 Appendix W

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H EGG-NTA-7224 1

k e

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l CONFORMANCE TO GENERIC LETTER 83-28

-ITEM.2.1 (PART 1) EQUIPMENT' CLASSIFICATION (RTS COMPONENTS)

CATAWBA 1 AND 2 '

MCGUIRE.1 AND 21 COMANCHE _ PEAK,1 AND 2 DIABLO CANYON 1 AND 2 R. Haroldsen Published May 1986 EG&G Idaho, Inc.

Idaho falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-761001570 FIN Nos. 0600 and 06002 Comanche Peak SSER 21 Appendix W o

ABS 1RACT l

l This EG&G Idaho, Inc. report provides a review of the submittals from selected operating and applicant Pressurized Water Reactor (PWR) plants for conformance to Generic Letter 83-28. Item 2.1 (Part 1). The following plants are included in this review.

Plant Name Docket Number TAC Number Catawba 1 50 413 57743 Catawba 2 50 414 OL McGuire 1 50 369 52852 McGuire 2 50 370 52853 Comanche Peak 1 50 445 OL Comanche Peak 2 50 446 OL Diablo Canyon 1 50 275 52832 Diablo Canyon 2 50 323 OL l

Comanche Peak SSER 21 ii Appendix W

o

)

FOREWORD ,

i This report is supplied as part of the program for evaluating licensee / applicant conformance to Generic Letter 83-28. " Required Actions Based on Generic Implications of Salem ATWS Events." This work is being conducted for the U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation, Division of PWR Licensing-A, by the EG&G Idaho, Inc.

The U.S. Nuclear Regulatory Commission funded this work under the authorization B&R 20-19-10-11-3 and 20-19-40-41-3, FIN Nos. 06001 and 06002.

Comanche Peak SSER 21 iii Appendix W

CONTENTS ABSTRACT ............ ................................................. 11 FOREWORD .............................................................. iii

1. INTRODUCTION AND

SUMMARY

......................................... 1

2. PLANT RESPONSE EVALUATIONS ....................................... 3 2.1 Catawba 1 and 2, McGuire 1 and 2 ........................... 3 2.2 Conclusion ................................................. 3 2.3 Comanche Peak 1 and 2 ...................................... 4 2.4 Conclusion ................................................. 4 2.5 Diablo Canyon 1 and 2 ...................................... 5 2.6 Conclusion ................................................. 5

3. GENERIC REFERENCES ............................................... 6 l

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Comanche Peak SSER 21 iv Appendix W

I

1. ' INTRODUCTION AND

SUMMARY

On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant failed to open upon an automatic reactor trip signal froa the reactor protection system. This incident was terminated manually by the operator about 00 seconds after the initiation of the

' automatic trip signal. The failure of the circuit breakers was determined to be related to the sticking of the undervoltage trip attachment. Prior to the incident,.on February 22. 1983, an automatic trip signal was generated at Unit 1 of the Salem Nuclear Power Plant based on steam generator low-low level during plant startup. In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip.

Following these incidents, on February 28, 1983, the NRC Executive Director of Operations (E00), directed the staff to investigate and report on the generic implications of these occurrences at Unit 1 of the Salem Nuclear Power Plant. The results of the staff's inquiry into the generic implications of the Salem Unit 1 incidents are reported in NUREG-1000

" Generic Implications of the ATWS Events at the Salem Nuclear Power Plant."1 As a result of this investigation, the Commission (NRC) requested (by Generic Letter 83-28, dated July 8, 1983) all licensees of cperating reactors. applicants for an operating license, and holders of construction permits to respond to generic issues raised by the analyses of these two ATWS events.

This report is an evaluation of the resoonses submitted from a group of similar pressurized water reactors for Item 2.1 (Part 1) of Generic Letter 83-28.

The results of the reviews of saveral plant responses are reported on in this document to enhance review efficiency. The specific plants reviewed in this report were selected based on the similarity of plant design and convenierice of review. The actual documents which were reviewed Comanche Peak SSER 21 1 Appendix W

for each evaluation are listed at the end of each plant evaluation. The generic documents referenced in this report are listed at the end of the report.

Part 1 of Item 2.1 of Generic Letter 83-28 requires the licensee or applicant to confirm that all reactor trip system components are identified, classified, and treated as safety-related as indicated in the following statement:

Licensees and applicant: shall confirm that all components whose functioning is required to trip the reactor are identified as safety-related on documents, procedures, and information handling systems used in the plant to control safety-related activities, including maintenance, work orders, and parts replacement.

Comanche Peak SSER 21 2 Appendix W

s

2. PLANT RESPONSE EVALUATIONS 2.1 Catawba 1. 50 413, TAC No. 57739. Catawba 2. 50 414 (OL)

McGuire 1 and 2. 50-369/370. TAC Nos. 53014/53015 The licensee / applicant for the Catawba 1 and 2 and McGuire 1 and 2 plants (Duke Power Co.) responded to the requirements of Item 2.1 (Part 1) with 2 separate but essentially identical submittals dated November 4, 1983. Both submittals state that all components of the reactor trip system whose functioning is required to trip the reactor are identified as safety-related on documents, procedures, and handling systems used in the plants to control safety-related activities.

2.2 Cenclusion Based on the review of the licensee's/ applicant's submittals, we find that the licensee's/ applicant's responses confirm that the components required to trip the reactor have been identified as safety-related and are identified un relevant plant documents. The'se responses,' therefore, meet the requirements of Item 2.1 (Part 1) of Generic Letter 83-28 and are acceptable.

REFERENCES

1. Letter H. B. Tucker, Duke Power Co., to D. G. Eisenhut NRC, " Catawba Nuclear Station, Docket Nos. 50-369, 50-370," November 4, 1983.

1 I

2. Letter, H. B. Tucker, Duke Power Co., to D. G. Eisenhut NRC, "McGuire Nuclear Station, Docket Nos. 50-369, 50-370," November 4, 1983.

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Comanche Peak SSER 23 3 Appendix W

2.3 Comanche Peak Units 1 and 2. 50 445/446 (OL Plants)

The applicant for Comanche Peak Units 1 and 2 (Texas Utilities Generating Co.) responded to the requirements of Items 2.1 (Part 1) in submittals dated November 3, 1983 and November 21, 1983. The first submittal requested a time extension for Items 2.1 and 2.2.

. The second submittal stated that the applicant had completed a review of the reactor trip system and its components to ensure proper )

classification as part of the 0-list development. The reactor trip system and its components are stated to be classified as class IE (safety-related) j and are included in the Q-list. Future maintenance work orders and replacement parts are procedurally required to be evaluated for proper

• safety classifications prior to approval.

2.4 Conclusion Based on the review of the licensee's submittals, we find that the licensee has verified that the components that are necessary to perform reactor trip are classified as safety-related and that acti'atties relating to the safety-related components are controlled by procedures which reflect the special requirements for handling safety-related compor:ents. We, therefore, find that the licensee's responses meet the requirements of Item 2.1 (Part 1) and are ac eptable.

REFERENCES

1. Letter, 8. R. Clements, Texas Utilities Generating Co., to D. G. Eisenhut, November 3, 1983.

2. Letter, B. R. Clements Texas Utilities Generating Co., to D. G. Eisenhut, November 21, 1983.

Comanche Peak SSER 21 4 Appendix W

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2.5 Diablo Canyon 1. 50-225. TAC No. 52832. Diablo Canvan 2. 50-323 (OL)

The licensee / applicant of Diablo Canyon Units 1 and 2 (Pacific Gas and l- Electric Co.) responded to the requirements of Item 2.1 (Part 1) in submittals dated November 7, 1983, January 24, 1985 and April 18, 1985.

The first 2 submittals provided schedule and status reports on the item.

The third submittal stated that the licensee /applicent had developed a list of components that trip the reactor and had confirmed that the appropriate components were classified as safety-related. The submittal also states that a list of documents had been developed that identify and control the activities that may affect the reactor trip system components.

2.6 Conclusion Based on the review of the licensee's submittals, we find that the l

licensee has verified that the components that are necessary to perform reactor trip are classified as safety-related and that activities relating to the safety-related components are controlled by procedures which reflect the special requirements for handling safety-related components. We, therefore, find that the licensee's responses meet the requirements of Item 2.1 (Part 1) and are acceptable.

REFERENCES

1. Letter, J. O. Schuyler, Pacific Gas and Electric Co., to D. G. Eisenhut, NRC, November 7, 1983.

2. Letter, J. D. Shiffer, Pacific Gas and Electric Co., to G. W. Knighton, NRC, January 24, 1985.

3. Letter, J. D. Shiffer, Pacific Gas and Electric Co., to G. W. Knighton, NRC, April 18, 1985.

Comanche Peak SSER 21 5 Appendix W

', 'l

3. GENERIC REFERENCES

1. Generic Implications of'ATWS Events at the Salem Nuclear power Plant, NUREG-1000, Volume 1 April 1983; Volume 2, July 1983.

2. NRC. Letter, D. G. Eisenhut to all Licensees of Operating Reactors,

' Applicants for Operating License, and Holders of Construction Permits,

" Required Actions Based on Generic Implications of Salem ATWS Events L

(Generic Letter 83-28)," July 8, 1983.

L Comanche Peak'SSER 21 6 Appendix W

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APPENDIX X l

CONFORMANCE TO GENERIC LETTER 83-28 ITEM 2.1 (PART 2) - REACTOR TRIP SYSTEM VENDOR INTERFACE ll 1

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l Appendix X Comanche Peak SSER 21

EGG-NTA-7447  !

INPUT FOR SAFETY EVALUATION REPORT BEAVER VALLEY POWER STATION UNIT 1 BEAVER VAL.EY POWER STATION UNIT 2 BYRON STATION UNITS 1 AND 2 CALLAWAY PLANT UNIT 1 CATAWBA NUCLEAR STATION UNITS 1 AND 2 COMANCHE PEAK STEAM ELECTRIC STATION UNITS 1 AND 2 l JOSEPH M. FARLEY UNITS 1 AND 2 INDIAN POINT UNIT 2 INDIAN POINT UNIT 3 REACTOR TRIP SYSTEM VENDOR INTERFACE <

ITEM 2.1 (PART 2) 0F GENERIC LETTER 83-28 F. G. Farmer Published November 1986 Idaho National Engineering Laboratory EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 l

J Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-76100lS70 FIN Nos. 06001 and D6002 Comanche Peak SSER 21 Appendix X

ABSTRACT This EG&G Idaho, Inc. report provides a review of the submittals for some of the Westinghouse (W) nuclear plants for conformance to Generic Letter 83-28, Item 2.1 (Part 2). The report includes the following plants, all Westinghouse, and is in partial fulfillment of the following TAC Nos.:

Plant Docket Number TAC Number {

Beaver Valley Power Station Unit 1 50-344 52818 )

Beaver Valley Power Station Unit 2 (OL) 50-412 62495 Byron Station Unit 1 50-454 56276 Byron Station Unit 2 (0L) 50-455 N/A Callaway Plant Unit 1 50-483 55196 Catawba Nuclear Station Unit 1 50-413 57743 Catawba Nuclear Station Unit 2 (OL) 50-414 N/A Comanche Peak Steam Electric Station Unit 1 (OL) 50-445 N/A Comanche Peak Steam Electric Station Unit 2 (OL) 50-446 N/A Joseph M. Farley Unit 1 50-348 52836 Joseph M. Farley Unit 2 50-364 52837 Indian Point Unit 2 50-247 52846 Indian Point Unit 3 50-286 52947 l

Comanche Peak SSER 21 ii Appendix X l

__ J

i,

! FOREWORD This report is provided as part of the program for evaluating licensee / applicant conformance to Generic Letter 83-28, " Required Actions Based on Generic Implications of Salem ATWS Events." This work is conducted for the U..S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of PWR Licensing-A by EG&G Idaho, Inc.

The U. S. Nuclear Regulatory Commission funded the work under the authorization, B&R 20-19-19-11-3, FIN Nos. 06001 and 06002.

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Comanche Peak SSER 21 iii Appendix X

I CONTENTS i

ABSTRACT ............... .......................................... ii FOREWORD ............................................................. iii

1. INTRODUCTION .................................................... 1

2. REVIEW REQUIREMENTS ............................................. 2

3. GROUP REVIEW RESULTS ............................................ 3

4. REVIEW RESULTS FOR BEAVER VALLEY POWER STATION UNIT 1 ........... 4 4.1 Evaluation ................................................ 4 4.2 Conclusion ................................................ 4

5. REVIEW RESULTS FOR BEAVER VALLEY POWER STATION UNIT 2 ........... 5 5.1 . Evaluation ................................................ 5 5.2 Conclusion ................................................ 5

6. REVIEW RESULTS FOR BYRON STATION UN ITS 1 AND 2 . . . . . . . . . . . . . . . . . . 6 6.1 Evaluation ................................................ 6

.I

'1 6.2 Conclusion ................................................ 6 1

7. REVIEW RESULTS FOR CALLAWAY PLANT UNIT 1 ........................ 7 l

1 7.1 Evaluation ................................................ 7 l

l 7.2 Conclusion ................................................ 7

8. REVIEW RESULTS FOR CATAWBA NUCLEAR STATION UNITS 1 AND 2 ........ 8 8.1 Evaluation ................................................ 8 8.2 Conclusion ................................................ 8

9. REVIEW RESULTS FOR CDMANCHE PEAK STEAM ELECTRIC STATION UNITS 1 AND 2 ................................................... 9 9.1 Evaluation ................................................ 9 9.2 Conclusion ................................................ 9

10. REVIEW RESULTS FOR JOSEPH M. FARLEY UNITS 1 AND 2................ 10 10.1 Evaluation ................................................ 10 Comanche Peak SSER 21 iv Appendix X

10.2 Conclusion ................................................ 10

11. REVIEW RESULTS FOR INDIAN POINT UNIT 2 .......................... 11 11.1 Evaluation ................................................ 11 11.2 Conclusion ................................................. 11

12. REVIEW RESULTS FOR IND IAN POINT UN IT 3 . . . . . . . . . . . . . . . . . . . . . . . . . . 12 12.1 Evaluation ................................................ 12 12.2 Conclusion................................................ 12

13. GROUP CONCLUSION ................................................ 13

14. REFERENCES ...................................................... 14 Comanche Peak SSER 21 Appendix X

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CONFORMANCE TO GENERIC LETTER 83-28 ITEM 2.1 (PART 2)

BEAVER VALLEY POWER STATION UNIT 1 1

BEAVER VALLEY POWER STATION UNIT 2 BYRON STATION UNITS 1 AND 2 CALLAWAY PLANT UNIT 1 CATAWBA NUCLEAR STATION UNITS 1 AND 2 0

C,MANCHE PEAK STEAM ELECTRIC STATION UNITS 1 AND 2 JOSEPH M FARLEY UNITS 1 AND 2 INDIAN POINT UNIT 2 INDIAN POINT UNIT 3

1. INTRODUCTION On July 8,1983, Generic Letter 83-2S Iwas issued by D. G. Eisenhut, 1

Of rector of the Division of Licensing, Office of Nuclear Reactor l Regulation, to all licensees of operating reactors, applicants for operating licenses, and holders of construction permits. This letter included required actions based on generic implications of the Salem ATWS events. These requirements have been published in Volume 2 of NUREG-1000,

" Generic Implications of ATWS Events at the Salem Nuclear Power Plant."2 This report documents the EG&G Idaho, Inc. review of the submittals of a group of Westinghouse plants including Beaver Valley Units 1 and 2, Byron Units 1 and 2, Callaway Unit 1, Catawba Units 1 and 2, Comanche Peak Units 1 and 2, Farley Units 1 and 2 and Indian Point Units 2 and 3 for j conformance to Item 2.1 (Part 2) of Generic Letter 33-28. The submittals l from the licensees and applicants utilized in these evaluations are ]'

referenced in Section 14 of this report.

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Comanche Peak SSER 21 1 Appendix X

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2. REVIEW REQUIREMENTS l Item 2.1 (Part 2) (Reactor Trip System - Vendor Interface) requires licensees and aoplicants to establisn, implement and maintain a continuing program to ensure that vendor' information on Reactor Trip System (RTS) components is complete, current and controlled throughout the life of the plant, and appropriately referenced or incorporated in plant instructions and procedures._ The vendor interface program is to include periodi::

conrnunications with vendors to assure that all applicable information has been received, as well as a system of positive feedback with vendors for l mailings containing technical information, e. g., licensee / applicant acknowledgement for receipt of technical information.

That part of the vendor interface program which ens'ures that vendor information on RTS components, once acquired, is appropriately controlled, referenced and incorporated in plant instructions and procedures, will be evaluated as part of the review of Item 2.2 of the Generic Letter.

~

Because the Nuclear Steam System Supplier (NSSS) is ordinarily alt

~

the supplier of the entire RTS, the NSSS is also the principal source of information on the components of the RTS. This review of the licensee and applicant submittals will:

1. Confirm that the licensee /apolicant has identified an interface with either the NSSS or with the vendors of each of the comoonents of the Reacto- Trip System.

I

~2. Confirm that the interface identified by licensees / applicants includes periodic communication with the NSSS or with the vendors of each of i the components of the Reactor Trip System.

3. Confi'm that the interface identified by licensees / applicants includes a system of positive feedback to confirm receipt of transmittals of technical information.

Comanche Peak SSER 21 2 Appendix X

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3. GROUP REVIEW RESULTS The relevant submittals from each of the included reactor plants were reviewed to determine compliance with Item 2.1 (Part 2). First, the submittals from each plant were reviewed to establish that Item 2.1 (Part

2) was specifically addressed. Second, the submittals were evaluated to determine the extent to which each of the plants complies with the staff guidelines for Item 2.1 (Part 2).

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h Comanche Peak SSER 21 3 Appendix X

4 REVIEW RESULTS FOR BEAVER VALLEY POWER STATION, UNIT i 4.1 Evaluation Ouquesne Light, the licensee for Beaver Valley 1, provided their response to Item 2.1 (Part 2) of the Generic Letter on November 4, 1983.

In that response, the licensee confirms that the NSSS for Beaver Valley 1 is Westinghouse and that the RTS for Beaver Valley 1 is included as a part of the Westinghouse interface progran established for the Beaver Valley 1 NSSS.

The Westinghouse interface program for the NSSS includes both periodic communication between Westinghouse and licensees / applicants and positive feedback from licensees / applicants in the form of signed receipts for technical information transmitted by Westinghouse.

? Conclusion The staff finds the licensee's confirming statement that Beaver Valley e is a participant in the Westinghouse interf ace program for the RTS meets the staff position on Item 2.1 (Part 2) of the Generic Letter and is, therefore, acceptable.

. Comanche Peak SSER 21 4 Appendix X

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5. REVIEW RESULTS FOR BEAVER VALLEY POWER STATION, UNIT 2 5.1 Evaluation Duquesne Light, the applicant for Beaver Valley 2, provided their '

response to Item 2.1 (Part 2) of the Generic Letter on March 30, 1984 In that response, the applicant confirms that the NSSS for Beaver Valley 2 is -)

Westinghouse and that the RTS for Beaver Valley 2 is included as a part of the Westinghouse interface program established for the Beaver Valley 2 NSSS.

The Westinghouse interface program for the NSSS includes both periodic communication between Westinghouse and licensees / applicants and positive ,

feedback from licensees / applicants in the form of signed receipts for technical information transmitted by Westinghouse. i l

l 5.2 Conclusion The staff finds the applicant's confirming stateme,nt that Beaver Val, ley' 2 is a participant in the Westinghouse interf ace program for the RTS meets the staff position on Item 2.1 (Part 2) of the Generic Letter and is, therefore, acceptable.

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6. REVIEW 9ESULT:i FOR SYRON STATION UNITS 1 AND 2 6.1 Evaluation Commonwealth Edison, the licensee for Byron 1 and applicant for Byron 2, responded to Item 2.1 (Part 2) of the Generic Letter on May 7,1985. In that response, the licensee / applicant confirms that the NSSS for Byron is Westinghouse and that the RTS for Byron is included as a part of the Westinghouse interface program established for the Byron 1 and 2 NSSS.

The Westinghouse interface program for the NSSS includes both periodic communication between Westinghouse and licensees / applicants and positive feedback from licensees / applicants in the form of signed receipts for technical information transmitted by Westinghouse.

6.2 Conclusion The staff finds the licensee's/ applicant's confirming statement that Syron is a participant in the Westinghouse interface program for the RTS meets the staff position on Item 2.1 (Part 2) of the Generic Letter and is, therefore, acceptable, o

Comanche Peak SSER 21 6 Appendix X

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7. REVIEW RESULTS FOR CALLAWAY PLANT UNIT 1 l

7.1 Evaluation Union Electric Company, the licensee for Callaway, responded to Item 2.1 (Part 2) of the Generic Letter on November 18, 1983. In that response the licensee confirms that the NSSS for Callaway is Westinghouse and that the RTS for Callaway is included as a part of the Westinghouse interface program established for the Callaway NSSS.

The Westinghouse interf ace program for the NSSS includes both periodic communication between Westinghouse and licensees / applicants and positive feedback from licensees / applicants in the form of signed receipts for technical information transmitted by Westinghouse.

7.2 Conclusion The staff finds the licensee's confirming st'atement that Callaway is a participant in the Westinghouse interf ace program for the RTS meets the staff position on Item 2.1 (Part 2) of the Generic Letter and is, therefore, acceptable.

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8. REVIEW RESULTS FOR CATAWBA NUCLEAR STATION UNITS 1 AND 2

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8.1 Evaluation Duke Power Company, the applicant for Catawba Units 1 and 2, responded to Item 2.1 (Part 2) of the Generic Letter on November 4, 1983. In that response, the applicant confirms that the NSSS for Catawba is Westinghouse and that the RTS for Catawba is included as a part of the Westinghouse I interf ace program established for the Catawba NSSS.

The Westinghouse interf ace program for the NSSS includes both periodic communication between Westinghouse and licensees / applicants and positive feedback from licensees / applicants in the form of signed receipts for technical information ' ansmitted by Westinghouse.

8.2 Conclusion The staff finds the applicant's confirming statement that Catawb.a.is a I participant in the Westinghouse interf ace program for the RTS meets the staff position on Item 2.1 (Part 2) of the Generic Letter and is, therefore, acceptable.

Comanche Peak SSER 21 8 Appendix X

9. REVIEW RESULTS FOR COMANCHE PEAK STEAM ELECTRIC STATION UNITS 1 AND 2 9.1 Evaluation Texas Utilities Generating Company, the applicant for Comanche Peak Units 1 and 2, responded to Item 2.1 (Part 2) of the Generic letter on November 21, 1983. In that response, the applicant confirms that the NSSS for Comanche Peak is Westinghouse and that the RTS for Comanche Peak is included as a part of the Westinghouse interface program established for the Comanche Peak NSSS.

The Westinghouse interf ace program for the NSSS includes both periodic communication between Westinghouse and licensees / applicants and positive feedback from licensees / applicants in the form of signed receipts for technical information transmitted by Westinghouse.

5.2 Conclusion The staff finds the applicant's confirming statement that Comanche Peak is a participant in the Westinghouse interfa e program for the RTS meets the staff position on Item 2.1 (Part 2) of the Generic Letter and is, therefore, acceptable.

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10. REVIEW RESULTS FOR JOSEPH M. FARLEY NUCLEAR PLANT UNITS 1 AND 2 10.1 Evaluation '

Alabama Power, the licensee for Farley 1 and 2, provided their response to Item 2.1 (Part 2) of the Generic letter on November 4,1983.

In that response, the licensee confirms that the NSSS for Farley 1 and 2 is l Westinghouse and that the RTS for Farley 1 and 2 is included as a part of I the Westinghouse interface program established for the Farley NSSS.

The Westinghouse interf ace program for the NSSS includes both periodic communication between Westinghouse and licensees / applicants and positive feedback from licensees / applicants in the form of signed receipts for technical information transmitted by Westinghouse.

10.2 Conclusion The staff finds the licensee's confirming statement that Farley 1 and 2 is a participant in the Westinghouse interf ace program for the RTS meets the staff position on Item 2.1 (Part 2) of the Generic Letter and is, therefore, acceptable.

l Comanche Peak SSER 21 10 Appendix X

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11. REVIEW RESULTS FOR INDIAN POINT UNIT 2 11.1 Evaluation Consolidated Edison Company, the licensee for Indian Point 2, provided-their response to Item 2.1 (Part 2) of the Generic Letter on November 4,

,1983. lIn.that response, the licensee confirms th a the NSSS for Indian Point 2 is Westinghouse and that the RTS for Indian Point 2 is included as a part of the Westinghouse interface program established for the Indian Point 2 NSSS. '

The Westinghouse interface program for the NSSS includes both periodic communication between Westinghouse and licensees / applicants and positive feedback from licensees / applicants in the form of signed receipts for l technical information transmitted by Westinghouse.

11.2 Conclusion l

.2.is pa t Da n the et ou t c pro r fo t TS me the staff position on Item 2.1 (Part 2) of the Generic Letter and is,  !

therefore, acceptable.

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12. REVIEW RESULTS FOR INDIAN POINT UNIT 3 12.1 Evaluation The New York Power Authority, the licensee for Indian Point 3, provided their response to Item 2.1 (Part 2) of the Generic Letter on '

November 7,1983. In that response, the licensee confirms that the NSSS for Indian Point 3 is Westinghouse and that the RTS for Indian Point 3 is t included as a part of the Westinghouse interface program established for the Indian Point 3 NSSS.

The Westinghouse interface program for the NSSS includes both periodic j communication between Westinghouse and licensees / applicants and positive feedback from licensees / applicants in the form of signed receipts for technical-information transmitted by Westinghouse.

11.2 Conclusion The staff finds the licensee's confirming statement that Indian Poin't 3 is a participant in the Westinghouse interface program for the RTS meets the staff position on Item 2.1 (Part 2) of the Generic Letter and is, therefore, acceptable.

l Comanche Peak SSER 21 12 Appendix X L - - - - - - - - _ _ - - - _ - - _

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13. GROUP CONCLUSION

. , '; c . . The staff. concludes that the licensee / applicant responses'for the '

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listed. Westinghouse plants for Item.4.5.2 of Generic Letter 83-28 are

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14 REFERENCES

1. .NRC Letter, D. G. Eisenhut to all licensees of Operating Reactors, Aoplicants for Operating License, and Holders of Construction Permits,

" Required Actions Based on Generic Implications of Salem ATWS Events (Generic Letter 83-28)," July 8,1983.

2. Generic Implications of ATWS Events at the Salem Nuclear Power Plant

NUREG-1000, Volume 1, April 1983; Volume 2, July 1983.

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3. Duquesne Light letter to NRC, J. J. Carey to D. G. Eisenhut, Director, Division of Licensing, NRC, " Response to Generic Letter 83-28,"

November 4, 1983.

4 Duquesne Light letter to NRC, E. J. Woolever to D. G. Eisenhut, Director, Division of Licensing, NRC, " Response to Generic Letter 83-25," March 30, 1984

5. Conrnonwealth Edison letter to NRC, G. L. Alexander to Harold R.

Denton, Director, Office of Nuclear Recctor Regulation, May 7,1985. i

6. Union Electric Company letter to NRC, D. F. Schnell to Harold R.

Denton, Director of Nuclear Reactor Regulation, NRC, " Response to j Generic Letter 83-28," November 18, 1983. 1 i 7.. Duke Power Comoany letter to NRC, H. B. Tucker to D. G. Eisenhut, Director, Division of Licensing, November 4, 1983.

, 8. Texas Utilities Generating Comoany letter to NRC, Billy R. Clements to D. G. Eisenhut, Director, Division of Licensing, " Supplement 1 to the CPSES Response to Generic Letter 83-28," Novemoer 21, 1983.

9. Alabama Power letter to NRC, F. L. Clayton to Director, Nuclear Reactor Regulation, " Response to Generic Letter 83-28," November 4, 1983.

10. Consolidated Edison Coq any letter to NRC, John D. O'Toole to D. G.

Eisenhut, Director, E e,sion of Licensing, November 4, 1983.

11. New York Power Authority letter to NRC, J. P. Bayne to D. G. Eisenhut, Director, Division of Licensing, " Required Actions Based on Generic Implications of Salem ATWS Events (Generic Letter 83-28, dated July 8, j 1983)," November 7, 1983, 1

Comanche Peak SSER 21 14 Appendix X

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APPENDIX Y CONFORMANCE TO GENERIC LETTER 83-28 ITEMS 3.1.3 AND 3.2.3 - POST-MAINTENANCE TESTING OF REACTOR TRIP SYSTEM COMPONENTS AND OTHER SAFETY-RELATED COMPONENTS Comanche Peak SSER 21 Appendix Y

EGG-NTA-7206 l

CONFORMANCE TO GENERIC LETTER 83-28 ITEMS 3.1.3 and 3.2.3 COMANCHE PEAK UNITS 1 AND 2 R. Haroldsen 4

Published April 1986 EG&G Idaho, Inc.

Idaho Falls. Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. ~20555 Under DOE Contract No. DE-AC07-761001570 FIN No. D6002 Comanche Peak SSER 21 Appendix Y

l ABSTRACT This EG&G Idaho, Inc., report provides a review of the submittals from comanche Peak Units 1 and 2, for conformance to Generic Letter 83-28 Items 3.1.3 and 3.2.3.

I Docket Nos. 50-445 and 50-446 Comanche Peak SSER 21 ii Appendix Y

FOREWORD This report is supplied as part of the program for evaluating licensee / applicant conformance to Generic Letter 83-28 " Required Actions Based on Generic Implications of Salem ATWS Events." This work is being conducted for the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of PWR Licensing-A by EG&G Idaho, Inc., NRR l and I&E Support Branch.

The U.S. Nuclear Regulatory Commission funded the work under the authorization, B&R 10-19-19-11-3, FIN No. D6002.

4' Docket Nos. 50 445 and 50-446.

l Comanche Peak SSER 21 iii Appendix Y

CONTENTS ABSTRACT ............................................... .............. ii FOREWORD .............................................................. iii

1. INTRODUCTION ..................................................... 1

2. REVIEW REQUIREMENTS .............................................. 2

3. REVIEW RESULTS ................................................... 3 3.1 Evaluation ................................................. 3 3.2 Conclusion ................................................. 3

4. REFERENCES ....................................................... 4 I

Comanche Peak SSER 21 iv Appendix Y i 1

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l CONFORMANCE TO GENERIC LETTER 83-28 l ITEMS 3.1.3 AND 3.2.3 COMANCHE PEAK UNITS 1 AND 2 i

1. INTRODUCTION I

On July 8, 1983, Generic Letter No. 83-28 was issued by D. G. Eisenhut, Director of the Division of Licensing, Nuclear Reactor Regulation, to all licensees of operating reactors, applicants for operating licenses, and holders of construction permits. This letter included required actions based on the generic implications of the Salem ATWS events. These requirements have been published in Volume 2 of NUPEG-1000, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant".

This report documents the EG&G Idaho, Inc., review of the submittals from Comanche Peak Units 1 and 2, for conformance to items 3.1.3 and 3.2.3 of Generic Letter 83-28. The submittals and other documents utilized in this evaluation are referenced in Section 4 of this report.

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2. REVIEW REQUIREMENTS item 3.1.3 (Post-Maintenance Testing of Reactor Trip System Components) requires licensees and applicants to identify, if applicable, any post-maintenance test requirements for the reactor trip system (RTS) in existing technical specifications that can be demonstrated to degrade rather-than enhance safety. Item 3.2.3 applies this same requirement to all other safety-related components. Any proposed technical specification changes resulting from this action shall receive a pre-implementation review by the NRC.

The relevant submittals for Comanche Peak Units 1 and 2, were reviewed to determine compliance with Items 3.1.3 and 3.2.3 of the generic letter.

First, the submittals from this plant were reviewed to determine that these two items were specifically addressed. Second, the submittals were checked to determine if any post-maintenance test items specified in the technical specifications were identified that were suspected to degrade rather than l enhance safety. Last, the submittal was reviewed,for evidence of special conditions or other significant information relating to the two items of concern.

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3. REVIEW RESULTS FOR COMANCHE PEAK UNITS 1 AND 2 j l

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.3.1 Evaluation '

The Texas Utility Generating Company, the applicant for Comanche Peak Units 1 and 2 provided responses to Items 3.1.3 and 3.2.3 of Generic Letter- I 83-28 in submittals dated November 3, 1983 June 7, 1985 and August 19, 1985.5 At the time of the initial submittal the technical specifications were undergoing review in preparation for licensing. The applicant stated that a review was being conducted to identify testing or maintenance which could degrade rather than enhance safety.

In the subsequent submitta k the applicant provided a broad discussion of surveillance testing rather than the more limited specific concerns of post-maintenance testing identified by Items 3.1.3 and 3.2.3. However, the last submittal did provide a confirmation that no post-maintenance test  ;

program for reactor trip systems components.had been identified that may degrade safety, i

3.2 Conclusions The applicants confirmation that no post-maintenance test requirements i have been identified in the technical specifications that degrade safety meets the requirements of Items 3.1.3 and 3.2.3 of Generic Letter 83-29.

We therefore, find the applicant's submittals acceptable.

Comanche Peak SSER 21 3 Appendix Y i

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4. REFERENCES

1. NRC Letter, D. G. Eisenhut to all Licensees of Operating Reactors, Applicants for Operating License, and Holders of Construction Permits,

" Required Actions Based on Generic Implications of Salem ATWS Events (Generic Letter 83-28)", July 8, 1983.

2. Generic Implications of ATWS Events at the Salem Nuclear Power Plant, NUREG-1000, Volur:e 1, April 1983; Volume 2, July 1983.

3. Letter, 8. R. Clements, Texas Utilities Generating Co., to D. G. Eisenhut, NRC, November 3, 1983.

4. Letter, J. W. Beck, Texas Utilities Generatinn Co., to V. S. Noonan, j NRC, June 7, 1985.

5. Letter, W. G. Counsil, Texas Utilities Generating Co. to V. S. Noonan, NRC, August 19, 1985.

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APPENDIX Z CONFORMANCE TO GENERIC LETTER 83-28 ITEM 4.5.2 - REACTOR TRIP SYSTEM RELIABILITY l

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INPUT FOR SAFETY EVALUATION REPORT BEAVER VALLEY POWER STATION UNIT 1 1 b: AVER VALLEY POWER STATION UNIT 2 BRAIDWOOD STATION UNITS 1 AND 2 BYRON STATION UNITS 1 AND 2 CALLAWAY PLANT UNIT 1 CATAWBA NUCLEAR STATION UNITS 1 AND 2 COMANCHE PEAK STEAM ELECTRIC STATION UNITS 1 AND 2 DONALD C. COOK NUCLEAR PLANT UNITS 1 AND 2 DIABLO CANYON UNITS 1 AND 2 REACTOR TRIP SYSTEM RELIABILITY ITEM 4.5.2 0F GENERIC LETTER 83-28 F. G. Farmer l

Published November 1986 Idaho National Engineering Laboratory EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 l

r Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-76ID01570 FIN Nos. 06001 and 06002 Comanche Peak SSER 21 Appendix Z

ABSTRACT This EG&G Idaho, Inc. report provides a review of the submittals for some of the Westinghouse (W) nuclear plants for conformance to Generic Letter 83-28, Item 4.5.2. The report includes the following plants, all Westinghouse, and is in partial fulfillment of the following TAC Nos.:

Plant pocket Number TAC Number l

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Beaver Valley Power Station Unit 1 50-334 53962 Beaver Valley Power Station Unit 2 (0L) 50-412 62958 Braidwood Statien Unit 1 (OL) 50-456 N/A Braidwood Station Unit 2 (OL) 50-457 N/A Byron Station Unit i 50-454 56288 Byron Station Unit 2 (OL) 50-455 63254 Callaway Plant Unit 1 50-483 55206 Catawba Nuclear Station Unit 50-413 57744 Catawba Nuclear Station Unit 2 (0L) 50-414 N/A Comanche Peak Steam Electric Station Unit 1 (0L) 50-445 N/A Comanche Peak Steam Electric Station Unit 2 (0L) 50-446 N/A Donald C. Cook Nuclear Plant Unit 1 50-315 53971 Donald C. Cook Nuclear Plant Unit 2 50-316 53972 Diablo Canyon Unit 1 50-275 53976 Diablo Canyon Unit 2 (OL) 50-323 61723 l

I Comanche Peak SSER 21 ii Appendix Z

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i FOREWORD This report is provided as part of the program for evaluating licensee / applicant conformance to Generic Letter 83-28, " Required Actions l

Based on Generic Implications of Salem ATWS Events." This work is conducted for the U. S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of PWR Licensing-A by EG&G Idaho, Inc.

The U, S. Nuclear Regulatory Commission funded the work under the authorization, B&R 20-19-19-11-3, FIN Nos, 06001 and 06002.

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CONTENTS:

1 ABSTRACT ............................................................. 11- 11 FOREWORD ....................... ..................................... iii

1. - INTR 000CTION .................................................... 1

2. REVIEW REQUIREMENTS ............................................. 2

3. ' GROUP REVIEW RFSULTS ............................................ 5 'l

, 4 .

4. REVIEW RESULTS F3R BEAVER VALLEY POWER STATION UNIT 1 ........... -6

.i 4.1 Evaluation ................................................ 6 i 4.2 Conclusion ................................................ 6

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5. REVIEW RESULTS FOR BEAVER VALLEY P0WER-STATION UNIT 2 ........... 7 5.1 Evaluation ................................................. 7 5.2. Conclusion ........................ ....................... 7

6. REVIFW RESULTS FOR BRAIDWOOD STATION UNITS 1 AND -2 . . . . . . . . . . . . . . 8-6.1 Evaluation .........s...................................... 8 6.2 Conclusion ................................................ 8

7. REVIEW RESULTS FOR BYRON STATION UNITS 1 AND 2 .................. 9 7.1 Evaluation ................................................ 9-7.2 Conclusion ................................................ 9

8. REVIEW RESULTS FOR CALLAWAY PLANT UNIT 1 .................-....... 10

, 8.1 Evaluation ................................................ 10 i

i-o 8.2 Conclusion ................................................ 10

9. REVIEW RESULTS FOR CATAWBA NUCLEAR STATION UNITS 1 AND 2 ........ 11 9.1 Evaluation ................................................ 11 9.2 -Conclusion .............................................. 11

10. REVIEW RESULTS FOR COMANCHE' PEAK STEAM ELECTRIC STATION UNITS 1 AND 2 .....................................'.............. 12 10.1 Evaluation ................................................ 12 Comanche Peak SSER 21 iv Appendix Z t __ ______ _ _ _

10.2 Conclusion .............. ................... ..... ....... 12

11. REVIEW RESULTS FOR DONALD C. COOK NUCLEAR PLANT UNITS 1 AND 2 ................................................... 13 11.1 Evaluation ................................................ 13 11.2 Conclusion ............................. .................. 13

12. REVIEW RESULTS FOR DIABLO CANYON UNITS 1 AND 2................... 14 9

12.1 Evaluation ................................................ 14 i 12.2 Conclusion ................................................ 14

13. GROUP CONCLUSION ................................................ 15 !

14. REFERENCES ...................................................... 16 i

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i CONFORMANCE TO GENERIC LETTER 83-28 4'

ITEM 4.5.2 BEAVER VALLEY POWER STATION UNIT 1 8EAVER VALLEY POWER ">TATION UNIT 2 BRAIDWOOD STATION UNITS 1 AND 2 l BYRON STATION UNITS 1 AND 2 CALLAWAY PLANT UNIT 1 CATAWBA NUCLEAR STATION UNITS 1 AND 2 COMANCHE PEAK STEAM ELECTRIC STATION UNITS 1 AND 2 DONALD C. COOK NUCLEAR PLANT UNITS 1 AND 2 DIABLO CANYON UNITS 1 AND 2

1. INTRODUCTION On July 8, 1983, Generic Letter 83-28 was issued by D. G. Eisenhut, Director of the Division of Licensing, Office of Nuclear Reactor Regulation, to all licensees of operating reactors, applicants.for operating licenses, and holders of construction permits. This letter included required actions based on generic implications of the Salem ATWS events, These requirements have been published in Volume 2 of NUREG-1000,

" Generic Implications of ATWS Events at the Salem Nuclear Power Plant."

This report documents the EG&G Idaho, Inc. review of the submittals of some of the Westinghouse plants including Beaver Valley Units 1 and 2.

Braidwood Units 1 and 2, Byron Units 1 and 2, Callaway Unit 1, Catawba i Units 1 and 2, Comanche Peak Units 1 and 2, D. C. Cook Units 1 and 2 and Diablo Canyon Units 1 and 2 for conformance to Item 4.5.2 of Generic Letter 83-28. The submittals from the licensees utilizeo in these evaluations are referenced in Section 14 of this report.

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2. REVIEW REQUIREMENTS Item 4.5.2 (Reactor Trip System Reliability - System Functional Testing - On-Line Testing) requires licensees and. applicants with plants not currently designed to permit on-line testing to justify not making modifications to permit such testing. Alternatives to on-line testing will be considered where special circumst'ances exist and where the objective of high reliability can be met in another way. Item 4.5.2 may be ,

interdependent with Item 4.5.3 when there is a need to justify not performing on-line testing because of the peculi.arities of a particular design.

All p*ortions of the Reactor Trip System that do not have on-line l testing capability will be reviewed under the guidelines for this item.

However, the existence of on-line testability for the Reactor Trip Breaker undervoltage and shunt trip attachments on Westinghouse, B&W and CE plants; the silicon controlled rectifiers in the CRDCS on B&W plants; and the scram pilot and backup scram valves on GE plants' will only be confirmed here since they are specifically addressed in Items 4.4 and 4.5.1. Maintenance

' and testing of the Peactor Trip Breakers are also excluded from this review, as they are evaluated under Item 4.2. This review of the r, licensee / applicant sebmittals will:

1. Confirm that the licensee / applicant has identified those portions of the Reactor Trip System that are not on-line testable. If the entire Reactor Trip System is verified to be on-line testable, with those exceptions addressed above, no further review is required.

2. Evaluate modifications proposed by licensees / applicants to permit on-line testing against the existing criteria for the design of the protection systems for the plant being modified.

3. Evaluate proposed alternatives to on-line testing of the Reactor Trip System for ace otability based on the following:

Comanche Peak SSER 21 2 Appendix Z

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a. The licensee / applicant submittal substantiates the impracticality of the modifications necessary to permit on-line testing, and l

b. High Reactor Trip System availability (comparable to that which I would be possible with on-line testing) is achieved in another l way. Any such proposed alternative must be described in detail sufficient to permit an independent evaluation of the basis and analysis provided in lieu of performing on-line testing. Methods that may be used to demonstrate that the objective of high reliability has been met may include the following:

1. Demonstration by systematic analysis that testing at shutdown intervals provides essentially equivalent reliability to that obtained by on-line testing at shorter intervals.

it. Demonstration that reliability equivalent to that obtained ,

by on-line testing is accomplished by additional redundant and diverse components or by other features, iii. Development of a maintenance program based on early

. replacement of critical components that compensates for the 1

) lack of on-line testing. Such a program would require analytical justification supported by test data, iv. Development of a test program that compensates for the lack of on-line testing, e. g., one which uses trend analysis and identification of safety margins for critical parameters of safety-related components. Such a program would require analytical justification supported by test data.

4. Verify the capability to perform independent on-line testing of the  ;

reactor trip system breaker undervoltage and shunt trip attachments on Comanche Peak SSER 21 3 Appendix Z

t-CE plants.. Information from licensees and applicants with CE plants will be reviewed to verify that they require independent on-line testing of the reactor trip breaker undervoltage and shunt trip attachments.

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3. GROUP REVIEW RESULTS The relevant submittals from each of the Westinghouse reactor plants were reviewed to determine compliance with Item 4.5.2. First, the submittals from each plant were reviewed to establish that Item 4.5.2 was specifically addressed. Second, the submittals were evaluated to determine the extent to which each of the Westinghouse plants complies with the staff guidelines for Item 4.5.2.

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4. REVIEW RESULTS FOR BEAVER VALLEY POWER STATION, UNIT 1 4.1 Evaluation Duquesne Light, the licensee for Beaver Valley 1, provided their response to Item 4.5.2 of the Generic letter on November 4, 1983. In that

' response, the licensee states that Beaver Valley 1 performs on-line testing of the Reactor Trip System, with the exceptions of independent on-line testing of the shunt trip attachment and on-line testing of the reactor trip bypass breakers.

The licensee states that performance of independent on-line testing of the shunt trip attachment is contingent on implementation of an automatic shunt trip modification (Item 4.3 of the Generic Letter), which is scheduled to be installed and tested by June 3, 1986.

The licensee states that on-line testing of the bypass breakers during power' operation is not justified because only one bypass breaker can be in service at a time (for less than two hours per month), and that when a bypass breaker is in service the RTS will initiate a trip signal to one of the trip breakers which is tested bimonthly. Also, the addition of comoonents necessary to eliminate the need to' lift leads and install jumpers (currently required to test the bypass breakers) could decrease the reliability of the bypass breaker system.

4.2 Conclusic, <

The staff finds the licensee's statement of the extent to which they currently perform on-line testing of the RTS meets the staff position on Item 4.5.2 of the Generic Letter and is, therefore, acceptable; however, the licensee should confirm that Beaver Valley I has the capability to perf orm independent on-line testing of the ". hunt trip attachment. The staff finds the licensee justi.fication for not performing on-line testing of the bypass breakers sufficient and acceptable.

4 Comanche Peak SSER 21 6 Appendix Z

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5. REVIEW RESULTS FOR BE' AVER VALLEY POWER STATION, UNIT 2 5.1 Evaluation Duquesne Light, the applicant for Beaver Valley 2, prgvided their response to Item 4.5.2 of the Generic Letter on March 30, 1983'. In that response, the applicant states that Beaver Valley 2 will perform on-line testing of the Reactor Trip System, with the exception of on-line testing of the reactor trip bypass breakers. 1 The applicant states that on-line testing of the bypass _ breakers during power operation is not justified because only one bypass breaker can be in service at a time (for less than two hours per month), and that when i a bypass breaker is in service the RTS will initiate a trip signal to one of the trip breakers which is tested bimonthly. Also, the addition of components necessary to eliminate the need to lift leads and install jumpers (currently required to test the bypass breakers) could decrease the reliability of the bypass breaker system.

5.2 Conclusion The staff finds the applicant's statement of the extent to which they will_ perform on-line testing of the RTS meets the staff position on Item 4.5.2 of the Generic Letter and is, therefore, acceptable. Tne staff also finds the applicant's justification for not performing on-line testing of the bypass breakers sufficient and acceptable.

Comanche Peak SSER 21 7 Appendix Z

6. REVIEW RESULTS FOR 3RAIDW000 STATION UNITS 1 AND 2 6.1 Evaluation Commonwealth Edison, the applicant for Braidwood, responded to Item 4.5.2 of the Generic Let ter on November 5,1983. In that response, the applicant states that Braidwood will be modified to permit performance of on-line testing of the Reactor Trip System, and that verification of the undervoltage and shunt trip' features will be performed during the preoperational testing of these components. It is not clear from this response that Braidwood will have the capability to perform independent on-line verification of operability of the shunt and undervoltage trips.

6.2 Conclusion The staff finds the applicant's statement that they will perform on-line testing of the RTS. meets the staff position on Item 4.5.2 of the Generic Letter and is, therefore, acceptable; however, the applicant should confirm that Braidwood will have the capability to perform independent on-line verification of operability of the shunt and undervoltage trips.

Comanche Peak SSER 21 8 Appendix Z

7. REVIEW RESULTS FOR BYRON STATION UNITS 1 AND 2 7.1 Evaluation Commonwealth Edison, the licensee for Byron, responded to Item 4.5.2 of the Generic Letter on November 5, 1983. In that response, the licensee states that Byron will be modified to permit performance of on-line testing of the Reactor Trip System. It is not clear fr'om the response that independent on-line testing of the shunt and undervoltage attachments to the Reactor Trip Breakers will be possible following the modifications.

7.2 Conclusion The staff finds the licensee's statement tnat they perform on-line testing of the RTS meets the staff position on Item 4.5.2 of the Generic Letter and is, therefore, acceptable; however, the licensee should confirm that the shunt and undervoltage attachments can be independently tested on-line.

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l Comanche Peak SSER 21 9 Appendix Z f

8. REVIEW RESULTS FOR CALLAWAY PLANT UNIT 1 8.1 Evaluation Union Electric Company, the licensee for Callaway, responded to item 4.5.2 of the Generic Letter on November 18, 1983. In that response the licensee states that Callaway is designed to permit performance of on-line testing of the Reactor Trip System and commits to on-line testing of the reactor protection system, including independent verification of operability of the diverse trip features.

8.2 Conclusion The staff finds the licensee's statement that they will perform on-line testing of the RTS and independent verification of operability of the diverse trip features meets the staff position on Item 4.5.2 of the Generic Letter and is, therefore, acceptable, l

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Comanche Peak SSER 21 10 Appendix Z

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9. REVIEW RESULTS FOR CATAWBA NUCli AR SI A110N UNITS 1 AND 2 l

9.1 Evaluation Duke Power Company, the licensee / applicant for Catawba Units 1 and 2, responded to Item 4.5.2 of the Generic Letter on November 4, 1983. In that response, the licensee / applicant states that the Catawba design allows performance of on-line testing of the Reactor Trip System, including independent testing of the reactor trip breaker shunt and undervoltage trip attachments.

9.2 Conclusion The staff finds the licensee's/ applicant's statement that the Catawba design permits performance of on-line testing of the RTS meets the staff position on Item 4.5.2 of the Generic Letter and is, therefore, acceptable.

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Comanche Peak SSER 21 11 anoendix Z

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10. REVIEW RESULIS FOR COMANCHE PEAK STEAM ELECTRIC STATION UNITS 1 AND 2 f 10.1 Evaluation Texas Utilities Generating Company, the applicant for Comanche Peak )

Units 1 and 2, responded to Item 4.5.2 of the Generic Letter on November 3, 1983. In that response, the applicant states that the Comanche Peak 'aesign is being modified to, allow performance of on-line testing of the Reactor Trip System, tnat on-line testing of the RTS will be performed at Comanche Peak and that such testing will include independent testing of the reactor 1 trip breaker shunt and undervoltage trip attachments.

10.2 Conclusion .

}

The staff finds the applicant's statement of the extent to which they will perform on-line testing of the RTS meets the staff. position on Item 4.5,2 of the Generic Letter and is, therefore, acceptable. -1 Comanche Peak SSER 21 12 Appendix Z

11. REVIEW RESULTS FOR DONALD C. COOK NUCLEAR STATION UNITS 1 AND 2 i

11.1 Evaluation Indiana and Michigan Electric Company, the licensee for D. C. Cook Units 1 and 2, responded to the Generic Letter on November 4, 1983. The licensee's response confirms that both Cook units perform on-line testing of the Reactor Trip System, with the exception of the diverse trip feature, and that the shunt and undervoltage trip attachments will be independently tested on-line as soon as a pending modification is incorporated. '

11.2 Conclusion The staff finds the licensee's statement that they currently perform on-line testing of the RTS and their commitment to perform independent on-line testing of the shunt and undervoltage trip attachments meet the staff position on Item 4.5.2 of the Generic Letter and is, therefore, acceptable.

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Comanche Peak SSER 21 13 Appendix 2

12. REVIEW RESULTS FOR DIABLO CANYON UNITS 1 AND 2 12.1 Evaluation Pacific Gas and Electric Company, the licensee for Diablo Canyon Units 1 and 2 submitted a response to Item 4.5.2 of the Generic Letter on November 7, 1983. In that response, the licensee states that Diablo Canyon is designed to permit on-line testing of the Reactor Trip System, and that the applicable procedures are being revised to include the required on-line functional testing of the diverse trip features. The licensee's letter of June 27, 1984 confirms revision of those procedures; the procedures, which were included, establish that the shunt and undervoltage trips are tested independently.

12.2 Conclusion The staff finds the licensee's statement that Diablo Canyon Units 1 and 2 are designed to. permit on-line testing of the RTS and independent verification of operability of the diverse trip features meets the staff position on Item,4.5.2 of the Generic Letter and is, therefore, acceptable, i

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Comanche Peak SSER 21 14 Appendix Z

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13. GROUP CONCLUSION The staff concludes that the licensee / applicant responses for the listed Westinghouse plants for Item 4.5.2 of Generic Letter 83-28 are acceptable, with the exception of those for Beaver Valley Unit 1, Braidwood Units I and 2 and Byron Units I and 2, which were found to be incomplete as indicated in the plant specific review results.

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l Comanche Peak SSER 21 15 Appendix Z

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14. REFERENCES

1. NRC Letter, D. G. Eisenhut to all licensees of Operating Reactors, Applicants for Operating License, and Holders of Construction Permits,

" Required Actions Based on Generic Implications of Salem ATWS Events (Generic Letter 83-28)," July 8, 1983.

2. Generic Implications of ATWS Events at the Salem Nuclear Power Plant NUREG-1000, volume 1, April 1983; Volume 2, July 1983.

3. Duquesne Light letter to NRC, J. J. Carey to D. G. Eisenhut, Director, Division of Licensing, NRC, " Response to Generic Letter 83-28,"

November 4, 1983.

4 Duquesne Light letter to NRC, E. J. Woolever to D. G. Eisenhut, Director, Division of Licensing, NRC, " Response to Generic Letter 83-28," March 30, 1984.

5. Commonwealth Edison letter to NRC, P. L. Barnes to Harold R. Denton, Director, Office of Nuclear Reactor Regulation, November 5, 1983.

1

6. Union Electric Company letter to NRC, D. F. Schnell to Harold R.  !

Denton, Director of Nuclear Reactor Regulation, NRC, " Response to I Generic Letter 83-28," November 18, 1983.

7. Duke Power Company lett.er to NRC, H. 8. Tucker to D. G. Eisenhut, Director, Division of Licensing, Novemoer 4, 1983.

8. Texas Utilities Generating Company letter to NRC, Billy R. Clements to D. G. Eisenhut, Director, Division of Licensing, " Response to Generic Letter 83-28," November 3, 1983.

9. Indiana and Michigan Electric Company letter to NRC, M. P. Alexich to D. G. Eisenhut, Director, Division of Licensing, " Required Actions Based on Generic Implications of Salem ATWS Events (Generic Letter 83-28)," November 4, 1983.

10. Pacific Gas and Electric Company letter to NRC, J. O. Schuyler, to D.

G. Eisenhut, Director, Division of Licensing, " Generic letter 83-28 Required Actions Based on Generic Implications of ATWS Events,"

November 7, 1983.

11. Pacific Gas and Electric Company letter to NRC, J. O. Schuyler, to G.

W. Knighton, Chief, Division of Licensing, " Generic Letter 83-28 i Required Actions Based on Generic Implications of ATWS Events " June )

27, 198^.

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Comanche Peak SSER 21 16 Appendix Z

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APPENDIX AA  :

TECHNICAL EVALUATION REPORT - TMI ACTION - NUREG-0737 (II.D.1) i l

I o

Comanche Peak SSER 21 Annendix AA

1 I

I

. EGG-NTA-7701 1 m

i TECHNICAL EVALUATION REPORT TMI ACTION--NUREG-0737-(II.D.1)

COMANCHE PEAK STEAM ELECTRIC STATION, UNIT 1 DOCKET NO. 50-445 s i

N. E. Pace C. Y. Yuan C. L. Nalezny May 1987 Idaho National Engineering Laboratory EG&G Idaho, Inc. 1 Idaho Falls, Idaho 83415 l l

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[

t Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-76100157&

FIN No. D6005 Comanche Peak SSER 21 Appendix AA

k l

l ABSTRACT Light water reactors have experienced a number of occurrences of improper performance of safety and relief valves installed in the primary coolant system. As a result, the authors of NUREG-0578 (TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations) and subsequentij NUREG-0737 (Clarification of TMI Action Plan Requirements) recommended that programs be developed and completed which would reevaluate the functional performance capabilities of Pressurized Water Reactor (PWR) safety, relief, and block valves and which would verify the integrity of the piping systems for normal, transient, and accident conditions. This report documents the review of these programs by the Nuclear Regulatory Commission (NRC) and their consultant, EG&G Idaho, Inc. Specifically, this review examined the response of the Licensee for the Comanche Peak Steam Electric Station, Unit 1, to the requirements of NUREG-0578 and NUREG-0737 and finds that the Licensee has provided an acceptable response, reconfirming that the General Design Criteria 14, 15, and 30 of Appendix A to 10 CFR 50 have been met.

FIN No. 06005--Evaluation of CW Licensing Actions-NUREG-0737, II.D.1 Comanche Peak SSER 21 ii Appendix AA

CONTENTS ii ABSTRACT ..............................................................

1

1. INTRODUCTION .....................................................

1 1.1 Background .................................................

1.2 General Design Criteria and NUREG Requi rements . . . . . . . . . . . . . 1 PWR OWNERS' GROUP RELIEF AND SAFETY VALVE PROGRAM ................ 4 2.

3. PLANT SPECIFIC SUBMITTAL ......................................... 6

4. REVIEW AND EVALUATION ............................................ 7 4.1 Valves Tested .............................................. 7 4.2 Test Conditions ............................................ 7 4

4.2.1 FSAR Steam Transients .............................. 8 4.2.2 FSAR Liquid Transients ............................. 9 4.2.3 Extended High Pressure Injection Event ............. 12 4.2.4 Low Temperature Overpressure Transient ............. 12 4.2.5 PORV Block Val ve Fl uid Conditions . . . . . . . . . . . . . . . . . . 13 ii l 4.3 Operability........I.j...................................... 13 4.3.1 Safety Valves ...................................... 13 4.3.2 Power Operated Relief Valves ....................... 16 4.3.3 Electric Control Circuitry ......................... 17 17 4.3.4 PORV Block Valves ..................................

4.4 Piping and Support Evaluation .............................. 17 4.4.1 Thermal Hydraulic Analysis ......................... 18 4.4.2 Stress Analysis .................................... 21

5. EVALUATION

SUMMARY

............................................... 24

6. REFERENCES ....................................................... 26 TABLES 4.2.1

SUMMARY

OF TEST DATA FOR CROSBY 6M6 SAFETY VALVE AND COMPARISON WITH COMANCHE PEAK REQUIREMENTS ................ 10 4.3.1 EPRI TESTS ON CROSBY HP-BP-86 6M6 SAFETY VALVE ............ 15 Comanche Peak SSER 21 iii Appendix AA L ___-____--_- - -- __

TECHNICAL EVALUATION REPORT TMI ACTION--NUREG-0737 (II.D.1)

COMANCHE PEAK STEAM ELECTRIC STATION. UNIT 1 DOCKET NO. 50-445

1. INTRODUCTION i

1.1 Background

Light water reactor experience has included a number of instances of improper performance of relief and safety valves installed in the primary coolant system. There have been instances of valves opening below set pressure, valves opening above set pressure, and valves failing to open or reseat. From these past instances of improper valve performance, it is not known whether they occurred because of a limited qualification of the valve or because of a basic unreliability of the valve design. It is known that the failure of a power-operated relief valve (PORV) to reseat was a significant contributor to the Three Mile Island (TMI-2) sequence of events. These facts led the task force which prepared NUREG-0578 (Reference 1) and, subsequently, NUREG-0737 (Reference 2) to recommend that programs be developed and executed which would reexamine the functional performance capabilities of Pressurized Water Reactor (PWR) safety, relief, and block valves and which would verify the integrity of the piping systems for normal, transient, and accident conditions. These programs were deemed necessary to reconfirm that the General Design Criteria 14, 15, and 30 of l

Appendix A to Part 50 of the Code of Federal Regulations,10 CFR, are indeed satisfied.

1.2 General Design Criteria and NUREG Requirements i

General Design Criteria 14, 15, and 30 require that (a) the reactor primary coolant pressure boundary be designed, fabricated, and tested so as to have an extremely low probability of abnormal leakage, (b) the reactor coolant system and associated avr411ary, control, and protection systems be designed with sufficient margin to assure that the design conditions are Comanche Peak SSER 21 1 Appendix AA

= - - - - _ _ _ . - - __ _ - _ _ - - - _ _ - _ _ __ - _

not exceeded during normal operation or anticipated transient events, and (c) the components which are part of the reactor coolant pressure boundary shall be constructed to the highest quality standards practical.

To reconfirm the integrity of overpressure protection systems and thereby assure that the General Design Criteria are met, the NUREG-0578 position was issued as a requirement in a letter dated September 13, 1979 by the Dtvision of Licensing (DL), Office of Nuclear Reactor Regulation (NRR),

to ALL OPERATING NUCLEAR POWER PLANTS. This requirement has since been incorporated as Item II.D.1 of NUREG-0737, Clarification of TMI Action Plan Requirements (Reference 2), which was issued for implementation on October 31, 1980. As stated in the NUREG reports, each pressurized water reactor Licensee or Applicant shall:

1. Conduct testing to qualify reactor coolant system relief and safety valves under expected operating conditions for design basis transients and accidents.

2. Determine valve expected operating conditions i.nrough the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Rev. 2.

3. Choose the single failures such that the dynamic forces on the safety and relief valves are maximized.

4. Use the highest test pressures predicted by conventional safety analysis procedures.

5. Include in the relief and safety valve qualification program the qualification of the associated control circuitry.

6. Provide test data for Nuclear Regulatory Commission (NRC) staff review and evaluation, including criteria for success or failure of valves tested.

Comanche Peak SSER 21 2 Appendix AA

7. Submit a correlation or other evidence to substantiate that the valves tested in a generic test program demonstrate the functionability of as-installed primary relief and safety valves.

This correlation must show that the test conditions used are equivalent to expected operating and accident conditions as prescribed in the Final Safety Analysis Report (FSAR). The effect of as-built relief and safety valve discharge piping on valve operability must be considered.

8. Qualify the plant specific safety and relief valve piping and supports by comparing to test data and/or performing appropriate analysis.

1 i

i Comanche Peak SSER 21 3 Appendix AA

2. PWR OWNERS' GROUP RELIEF AND SAFETY VALVE PROGRAM In response to the NUREG requirements previously listed, a group of 1 utilities with PWRs requested the assistance of the Electric Power Research k Institute (EPRI) in developing and implementing a generic test

• program for pressurizer power operated relief valves, safety valves, block valves, and

' associated piping systems. The Texas Utilities Generating Co. (TUGC), owner  !

of the Comanche Peak Steam Electric Station (CPSES), Unit 1, was one of the utilities sponsoring the EPRI Valve Test Program. The results of the program are contained in a group of reports which were transmitted to the NRC by Reference 3. The applicability of these reports is discussed below.

EPRI developed a plan (Reference 4) for testing PWR safety, relief, and block valves under conditions which bound actual plant operating conditions. EPRI, through the valve manufacturers, identified the valves used in the overpressure protection systems of the participating utilities.

Representative valves were selected for testing with a sufficient number of the variable characteristics that their testing would adequately demonstrate the performance of the valves ,used by utilities (Reference 5,). EPRI, through the Nuclear Steam Supply System (NSSS) vendors, evaluated the FSARs of the participating utilities and arrived at a test matrix which bounded j the plant transients for which overpressure protection would be required (Reference 6). j EPRI contracted with Westinghouse Electric Corp. to produce a report on the inlet fluid conditions for pressurizer safety and relief valves in Westinghouse designed plants (Reference 7). Since CPSES, Unit 1, was designed by Westinghouse this report is relevant to this evaluation.

Several test series were sponsered by EPRI. PORVs and block valves were tested at the Duke Power Company Marshall Steam Station located in Terrell, North Carolina. Additional PORV tests were conducted at the Wyle Laboratories Test Facility located in Norco, California. Safety valves were tested at the Combustion Engineering Company, Kressinger Development Laboratory, located in Windsor, Connecticut. The results for the relief Comanche Peak SSER 21 4 Appendix AA

and safety valve tests are reported in Reference 8. The results for the block valves tests are reported in Reference 9. A guide for applying these test results to the specific plants is presented in Reference 10.

The primary objective of the EPRI/C-E Valve Test Program was to test each of the various types of primary system safety valves used in PWRs for the full range of fluid conditions under which they may be required to operate. The conditions selected for test (based on analysis) were limited to steam, subcooled water, and steam to water transition. Additional ,

1 objectives were to (a) obtain valve capacity data,' (b) assess hydraulic and structural effects of associated piping on valve operability, and (c) obtain piping response data that could ultimately be used for verifying analytical piping models.

Transmittal of the test results meets the requirement of Item 6 of Section 1.2 to provide test data to the NRC.

l 1

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i Comanche Peak SSER 21 5 Appendix AA

3. PLANT SPECIFIC SUBMITTAL A preliminary plant specific evaluation of the adequacy of the overpressure protection system for CpSES, Unit 1, was submitted by TUGC to the NRC on July 8, 1981 (Reference 11) and March 31, 1982 (Reference 12).

The information provided was reviewed and a request for additional information was sent on July 5, 1985 (Reference 13), to which the licensee responded on June 13, 1986 (Reference 14). On March 27, 1987 (Reference 15) the NRC transmitted a second request for information to TUGC, and the ,

licensee responded on April 15, 1987 (Reference 16).

The response of the overpressure protection system to Anticipated Transients Without Scram (ATWS) and the operation of the system during feed and bleed decay heat removal are not considered in this review. Neither the Licensee nor the NRC have evaluated the performance of the system for these events.

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Comanche Peak SSER 21 6 Appendix AA

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4. REVIEW AND EVALUATION l

4.1 Valves Tested CPSES, Unit 1, is a four-loop PWR designed by the Westinghouse Electric Co. It is equipped with three (3) safety valves, two (2) PORVs, and two (2)

PORV block valves in its overpressure protection system. The safety valves are 6-in. Crosby Model HB-BP-86, 6M6, spring loaded valves with loop seal internals. The design set pressure is 2485 psig and the rated steam flow capacity is 420,000 lbm/h. The PORVs are 3-in. Copes-Vulcan Model D-100-160 globe valves with 316 S.S. stellited plugs and 17-4 PH cages. The PORV opening set pressure is 2335 psig and the rated steam flow capacity is 210,000 lbm/h. The inlet pipe to the safety valve and PORVs include loop seals. The PORV block valves are 3-in. Westinghouse Model 3GM88 gate valves with Limitorque SB-00-15 motor operators; the operator gear ratio has been modified and the wiring has been changed for limit control to assure valve operability as recommended by Westinghouse.

Safety valves, PORVs, and PORV block valves identical to those used at l

CPSES, Unit 1, were included in the EPRI tests. Since there is no difference between the valves tested and the valves installed at the plant, the test results for these valves are directly applicable to CPSES, Unit 1. ,

Therefore, those parts of the criteria of Items 1 and 7 as identified in Section 1.2 of this report regarding applicability of the test valves are fulfilled.

4.2 Test Conditions As stated above, CPSES, Unit 1, is a four-loop PWR designed by the Westinghouse Electric Corp. The valve inlet fluid conditions that bound the overpressure transients for Westinghouse designed PWR Plants are identified in Reference 7. The transients considered in this report include FSAR, extended high p~ressure injection, and low temperature overpressurization events. The expected fluid conditions for each of these events and the applicable EPRI tests are discussed in this section.

Comanche Peak SSER 21 7 Appendix AA 1

\ _ _ _ _ . _--- . _ _ _ . _ . _ _ _ _ _ _ - - - - _ -

i l

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4.2.1 FSAR Steam Transients For the CPSES, Unit 1, PWR, the limiting FSAR transients resulting in steam discharge through the safety valves alone and in steam discharge  !

through both the safety and relief valves are the loss of load event (for maximum pressurizer pressure) and the locked rotor event (for the maximum pressurization rate).

In the case when the safety valves actuate alone, the maximum pressu'rizor pressure and maximum pressurization rate are predicted to be 2555 psia and 144 psi /s, respectively. The maximum developed backpressure in the outlet piping is 500 psig (Reference 15). The loop seal is insulated such that the valve inlet temperature is 3000F (Reference 15).

/

EPRI tests representative of the valve inlet fluid conditions for the limiting transient were selected for the plant specific evaluation. In I selecting the EPRI tests, the safety valve ring settings and the pressure j drop through the inlet pipe were also considered. For' steam flow conditions, four loop seal discharge tests (Test No. 929, 1406, 1415, 1419) were applicable to CPSES, Unit 1. These tests were performed with valve ring settings representative of the typical ring settings used in Westinghouse PWRs including CpSES. The ring settings used in these tests were (-71, -18) or (-77, -18). These represent the upper and lower ring positions measured from the level position referenced to the bottom of the disc ring. The relative ring settings used at CPSES Unit 1 are -82 to -103, and -18 relative to the level position. Since both the test ring settings and the in-plant ring settings were determined by the valve manufacturer, the Crosby Valve and Gage Co., using the same methods and the same standard of performance, these ring settings are considered comparable.

Both the inlet pipe length and the water seal volume in the EPRI tests were greater than those employed at the plant. Therefore, the pressure drop through the inlet pipe would be higher in the tests than that at CPSES, Unit 1. The Licensee has provided calculated values for the inlet pressure drop on valve opening and closing which are comparable to the test Comanche Peak SSER 21 8 Appendix AA

i values. The loop seal temperature measured in the tests ranged from 90 to 3600F at the valve inlet. The maximum pressurizer (tank 1) pressures were in the range of 2675 to 2760 psia and the pressurization rate was 90 to 360 psi /s. The backpressure developed in the tests were 245 to 710 psia.

The above data, which is summarized in Table 4.2.1, show that the inlet fluid conditions and backpressure of these tests envelop the corresponding fluid data predicted for the CPSES, Unit 1, safety valves.

When both the safety valves and PORVs are actuated, the maximum pressurizer pressure is predicted to be 2532 psia and the maximum pressurization rate is 130 psi /s. In the EPRI tests on the Copes-Vulcan PORV, the maximum steam pressure at valve opening was 2715 psia, which bounds the predicted pressure at CPSES, Unit 1. The backpressure developed at the outlet of the PORVs is not an important consideration, since the air operated PORVs used at the CPSES, Unit 1, plant are not sensitive to backpressure (Reference 6). Therefore the EPRI test inlet fluid conditions for the PORV in steam discharge are representative of the plant specific transisnt conditions.

l 4.2.2 FSAR Liquid Transients The limiting FSAR transient resulting in liquid discharge through the PORVs and safety valves is the main feedline break accident (Reference 7).

From a review of feedwater line break analysis for CPSES, Unit 1, it is clear that the feedwater line break is most likely to be the limiting transient for providing high pressure liquid to tne safety valves, a fluid for which they were not originally designed. Therefore, in accordance with the NUREG requirements, the safety valves and PORVs should be qualified for inlet conditions typical of the feedline break event. Therefore, the valve operability will be reviewed using the feedline break data provided in Reference 7.

Reference 7 showed that, in a feedline break accident at CPSES, Unit 1, )

the maximum pressure at the safety valve inlet during liquid discharge was l

l Comanche Peak SSER 21 9 Appendix AA i l

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calculated to be 2503 psia, the pressurization rate was 5 psi /s and the maximum pressurizer surge rate was 1109.5 gpm (-369,000 lbm/hr) liquid at 608-6150F.

In a feedline break accident resulting in safety valve actuation, water discharge is always preceded by steam and steam to water transition flows.

Among the EPRI tests performed on the 6M6 valve, Tests 931a and 931b were performed for loop seal steam, steam to water transition, and water discharge conditions. The valve ring settings and inlet pipe configuration used in these tests were comparable to those of the in-plant safety valves.

In Test No. 931a, the maximum inlet pressure was 2578 psia. The pressurization rate was 2.5 psi /s, the inlet fluid temperature was 1170F and the tank fluid temperature was 6350F. After the valve closed in Test 931a, the system was allowed to repressurize and the valve cycled on approximately 6400F water for Test 931b. Since the inlet temperature and pressure of the tests bounded the predicted in-plant condition, the results of these tests are considered representative of the CPSES, Unit 1, safety I valves. The inlet fluid conditions and corresponding test data for liquid discharge are also summarized in Table 4.2.1.

The expected fluid conditions at the inlet of the safety valve were based on a Westinghouse analysis that assumed the PORVs were not operable during the feedline break transient. If the PORVs are operable, the same fluid conditions postulated for the safety valve inlet can also be expected at the p0RV inlet (Reference 6). In the EPRI tests, high temperature water 1 6

discharge and steam to water transition tests were performed with the j Copes-Vulcan PORV. In the water discharge test, Test No. 76-CV-316-2W, the maximum pressure at the valve inlet was 2535 psia and the temperature was 6470F. In the transition test, Test No. 77-CV-316-7S/W, the maximum inlet pressure was 2532 psia and the water temperature was 6570F. The inlet fluid conditions for these tests compare well with the predicted maximum pressure and temperature of 2505 psia and 6150F for the CPSES, Unit 1, plant. Therefore this test is adequate to represent the in-plant PORV performance in the feedline break event.

Comanche Peak SSER 21 11 Appendix AA

-4.2.3 Extended High Pressure Injection Event The limiting extended high pressure injection event is the spurious actuation of the safety injection system at power (Reference 7). For a four-loop plant, both the safety valves and PORVs will be challenged. Both steam and water discharge are expected. For safety valve actuation, the maximum pressure at valve ' inlet is predicted to be 2507 psia and the pressurization rate is within 4 psi /s. The inlet' temperature ranges from 567 to 5720F. For the PORV, the maximum pressure is predicted to be 2353 psia and the pressurization rate is within 4 psi /s and the inlet {

temperature ranges from 565 to 5690F. In this event, however, the safety valves or PORVs open on steam and liquid discharge would not be observed until the pressurizer becomes water solid. According to Reference 7, this would not occur until at least 20 minutes into the event which allows ample  !

time for operator action. Thus the potential for liquid discharge in extended HPI events can be disregarded.

4.2.4 Low Temperature Overpressurization Transient The PORV is used for overpressure protection during the low temperature stages in reactor startup and shutdown operations. The low pressure set point of the PORVs vary with valve and temperature and range from 445 psig to 2350 psig (Reference 14), The expected inlet fluid conditions for low temperature overpressurization transients are identified in Reference 7 and range from cold water to steam.

For steam discharge through the PORV, the high pressure steam tests discussed in Section 4.2.1 would cover the low pressure steam conditions predicted for low temperature overpressurization transients. For water discharge conditions, there were two low pressure and low temperature water tests performed on the Copes-Vulcan PORV with stellited plug and 17-4 PH cage. The tests were conducted at an inlet pressure of 675 psia and water temperatures of 1050F and 4420F, respectively. These conditions are representative of those at CPSES, Unit 1. Therefore, the EPRI tests can be used to evaluate the performance of the CPSES, Unit 1, PORV for low temperature overpressurization transients.

Comanche Peak SSER 21 12 Appendix AA

I 4.2.5 PORV Block Valve Fluid Conditions The block valves used are 3-inch Westinghouse Model 3GM88 gate valves with Limitorque SB-00-15 operators. These valves were tested for full flow and pressure conditions in the EPRI tests and, as modified, qualify for operation at these conditions.

The test sequences and analyses described above demonstrate that the test conditions bound the conditions for the plant values. They also verify that Items 2 and 4 of Section 1.2 have been met, in that conditions for the operational occurrences have been determined and the highest predicted pressures were chosen for the test. The part of Item 7, which requires showing that the test conditions are equivalent to conditions prescribed in the FSAR, is also met.

4.3 Operability 4.3.1 Safety Valves The EPRI tests representative of the steam discharge condition for the CPSES, Unit 1, safety valves are the loop seal tests on the Crosby 6M6 valve, Test No. 929, 1406, 1415, 1419. In all these tests (except Test No. 1415), the valve fluttered or chattered during loop seal discharge and stabilized when steam flow started. The valve opened within +4% of the design set pressure and closed with 5.1 to 9.4% blowdown. Up to 111% of I

rated flow was achieved at 3% accumulation with valve lift positions at 92 to 94% of rated lift. These tests demonstrated that the valve performed its function in spite of the initial chatter during loop seal discharge.

In Test 1419, the valve chattered on closing and the test was terminated after the valve was manually opened to stop the chatter. This result does not indicate a valve closing problem for the CPSES, Unit 1, safety valve since an identical test (Test 1415) had already demonstrated that the valve performed satisfactorily and exhibited no sign of <

instability. The closing chatter in Test 1419 may possibly be a result of Comanche Peak SSER 21 13 Appendix AA

l the repeated actuation of the valve in loop seal and water discharge tests.

As shown in Table 4.3.1 the 6M6 test valve was subjected to seventeen steam, water, and transition tests. In the first four or five tests, the valve fluttered and chattered during loop seal discharge but stabilized and closed successfully. After Test 913, there were four instances in which the test was terminated due to chattering on closing. Galled guiding surfaces and damaged internal parts were found during inspection and the damaged parts were refurbished or replaced before the next test started. The test results showed that the valve performed well after each repair, but the closing chatter recurred in the subsequent test. Test 1415 was performed immediately after valve maintenance and the valve performed stably. The next test (Test 1419) encountered chatter in closing even though it was a repeat of Test 1415 at similar fluid conditions. This suggests that inspection and maintenance are important to the continued operability of the valves. The Licensee should develop a formal procedure requiring that the safety valves be inspected after each actuation and the procedure should be incorporated into the plant operating procedures or licensing documents such as the plant technical specifications. I The blowdown in these tests (5.1 to 9.4%) were in excess of the 5%

value specified by the valve manufacturer and the ASME Code. Westinghouse performed an analysis, " Safety Valve Contingency Analysis in Support of the EPRI Safety / Relief Valve Testing Program--Volume 3: Westinghouse Systems,"

EPRI NP-2047-LD, October 1981, on the effects of increased blowdown and concluded that no adverse effects on plant safety occurred in that the reactor core remained covered. Therefore, the amount of increased blowdown which occurred in the Crosby 6M6 steam tests is considered acceptable.

As discussed in Section 4.2.2, the limiting FSAR transient resulting in liquid discharge is the main feedline break accident. Tests 931a and 931b with typical plant ring settings of (-71, -18) simulate the expected CPSES, Unit 1, feedwater line break conditions. Test 931a was a loop seal / steam / water transition test. The 6M6 valve initially opened, fluttered or chattered in a partial lift position during loop seal discharge, then Comanche Peak SSER 21 14 Appendix AA

. TABLE 4.3.1. EPRI TESTS ON CROSBY HB BP-86 6M6 SAFETY VALVE Leakage Sean Ring Pro Post No. Test No. Setting Test Type Stability igpml (gpm) 1 903 1 Steam Stable 0 0 Inspection / Repair 2 906a,b,c 1 L.S. Stable 0 0 3 908 1 L.S. f/c 0 0 4 910 L.S. f/c 0 0 Inspection / Repair 5 913 2 L.S. f/c 0 1.0 6 914a,b c 2 L.S. Transition Terminated 0 Large Inspection / Repair 7 917 3 L.S. f/c 0 0 8 920 3 L.S. Terminated 0 0 Inspection / Repair 9 923 3 L.S. f/c 0 10 926a,b c,d 3 Transition Stable 0.36 0.08 Inspection / Repair 11 929 4 L.S. f/c 0 0 12 931a,b 4 L.S. Transition c 0 0 13 932 4 Water Terminated 0 --

Inspection / Repair 14 1406 4 L.f. f/c. 0 0.63 Inspection / Repair 15 '1411 4 Steam Stable 0.76 0.37 Insegetion/ Repair l 16 1415 4 L.S. Stable 0 0 17 1416 4 L.S. Terminated 0 1.5 Inspection / Repair c--chatter f/c--flutter / chatter L.S.--loop seal Ring setting--four different sets of ring settings were tested. Actual ring positions not shown.

Terminated--Test terminated after valve was manually opened to stop chatter.

Comanche Peak SSER 21 15 Appendix AA

popped open, stabilized on steam, and closed with a 12.7% blowdown.

Test 931b was a saturated water test. The 6M6 valve opened on 6400F water, chattered, and then stabilized. The valve closed with 4.8%

blowdown. For these tests the valve opened within -1% and +3% of the set pressure. The maximum calculated surge rate at CPSES, Unit 1, during the feedline break transient is 1109.5 gpm. The 6M6 valve tested by EPRI passed 2355 gpm at 2415 psia and 641 0 F which is much higher than the predicted flow rate for CPSES, Unit 1. The above results demonstrate that the Crosby 6M6 safety valves would be adequate to perform the required water relief function.

The loads induced on the safety valves tested by EPRI exceed the loads for the CPSES, Unit 1, safety valves. The maximum moment tested for the 6M6 valve was during test 908 and was 298,750 in.-lb. The largest moment predicted for the safety valve inlet at CPSES, Unit 1, is 178,175 in.-lb.

This demonstrates functionability for the CPSES, Unit 1 safety valves.

4.3.2 Power Operated Relief Valves The EPRI tests on the Copes-Vulcan PORV with 316 S.S. stellited plug and 17-4 PH cage demonstrated that the valve opened and closed on demand in steam, water, and steam to water transition conditions. The opening and closing time were within the 2.0 second opening and closing time normally required for Westinghouse PWRs. The lowest steam flow rate observed in the tests was 232,000 lb/h which exceeded the rated flow of 210,000 lb/h for the CPSES, Unit 1, PORVs.

A bending moment of 43,000 in.-lb was induced in the inlet of the  ;

Copes-Vulcan PORV test valve per EPRI 64-CV-174-25. The largest moment predicted for the PORV inlet in the CPSES, Unit 1, valves is 28,708 in.-lb.

This demonstrates functionability of the CPSES, Unit 1, relief valves.

Comanche Peak SSER 21 16 Appendix AA l - _ _ _ - _ -

4.3.3 Electric Control Circuitry NUREG-0737 II.D.1 requires qualification of associated control circuitry as part of the safety / relief valve qualification. The Licensee addressed (Reference 14) the qualification of the PORV control circuitry required by NUREG-0737, Item II.D.1. The information contained a list of equipment and the abnormal events for which they have been qualified. This submittal included the tests required by NUREG-0737, Item II.D.1, thereby assuring the functionality of the control circuitry.

4.3.4 p0RV Block Valves The Westinghouse 3-inch Model 3GM88 block valves that are used in CPSES, Unit 1, are the same as those finally tested by EPRI (modified the same as the test valves) and the test valves opened and closed fully under the full range of operating conditions. Therefore, the plant valves are expected to function acceptably when required to do so.

The ,above discussion, demonstration,g that the valves operated satisfactorily, verifies that the part of Item 1 of Section 1.2 which requires conducting tests to qualify the valves and that part of Item 7 which requires the effect of discharge piping on operability be considered have been met provided that the Licensee documents formal procedure for the inspection of safety valves as discussed in Section 4.3.1. Also, the qualification testing of the control circuitry satisfies the requirements of Item 5 of Section 1.2.

4.4 piping and Support Evaluation This evaluation covers the piping and supports upstream and downstream of the safety valves and PORVs extending from the pressurizer nozzles to the pressurizer relief tank. The piping was designed for dead weight, internal pressure, thermal expansion, earthquake, and safety and rel'ief valve discharge conditions. The calculation of the time histories of hydraulic forces due to valve discharge, the method of structural analysis, and the i load combinations and stress evaluation are discussed below.

Comanche Peak SSER 21 17 Appendix AA

l-L 4.4.1 Thermal Hydraulic Analysis l

Pressurizer fluid conditions were selected for use in the thermal hydraulic analysis such that the calculated pipe discharge forces would I

bound the forces for any of the FSAR, HPI, and cold pressurization events, including the single failure that would maximize the forces on the valve.

In the analysis, the safety valve and PORV discharge transients were treated as two separate events, that is, the safety valves were assumed to actuate simultaneously with the PORVs closed, and the PORVs were assumed to actuate simultaneously with the safety valves closed. This approach is acceptable, since the safety valves and PORVs have different set points; they will not lift at the same time. A valve operating condition which is more likely to occur would be a discharge of the PORVs followed by discharge of the safety valves at their respective set points. Since the PORVs have a lower set point, they will open ahead of the safety valves. In this case, the PORV piping loads would be the same as those calculated from the simultaneous PORV actuation case above, but the safety valve discharge forces would be reduced due to the build-up of backpressure in the discharge piping resulting from the preceding PORV actuation. Therefore, this condition needs not to be analyzed.

The steam discharge transients are potentially the worst loading conditions for the safety valve and PORV piping. Both the safety valves and PORVs at Comanche Peak I have loop seals upstream of the valve inlets. When the safety valve or PORV actuates, the water slug in the loop seal driven by the high steam pressure imposes the highest hydrodynamic loads on the piping and supports. The piping loads due to water discharge are not expected to exceed the steam discharge conditions. The loop seal temperature is 3000F at the safety valve inlet and 6500F at the steam / water interface at the upstream end of the loop seal (Reference 17). The average temperature of the loop seal water would be in the range of 3500F to 4000F. The limiting event for water discharge through the safety valves is the feedline break accident. The water temperature was predicted to be 6080F to 6150F (Reference 7). This temperature is considerably higher than the Comanche Peak SSER 21 18 Appendix AA

loop seal temperature. Therefore, more flashing is expected in the high 3 temperature water discharge case and the hydrodynamic forces on the piping would be less severe than the water slug discharge condition.

For the PORVs, the steam discharge also represents the limiting l condition for the pipe loads. The PORV inlet piping has a cold loop seal with 1500F water (Reference 14). The thrust of the cold water slug under high steam pressure generates the highest piping loads of all steam and water discharge transients including the cold overpressurization events.

Therefore, the valve discharge conditions selected for the piping and support stress evaluations were the steam discharge conditions resulting from the simultaneous actuation of the safety valves and the simultaneous actuation of the PORVs respectively.

In the thermal hydraulic analysis, fluid conditions were assumed to bound all limiting transients discussed in Section 4.2. For the safety valve analysis, the initial pressure of the saturated steam upstream of the loop seals was assumed to be 2575 psia and the inifial downstream pressure was assumed to be 18 psia. The pressurizer conditions were held constant for the entire transient at 2575 psia and 1100 Btu /lb. The loop seal water temperature was assumed to be 300 0 F at the safety valve inlet. For the l PORV analysis, the initial upstream pressure of the saturated steam was assumed to be 2350 psia and the downstream pressure was assumed to be 18 psia. The pressurizer conditions were held constant for the entire transient at 2350 psia and 1162 Btu /lb. The temperature of the liquid upstream of the PORV was assumed to be a constant 1500F. The pressurizer pressure used in the analysis is lower than the maximum pressure of 2532 psia predicted by Westinghouse. This slight nonconservatism is more than compensated by assuming a constant loop seal temperature of 1500F, since the loop seal water at the steam-water interface would be at saturation and the temperature along the loop seal would be hotter than at the PORV inlet. The effect on calculated discharge loads of slightly lower upstream pressure is less than the the effect of lower loop seal temperature.

Comanche Peak SSER 21 19 Appendix AA

The thermal hydraulic analysis was performed using the Westinghouse computer code, ITCHVALVE. ITCHVALVE calculates the fluid parameters as a function of time. The unbalanced forces or wave forces in the piping segments are calculated from the fluid properties obtained from the ITCHVALVE analysis using another Westinghouse program, FORFUN. The forcing functions (' the piping system resulting from the fluid transients are obtained from these calculations.

The adequacy of the ITCHVALVE/FORFUN programs for the thermal-hydraulic analyses was verified by comparing the analytical and test results for thermal hydraulic loadings in safety valve discharge piping for two EPRI tests (Test Nos. 908 and 917). The detailed comparisons of the ITCHVALVE predicted force time-histories and the EPRI test results are presented in the submittal (Reference 3) and results of these comparisons are considered satisfactory.

The thermal hydraulic and stress ana'ysis of the Comanche Peak 1 safety valve and PORV piping and supports were performed by the Westinghouse Electric Co. as a consultant to the Licensee. The typical Westinghouse analysis for such piping systems has been fully reviewed in previous submittals for similar PWR plants such as the Diablo Canyon Units 1 and 2 (References 18 and 19). The method of analysis used by Westinghouse including the analysis assumptions, the structural modeling as well as the key parameters used in the computer inputs such as the node spacing, calculation time interval, valve opening time, etc. has been examined and fourd to be acceptable. The Comanche Peak 1 piping analysis followed the same method and procedure used in previous Westinghouse analyses. Therefore the Comanche Peak 1 analysis method is considered acceptable. The flow rate of the safety valve used in the analysis was 120% of the rated flow for the Crosby, 6M6, safety valves. The conservative factor contained in this flow rate is greater than needed to account for the 10% derating_for the safety j valve required by ASME Code and the allowance for uncertainties or errors. '

The PORV flow rate used in the analysis was 139% of the rated flw for the Copes-Vulcan valve, which again is amply conservative.

I l

Comanche Peak SSER 21 20 Appendix AA l 1

4.4.2 Stress Analysis The structural responses of the piping system due to the safety valve /PORV div.harge transients were calculated using the modal superposition method. The fluid force time histories generated from the FORFUN program in the thermal hydraulic analysis were used as forcing functions on the structural model . The Westinghouse series of structural analysis programs, namely WESTDYN7, FIXFM3 and WESTDYN2 were used to calculate the piping natural frequencies and mode shapes, the nodal displacements and the internal forces and support reactions. The FIXFM3 code calculates the displacements at the structural node points, using the forcing functions generated by FORFUN and the modal data from WESTDYN7. The structural displacements were then used by WESTDYN2 to compute the piping internal loads and support loads.

The WESTDYN series of structural programs mentioned above was previously reviewed and approved by the NRC (Reference 20). The adequacy of, these programs for piping discharge analysis was further verified by comparing the solutions generated by these programs with the EPRI safety valve test results (Reference 21).

l The piping stress analysis was performed in accordance with the requirements of the ASME Boiler and Pressure Vessel Code,Section III, Subsection NB, 1977 Edition, with addenda to and including Summer 1979. The j piping supports were analyzed in accordance with the ASME Boiler and Pressure Vessel Cnde Section III, Subsection NF, 1974 Edition, with addenda to and including Winter 1979. The load combination equations and stress  ;

limits used for the evaluation of the piping and support stresses are identical to those recommended by the piping subcommittee of the PWR Pressurizer safety and relief valve test program (Reference 10). The piping stress summaries presented by the Licensee (Reference 16) contain a comparison of the highest stresses in the piping with the applicable stress limits for the load combinations defined above. The piping stress results were reviewed and all the ; tresses were found to be within the applicable stress limits.

Comanche Peak SSER 21 21 Appendix AA i

I J

According to results of EPRI tests performed on the Crosby 6M6 safety valve, high frequency pressure oscillations of 170-260 Hz occurred in the piping upstream of the safety valve as a loop seal water slug passed through the valve. This raises a concern that these oscillations could potentially excite high frequency vibration modes in the inlet piping that could contribute to higher bending moments in the piping. This phenomenon was not accounted for in the structural analysis of the system. The piping between the pressurizer and safety valves in the EPRI tests, however, was composed l of 8-in. Schedule 160 and 6-in. Schedule XX while that at Comanche Peak 1, is 6-in. Schedule 160. Since the test piping did not sustain any discernible damage during pressure oscillations occurring in the tests, it is expected that the plant piping also would not incur damage during similar oscillations. Thus, a specific analysis for these pressure oscillations is not necessary for this plant.

i Reference 16 presented the worst case load / stress versus the allowables for representative piping supports. The results all showed that the load / stresses were within their respective allowablas.

The selection of a bounding case of the piping evaluation and the piping and support stress analysis demonstrate that the requirements of Item 3 and Item 8 of Section 1.2 outlined in this report have been met.

t l

Comanche Peak SSER 21 22 Appendix AA

1 li

5. EVALUATION

SUMMARY

The Licensee for CPSES, Unit 1, has provided an acceptable response to the requirements of NUREG-0737, and thereby reconfirmed that the General Design Criteria 14, 15, and 30 of Appendix A to 10 CFR 50 have been met. I The rationale .for this conclusion is given below.

The Licensee participated in the development and execution of an acceptable Relief and Safety Valve Test Program designed to qualify the operability of prototypical valves and to demonstrate that their operation would not invalidate the integrity of the associated equipment and piping.

The subsequent tests were successfully completed under operating conditions which by analysis bounded the most probable maximum forces expacted from anticipated design basis events. The generic test results and piping analyses showed that the valves tested functioned corr'ectly and safely for all relevant steam discharge events specified in the test program and that the pressure boundary component design criteria were'not exceeded. Analysis andreviewofthetestresultsandtheLicensee'sjudtficationsindicated I direct'applicabilityoftheprototypicalvalveandvalqeperformancesto.the in-plant valves and systems intended to be covered by the generic test program.

Thus, the requirements of Item II.D.1 of NUREG-0737 have been met (Items 1-8 in Paragraph 1.2) and, thereby demonstrate by testing and analysis, that the reactor primary coolant pressure boundary will have a low probabilityofabnormalleakage(GeneralD$signCriterionNo.14)andthat the reactor primary coolant pressure boundary and its associated components (piping, valves, and supports) have been designed with sufficient margin such that design conditions are not exceeded during relief / safety valve events (General Design Criterion No. 15).

Furthermore, the prototypical tests and the successful performance of the valves and associated components demonstrated that this equipment has d been constructed in accordance with high quality standards (General Design Criterion No. 30). I l

I Comanche Peak SSFR 21 23 Appendix AA i

REFERENCES

1. TMI Lessons Learned Task Force Status Report and Short-Term Recommendations, NUREG-0578, July 1979.

2. Clarification of TMI Action Plan Requirements, NUREG-0737, November 1980.

3. D. P. Hoffman, Consumers Power Co., letter to H. Denton, NRC,

" Transmittal of PWR Safety and Relief Valve Test Program Reports,"

September 30, 1982.

4. EPRI Plan for Performance Testing of PWR Safety and Relief Valves, July 1980.

5. EPRI PWR Safety and Relief Valve Test Program Valve Selection / Justification Report, EPRI NP-2292, January 1983.

6. EPRI PWR Safety and Relief Valve Test Program Test Condition Justification Report, EPRI NP-2460, January 1983.

l

7. Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves in Westinghouse-Designed Plants, EPRI NP-2296, January 1983.

8. EPRI PWR Safety and Relief Test Program Safety and Relief Valve Test Report, EPRI NP-2628-SR, December 1982.

9. R. C. Youngdahl, Consumers Power Co., letter to H. Denton, NRC,

" Submittal of PilR Valve Data Package," June 1, 1982.

10. EPRI PWR Safety and Relief Valve Test Program Guide for Application of Valve Test program Results to Plant-Specific Evaluations, Revision 2, Interim Report, July 1982.

11. P. Knight, Texas Utilities Services, Inc., memorandum for R. L. Tedesco, Assistant Director of Licensing, NRC, " Input to Comanche Peak SER," July 8, 1981.

12. H. C. Schmidt, Texas Utilities Services, Inc., letter to S. B. Burwell, Licensing Project Manager, NRC, " Comanche Peak Steam Electric Station Draft Response to Generic Letter No. 81-36," Log # TXX--3503, File #

10035, March 31, 1982. -

13. V. S. Noonan, NRC, letter to M. D. Spence, Texas Utilities Generating Co., " Performance Testing of Relief and Safety Valves," July 5, 1985.

l

14. W. G. Counsil, Texas Utilities Generating Co., letter to V. S. Noonan, Director, Comanche Peak Project, DL, NRC, " Comanche Peak Steam Electric i Station (CPSES), Docket Nos. 50-445 and 50-446, NUREG-0737, Item II.D.1--Performance Testing of Relief and Safety Valves,"

Log # TXX-4849, File # 10010, June 13, 1986.

Comanche Peak SSER 21 24 Appendix AA

15. C. Grimes, NRC, letter to W. G. Counsil, Texas Utilities Generating l Co., " Performance Testing of Relief and Safety Valves," March 27, 1987.

16. W. G. Counsil, Texas Utilities Generating Co., letter to V. S. Nuclear Regulatory Commission, " COMANCHE PEAK STEAM ELECTRIC STATION (CPSES),

DOCKET NOS. 50-445 and 50-446, REQUEST FOR ADDITIONAL INFORMATION -

NUREG-0737, ITEM II.D.1--PERFROMANCE TESTING 0F RELIEF AND SAFETY VALVES," Log # TXX-6398, File # 10010, April 15, 1987.

17. Amendment 40 to Comanche Peak FSAR, " Response to NRC Action Plan, II.D.1," May 10, 1983. .

18. L. C. Smith and T. L. Nazlett, Pressurizer Safety and Relief Line Evaluation Summary Report, Pacific Gas and Electric Corp., Diablo Canyon Unit 1 and Unit 2, Westinghouse AM-SSA-2534, November 1982.

19. J. O. Schuyler, Pacific Gas and Electric, letter to G. W. Knighton, NRC, " Response to NRC Request for Additional Information, NUREG-0737, Item II.D.1, Performance Testing,of Relief and Safety Valves,"

January 23, 1984.

20. R. L. Tedesco, NRC letter to T. M. Anderson, Westinghouse Electric Co. , '

" Acceptance for Referencing of Licensing Topical Report WCAP-8252, Revision 1," April 7, 1981.

l

21. L. C. Smith and T. M. Adams, " Comparison of Analytically Determined Structural Solutions with EPRI Safety Valve Test Results," 4th National Congress on Pressure Vessel and Piping Technology, Portland, Oregon, June 19-24, 1983, PVP-Volume 74, pp. 193-199.

1 l

Comanche Peak SSER 21 25 Appendix AA

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Supplement 21 to the Safety Evaluation Report related to the operation of the Comanche Peak Steam Electric Station, Units 1 and 2 (NUREG-0797), has been prepared by the Office of Nuclear Reactor Regulation of the U. S. Nuclear Regulatory Commission. The facility is located in Somervell County, Texas, approximately 40 miles southwest of Fort Worth, Texas.

This supplement reports the status of certain issues that had not been resolved at the time of publication of the Safety Evaluation Report and Supplements 1, 2, 3, 4, 6, and 12 to that report. This supplement also lists the new issues that have been identified since Supplement 12 was issued and includes the evaluations for licensing items resolved in this interim period.

Supplement 5 has not been issued. Supplements 7 through 11 were limited to the staff evaluation of- allegations investigated by the NRC Technical Review Team. Supplement 13 presented the staff's evaluation of the Comanche Peak Response Team (CPRT) Program Plan, which was formulated by the applicant to resolve various construction and design issues raised by sources external to the applicant. Supplements 14 through 20 presented the staff's evaluation of the applicant's Corrective Action Program and CPRT activities.

Items identified in Supplements 7, 8, 9, 10, 11, 13, 14, and 15 through 20 are not included in this supplement, except to the extent that they affect the applicant's Final Safety Analysis Report.

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