ML20036A069

Safety Evaluation Report Related to the Operation of Comanche Peak Steam Electric Station, Unit 2.Docket No. 50-446.(Texas Utilities Electric Company,Et Al.)

ML20036A069
Person / Time
Site:	Comanche Peak
Issue date:	04/30/1993
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0797, NUREG-0797-S27, NUREG-797, NUREG-797-S27, NUDOCS 9305100006
Download: ML20036A069 (49)

Text

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a .b-I NUREG-0797: Supplement No. 27 j Safety Evaluation Report; related to the operation of j Comanche. Peak Steam Electric. Station, l Unit 2 j ' Docket No. 50-446 l Texas Utilities Electric Company, et al.- q l i .i l .U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation April 1993 l t "afo h p ug .g 4, .E E <[. 4!! l 9305100006 930430 ) PDR ADOCK 05000446 I - E PDR l ,,,.,m_,, .m.,

F-1 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: 1. The NRC Public Document Room 2120 L Street, NW., Lower Level, Washington, DC f 20555 2. The Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, - Washington, DC 20013-7082 3. The National Technical information Service, Springfield, VA 22161 i' - Although the listing that follows represents the majority of documents cited in NRC publica- ' 7 tions, it is not intended to be exhaustive. l Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda; NRC bulletins, circulars, information notices, inspection and investigation notices; licensee event reports: l vendor reports and correspondence: Commission papers; and applicant and licensee docu. ments and correspondence. t The following documents in the NUREG series are available for purchase from the GPO Sales f Program: formal NRC staff and contractor reports, NRC-sponsored conference proceed-ings, international agreement reports, grant publications, and NRC booklets and brochures. Also available are regulatory guides, NRC regulations in the Code of. Federal Regulations, and Nuclear Regulatory Commission Issuances. l 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 pub!ic and special technical libraries include all open literature items, such as books, journal articles, and transactions. Federal Register notices, Federal t and State :egislation, and congressional reports can usually be obtained from these i libraries. Documents such as theses, dissertations, foreign reports and translations, and non-NRC { conference proceedings are available for purchase from the organization sponsoring the publication cited. Single copies of NRC draft reports are available free, to the extent of supply, upon written request to the Office of Administration, Distribution and Mail Services 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 Avenue, Bethesda, Maryland, for use by the public. Codes and standards are usually copyrighted and may_be purchased i l from the originating organization or, if they are American National Standards, from the. l American National Standards Institute,1430 Broadway, New York, NY 10018. I l I i ,m. ....,_e y-,g- ,e,. ,-w - + - 3,-,

NUREG-0797 Supplement No. 27 Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Unit 2 Docket No. 50446 Texas Utilities Electric Company, et al. U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation April 1993 l on aisu k-I 1 1 l l

ABSTRACT Supplement 27 to the Safety Evaluation Report related to the operation of the Comanche Peak Steam Electric Station (CPSES), Unit 2 (NUREG-0797), has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission (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, 12, 21, 22, 23, 24, 25, and 26 to that report were published. This supplement deals primarily with Unit 2 issues. Supplement 5 was cancelled. Supplements 7, 8, 9,10, and 11 were limited to the staff's 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 TV Electric (applicant). Supplements 14 through 19 presented the staff's evaluation of the CPSES Corrective Action Program: 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). Supplement 20 presented the staff's evaluation l of the CPRT implementation of its 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. 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 licensee's Final Safety Analysis Report. This twenty-seventh supplement, which is in support of the full-power license for Unit 2, provides updated information on the issues that had been considered previously, as well as the evaluation of issues that have arisen since the twenty-sixth supplement was issued. This evaluation addresses all of the issues necessary to support the issuance of a full-power license for Unit 2. Comanche Peak SSER 27 iii

TABLE OF CONTENTS i i U i ABSTRACT................................ iii 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT............ 1-1 1.1 Introduction........................ 1-1 i 1.7 Summary of Outstanding Issues................ 1-4 1.8 Confirmatory Issues..................... 1-4 1.9 License Conditions..................... 1-4 i 9 AUXILIARY SYSTEMS..................,...... 9-1 t 9.5 Other Auxiliary Systems................... 9-1 9.5.1 Fire Protection................... 9-1 13 CONDUCT OF OPERATIONS....................... 13-1 13.1 Organizational Structure and Qualifications......... 13-1 13.1.1 Management and Technical Resources......... 13-1 I 13.1.2 Operating Organization............... 13-2 i 13.3 Emergency Pl anni ng..................... 13-2 14 INITIAL TEST PROGRAM........................ 14-1 16 TECHNICAL SPECIFICATIONS..................... 16-1 i 17 QUALITY ASSURANCE......................... 17-1 17.2 Organization of the QA Program............... 17-1 l 22 TMI-2 REQUIREMENTS........................ 22-1 l APPENDICES Appendix A Listing of Correspondence Since Last SSER Appendix B Bibliography; Listing of Reports, Codes, Standards GLs Referenced in Body Appendix C TMI Issues Appendix D List of Principal Contributors l i 1 Comanche Peak SSER 27 v Y ~.... _, ~ -,

1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1 Introduction j The Nuclear Regulatory Commission (NRC) Safety Evaluation Report (SER), NUREG-0797, on the application of the Texas Utilities Generating Company (TUGCO$ ghe applicant) for a license to operate the Comanche Peak Steam Electric Station I (CPSES), Units 1 and 2, was issued in July 1981. Since then the following supplements have been issued: Supplement 1 (SSER 1) was issued in October 1981. It described the i resolution of a large portion of the outstanding and confirmatory issues identified in the SER. Sucolement 2 (SSER 2) was issued in January 1982. It included the report of the Advisory Committee on Reactor Safeguards (ACRS) to the NRC Chairman by letter dated November 17, 1981, which was appended as Appendix F. Applicant and staff responses to comments by the ACRS were also included. Supplement 3 (SSER 3) was issued in March 1983. It addressed outstanding and confirmatory issues resolved since SSER 2 was issued. The staff's evaluation of the applicant's emergency plans was also described. 7 Supplement 4 (SSER 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 CPSES. Supplement 5 (SSER 5) has been canceled. 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 (14-19) for each corrective action program design workscope. Therefore, the planned supplement was never issued. Suoolement 6 (SSER 6) was issued in November 1984. It addressed outstanding and confirmatory issues resolved since SSER 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 l Federal Reaulations (10 CFR Part 50) deleting the requirement for installing jet impingement shields for the Unit 1 primary coolant loop l piping at postulated break locations.

• 0n January 16, 1987, TUGC0 informed the NRC that it had adopted a new corporate signature and would be known as TU Electric (Texas Utilities Electric Company).

l Comanche Peak SSER 27 1-1 -m ..e>- <we-.r a, .n -ee4,-.n~.,e.m..v e w e--,,-,-wes- - -..

l l Supolement 7 (SSER 7) was issued in January 1985.. It was limited exclusively to the staff's evaluation of allegations investigated by the I NRC's Technical Review Team (TRT) pertaining to plant electrical / instrumentation systems and testing programs. i Supplement 8 (SSER 8) was issued in February 1985. It was limited ~ exclusively to Me staff's evaluation of allegations investigated by the TRT pertaining to the plant's civil / structural and other miscellaneous construction and plant-readiness testing items. Supplement 9 (SSER 9) was issued in March 1983. 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 (SSER 10) was issued in April 1985. It was limited exclusively to the staff's evaluation of allegations investigated by the TRT pertaining to the mechanical and piping areas. Sucolement 11 (SSER 11) was issued in May 1985. It was limited exclusively to the staff's evaluation of allegations investigated by the TRT' pertaining to quality assurance / quality control (QA/QC) practices in the design and construction of CPSES. i Supplement 12 (SSER 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) amendments 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 SSER 6 was issued. SSER 12 also listed several new issues that had been identified since SSER 6 was published and that were unresolved. Supplement 13 (SSER 13) was issued in May 1986. It presented the staff's i evaluation of the Comanche Peak Response Team (CPRT) Program Plan, which was formulated by the applicant to resolve various design and construction i 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. Sucolement 14 (SSER 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 and 16 (SSERs 15 and 16) were issued in July 1988; i Supplements 17 throuah 19 (SSERs 17-19) 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 (SSER 15); conduit supports (SSER 16); the mechanical, civil / structural, electrical, and instrumentation and controls workscopes, and systems portions of the heating, ventilation, and air conditioning (HVAC) system workscope (SSER 17); HVAC structural design (SSER 18); and equipment qualification (SSER 19). Comanche Peak SSER 27 1-2

Supplement 20 (SSER 20) was issued in November 1988. It presented the = staff's evaluation of the 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. Sucolement 21 (SSER 21) was issued in April 1989. It updated the SER further by providing the results of the staff's review of information that the applicant submitted by letter and in FSAR amendments. It addressed 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 SSER 12 was issued. Of note from an administrative standpoint, SSER 21 renumbered items appearing in Sections 1.7,1.8, and 1.9, and deleted all items that were previously resolved but listed in SSER 12. Supplement 22 (SSER 22) was issued in January 1990. It updated 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 addressed 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 SSER 21 was issued. Supplement 23 (SSER 23) was issued in February 1990 with the low-power operating license for CPSES Unit 1. It documented resolution of the remaining outstanding issues appearing in Section 1.7 of SSER 22. Supplement 24 (SSER 24) was issued with the full-power operating license for CPSES Unit 1. Confirmatory issues remaining at the time of license issuance, as well as proposed license conditions, were listed in Sections 1.8 and 1.9, respectively. Supplement 25 (SSER 25) was issued in September 1992. It updated the SER and subsequent SSERs, by presenting the results of the staff's review of information that the applicant submitted by letter and in FSAR amendments; specifically documenting reviews in support of the licensing of Unit 2. The staff review also addressed the translation of the Unit I and common area Corrective Action Program to Unit 2. Supplement 26 (SSER 26) was issued in February 1993. It updated the SER and subsequent SSERs by presenting the results of the staff's review of information that the applicant submitted by letter and in FSAR amendments. Significant issues contained in this SSER included TU Electric's fire barrier qualification testing program, preservice inspection and inservice testing programs and relief requests, an optimized fuel assembly review, and the plant's dual-unit station blackout review. This evaluation addressed all of the issues necessary to support the issuance of a low-power license for Unit 2. SSER 27 updates the SER and subsequent SSERs by presenting the results of the staff's review of information that TV Electric has submitted by letter. It addresses all of the issues necessary to support the issuance of a full-power license for Unit 2. Each section or appendix of this supplement is numbered and titled so that it corresponds to the section or appendix of the SER that has Comanche Peak SSER 27 1-3

l l been affected by the staff's additional evaluations and, except where specifically noted, does not replace the corresponding SER section or appendix. Appendix A is a continuation of the chronology of correspondence between the NRC and TV Electric that updates the correspondence listed in the SER and in SSERs 1 through 26. Appendix B includes references other than NRC documents and correspondence cited in this supplement. Appendix C contains information concerning the status of Three Mile Island (TMI) issues for CPSES Unit 2. Appendix D contains a list of principal contributors to this supplement. No changes were made to SER Appendices E, F, G, H, I, J, K, L, M, N, 0, P, Q, R, S, T, U, V, W, X, Y, Z, AA, BB, CC, DD, EE, or FF by this supplement. Copies of this supplement are available for public inspection at the NRC's Public Document Room, the Gelman Building, 2120 L Street, N.W., Washington, D.C. 20555; and at the University of Texas at Arlington Library, Government Publications / Maps, 701 South Cooper, P.O. Box 19447, Arlington, Texas 76019. The NRC Project Manager for Comanche Peak Steam Electric Station, Unit 2, is Brian E. Holian. Mr. Holian may be contacted by calling (301) 504-1334 or by writing to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555. 1.7 Summary of Outstandina Issues Section 1.7 of the SER, as supplemented, identified ne open issues at the time SSER 26 was issued. Those issues that were resolved in previous supplements were not listed in SSER 26. 1.8 Confirmatory Issues Section 1.8 of the SER, as supplemented, identified no confirmatory issues at the time SSER 26 was issued. 1.9 License Conditions In Section 1.9 of SSER 26, the staff listed three proposed license conditions. Those license conditions that were resolved in previous supplements were not listed in SSER 26. License conditions discussed in previous SSERs that were included in the Unit I license, and are similarly included in the Unit 2 license, follow: (1) The applicant shall continue to control mineral exploration within the exclusion area; that is, at distances beyond 2250 feet from safety-related structures per GDC 4, 10 CFR Part 50, Appendix A.

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

l l Comanche Peak SSER 27 1-4

l l (2) The applicant must implement and maintain in effect all provisions of the i approved fire protection program, as described in the Final Safety Analysis i Report (as amended) and as approved in the SER and its supplements, subject to the following provision: "The applicant may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely af fect the ability to achieve and. maintain safe shutdown in the event of a fire." (3) The applicant shall fully implement and maintain in effect all provisions of the physical security, guard training and qualification, and safeguards contingency plans, previously approved by the Commission, and all l' amendments made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain safeguards information protected under 10 CFR 73.21, are entitled: " Comanche Peak Steam Electric Station Physical t Security Plan" with revisions submitted through January 14, 1993; " Comanche i Peak Steam Electric Station Security Training and Qualification Plan" with revisions submitted through June 10, 1991; and " Comanche Peak Steam Electric Station Safeguards Contingency Plan" with revisions submitted } through December 1988. F F l l i l l I i i 3 i Comanche Peak SSER 27 1 l l l l A.-,_ .~._

l l 9 AUXILIARY SYSTEMS ~ l 9.5 Other Auxiliary Systems 9.5.1 Fire Protection In Supplement 21 to the SER (SSER 21), which was issued in April 1989, the staff j reviewed the Comanche Peak fire protection program as documented in the FSAR through Amendment 71 and as described in Revision 1 to the Fire Protection Report. In Supplement 26 to the SER (SSER 26), the staff documented a review of fire protection-related changes and modifications made to the FSAR through Amendment 87 and through Revision 6 to the Fire Protection Report. In SSER 26, the staff concluded that the fire protection program for Unit 2 adheres to the guidance in Appendix A to Branch Technical. Position (BTP) APCSB 9.5-1, and with Sections G, J, and 0 of Appendix R to 10 CFR Part 50. The following TU Electric commitments related to fire protection issues were documented in SSER 26: 36-inch wide fire barrier: Perform a confirmatory test or provide additional information addressing the staff's concerns (SSER 26, pp. 9-10 and 9-23). This issue is discussed in this SSER (below). The NRC staff witnessed the fire test and reviewed the preliminary test results submitted by TU Electric. The staff will review the final fire test report and will prepare a safety evaluation. This action is tracked by NRC TAC No. M85998. Amoacity deratina testina: Perform plant-specific testing (SSER 26, pp. = 9-20 and 9-32). i l This issue is discussed in this SSER (below). TV Electric provided preliminary results from their plant-specific ampacity derating tests. The l staff observed a portion of the testing, and will review the final ampacity test report and will prepare a safety evaluation. This action is tracked by ) NRC TAC No. M85999. " Box enclosure" barriers: Establish qualification (SSER 26, pp. 9-22 and 9-23). This issue is closed in this SSER (below) based on NRC staff's onsite inspection and review of the fire barrier upgrades and a review of the engineering documentation associated with these upgrades. Comanche Peak SSER 27 9-1

Compensatorv measures: Provide fire watches in accordance with the CPSES = fire protection plan until the barriers have cured for 30 days, and until the box enclosures are qualified (SSER 26, pp. 9-22 and 9-26). This issue is closed in this SSER (below) based on the preliminary results of - the 36-inch-wide cable tray fire test, with a seven-day cure time, and the completion of " box enclosure" upgrades. Alternative shuidown system desion enhancements: Implement design changes, = as necessary, to ensure that the torque and limit switches in the affected motor-operated valve operators are electrically connected downstream of the contacts located in the motor control center (SSER 26, pp. 9-36). This issue is not addressed below. In SSER 26, the staff reported TV Electric's commitment to perform the necessary design changes before startup i from the next refueling outage (first refueling outage for Unit 2 and the third refueling outage for Unit 1). This action is tracked by NRC TAC No. M86000. l In letters of February 26,1993 (TV Electric letter TXX-93101 to NRC), March 10, 1 1993 (TV Electric letter TXX-93125 to NRC), and. March 23, 1993 (TU Electric ~ letter TXX-93136 to NRC), TU Electric submitted updated information regarding the fire barrier commitments discussed above. In addition to the information that TU Electric submitted to the staff in the three letters, the staff visited the site on February 19 and 26, 1993. Through these letters and visits, the staff gathered the information to update the material contained in SSER 26. Four of the issues are discussed separately bel ow. i j The fifth issue is not addressed below as TU Electric's commitment to perform-the necessary changes is recorded in SSER 26. 36-INCH CABLE TRAY TEST Backaround In a letter of October 29, 1992, the staff stated that TU Electric's proposed acceptance criteria, as supplemented, were acceptable. In summary, the approved fire test acceptance criteria were: (1) External conduit, cable tray rail, and cable jacket temperatures should ( L not exceed a temperature rise of 250 'F (139 *K) plus ambient (using I thermocouple averaging), and no single thermocouple reading should exceed l 30 percent above the specified average temperature rise. I (2) The fire barrier should not burn through or develop any openings through which either the test specimen raceway or cables were visible. (3) If the temperature rise criteria were not satisfied, the cables should be inspected for visible damage. The following attributes constitute cable damage: Jacket swelling, splitting, discoloration, hardening, blistering, Comanche Peak SSER 27 9-2 - - - - ~ - ~

l cracking, or melting; conductor insulation exposed, degraded, or l discolored; shield exposed; or bare copper conductor exposed. j (4) If the fire barrier burned through during the fire exposure, or if a visual l cable inspection revealed any of the damage attributes listed above, then the barrier was considered to have deviated from the acceptance criteria. l i Use of the fire test results to qualify a deviating fire barrier would i require that cable functionality'be demonstrated. Cable functionality test methodology and criteria were specified in the staff's October 29, 1992, letter. In a letter of October 29, 1992, the staff concluded that TU Electric's acceptance criteria, as supplemented by the conditions stated in the October 29, 1992 letter, ensured that adequate cable and barrier tests would be performed and that satisfactory results from these tests would constitute an acceptable l l basis for qualifying the CPSES Unit 2 fire barriers. } l In a letter of February 1,1993 (TU Electric letter TXX-93076 to NRC), TU Electric committed to either perform a confirmatory test of a 36-inch cable i tray, participate in an industry testing program to resolve concerns over a 36-inch-wide barrier, or submit additional information that adequately addresses the staff's concerns. TV Electric committed to perform one of these actions by the end of the first refueling outage for Unit 2. l UDdate { In a letter of February 26,1993 (TV Electric letter TXX-93101 to NRC), TU i Electric committed to perform the confirmatory fire endurance test. A 36-inch i cable tray " straight run" configuration was constructed using licensee-proposed upgrades for the Unit 1 plant Thermo-Lag fire barrier configurations (stress skin reinforcement on joints instead of stress skin and stitching as used on Unit 2). The test configuration was built with the application of Thermo-Lag topcoat material approximately 72 hours following completion of the raceway i envelope. The fire test was conducted four days after the topcoat was applied. Circuit integrity was not monitored during the test. j l The 36" x 4" ladderback cable tray (straight run with 90 degree sweeping bends) was tested on March 4, 1993; the staff observed the test. The cable tray was l protected with 1/2" (nominal) Thermo-Lag panels with longitudinal, vertical, and i bottom joints reinforced with stress skin and trowel-grade material. TU Electric summarized the test data in a_ letter of March 10,1993 (TV Electric letter TXX-93125 to NRC). Temperatures were below the acceptance criteria-(which allows a 250 *F rise above ambient). The proper conduct of the fog hose stream test was observed. The hose stream test did not damage.the barrier; no fire barrier burn-through was noted. Post-fire cable visual inspections were j satisfactory. There were no signs of ccble damage. 4 Comanche Peak SSER 27 9-3

i l l l Texas Utilities Fire Barrier Testing for Comanche Peak Unit 2 (Conducted March 4, 1993, at Omega Point Laboratories) i Maximum Average temperatures in Thermocouple temperatures in *F

• F l

locations (Ambient - 68 'F) (Ambient - 68 *F) l Power cable 241 277 Control cable 210 224 Instrument cable 217 240 Front tray rail 244 285 Rear tray rail 247 292 Scheme 15 36" Cable Tray i These preliminary test results meet the acceptance criteria and are indicative of a satisfactory test, subject to staff review of the final fire test report. l This test was conducted in an identical method as the previous upgraded testing (documented in SSER 26), with the exception of a shorter material cure time and the absence of circuit integrity measurements. The cure time difference will be discussed below. The circuit integrity measurements were not taken for this test since TU Electric considered them unnecessary. The staff does not consider circuit integrity measurements an adequate test of cable functionality, and has determined that post-fire megger testing (described in the acceptance criteria as appropriate cable functionality testing) should be conducted as soon as possible following the test. Therefore, TV Electric's minor change to their test methodology (not performing circuit integrity measurements) is acceptable. This 36" wide cable tray test was performed by TV Electric to satisfy the SSER 26 commitment regarding a confirmatory test of the widest cable tray. The staff will review the final test report when it becomes available and document the results of its review in a safety evaluation report. Staff actions will be tracked by NRC TAC No. M85998. AMPACITY DERATING Backaround Cables enclosed in electrical raceways protected with fire barrier materials are derated because of the insulating effect of the fire barrier material. Other factors that affect ampacity derating include cable fill, cable loading, cable type, raceway construction, and ambient temperature. The National Electrical Code, Insulated Cable Engineers Association (ICEA) publications, and other industry standards provide general ampacity derating factors for open air installations, but do not include derating factors for fire barrier systems. Historically, ampacity derating factors for raceways enclosed with fire barrier material have been determined for specific installation configurations by Comanche Peak SSER 27 9-4

testing. In SSER 26, the staff discussed its concerns with inconsistent ampacity derating test data, but recognized that the ampacity derating concern is an aging issue rather than an immediate operability issue. In SSER 26, the staff (1) documented TU Electric's commitment to complete plant-specific ampacity derating testing by the end of the first refueling outage and (2) concluded that the use of TU Electric's interim ampacity derating factors is acceptable. Update After SSER 26 was issued, TU Electric conducted a series of ampacity derating tests for Thermo-Lag fire barrier configurations at Omega Point Laboratories (OPL) in San Antonio, Texas from March 3 through March 13, 1993. The NRC staff observed test preparation and testing from March 2 to 7, 1993. The first test group, conducted from March 2, 1993 to March 3, 1993, consisted of a 3/4"- diameter conduit with a single 3/C #10 AWG 600-volt copper cable and a 2"- diameter conduit with a single 3/C #6 AWG 600-volt copper cable. The second test group, conducted from March 5 to March 8, 1993, consisted of a 24" x 4" cable tray filled to a 2.95-inch depth with 3/C #6 AWG 600-volt copper cables and a free air drop (small) made of a single 3/C #6 AWG 600-volt copper cable. The final test group, conducted from March 10 to 14, 1993, consisted of a 5"- diameter conduit with four 1/C 750MCM 600-volt copper cable and a free air drop (large) made of three 1/C 750McM 600-volt copper cable. The ampacity derating factor test results are summarized below. The TV Electric ampacity derating test methodology followed the guidance detailed in the proposed standard IEEE-P848, " Procedure for the Determination of the Ampacity Derating of Fire Protected Cables," Draft 11, dated April 6, 1992, except for the following changes described further in TU Electric's ampacity test plan, revision 3, dated March 3, 1993: (1) Conduit / air drop test articles were selected to be consistent with CPSES installation including the enhanced Thermo-Lag configurations. (2) Test articles were supported by wood blocks during the performance of the tests. (3) Type T special accuracy thermocouples were used for the conduit / air drop test articles and for all ambient temperature measurements. Type K thermocouples were used for tray configurations, with directions to make adjustments, if necessary, for the difference in accuracy. (4) Baseline tests may be run before or after the ampacity derating test. (5) Three thermocouples were installed at each location for the conduit / air drop test articles. (6) Both the baseline and ampacity derating test shall utilize measured current normalized as outlined in ICEA P-46-426 for final conductor and ambient temperatures (that were not 90 "C and 40 "C, respectively). Comanche Peak SSER 27 9-5 l

[ Note: By letter of March 23, 1993 (TV Electric letter TXX-93136 to NRC), TU Electric referenced Revision 4 of their ampacity test plan. The staff's review of this latest revision will be included in the staff's review of the final test reports, as discussed below). In addition, the subject test plan supplemented elements of the Draft IEEE-P848 document in the following manner: Use a clamp-on ammeter with an accuracy of 1 percent to take the final current measurements. Base the data interpretation of the ampacity derating factor on the measured values irrespective of the published ICEA values in accordance with the TV Electric letter of February 26, 1993 (TV Electric letter TXX-93101 to NRC). The ampacity derating test procedure used for the test articles was performed in two steps, as follows: (1) An ampacity product (or derating) test was conducted with the Thermo-Lag material configured around the test article. (2) Then the baseline test was conducted on the instrumented article without the Thermo-Lag product. Each ampacity test was performed by raising the conductor temperature from ambient (i.e., 40 *C) to its rated temperature limit (i.e., 90 "C), allowing the test article to reach thermal equilibrium, and then measuring the final current or ampacity value for the test article. The ampacity derating factor was calculated as follows: Ampacity derating factor I, where: I, - ampacity value for product test 1, - ampacity value for baseline test TV Electric performed a series of calculations to establish the existing design margin for cable ampacity derating. These calculations were performed for the cables fed from the various switchgear, as follows: Calculation Cables Calculated excess amoacity marain

1. EE-CA-0008-3097 From 6.9 kV Cable tray - 40%

l Conduit - 40%

1. EE-CA-0008-3038 From 480 V Cable tray - 38%

Conduit - 23%

1. 2-EE-053 All other Cable tray - 40%

Conduit - 35% Comanche Peak SSER 27 9-6

Calculation Cables Calculated excess amoacity marain 016345-EE(B)-140 Air drops Cable tray - 39% Conduit - 35% TV Electric letters of March 10, 1993 (TV Electric letter TXX-93125 to NRC) and March 23,1993 (TV Electric letter TXX-93136 to NRC), supplied preliminary information regarding both TV Electric's calculated excess ampacity margin and the test result data for the plant-specific ampacity derating tests. Based on its testing, TU Electric is revising its design basis document to reflect the following derate factors: 11% for cables in conduits; and, 32% for cables in trays and air drops. The following table summarizes the preliminary test data, and provides the ampacity derate margin based on the effects of the fire barrier (calculated excess ampacity margin minus the actual test data): Amnacity derate Excess amoacity Raceway test value derate marain 3/4" conduit 9.1% 25.9% 2" conduit 6.5% 28.5% 5" conduit 10.7% 12.3% 24" cable tray 31.4% 6.6% Small air drop 23.0% 12.0% Large air drop 31.7% 3.3% The NRC staff finds that the preliminary ampacity test derate factor data provided by TV Electric are bounded by the calculated (design) ampacity margins. However, the NRC staff is still reviewing TV Electric's plant-specific ampacity derating program and test results. The NRC staff will complete its review of the plant-specific test program and results after TU Electric submits the final test reports (consistent with the schedule published in SSER 26). Staff actions can be tracked under NRC TAC No. M85999. "B0X ENCLOSURE" UPGRADES Backaround In a letter of January 19,1992 (TV Electric letter TXX-93038 to NRC), TV Electric submitted engineering report ER-ME-082, " Evaluation of Unit 2 Thermo-Lag Configurations," to the staff for review in order to (1) establish the design basis for the Thermo-Lag fire barriers installed at CPSES Unit 2 that were configured differently from the tested configurations and (2) provide reasonable assurance that these Thermo-Lag fire barrier configurations will provide sufficient fire resistance to ensure that at least one train of safe shutdown systems will remain free of fire damage. TV Electric's fire testing progre established the technical and installation attributes for most of the Thermo-Lag fire barrier configurations installed at CPSES Unit 2. TV Electric documented about 180 cases in which the application of Thermo-Lag fire barrier materials used to protect electrical raceways and structural steel varied from the tested configurations. The staff recognized that there are actual field conditions that cause the application of fire i Comanche Peak SSER 27 9-7

1 l 1 barrier assemblies to differ from the tested configurations. These cases may require the creation of a unique fire barrier design to address structural steel, other raceways, or mechanical equipment interferences. The staff also recognized that it was not feasible to qualify all aspects of the in-plant fire barriers through configuration-specific fire endurance testing. In Generic Letter 86-10, the staff provided guidance for performing engineering evaluations l of raceway fire barrier systems that differed from the tested configurations. TU Electric used this guidance to establish its fire barrier evaluation criteria i for configurations that differed from the tested configurations. The following summarizes TV Electric's criteria: the continuity of the fire l barrier material applied was consistent with the tested configuration; the effective thickness of the fire barrier material applied to the unique l configuration was consistent with the thickness of the fire barrier material l that was tested; the nature and effectiveness of the fire barrier support l assembly was consistent with the tested configurations; and the application and l end use of the fire barrier material was consistent with the tested configuration. In its engineering report, TV Electric evaluated the following: l l unique fire barrier configurations, minor protected commodity deviations, protruding and interfering item coverage deviations, and structural steel deviations. I c In reviewing TV Electric's engineering report, the staff sampled those unique l configurations where the fire barrier installations on safe shutdown raceways were constructed differently from those raceway fire barrier configurations tested by TV Electric's fire test program. The staff reviewed the engineering report and selected approximately 27 configurations for onsite review. The sample represented typical and unique configurations that varied from the tested l configurations. In SSER 26, the staff documented specific reviews of six representative configurations from this sample; for three of these, the staff requested additional actions. Configurations 1 and 3 represented " box-type" configurations, which the staff determined were not adequately justified in the engineering report. Specifically, two layers of Thermo-Lag material had been used for the qualification testing of junction box barriers, and the staff determined that designs similar to Configurations 1 and 3 would be more appropriately bounded by that type of construction. The staff considered l Configuration 2, consisting of two parallel horizontal cable trays (18 and 12-inches wide), acceptable subject to the confirmatory resolution of staff i concerns regarding the 36-inch wide cable tray fire barrier. l Update As discussed, the preliminary results of the 36-inch cable tray fire test appear to have been satisfactory. Subject to staff review of the final fire test report, this confirmatory test satisfies staff concerns regarding the l appropriate testing of the widest span cable trays. On the basis of the preliminary test results, the staff has reasonable assurance that Configuration 2 is acceptable. Any questions arising from the staff's review of the final test report for the 36-inch cable tray will be tracked by NRC TAC No. M85998. Comanche Peak SSER 27 9-8

Regarding the " box-type" configurations (Configurations 1 and 3), in letters of February 26, 1993 (TV Electric letter TXX-93101 to NRC), March 10, 1993 (TV Electric letter TXX-93125 to NRC), and March 23, 1993, (TV Electric letter TXX-93136 to NRC), TV Electric discussed the upgrades and documented their completion. TV Electric verified that the staff's concern with " box-type" configurations was limited to 13 plant configurations. In a letter of February 26, 1993, (TV Electric letter TXX-93101 to NRC), TU Electric committed to either increase the material thickness or rework the configurations in accordance with designs bounded by the previous fire barrier testing. l The NRC staff performed "walkdowns" of the subject configurations on February 19 ) and 26, 1993. On February 19, 1993, various elevations in the auxiliary l building were walked down to ensure that TU Electric had properly selected the configurations for upgrade. No additional examples of cable tray box-type enclosures which would necessitate an additional layer of Thermo-Lag were 1dentified. On February 26, 1993, walkdowns of all 13 upgrades were conducted l while work was in progress. TV Electric " Minor Modification Forms" 93-123 through 93-126 were reviewed; these documented the upgrades by building elevation - 810' auxiliary, 832' auxiliary, 810' safeguards, and 790' auxiliary, j respectively. The minor modification forms were verified to include l l engineering-basis discussions addressing the acceptability of the upgrades in regard to ampacity derating and the added weight on the supports. TV Electric l redesigned one of the 13 upgrades instead of adding an additional layer of Thermo-Lag material. The previous box design had covered an airdrop from two cable trays to through-wall sleeves. The redesign incorporated three layers of flexi-blanket Thermo-Lag material covering the air drops, and the installation of an elastomer fire stop material. The redesign is appropriately bounded by laboratory-tested airdrop configuration, Scheme 11-1. In a letter of March 10, 1993 (TV Electric letter TXX-93125 to NRC), TV Electric certified the completion of these upgrades. The NRC staff concludes from its review of the engineering documents and walkdown of the specific configurations that the barriers are properly bounded by acceptable test schemes and are, therefore, acceptable. COMPENSATORY MEASURES Backaround In a letter to the staff of October 5, 1991, the vendor stated that Thermo-Lag trowel-grade material takes about 30 days to reach its optimum properties. In a letter of January 19, 1993 (TV Electric letter TXX-93038 to NRC), TV Electric stated that it considered its Thermo-Lag fire barriers to be functional (capable of performing their design function) immediately after completion of the barrier installation and inspection. In a letter of January 25, 1993 (TV Electric letter TXX-93060 to NRC), TV Electric submitted additional information regarding cure time, stating that its vendor concurred with TV Electric's recommendation on cure time. Comanche Peak SSER 27 9-9

i TU Electric cured its fire test specimens for at least 30 days preceding the conduct of the fire endurance tests. The staff was concerned that Thermo-Lag fire barriers are not functional until they are either cured for 30 days in accordance with the vendor's original recommendation or until the installed barriers reflect the tested conditions. In a letter of January 28,1993 (TV Electric letter TXX-93061 to NRC) TV Electric committed to provide fire watches as compensatory measures in accordance with the CPSES fire protection plan for the Thermo-Lag fire barriers installed in areas that contain fire-safe shutdown conduits or cable trays until the barriers have cured for 30 days, and where box enclosures are located, until this issue is adequately resolved with the staff. i The use of fire watches is consistent with the compensatory measures implemented l by TV Electric for the CPSES Unit 1 Thermo-Lag fire barriers in response to NRC Bulletin 92-01, " Failure of Thermo-Lag 330 Fire Barrier System to Maintain Cabling in Wide Cable Trays and Small Conduits Free From Fire Damage," June 24, 1992. The staff concluded in SSER 26, therefore, that TV Electric's commitment was acceptable and ensures that an adequate level of fire protection is provided at CPSES Unit 2 until the Thermo-Lag fire barriers are cured (1) to reflect the condition of the fire test specimens and (2) the box enclosure issue is resolved. Update On March 4, 1993, TV Electric tested a 36-inch cable tray with a seven-day cure time (topcoat was applied after the assembly had cured for three days; four days later the fire test was conducted). The utility performed this test for two reasons: to satisfy their commitment to perform a confirmatory cable tray test bounding their widest tray, and to perform a test with a shorter cure time in order to demonstrate TV Electric's position that the configurations can be considered operable less than 30 days after completion of the installation and inspection. The 36-inch cable tray passed its confirmatory test, demonstrating the operability of Comanche Peak fire barrier designs with a seven-day cure time. Accordingly, TU Electric informed the staff that compensatory measures such as fire watches are no longer required for configurations that have exceeded a seven-day cure time. The staff considers TV Electric's position on cure time acceptable based on preliminary results of the 36-inch-wide cable tray fire test. A majority of the fire tests were conducted with a 30-day cure time in order to ensure that the moisture content in the Thermo-Lag material had reached equilibrium, thereby providing conservative qualification fire test results. Additionally, a test was conducted with a shorter cure time, and the fire barrier did not exhibit any seam separation, i On the basis of its findings from the fire barrier testing program, TU Electric has demonstrated that its installed configurations are bounded by test results; therefore, the staff concludes that compensatory measures are not required for fire barrier installations that exceed a seven-day cure time. Comanche Peak SSER 27 9-10

13 CONDUCT OF OPERATIONS 13.1 Oraanizational Structure and Qualifications I 13.1.1 Management and Technical Resources In SSER 26 the staff recorded TV Electric's commitment to submit organizational changes resulting from the change to two-unit operation in a future FSAR amendment. In a letter of February 26, 1993 (TV Electric letter TXX-93102 to NRC), TU Electric submitted an advanced change to the FSAR to revise the TV Electric corporate structure. The organization was revised to include the following divisions: Operations, Production, Bulk Power and Technical Support, and Finance and Corporate Support. The Production Division retained corporate responsibility for the design, construction, and operation of CPSES. Within the Production Division, the nuclear group, redesignated the Nuclear i Production Group (formerly the Nuclear Engineering and Operations Group), provides the design, engineering, construction, licensing, operation, and fuel l management support for CPSES. The Nuclear Production Group was reorganized into four organizations to better focus resources on operation of the dual-unit CPSES facility. These four organizations are Nuclear Operations, Nuclear Engineering and Support, Nuclear Overview, and Regulatory Affairs. All previously assigned responsibilities and duties have been reassigned to appropriate positions within the new management structure. Positions associated l with construction activities have been deleted as part of the transition from Unit 2 construction completion to dual-unit operation. The description of the 4 responsibilities for the Manager, Administrative Services has been removed because this position does not perform a safety function and, therefore, does not need to be in the FSAR. The staff finds this acceptable. The new l organizational structure is shown in revised FSAR Figure 13.1-2, included as an attachment to the February 26, 1993, letter. The changes to the corporate organization made by the licensee primarily reflect an organizational restructuring to focus resources on dual-unit operation of the CPSES facility. The new lines of management authority and communication have been clearly defined. Other changes made to the corporate organization reflect i changes in name, not in function. Therefore, they do not change the staff's previous conclusion that the corporate level management structure is acceptable. The staff concludes that the revised organization continues to meet the acceptance criteria of Section 13.1.1 of the Standard Review Plan (SRP) (NUREG-0800) for appropriate lines of authority, and is, therefore, acceptable. l Comanche Peak SSER 27 13-1 l l

a 13.1.2 Operating Organization in SSER 26 the staff recorded TV Electric's commitment to submit organizational changes resulting from the change to two-unit operation in a future FSAR amendment. In a letter of February 26, 1993 (TV Electric letter TXX-93102 to NRC), TV Electric submitted an advanced change to the FSAR to revise the TU Electric operating organization structure. The Nuclear Operations organization, under the direction of the Vice President of Nuclear Operations, is responsible t for plant operations and operating support. The Vice President of Nuclear Operations has assumed the duties of the Plant Manager and is responsible for the operation and maintenance of CPSES. Reporting to the Vice President of Nuclear Operations are the Manager, Operations; Manager, Maintenance; Manager, Plant Support; Manager, Work Control; and the Radiation Protection Manager. The Manager, Nuclear Training has been reassigned from the Nuclear Operations organization and now reports to the Director of Nuclear Overview. All previously assigned responsibilities and duties have been reassigned to appropriate positions within the new operating organization. The Chemistry Manager (formerly the Chemistry and Environmental Manager) now reports to the Manager, Operations. Environmental responsibilities have been reassigned under the Manager of Design / Support Engineering, in the Nuclear Engineering and Support organization. The new organizational structure is shown in revised FSAR Figure 13.1-3, included as an attachment to the February 26, 1993, letter. The changes to the operating organization made by the licensee primarily reflect an organizational restructuring to focus resources on dual-unit operation of the CPSES facility. The new lines of management authority and communication have been clearly defined. These changes do not affect the staff's previous conclusion that the operations organization is acceptable. The staff concludes that the revised organization continues to meet the acceptance criteria of Section 13.1.2 of the SRP for appropriate lines of authority, and is, therefore, acceptable. 13.3 Emeraency Plannina The Federal Emergency Management Agency (FEMA) evaluated the offsite radiological emergency response plans site-specific to Comanche Peak during an exercise conducted on November 19, 1991, and a remedial drill conducted on February 6, 1992. In a letter of June 24, 1992, FEMA stated that on the basis of these evaluations, the offsite radiological emergency response plans and preparedness site-specific to Comanche Peak are adequate to give reasonable assurance that appropriate measures can be taken offsite to protect the health and safety of the public in the event of a radiological emergency at the site. Before issuing low-power and full-power licenses, the NRC confirmed with FEMA that there were no offsite emergency preparedness issues that would potentially affect startup of CPSES Unit 2. The NRC conducted a special inspection on May 18-21, 1992 of TV Electric's emergency preparedness program as it related to the licensing of Unit 2. No areas were identified that would preclude the licensing of Unit 2. Additionally, TV Electric successfully passed the last annual exercise evaluated Comanche Peak SSER 27 13-2

by the NRC (conducted on November 18, 1992). TU Electric has responded with a corrective action plan to three onsite exercise weaknesses identified during this inspection. These issues are being tracked by the NRC's Region IV staff and will be evaluated during a future inspection. The staff concludes that the overall state of emergency preparedness at Comanche Peak is adequate to support dual-unit operations. Comanche Peak SSER 27 13-3

14 INITIAL TEST PROGRAM Pre-operational Test Deferral In SSER 26, the staff documented its review of TU Electric's preoperational test program changes for Unit 2 (TV Electric letters of December 23, 1992, TXX-92586 to NRC; January 8, 1993, TXX-93011 to NRC; and January 25, 1993, TXX-93051 to NRC). TV Electric proposed to defer certain preoperational tests until after I fuel load. The staff verified that TV Electric's letters contained commitments for completing the tests at the appropriate plant power levels or plant milestones. The staff determined that the schedule for performing the deferred l testing ensured that systems required to prevent, limit, or mitigate the consequences of postulated accidents would be tested before the systems would be required to be operable and ensured that the safety of the plant would not be dependent on the performance of untested systems, structures, and components. Therefore, the staff considered that TU Electric's justification for deferred testing and its subsequent schedule for conducting the tests was acceptable. In a letter of March 22,1993 (TV Electric letter TXX-93140 to NRC), TV submitted additional information regarding the status of several deferred preoperational tests which had been reviewed by the staff before the low-power license was issued. The additional information contained updated test l methodology, results, schedules and deletions regarding plant computer, plant i i communication system, pressurizer spray valve, and steam dump valve testing. l The pressurizer spray valve re-test was completed; however, a maintenance item i is being tracked to correct a slightly higher valve leak-by rate. The steam l dump valves will be retested " hot," but with the downstream block valves closed (similar to Unit I preoperational testing). Additionally, the schedule for testing the availability of the safety parameter display system and submitting j the test report was clarified. The staff reviewed the additional information and determined that the 1 conclusions reached in SSER 26 are still valid; that is, the systems will be i adequately tested before they will be required to be operable, and the safety of the plant will not be dependent on the performance of untested systems, structures, and components. Therefore, TV Electric's additional information, including test methodology and scheduler changes, regarding the deferred testing is acceptable, i Comanche Peak SSER 27 14-1

16 TECHNICAL SPECIFICATIONS The NRC issued the " Final Draft Combined Technical Specifications for Comanche Peak Unit I and Unit 2" to TU Electric on September 9, 1992. TU Electric certified on November 4, 1992 (TV Electric letter TXX-92536 to NRC), that the final draft accurately reflects the as-built plant and the Final Safety' Analysis Report. TU Electric also noted certain minor corrections. The staff discussed the corrections with TU Electric and appropriate changes were made to the Final Draft Technical Specifications (TS). The staff issued the Final Draft TS to TU Electric in a letter of January 22, 1993. Editorial corrections were discussed and TU Electric recertified the TS in a letter of January 30, 1993 (TU Electric letter TXX-93001 to NRC). The " Combined Comanche Peak Unit I and 2 Technical Specifications" were included as Appendix A to the low-power license issued on February 2, 1993. The same TS were reissued with the full-power license. f Comanche Peak SSER 27 16-1

17 QUALITY ASSURANCE The staff reviewed TU Electric's operations phase quality assurance (QA) program organization in SSER 22. In a letter dated February 26, 1993 (TU Electric letter TXX-93102 to NRC), TU Electric submitted an advance FSAR change to update the organizational structure, as discussed in Section 13.1 of this SER supplement. These changes resulted in some revisions to the description of the QA program organization described in Section 17.2 of the FSAR. The staff's reevaluation of the licensee's revised QA organization is presented below. 17.2 Oraanization of the OA Proaram The Group Vice President, Nuclear Production, is responsible for the overall management and operation of CPSES, including the establishment of company nuclear policies. The Group Vice President has the overall responsibility for establishing and executing the CPSES QA program for operations. The Group Vice President has assigned to the Vice President of Nuclear Engineering and Support the overall responsibility for engineering and support of CPSES, and for implementation of the QA program for the nuclear engineering and support function at CPSES. The Group Vice President has assigned to the Vice President of Nuclear Operations the overall responsibility for operating CPSES and for implementing the QA program for operations at CPSES. The Vice President of Nuclear Operations is responsible to the Group Vice President, Nuclear Production for operating activities at CPSES. Duties and responsibilities of the Vice President of Nuclear Operations include technical and administrative direction of the Manager, Operations; the Manager, Maintenance; the Radiation Protection Manager; the Manager, Work Control; the Manager, Plant Support; and the technical and administrative direction for implementing QA controls at nuclear plants operated by the licensee. The Vice President of Nuclear Engineering and Support is responsible to the Group Vice President, Nuclear Production for providing engineering related technical services in support of CPSES operations. Duties and responsibilities of the Vice President of Nuclear Engineering and Support include technical support to the nuclear operations organization, and assistance in the procurement of equipment, materials, and services for the operation, maintenance, and modification of CPSES. The Director of Nuclear Overview reports directly to the Group Vice President, Nuclear Production and is responsible to the Group Vice President for ensuring effective implementation of the QA program. This reporting relationship ensures that the Director of Nuclear Overview has sufficient authority, organizational freedom, and independence from undue influence from, or responsibility for, costs and schedules to effectively ensure implementation of and compliance with the CPSES operations QA requirements and controls. Comanche Peak SSER 27 17-1

The Director of Nuclear Overview communicates directly with the Nuclear Production Group supervisory and management personnel and with appropriate management levels in consultant and contractor QA organizations to identify-quality problems; initiate, recommend, or provide solutions; and to verify implementation of solutions to quality problems. The Director has the authority to "stop work" during the operations phase. Specific duties of the Director of Nuclear Overview include the direction of Nuclear Overview Department personnel; technical and administrative direction of the Manager, Operations Quality Control; Manager, QA; Manager, Independent Safety Engineering Group; Manager, Plant Analysis; and Manager, Nuclear Training; verification that procedures for the control of quality-related activities comply with QA requirements; verification of the implementation of the QA program within the Nuclear Production Group; verification that consultants, contractors, and suppliers providing quality-related items or services have established and implemented an adequate QA program; and membership or representation on the Operations Review Committee. The Nuclear Overview Department, under the Director of Nuclear Overview, functions to ensure effective implementation of the QA program. The department performs internal and external audits, surveillances, and inspections. The audits, surveillances, and inspections are performed by qualified individuals j other than those who performed or directly supervised the work. On the basis of its review and evaluation, the staff concludes that the-applicant's QA organization has (a) sufficient-independence from cost and schedule, (b) authority to effectively carry out the operations QA program, and (c) access to management at a level necessary to perform the QA functions. The y staff concludes that the applicant's description of the QA organization is in r compliance with applicable NRC regulations and is acceptable for the operation of CPSES. i l i l i f Comanche Peak SSER 27 17-2 -w, vn. -,e-g- - e, n w . =. -

22 THI-2 REQUIREMENTS After the accident at Three Mile Island (THI) Unit 2, the NRC staff developed NUREG-0660, "NRC Action Plan Developed as a Result of the THI-2 Accident," to provide a comprehensive and integrated plan to improve safety at power reactors. The Commission approved specific items from NUREG-0660 for implementation at reactors. NUREG-0737, " Clarification of TMI Action Plan Requirements," was issued in November 1980; this document included items approved by the Commission and additional information about schedules, applicability, method of implementation review, submittal dates, and clarification of technical positions. In Chapter 22 of the SER and its supplements, the staff discussed TMI issues relative to CPSES. In the table of TMI action plan issues, included in Appendix C of this SSER, the staff summarizes each THI item, including the SER (or supplement) that documents issue resolution and the date of inspection verification and associated inspection report number (if applicable). The staff adhered to the TMI action plan numbering scheme in SSER 24 (except where items were consolidated for inspection activity performed in accordance with Revision 2 of Temporary Instruction (TI) 2515/065, "TMI Action Plan Requirement Follow-up"). l Comanche Peak SSER 27 22-1

l APPENDIX A CONTINUATION OF CHRONOLOGICAL LIST OF CORRESPONDENCE This appendix continues the chronological listing of routine licensing correspondence, regarding Unit 2 and Unit 1/ Unit 2 common issues, between the U.S. Nuclear Regulatory Commission (NRC) staff and the applicant (Texas Utilities Electric Company) since Supplement 26 was issued. January 4, 1993 Summary of November 4,1992, meeting with applicant regarding pressurizer surge line leak-before-break analysis. January 6, 1993 Summary of December 17, 1992, meeting with applicant regarding fire protection issues. January 11, 1993 Letter to applicant transmitting environmental assessment for exemption from 10 CFR 70.24. January 11, 1993 Letter to applicant transmitting environmental assessment for exemption from 10 CFR Part 50, Appendix J, Section III.D.2(b)(ii). January 18, 1993 Letter from applicant transmitting information regarding augmented inservice testing for CVCS valves. January 19, 1993 Letter from applicant transmitting response to Generic Letter 92-08. January 19, 1993 Letter to applicant transmitting safety evaluation regarding topical report RXE-1-002, " Reactivity Anonciy Events Methodology". January 20, 1993 Letter from applicant transmitting final response for Unit 2 to NRC Bulletin 88-08. January 21, 1993 Letter from applicant transmitting information regarding ASME IST and Inservice Test Program relief request. January 22, 1993 Letter to applicant transmitting final draft version of combined technical specifications. January 25, 1993 Letter from applicant forwarding information regarding Thermo-Lag testing data and engineering evaluations. January 25, 1993 Letter from applicant forwarding information regarding scheduled completion of primary plant ventilation system and plant computer testing. Comanche Peak SSER 27 1 Appendix A

January 28, 1993 Letter from applicant forwarding supplemental response to Bulletin 88-08. January 28, 1993 Letter from applicant forwarding clarifying information regarding test scheme 1, conduit support modifications, and use of test scheme 9 results. January 29, 1993 Letter from applicant transmitting interim change request to preservice program plan. January 29, 1993 Letter from applicant transmitting information regarding HVAC design validation. January 29, 1993 Letter to applicant transmitting significant findings of the Operational Readiness Assessment Team Inspection. February 1, 1993 Letter from applicant transmitting response to concerns i regarding turnover process, fire seals for piping penetrations and containment spray system nozzle completion. j February 2, 1993 Letter to applicant transmitting Facility Operating License No. NPF-88 for Comanche Peak Unit 2. i r February 2, 1993 Memo to File regarding request for stay of issuance of the low power operating license. February 3, 1993 Board Notification 93-01 regarding new information regarding Comanche Peak Unit 2. February 3, 1993 Letter from licensee forwarding Revision 11 to Technical Requirements Manual. February 4, 1993 Letter from licensee forwarding Revision 1 to IST plan for pumps and valves first interval. February 5, 1993 Letter to licensee transmitting correction to Appendix B of Facility Operating License No. NPF-88. February 9, 1993 Letter to licensee transmitting correction to. Indemnity Agreement No. B-96. February 18, 1993 Letter from licensee forwarding overview of self-assessment plans for power operation above 5 percent and i above 50 percent. e February 19, 1993 Letter from licensee forwarding results of engineering h review of plant record to address issues in NRC Bulletin 90-01. February 22, 1993 Letter to licensee forwarding NUREG-1275, Volume 8,- " Operating Experience Feedback Report - Human Performance in Operating Events." Comanche Peak SSER 27 2 Appendix A 5 m

February 24, 1993 Letter from licensee forwarding summary of personnel monitoring ending December 31, 1992. February 25, 1993 Letter from licensee forwarding documentation of discussions with NRC regarding planned method of treating DNB penalties. February 25, 1993 Letter from licensee forwarding documentation of sensitivity study performed to evaluate effect of variations in core noding on calculated peak cladding temperature. February 26, 1993 Letter from licensee forwarding revisions to Sections 13.1.17.1 and 17.2 to FSAR reflecting organizational changes. February 26, 1993 Letter from licensee forwarding clarification on ampacity derating test and Thermo-Lag fire endurance test. March 2, 1993 Summary of January 21, 1993, meeting concerning Comanche Peak Steam Electric Station fire protection issues. March 9, 1993 Letter to licensee forwarding operation readiness assessment team inspection report. March 10, 1993 Letter from licensee forwarding an updated status of open issues in Section 9.5 of SSER 26 regarding preliminary fire endurance and ampacity test results. March 10, 1993 Letter to licensee forwarding clarification of staff safety evaluation on Topical Report RXE-91-002, " Reactivity Anomaly Events Methodology." l March 11, 1993 Letter from licensee forwarding RXE-93-003, CPSES Unit 2 Cycle 1 Core Operating Limits Report. March 11, 1993 Letter to licensee forwarding " Toxicological Evaluation of the Combustion Products from a Thermal Barrier Material Decomposed under Flaming and Nonflaming Conditions." March 17, 1993 Letter from licensee forwarding supplemental information to include Unit 2 in license amendment requests 92-05, 92-06, 92-07, and 92-08. March 22, 1993 Letter from licensee describing approach for analysis of la ge break LOCA with mixed cores delineated in Topical Report RXE-90-007. March 23, 1993 Letter from licensee submitting results of ampacity testing of upgraded Thermo-Lag installations and provides information on box configurations. Comanche Peak SSER 27 3 Appendix A

March 28, 1993 Letter from licensee transmitting certification for readiness for full power operating license. March 31, 1993 Letter to licensee transmitting documents filed by the staff with the Commission relating to Comanche Peak Unit 2 full-power licensing. Comanche Peak SSER 27 4 Appendix A

APPENDIX B BIBLIOGRAPHY Miscellaneous Code of Federal Reaulations. Title 10, " Energy," U.S. Government Printing Office, Washington, D.C., 1991. National Electrical Code, Insulated Cable Engineers Association (ICEA) publications. IEEE-P848, " Procedure for the Determination of the Ampacity Derating of Fire Protected Cables," Draft 11, April 6,1992. NRC Bulletins Bulletin 92-01, " Failure of Thermo-Lag 330 Fire Barrier System to Maintain Cabling in Wide Cable Trays and Small Conduits Free From Fire Damage," June 24, 1992. NRC Generic letters Generic Letter 86-10, " Implementation of Fire Protection Requirements," April 28, 1986. NRC Letters See Appendix A. NRC NUREG-Series Reports NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident," May 1980. NUREG-0737, " Clarification of TMI Action Plan Requirements," October 1980. NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," July 1981. Comanche Peak SSER 27 1 Appendix B

APPENDIX C TMI ACTION PLAN ISSUES

TMI ACTION PLAN ISSUES SER/SSER resolved ITEM [Verif. rpt. no., If Applicable] I.A.1.1.1 Shift Technical Advisor; SSER 1 & 23 On Duty [11/14/90; 50-446/90-40] I.A.1.1.2 Shift Technical Advisor SSERs 1 & 23 & TS 6.2.4 1.A.I.l.3 Shift Technical Advisor; SSERs 1 & 23 Training [11/14/90; 50-446/90-40] I.A.I.1.4 Shift Technical Advisor; SSERs 1 & 23 Long-Term Program I.A.I.2 Shift Supervisor; SSER 1 Administrative Duties [11/14/90; 50-446/90-40] I.A.l.3.1 Shift Manning; Overtime SSER 1 & TS 6.2.2.f [11/14/90; 50-446/90-40] I.A.l.3.2 Shift Manning; Minimum SSER 1 & TS T6.2-1 Shift Crew [11/14/90; 50-446/90-40] I.A.2.1.1 Immediate Upgrading of SER & letter dated 3/8/85 Operator and Senior Operator Training and Qualifications; SR0 Experience I.A.2.1.2 Immediate Upgrading of SER & letter dated 3/8/85 Operator and Senior Operator Training and Qualifications; Training I.A.2.1.3 Immediate Upgrading of SER & letter dated 3/8/85 Operator and Senior Operator Training and Qualifications; Facility Certification and Fitness of Applicants I.A.2.1.4 Immediate Upgrading of SER & letter dated 3/8/85 Operator and Senior Operator [11/14/90; 50-446/90-40] Training and Qualifications; Modify Training Comanche Peak SSER 27 1 Appendix C

TMI ACTION PLAN ISSUES (Continued) SER/SSER resolved ITEM [Verif. rpt. no., If Applicable] I.A.2.1.5 Immediate Upgrading of SER & letter dated 3/8/85 Operator and Senior Operator Training and Qualifications; Facility Certification I.A.2.3 Administration of Training SER & SSER 23 Programs for Licensed Operators I.A.3.1.1 Revised Scope and Criteria for SER Licensing Examination - Increase Scope I.A.3.1.2 Revised Scope and Criteria for SER Licensing Examination - Increase Passing Grade I.A.3.1.3.A Revised Scope and Criteria for SER Licensing Examination - With Simulators I.B.1.2 Evaluation of Organization and SSER 1 Management Improvements of Near-Team Operating License Applicants I.C.1.1 Procedures for Transients and SSERs 6, 12, & 22 Accidents, Short-Term; Small- [3/11/93; 50-446/92-60] Break LOCA I.C.I.2.A Procedures for Transients and SSERs 6, 12, & 22 Accidents, Short-Term; Inadequate Core Cooling; Reanalyze Guidelines I.C.1.2.8 Procedures for Transients and SSERs 6, 12, & 22 Accidents, Short-Term; [3/11/93; 50-446/92-60] Inadequate Core Cooling; Revise Procedures Comanche Peak SSER 27 2 Appendix C i l u__________ __j

g TMI ACTION PLAN ISSUES (Continued) SER/SSER resolved ITEM [Verif. rpt. no., If Applicable] I.C.I.3.A Procedures for Transients and SSERs 6, 12, & 22 Accidents, Short-Term; Transients and Accidents; Reanalyze Guidelines I.C.I.3.8 Procedures for Transients and SSERs 6, 12, & 22 Accidents; Short-Term; .[3/11/93; 50-446/92-60] Transients and Accidents I.C.2 Shift Relief and Turnover SSER 6 Procedures [11/14/90; 50-446/90-40] I.C.3 Shift Supervisor SSER 1 Responsibilities [11/14/90; 50-446/90-40] I.C.4 Control Room Access SSER 1 [11/14/90; 50-446/90-40] I.C.5 Procedures for Feedback of SSERs 6, & 23 Operating Experience to Plant [11/14/90; 50-446/90-40] Staff I.C.6 ' Procedures for Verification of SSER 1 Current Performance of [11/14/90; 50-446/90-40] Operating Activities I.C.7.1 NSSS Vendor Review of SER & SSER 23 Procedures; Low Power Test [11/14/.90; 50-446/90-40] Program I.C.7.2 NSSS Vendor Review of SER & SSER 23 Procedures; Low Power, Power [11/14/90; 50-446/90-40] Ascension and Emergency Procedures I.C.8 Pilot Monitoring of Selected SSER 6 Emergency Procedures for NT0L Applicants I.D.1 Control Room Design Reviews SSER 22 (Unit 1) SSER 26 (Unit 2) Comanche Peak SSER 27 3 Appendix C

7. I i i TMI ACTION PLAN ISSUES (Continued) SER/SSER resolved l j ITEM [Verif. rpt. no., If Applicable] I I.D.2.1 Plant Safety Parameter Display SSER 22 ) Console; Description I.D.2.2 Plant Safety Parameter Display SSER 22 (Unit 1) Console; Installed SSER 26 (Unit 2) [3/11/93; 50-446/92-60] I.D.2.3 Plant Safety Parameter Display SSER 22 (Unit 1) l Console; Fully Implemented .SSER 26 (Unit 2) [3/11/93; 50-446/92-60, i final verification i tracked as insp. item] l l I.G.I.1 Training During Low-Power SSER 6 l l Testing; Proposed Tests I.G.I.2 Training During Low-Waer 'SSER 6 i Testing; Submit Analysis and i Procedures I.G.I.3 Training During Low-Power SSER 6 Testing; Training and Results ] 11.B.1.1 Reactor Coolant System Vents; SSER 6 Design II.B.I.2 Reactor Coolant System Vents SSER 6 Install [3/11/93; 50-446/92-60] j 11.B.1.3 Reactor Coolant System Vents; SSER 6 & TS 3.6.1.7 Procedures [3/11/93; 50-446/92-60] II.B.2.1 Plant Shielding to Provide SSERs 2 & 22 Access to Vital Areas and Protect Safety Equipment for Post-Accident Operation; Design l II.B.2.2 Plant Shielding to Provide SSERs 2 & 22 Access to Vital Areas and [Incorp. in II.B.2.3] Protect Safety Equipment for Post-Accident Operation; Corrective Actions Comanche Peak SSER 27 4 Appendix C

m.. m- .m m l TMI ACTION PLAN ISSUES (Continued) SER/SSER resolved ITEM [Verif. rpt. no., If Applicable] II.B.2.3 Plant Shielding to Provide SSERs 2 & 22 Access to Vital Areas and -[8/22/91; 50-446/91-21] l Protect Safety Equipment for [l/27/93; 50-446/93-05] Post-Accident Operation; Modifications II.B.2.4 Superseded by 10 CFR 50.49 [ i II.B.3.1 Post-Accident Sampling; SSERs 6, 22, & 23 Interim System [11/14/90; 50-446/90-40] i II.B.3.2 Post-Accident Sampling; SSERs 6 & 22 Corrective Actions [Incorp. in II.B.3.4] II.B.3.3 Post-Accident Sampling; SSERs 6 & 22 Procedures (3/11/93; 50-446/92-60] j II.B.3.4 Post-Accident Sampling; Plant SSERs 6, 22, & 23 Modifications [3/11/93; 50-446/92-60] II.B.4.1 Training for Mitigating Core SER & SSER 23 Damage; Develop Training II.B.4.2.A Training for Mitigating Core SER & SSER 23 Damage; Initial -[11/14/90; 50-446/90-40] II.B.4.2.B Training for Mitigating Core SER & SSER 23 i Damage; Complete [11/14/90; 50-446/90-40] II.D.l.1 Relief and Safety Valve SSER 21 (Unit 1) I Testing Requirements; Submit SSER 26 (Unit 2) i II.D.I.2.B Relief and Safety Valve SSER 21 (Unit 1) Testing Requirements; Plant-SSER 26 (Unit 2) Specific Report II.D.1.3 Relief and Safety Valve SSER 21 (Unit 1) Testing Requirements; Block SSER 26 (Unit 2) Valve Testing II.D.3.1 Valve Position Indication; SER Install Direct Indicators of [3/11/93; 50-446/92-60] Valve Position Comanche Peak SSER 27 5 Appendix C m-.. ---r m e + -c-yFm W w +r-

TMI ACTION PLAN ISSUES (Continued) SER/SSER resolved ITEM [Verif. rpt. no., If Applicable] II.D.3.2 Valve Position Indication; SER & TS 3.4.4 & 4.0.5 Technical Specifications I II.E.1.1.1 Auxiliary Feedwater System; SER & SSER 21 Analysis II.E.1.1.2 Auxiliary Feedwater System SER & SSER 21 Evaluation; Short-Term [3/11/93; 50-446/92-60, Modifications Item II.E.1.2] II.E.1.1.3 Auxiliary Feedwater System SER & SSER 21 Evaluation; Lor.g-Term [3/11/93; 50-446/92-60, Modifications Item II.E.1.3] II.E.1.2.1.A Auxiliary Feedwater System SSER 21 Initiation and Flow; Control [4/24/89; 50-446/89-17] Grade II.E.1.2.1.8 Auxiliary Feedwater System SER Initiation and Flow; Safety [4/24/89; 50-446/89-17] Grade II.E.1.2.2.A Auxiliary Feedwater System SER Flow Indication; Control Grade [4/24/89; 50-446/89-17] II.E.1.2.2.B Auxiliary Feedwater System SER & TS 3.7.1.2 Flow Indication; LL Cat A Technical Specifications II.E.1.2.2.C Auxiliary Feedwater System SER Flow Indication; Safety Grade [3/11/93; 50-446/92-60] II.E.3.1.1 Emergency Power for SER & SSER 22 Pressurizer Heaters; Upgrade [3/11/93; 50-446/92-60] Power Supply II.E.3.1.2 Emergency Power for SER & TS 3.4.3 Pressurizer Heaters; Technical Specifications II!:E.4.2.1-4 Containment Isolation SSER 23 Dependability; Diverse [4/24/89; 50-446/89-17] Isolation Comanche Peak SSER 27 6 Appendix C

~ ~ l l i l TMI ACTION PLAN ISSUES (Continued) i i SER/SSER resolved ITEM [Verif. rpt. no., If Applicable] t II.E.4.2.5.A Containment Isolation SSER 22 l Dependability; Containment Pressure Setpoint; Specify Pressure II.E.4.2.5.8 Containment Isolation SSER 22 Dependability; Containment [8/21/91; 50-446/91-46] Pressure II.E.4.2.6 Containment Isolation SSER 23 i Dependability; Containment [12/31/92; 50-446/92-51] Purge Valves II.E.4.2.7 Containment Isolation SSER 23 & TS.3.6.1.7 i Dependability; Radiation [4/24/89; 50-446/89-17]- l Signal on Purge Valves II.E.4.2.8 Containment Isolation SSER 23 & TS T-3.3-4.2 l Dependability; Technical Specifications II.F.1.1 Accident Monitoring; SSER 3 Procedures [ Refer to II.F.1 items below] II.F.1.2.a Accident Monitoring; Noble Gas SSER 3 & TS T-3.3-4.1.b Monitor [4/24/89; 50-446/89-17; (interim) [9/20/89; 50-446/89-67 (long term); i 3/11/93; 50-446/92-60] II.F.1.2.b Accident Monitoring SSER 3 & TS T-3.3-4.1.a Particulate Sampling [5/18/89; 50-446/89-24 (long term); 2/1/93; 50-446/92-54] II.F.1.2.c Accident Monitoring; SSER 3 & TS T-3.3-6.10 Containment High-Range [3/11/93; 50-446/92-60] Monitors II.F.1.2.d Accident Monitoring; SSER 3 & TS T-3.3-6.1 Containment Pressure [3/11/93; 50-446/92-60] Comanche Peak SSER 27 7 Appendix C we v y-r--Aw mv-.m g-wpu g- +v=> -TF+-*

TMI ACTION PLAN ISSUES (Continued) SER/SSER resolved ITEM [Verif. rpt. no., If Applicable] 1 i II.F.1.2.e Accident Monitoring; SER & TS T-3.3-6.8 Containment Water Level [3/11/93; 50-446/92-60] 1 i l II.F.1.2.f Accident Monitoring; SER & TS T-3.3-7 & l Containment Hydrogen 3.6.4.1 l [3/11/93; 50-446/92-60] II.F.2.1 Instrumentation for Detection SSERs 6, 21, & 23 of Inadequate Core Cooling; [Incorp. in II.F.2.2] Procedure j II.F.2.2 Instrumentation for Detection SSERs 6, 21, 23 of Inadequate Core Cooling; [3/11/93; 50-446/92-60] j Subcool Meter; Install II.F.2.4 Instrumentation for Detection SSER 6, 21, & 23 & TS T-of Inadequate Core Cooling; 3.3-6 i Additional Instruments [3/11/93; 50-446/92-60] ) II.G.1.1 Power Supply for Pressurizer SER Relief Block Valves and Level [3/11/93; 50-446/92-60] Indication; Upgrade II.G.I.2 Power Supply for Pressurizer SER & TS 3.4.4 Relief Block Valves and Level Indication; Technical Specifications II.K.l.5 Measures to Mitigate Small-SER Break LOCA and Loss-of-Feedwater Accidents; IE Bulletins; Review ESF Valves II.K.1.10 Measures to Mitigate Small-SER Break LOCA and Loss-of-Feedwater Accidents; IE Bulletins; Operability Status II.K.l.17 Measures to Mitigate Small-SER Break LOCA and Loss-of-Feedwater Accidents; IE Bulletins; Trip per Pressurizer low Level Comanche Peak SSER 27 8 Appendix C

TMI ACTION PLAN ISSUES (Continued) SER/SSER resolved ITEM [Verif. rpt. no., l If Applicable] II.K.2.13 Thermal-Mechanical Report on SSER 6 Effect of HPI on Vessel Integrity for Small-Break LOCA with no AFW II.K.2.17 Analysis of Potential Voiding SSER 6 in RCS During Anticipated l Transients j II.K.3.1.A Automatic PORV Isolation SSER 6 System; Design II.K.3.1.B Automatic PORV Isolation SSER 6 System; Test / Install [5/18/89; 50-446/89-24] II.K.3.10 Anticipatory Trip H: Power SSER 25 l [5/18/89; 50-446/89-24] II.K.3.11 Justification for Use of SSER 6 Certain PORVs II.K.3.12.A Confirm Existence of SER & SSER 22 Anticipatory Trip Upon Turbine [5/18/89; 50-446/89-24] Trip; Proposed Modifications II.K.3.12.8 Confirm Existence of SER & TS T-3.3.1-16 Anticipatory Trip Upon Turbine [5/18/89; 50-446/89-24] Trip; Modify II.K.3.17 Report on Outage of ECCS SER II.K.3.2 Report on Overall Safety PORV SER Isolation System II.K.3.25.A Effect of Loss of AC Power on SER Pump Seals; Proposed ) Modifications l II.K.3.25.B Effect of Loss of AC Power on SER Pump Seals; Modifications [5/18/89; 50-446/89-24] Comanche Peak SSER 27 9 Appendix C

THI ACTION PLAN ISSUES (Continued) SER/SSER resolved ITEM [Verif rpt. no., If Applicable] II.K.3.3 Report Safety and Relief Valve SER & TS 6.9.1.2 & Failures Promptly and 6.9.1.5 Challenges Annually II.K.3.30.A Schedule for Outline of Small-SSERs 6 & 12 Break LOCA Model 1 II.K.3.30.B Small-Break LOCA Model; SSERs 6 & 12 Justification II.K.3.30.C Small-Break LOCA Model; New SSERs 6 & 12 Analysis II.K.3.31 Plant-Specific Calculations to SSERs 6, 12, & 21 Show Compliance with 10 CFR 50.46 II.K.3.5.A Automatic Trips of Reactor SSER 22 Coolant Pumps; Proposed Modifications II.K.3.5.B Automatic Trips of Reactor SSER 22 Coolant Pumps; Modifications [9/6/91; 50-446/91-38] ) l II.K.3.9 Proportional Integral SER & SSER 22 Derivative Controller [12/31/92; 50-446/92-51] Modification III.A.l.1 Improve Emergency Preparedness SSERs 6, 12, 22, & 24 III.A.I.2.1 Upgrade Emergency Support SSERs 3 & 22 Facilities; Interim TSC, OSC, [11/14/90; 50-446/90-40] and E0F III.A.1.2.2 Upgrade Emergency Support SSERs 3 & 22 Facilities; Design (Superseded by MPAs F063, F064 and F065) III.A.I.2.3 Upgrade Emergency Support SSERs 3 & 22 Facilities; Modifications [11/14/90; 50-446/90-40] (Superseded by MPAs F063, F064 and F065) Comanche Peak SSER 27 10 Appendix C

1 1 TMI ACTION PLAN ISSUES (Continued) SER/SSER resolved ITEM [Verif. rpt. no., If Applicable] III.A.2.1 Upgrade Preparedness; SSERs 3 & 6 (App. G., Emergency Plans Sec. 4) III.A.2.2 Upgrade Preparedness; SSERs 3 & 6 (App. G., Meteorological Data Sec. 4) III.A.2.3 Upgrade Preparedness; SSERs 3 & 6 (App. G., Implement Plans Sec. 4) III.D.1.1.1 Integrity of Systems Outside SSERs 4, 22, & 23 Containment; Leak Reduction [10/21/91; 50-446/91-46] III.D.I.1.2 Integrity of Systems Outside SSERs 4 & 23 & TS 6.8.3 l Containment; Technical Specifications Ill.D.3.3.1 Improved Plant Iodine SER & SSERs 6 & 22 Instrumentation Under Accident [4/1/91; 50-446/91-07] Conditions; Determine Presence of Radiciodine III.D.3.3.2 Improved Plant Iodine SER & SSERs 6 & 22 Instrumentation Under Accident [4/1/91; 50-446/91-07] Conditions; Modification to Accurately Measure Iodine III.D.3.4.1 Control Room Habitability; SER Review [11/14/90; 50-446/90-40] III.D.3.4.2 Control Room Habitability; SER Schedule Modifications [11/14/90; 50-446/90-40] III.D.3.4.3 Control Room Habitability; SER & TS T-3.3-4 & 3.7.7 Implement Modifications [11/14/90; 50-446/90-40] Comanche Peak SSER 27 11 Appendix C

t APPENDIX D LIST OF PRINCIPAL CONTRIBUTORS Contributor Oraanization E. Baker Office of Nuclear Reactor Regulation Project Directorate IV-2 D. Graves Senior Resident Inspector Region IV B. Holian Office of Nuclear Reactor Regulation Project Directorate IV-2 R. Jenkins Office of Nuclear Reactor Regulation Electrical Engineering Branch P. Madden Office of Nuclear Reactor Regulation Plant Systems Branch I. Miller Office of Nuclear Reactor Regulation Plant Systems Branch E. Peyton Office of Nuclear Reactor Regulation Project Directorate IV-2 R. Schaaf Office of Nuclear Reactor Regulation Project Directorate IV-2 D. Skay Office of Nuclear Reactor Regulation Project Directorate IV-2 S. West Office of Nuclear Reactor Regulation Plant Systems Branch i l l l Comanche Peak SSER 27 1 Appendix D

u.s. NUCLEAR REGULATORY COMMISSION 1. T E,R NRgFoau sas ~ ~ ~"~ E2E22E BIBLIOGRAPHIC DATA SHEET NUREG-0797 (See instructions on the reverse) Supplement No. 27

2. TITLE AND SUBMLE Safety Evaluation Report related to the Operation 3.

DATE REPORT PUBLISHED of Comanche Peak Steam Electric Station, Unit 2 j uoe veas April 1993

4. F IN OR GR ANT NUMBE R

5. AUTHOR (S)

6. TYPE OF REPORT Regulatory

7. Pt RIOD COVER ED Isnctuseve paresi February 1993 March 1993 B. PE RF ORMING ORG ANIZAT10N - N AME AND ADDR ESS tar Nac. provier Omwon. ortwo or Region. v.s. Nucarar Reevierary commossoon. ano martma ndereu. ts contractor. cmerce name end meetine nedreasJ Division of Reactor Projects, III/IV/V Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

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10. SUPPLEMENTARY NOTES Docket No. 50-446

11. ABSTR ACT (200 eens or em>

Supplement No. 27 to the Safety Evaluation Report related to the operation of the Comanche Peak Steam Electric Station (CPSES), Unit 2, has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission (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, 12, 21, 22, 23, 24, 25, and 26 to that report were published. This supplement deals primarily with Unit 2 issues. )

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