ML20035D439
| ML20035D439 | |
| Person / Time | |
|---|---|
| Issue date: | 04/06/1993 |
| From: | NRC COMMISSION (OCM) |
| To: | |
| References | |
| REF-10CFR9.7 NUDOCS 9304130214 | |
| Download: ML20035D439 (8) | |
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i UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMIS SION I 3l AFFIRMATION / DISCUSSION AND VOTE LOCition ROCKVILLE, MARYLAND r
bdI6 APRIL 6, 1993 23963 4 PAGES s
I NEALR.GROSSANDC0.,INC.
COURT REPORTERS AND TRAWSCRIBERS 1323 Rhode Island Avenue, Northwest Washington, D.C.
20005 (202) 234-4433 2
DISCLAIMER This is an unofficial transcript of a meeting of the United States Nuclear Regulatory Commission held on April 6, 1993.
in the Commission's office at One White Flint North, Rockville, Maryland.
The meeting was open to public attendance and observation.
This transcript has not been reviewed, corrected or edited, and it may contain inaccuracies.
The transcript is intended solely for general informational purposes.
As provided by 10 CFR 9.103, it is not part of the formal or informal record of decision of the matters discussed.
Expressions of opinion in this transcript do not necessarily reflect final determination or beliefs.
No pleading or other paper may be filed with the Commission in any proceeding as the result of, or addressed to, any statement or argument contained herein, except as the Commission may authorite.
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HEAL R. GROS $
COURT RtPoRTERS AND TRANSCRtttRS 1323 RHoDt ISLAND AYtHUt, N.W.
(202) 234-4433 WASHtNGioN. D.C.
20005
' (202) 232-6600
1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION AFFIRMATION / DISCUSSION AND VOTE PUBLIC MEETING l
Nuclear Regulatory Commission One White Flint North Rockville, Maryland Tuesday, April 6,
1993 The Commission met in open
- session, j
pursuant to notice, at 12:04 p.m.,
Ivan
- Selin, l
Chairman, presiding.
l COMMISSIONERS PRESENT:
IVAN SELIN, Chairman of the Commission
)
KENNETH C.
ROGERS, Commissioner JAMES R.
CURTISS, Commissioner FORREST J.
REMICK, Commissioner Is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVENUE, N W.
(202) 234 4433 WASHINGTON. D C. 20005 (202) 2344433
2 i
STAFF SEATED AT THE COMMISSION TABLE:
SAMUEL J.
CHILK, Secretary WILLIAM C.
PARLER, General Counsel i
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l NEAL R. GROSS l
COURT REPORTERS AND TRANSCRIBERS I
j 1323 RHODE ISLAND AVENUE. N W.
l (202) 234-4433 WASHINGTON. D C. 20005 (202) 23444?1
3 1
P-R-O-C-E-E-D-I-N-G-S 2
12:04 p.m.
3 CHAIRMAN SELIN:
Good afternoon, ladies 4
and gentlemen.
5 This is an affirmation session and we have 6
two items to come before us.
7 Mr.
- Chilk, would you please lead us 8
through the session?
9 SECRETARY CHILK:
The first item, Mr.
10 Chairman, is SECY-93-083.
It's a Motion for Stay of 11 Full Power License for Comanche Peak Unit 2 and 12 Petition to Intervene in Comanche Peak Proceeding.
13 The Commission is being asked to act on an 14 order that responds to two motions to stay the 15 issuance of the full power licence by B. Irene Orr and 16 D.I.
Orr and a petition to intervene'by R. Micky Dow 17 and Sandra Long Dow.
18 All Commissioners have approved an order 19 which denies both a stay motion and a petition to 20 intervene.
Commissioner de Plangue, although not 21 present, has indicated her approval of the order.
22 Would you please affirm your votes?
23 CHAIRMAN SELIN:
Aye.
24 COMMISSIONER ROGERS:
Aye.
25 COMMISSIONER CURTISS:
Aye.
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVENUE, N W (202) 234 4433 WASHINGTON. D.C. 20005 (202) 234-4433
4 1
COMMISSIONER REMICK:
Aye.
2 SECRETARY CHILK:
The second item is the 3
subject of Comanche Peak Unit 2 Full Power Operating 4
License.
5 The Commission is being asked to act on a 6
recommendation that the Director of the Office of 7
Nuclear Regulatory Regulation be authorized to issue 8
a full power license for Comanche Peak Unit 2.
9 All Commissioners have voted to approve 10 the authorization.
Commissioner de Planque, although 11 not present, has approved this authorization.
12 Would you please affirm your votes?
13 CHAIRMAN SELIN:
Aye.
14 COMMISSIONER ROGERS:
Aye.
15 COMMISSIONER CURTISS:
Aye.
16 COMMISSIONER REMICK:
Aye.
17 SECRETARY CHILK:
I have nothing further.
18 CHAIRMAN SELIN:
I would like to point out 19 that my vote and perhaps those of others was 20 conditioned by the later staff response of specific 21 question on April 2nd in addition to the original 22 submission from the staff.
Thank you.
23 This session is completed.
24 (Whereupon, at 12:05 p.m.,
the above-25 entitled matter was concluded.)
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVENUE, N W.
(202) 2344433 WASHINGTON. D.C. 20005 (202) 234 4433
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CERTIFICATE OF TRANSCRIBER This is to certify that the attached events of a meeting f
of the United States Nuclear Regulatory Commission entitled:
l t
TITLE OF MEETING: AFFIRMATION / DISCUSSION AND VOTE l
i PLACE OF MEETING: ROCKVILLE, MARYLAND DATE OF MEETING:
APRIL 6, 1993 l
were transcribed by me. I further certify that said transcription is accurate and complete, to the best of my ability, and that the
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transcript is a true and accurate record of the foregoing events.
- /2 44%
A.,
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I Reporter's name:
Peter Lynch i
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HEAL R. GROSS COURT REPORTSRS AMO TRAN$CRISER$
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[*g UNITED STATES NUGLLAH REGULATORY COMMISSION 3,i g j' g
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April 5, 1993 l
NOTE T0: The Chairman Comissioner Rogers Comissioner Curtiss Comissioner Remick Comissioner de Planque FROM:
James M. Taylor Executive Director for Operations
SUBJECT:
COMANCHE PEAK UNIT 2 FULL POWER LICENSE BRIEFING BOOK On March 8, 1993, the subject briefing book was provided to you in support of the March 16, 1993 Comission Meeting to consider authorizing a full power license for Comanche Peak Steam Electric Station Unit 2.
The March 8, 1993 note included information regarding fire barrier and ampacity derating testing recently conducted by TV Electric.
You were informed that updates to this information would be described in SSER 27, to be published upon full power license issuance. SSER 27 (Draft) is enclosed for your information. The final ORAT Inspection Report was issued on March 9,1993.
The changes from the draft, which was included in the briefing book, were editorial in nature.
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James H. Taylor Executive Director for Operations i
Enclosure-i
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SSER 27 (Draft) cc: SECY (w/ enclosure)
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..,_.__-m w_a-i d Supplement No. 27 Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Unit 2 Docket No. 50-446 i
Texas Utilities Electric Company, et al.
U.S. Nuclear Regulatory Commission Omce of Nuclear Reactor Regulation April 1993
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ADSTRACT Supplement 27 to the Safety Evaluation Report related to the operation of the Comanche Peak Steam E actric 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, i
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 20 to thit rcr t wcrc published.
This supplement deals primaril; with Unit 2 issues.
Supplement 5 was ctr.:t'ltd.
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 ev.aluation 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 ev11uation 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 Page ABSTRACT................................ iii 1
INTRODUCTION AND GENERAL DESCRIPTION OF PLANT...........
1-1 1.1 Introduction........................
1-1 1.7 Summary of Outstanding Issues................
1-4 1.P Confirratn y Issues.
1-4 1.9 License Conditions.....................
1-4 9
AUXILIARY SYSTEMS.........................
9-1 9.5 Other Auxiliary Systems...................
9-1 9.5.1 Fire Protection...................
9-1 13 CONDUCT OF OPE RAT IONS.......................
13-1 13.1 Organizational Structure and Qualifications.........
13-1 13.1.1 Management and Techn; cal Resources.........
13-1 13.1.2 Operating Organization...............
13-2 13.3 Emergency Pl anning.....................
13-2 14 I N I T I A L T E ST PROC RAM.......................
14 - 1 16 TECHNICAL SPECIFICATIONS.....................
16-1 17 QU AL IT Y AS SURANC E.........................
17-1 17.2 Organization of the QA Program...............
17-1 22 TMI-2 REQUI REMENTS........................
22-1 APPENDICES i
Appendix A Listing of Correspondence Since Last SSER Appendix B Bibliography; Listing of Reports, Codes, Standards GLs Referenced in Body Appendix C TM1 Issues Appendix D List of Principal Contributors Comanche Peak SSER 27 v
1 INTRODUCTION AND GENEPAL 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 (TUGCO)* (the appiicant) for a license to operate the Comanche Peak Steam Electric Station (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 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.
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 05 steam generators installed at CPSES.
Supolement 5 (SSER S) 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.
Supplement 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 Federal Reculations (10 CFR Part 50) deleting the requirement for installing jet impingement shields for the Unit 1 primary coolant loop 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).
Comanche Peak SSER 27 1-1
i Supplement 7 (SSER 7) was issued in January 1985.
It was limited exclusively to the staff's evaluation of allegations investigated by the NRC's Technical Review Team (TRT) pertaining to plant electrical /
instrumentation systems and testing programs.
Supplement 8 (SSER 8) was issued in February 1985.
It was limited exclusively to the 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 91 was issued in March 1985.
It was limited exclusive 1v to the staff's evaluation of coating requirements inside containment and allegations of coating deficiencies investigated by the TRT.
Sucolement 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.
Supplement 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.
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 evaluation of the Comanche Peak Response Team (CPRT) Program Plan, which was formulated 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.
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.
Supolements 15 and 16 (SSERs 15 and 16) were issued in July 1988; Supplements 17 throuch 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 m
Supplement 20 (SSER 20) was issued in November 1988.
It presented the staff's evaluaticn cf 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 haroware at CPSES.
Supplement 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.? cf it: f7 tt:t v re unresolved at the time SSEP 12 wts issued.
Of note from an administrative standpoint, SSER 21 renumbered items appearing in Sections 1.7,1.5, and 1.9, and deleted all items that were previously resolved but lirtrd ir. SSEP 12.
Sucolement 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.
Sucolement 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 TU 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
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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
'i and TU Electric that updates the correspondence listed in the SER and in SSERs 1 through 26. Appendix E 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 Soutn 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 Issuei Section 1.7 of the SER, as supplemented, identified no 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 1 license, and are similarly included in the Unit 2 license, follow:
(1) The applicant shall continue to control mineral exploratian 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.
Comanche Peak SSER 27 1-4 l
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(2) The applicant must implement and maintain in effect all provisions of the approved fire protection program, as described in the Final Safety Analysis j
Report (as amende?) and as approved in the SER and its supplemer.ts, 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 affect 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 piiysicci security, guard training and qualification, and safeguu m.
contingency plans, previously approved by the Commission, and all amendments made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).
The pians, wiich contain safeguards information protected ur.dcr i
10 CFR 73.21, are entitled:
" Comanche Peak Steam Electric Station Physical i
Security Plan" with revisions submitted through July 21, 1992; " Comanche 1
Peak Steam Electric Station Security Training and Qualification Plan" with 4
revisions submitted through June 10, 1991; and " Comanche Peak Steam Electric Station Safeguards Contingency Plan" with revisions submitted through December 1988.
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s Comanche Peak SSER 27 1-5
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t 9 AUXILIARY SYSTEB 9.5 Other Auxiliary Systers l
9.5.1 Fire Protection In Supplement 21 to the SEP (SSER 21), which was issued in April 1989, the steff reviewed the Comanche Peak fire protection program as documented in the FSAR i
through Amendment 71 and as described in Revision 1 to the Fire Protection Report.
l In Supplement 26 to the SER (SSER 26), the staff documented a review of fire I
protection-related changes and modifications made to the FSAR through Amendment j
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, l
J, and 0 of Appendix R to 10 CFR Part 50.
i l
The following TU Electric commitments related to fire protection issues were documented in SSEh 26:
l 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).
i i
This issue is discussed in this SSER (below). The NRC staff witnessed the l
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.
i Amoacity deratina testina:
Perform plant-specific testing (SSER 26, pp.
9-20 and 9-32).
This issue is discussed in this SSER (below).
TU Electric provided preliminary results from their plant-specific ampacity derating tests. The staff observed a portion of the testing, and will review the final ampacity i
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
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Comensatory measurer:
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 f
the 36-inch-wide cable tray fire test, with a seven-day cure time, and the completion of
- box enclosure" upgrades.
l t
Alternative shutdown system desian 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).
i This issue is not addressed below.
In SSER 26, the staff reported TU Electric's commitment to perform the necessary design changes before startup 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.
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, 1903 (TV Electric letter TXX-93136 to NRC), TV 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 below.
i 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 not
)
exceed a rise of 250 *F (139 'K) plus ambient temperature (using thermocouple averaging), and no single thermocouple reading should exceed i
30 percent above the specified average temperature rise.
(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
cracking, or meitu g; conductor insulation exposed, degraded, or discolored; shield exposed; or bare copper conductor exposed.
(C If the fire ba H turred through during the fire exposure, or if a visual cable inspection revealed any of the damage attributes listed above, then the barrier was considered to have deviated from the acceptance criteria.
Use of the fire test results to qualify a deviating fire barrier would 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 TV Electric's acceptance criteria, as supplemented by the conditions stated in the October 29.
1992, letter, ensureo inat acequate cable and barrier tests would be performed and that satisfactory results from these tests would constitute an acceptable basis for qualifying the CPSES Unit 2 fire barriers.
In a letter of February 1,1993 (TV Electric letter TXX-93076 to NRC), TU Electric comitted to either perform a confirmatory test of a 36-inch cable 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.
TU Electric comitted to perform one of these actions by the end of the first refueling outage for Unit 2.
Update In a letter of February 26,1993 (TV Electric letter TXX-93101 to NRC), TU Electric comitted to perform the confirmatory fire endurance test. A 36-inch 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 or Unit 2). The test configuration was built with the application of Thermo-Lag topcoat material approximately 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following completion of the raceway envelope. The fire test was conducted four days after the topcoat was applied.
Circuit integrity was not monitored during the test.
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 protected with 1/2" (nominal) Thermo-Lag panels with longitudinal, vertical, and bottom joints reinforced with stress skin and trowel-grade material. TU Electric summarized the test data in a letter of March 10. 1993 (TU 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 satisfactory. There were no signs of cable damage.
Comanche Peak SSER 27 9-3
Texas Utilitic: P-c Barrier Testing for Comanche Peak Unit 2 (Conducted March 4, 1993, at Omega Point Laboratories)
Maximum Average temperatures in Thermocouple temperatures in 'F
- F locations (Ambient - 68 'F)
(Ambient - 68 'F)
Power cable 241 277 Cor.tr c. cc. t 210 224 Instrument cable 217 240 i
Front tray ru i 244 285 l
Rear tray rail 247 292 Scheme 15 36* Cable Tray 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.
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 TV 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, TU 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. MS5998.
AMPACITY DERATING BackQround 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 SSEP 2f tN tteff discussed its concerns with inconsistent ampacity derating test data, but recognized that the ampacity derating concern is an aging issue rather than an imediate operability issue.
In SSER 26, the staff (1) documented TV 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 SSEF,2C
,,o
.~., TU Electric conducted a series of ampacity derati%
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 cbservcd tc:t prc;..:.th, ar.d testing from March 2 tc 7, 1953. The first test group, conducted from March 2, 1993 to March 3, 1993, consisted of a 3/4*- diameter conduit witn 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 f actor test results are sumarized 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 i
drop test articles.
1 1
(6) Both the baseline and ampacity derating test shall utilize measured current I
normalized as outlined in ICEA P-46-426 for final conductor and ambient j
temperatures (that were not 90 *C and 40
- C, respectively).
]
Comanche Peak SSER 27 9-5
[ Note:
By letter of March 23,1993 (TU Electric letter TXX-93136 to NRC), TU Electric reference; ise,-ision 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 ameter 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 irrespectivt cf the published ICEA values in accordance with the TU Electric letter of February 26,1993 (TV Electric letter TXX-93101 to NRC).
The ampacity derating test procedure used for all 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 same 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 cair lated as follows:
Ampacity derating factor = 1 -
where:
I, = ampacity value for product test I, = ampacity value for baseline test TU 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 maroin
- EE-CA-0008-3097 From 6.9 kV Cable tray - 40%
Conduit
- 40%
- EE-CA-0008-3038 From 480 V Cable tray - 38%
Conduit
- 23%
- 2-EE-053 All other Cable tray - 40%
Conduit
- 35%
Comanche Peak SSER 27 9-6
e i
i Calculation Cables Calculated excess ampacity marain
- 16345-EE(B)-140 Air drops Cable tray - 39%
Conduit
- 35%
TU Electric letttrs of March 10, 1993 (TU Electric letter TXX-93125 to NRC) and March 23,1993 (IU risctric letter TXX-93136 to NRC), supplied preliminary information regar.iing both TU Electric's calculated excess ampacity margin and i
the test result dita for the plant-specific ampacity derating tests.
Based on its testing, TU Elt;tric 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 provider the ampt 'ty derate margin based on the effects of the fire barrier (calculated excess ampacity margin minus the actual test data):
Ampacity derate Excess ampacity i
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%
i The NRC staff finds that the preliminary ampacity test results provided by TU Electric are acceptable since the test derate factor data are bounded by the calculated (design) ampacity margins.
However, the NRC staff is still reviewing TV Electric's plant-specific ampacity derating program. The NRC staff will complete its review of the plant-specific test program and results after TU Electric's submits the final test reports (consistent with the schedule published in SSER 26). Staff actions can be tracked under NRC TAC No. MS5999.
" BOX ENCLOSURE" UPGRADES BBCkorcund i
In a letter of January 19,1992 (TV Electric letter TXX-93038 to NRC), TU 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 oesign basis for the Thermo-Lag fire barriers installed at CPSES Unit 2 that were configured differently f rom 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.
TU Electric's fire testing program established the technical and installation attributes for most of the Thermo-Lag fire barrier configurations installed at CPSES Unit 2.
TU 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 Comanche Peak SSER 27 9-7
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 configarnion-specific fire endurance testing.
In Generic Letter 86-10, the staff provided guidance for performing engineering evaluations of raceway fire barrier systems that differed from the tested configurations.
TU Electric used this guidance to establish its fire barrier evaluation criteria for configurations that differed from the tested configurations.
The following su r r i:n TV Electric's criteria:
the continuity of the fire barrier material applied was consistent with the tested configuration; the effective thickness of the fire barrier material applied to the unique configuration was consistant with the thickness of the fire barrier material that was tested; the nature and effectiveness of the fire barrier support l
assembly was consistent with the tested configurations; and the application and i
end use of the fire barrier material was consistent with the tested I
configuration.
In its engineering report, TU Electric evaluated the following:
unique fire barrier configurations, minor protected commodity deviations, I
protruding and interfering item coverage deviations, and structural steel deviations.
In reviewing TU Electric's engineering report, the staff sampled those unique configurations where the fire barrier installations on safe shutdown raceways were constructed differently from those raceway fire barrier configurations tested by TU 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 configurations.
In SSER 26, the staff documented specific reviews of six representative 2.
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.
Configuration 2, consisting of two parallel horizontal cable trays (18 and 12-inches wide), acceptable subject to the confirmatory resolution of staff concerns regarding the 36-inch wide cable tray fire barrier.
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 i
report, this confirmatory test satisfies staff concerns regarding the 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 m
Regarding the " box-type" configurations (Configurations 1 and 3), in letters of February 26,1993 (lu Electric letter TXX-93101 to NRC), March 10, 1993 (TV Electric letter TXY-931?5 to NRC), and March 23,1993, (TV Electric letter TXX-93136 to NRC), TU Electric discussed the upgrades and documented their completion. TU 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 comitted to either increase the material thickness or rework the configurations in accordance with designs bounded by the previous fire barrier testing.
The NRC staff perfomed "walkdowns" of the subject configurations on February 19 and 26, 1993. On February 19, 1993, various elevations in the auxiliary building were walkw -. to ensure that TU Electric had properly selected tim configurations for upgrade.
No additional examples of cable tray box-type enclosures which would necessitate an additional layer of Thermo-Lag were identified. On February 26, 1993, walkdowns -of all 13 upgrades were conducted while work was in orcaress. TU Electric
- Minor Modification Forms"93-123 through 93-126 wer'e reviewed; these documented the upgrades by building elevation - 810' auxiliary, 832' auxiliary, 810' safeguards, and 790' auxiliary, respectively. The minor codification forms were verified to include engineering-basis discussions addressing the acceptability of the upgrades in regard to ampacity derating and the added weight on the supports. TU Electric 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 (TU Electric letter TXX-93125 to NRC), TU 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 Backcround i
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), TU 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 (TU Electric letter TXX-93060 to NRC), TU Electric submitted additional ir. formation regarding cure time, stating that its vendor concurred with TU Electric's recommendation on cure time.
Comanche Peak SSER 27 9-9
___m
____m_
TU Electric cured it: fire test specimens for at least 30 days preceding the conduct of the fire endurance tests. The staff was concerned that Themo-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 j
barriers reflect the tested conditions.
In a letter of January 28,1993 (TV Electric letter TXX-93061 to NRC) TU Electric comitted to provide fire watches as compensatory measures in accordance with the CPSES fire protection plan for the Themo-Lag fire barriers installed in areas that contain fire-safe shutdown conduits or cable trays until the barriers have curc; ict 30 days, and where box enclosures are located, until this issue is adequately resolved with the staff.
The use of fire watches is consistent with the compensatory measures implerented by TU Electric for the CPSES Unit 1 Themo-Lag fire barriers in response to NRC Bulletin 92-01, " Failure of Themo-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 TU Electric's commitment was acceptable and ensures that an adequate level of fire protection i
is provided et CPSES Unit 2 until the Thermo-Lag fire barriers are cured (1) to i
reflect the condition of the fire test specimens and (2) the box enclosure issue is resolved.
Update On March 4,1993, TU 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 l
order to demonstrate TU 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.
1 2
The staff considers TU 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 i
moisture content in the Thermo-Lag material had reached equilibrium, thereby providing conservative qualification fire test results. Additionally, a test i
was conducted with a shorter cure time, and the fire barrier did not exhibit any seam separation.
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 Oroanizational Structure and Oualifications 1
13.1.1 Management and Technical Resources In SSER 26 the st.fi recorded TU Electric's comitment 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 IGC), TU Electric sub:itted an advanced change to the FSAR to revise the TU 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 Production Group (formerly the Nuclear Engineering and Operations Group),
provides the design, engineering, construction, licensing, operation, and fuel management support for CPSES. The Nuclear Production Group was reorganized into i
i four organizations to better focus resources on operation of the dual-unit CFSES 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 with construction activities have been deleted as part of_the transition from Unit 2 construction completion to dual-unit operation. The description of the responsibilities for the Manager, Administrative Services has been removed because this position does not perfonn a safety function and, therefore, does l
not need to be in the FSAR. The staff finds this acceptable. The new organizational structure is shown in revised FSAR Figure 13.1-2, included as an attachment to the February 26, 1993, letter.
i The changes to the corporate organization made by the licensee primarily reflect an organizational restructuring to focus resources on dual-unit operation of the j
CPSES facility. The new lines of management authority and communication have been clearly defined. Other changes made to the corporate organization reflect changes in name, not in function. Therefore, they do not change the staff's l
previous conclusion that the corporate level management structure'is acceptable.
l 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-
__., _ _ _ J
~
1 i
l 13.1.2 Operating Organization i
i In SSER 26 the staff recorded TU 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 (TU Electric letter TXX-93102 to NRC), TV Electric submitted an advanced change to the FSAR to revise the TV Electric operating organization structure. The Nuclear Operations organization, under the direction of the Vice President of Nuclear Operations, is responsible 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 ard mairienance 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, Nuclee-TreiHr-W 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 Emercency planninc 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 TU Electric's emergency preparedness program as it related to the licensing of Unit 2.
No l
areas were identified that would preclude the licensing of Unit 2.
1 Additionally, TU 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 concluder that the overall state of emergency preparedness at Comanche Peak is adequate to support dual-unit operations.
I Comanche Peak SSER 27 13-3
6 14 INITIAL TEST PROGRAM Pre-operational Test Deferral 1
In SSER 26, the staff documented its review of TU Electric's preoperational test i
program changes for Unit 2 (TV Electric letters of December 23, 1992, TXX-92586 1
i to NRC; January 8, 1993, TXX-93011 to NRC; and January 25, 1993, Electric proposed to defer certain preoperational tests T"
TXX-93051 to N N.
until after fuel load. The staff verified that TV Electric's letters contained commitments for completing the tests at the appropriate plant power levels or pl e-+ rilestenes. The ste " deterrined that the schedule for performing the deferred testing ensured that systems required to prevent, limit, or mitigate l
the consequences of postulated accidents would be tested before the systems I
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), TU submitted additional ir. formation 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
{
methodology, results, schedules and deletions regarding plant computer, plant com.unication 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 dump valves will be retested " hot," but with the downstream block valves closed (similar to Unit 1 preoperational testing). Additionally, the schedule for I
testing the availability of the safety parameter display system and submitting the test report was clarified.
l The staff reviewed the additional information and determined that the l
conclusions reached in SSER 26 are still valid; that is, the systems will be adequately tested before they will be required to be operable, and the safety of the plant will not be dependent on the performance of untected systems, structures, and components. Therefore, TU Electric's additional information, including test methodology and scheduler changes, regarding the deferred testing i
is acceptable.
Comanche Peak SSER 27 14-1 l
].
l il 16 TECHNICAL SPEClilCATIONS The NRC issued the " Final Draft Combined Technical Specifications for Comanche Peak Unit I and Unit 2" to TV Electric on September 9, 1992. TU Electric certified on November 4,1992 (TU Electric letter TXX-92536 to NRC), that the final draft accurately reflects the as-built plant and the Final Safety Analysis i
TV Electric alte noted certain minor corrections. The staff discussed Peport.
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 l
f r January 22, 1993.
Editorial corrections were discussed r
Electric in a letter e
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.
l n
i l
1 I
l v
i Comanche Peak SSER 27 16-1 m
I L
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 (TV 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
- plement. There cherg
- resulted in some revisions to the description of the i
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 Oroanization 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 l
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 tne 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 qu:lity-rc'tted 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 ite r 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 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 staff concludes that the applicant's description of the QA organization is in compliance with applicable NRC regulations and is acceptable for the operation of CPSES.
l 1
Comanche Peak SSER 27 17-2 w
v
22 TMI-2 REQUIREMEN'"
After the accident at Three Mile Island (TMI) Unit 2, the NRC staff developed i
4 NUREG-0660, "NRC Action Plan Developed as a Result of the THI-2 Accident," to i
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 Requirerents," was issued in hovemoer 19c0; inis document included items approved bj 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 i
Appendix C of this SSER, the staff summarizes each TMI 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 i
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 l
Comanche Peak SSER 27 22-1 see-a-vea"-raw er +r eew+.
s--wweme s wowW
'T m
I, l.
1 APPENDIX A CONTINUATION OF CHRONOLOGICAL LIST OF CORRESPONDENCE This appendix continues the chronological listing of routine licensing correspondence, regaroing unit 2 and Unit 1/ Unit 2 comon issues, between the U.S. Nuclear Regulate y Comission (NRC) staff and the applicant (Texas Utilities Eiectric Company) since Supplement 26 was issued.
January 4, 1993 Sumary of November 4,1992, meeting with applicant regarding pressurizer surge line leak-before-break analysis.
January 6, 1993 Sumary of December 17, 1992, meeting with applicant regarding fire protection issues.
j 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 Anomaly 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
l January 28, 1993 Letter from applicant forwarding supplemental response to buiietin 88-08.
January 28, 1993 Letter from applicant forwarding clarifying information regarding test scheme 1, conduit support modifications, l
and use of test scheme 9 results.
January 29, 1993 Letter from applicant transmitting interim change request to preservice program plan.
January 29, 199?
t ette-from applicant transmitting information regardin; HVAC design validation.
January 29, 199?
Let+er to applicant transmitting significant findings of the Operational Readiness Assessment Team Inspection.
February 1, 1993 Letter from applicant transmitting response to concerns regarding turnover process, fire seals for piping penetrations and containment spray system nozzle completion.
February 2, 1993 Letter to applicant transmitting Facility Operating License No. NPF-88 for Comanche Peak Unit 2.
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 above 50 percent.
February 19, 1993 Letter from licensee forwarding results of engineering 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
February 24, 1993 Letter from licensee forwarding summary of personnel mcr.itcrir.; ending December 31, 1992.
february 25, 1993 Letter from licensee forwarding documentation of discussions with NRC regarding planned method of treating DNB penalties, l
February 25, 1993 Letter from licensee forwarding documentation of sensitivity study performed to evaluate effect of v : " t t i c t,: in core noding on calculated peak cladding temperature.
February 2C, 1993 Lciter 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, l
" Reactivity Anomaly Events Methodology."
March 11, 1993 Letter from 1.ca;see forwarding RXE-93-003, CPSES Unit 2 Cycle 1 Core Operating Limits Report.
l March 11, 1993 Letter to licensee forwarding " Toxicological Evaluation l
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 large 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 l
information on box configurations.
i l
Comanche Peak SSER 27 3
Appendix A PJ
March 28, 1993 Letter from licensee transmitting certification for rea::iness 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.
l l
I i
Comanche Peak SSER 27 4
Appendix A l
i
I i
l 1
APPENDIX B j
BIBLIOGRAPHY Miscellaneous Code of Federal Reculations. Title 10, " Energy," U.S. Government Printing Office, Washington, D.C.,
1991.
l National Electrical Code, insulated Cable Engineers Association (ICEA) publications.
i IEEE-P848, " Procedure for the Determination of the Ampacity Derating of Fire Protected Cables," Draft 11, April 6,1992.
i NRC BulletinsBulletin 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.
J t
i NRC Generic Letters Generic Letter 85-10, " Implementation of Fire Protection Requirements,"
April 28, 1986.
i NRC letters l
See Appendix A.
NRC NUREG-Series Reports
}
1 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.
J I
Comanche Peak SSER 27 1
Appendix B s* *w w + t v vw w -- -v -$
o--
-.e
,...,.-,,,_me,,,..
.m.,-o-r
.w,. -,.
,-er,,
%m.w
-....,cr.-e,.
m,-w..w--m,..w-c,,-c.m.,,.,..,ey-,*,w
G 1
l APPENDIX C IMI ACTION PLAN ISSUES 1
I
TM1 ACTION PLAN ISSUES SER/SSER resolved ITEM
[Verif. rpt. no.,
If Applicable]
I.A.I.l.1 P"t Technical Advisor; SSEP. 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 I.A.I.1.3 Shift Technical Advisor; SSERs 1 & 23 Training
[11/14/90; 50-446/90-40]
I.A.I.l.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]
1.A.I.3.1 Shift Manning; Overtime SSER 1 & TS 6.2.2.f
[11/14/90; 50-446/90-40]
1.A.I.3.2 Shift Manning; Minimum SSER 1 & TS T6.2-1 Shift Crew (11/14/90; 50-446/90-40]
1.A.2.1.1 Imediate Upgrading of SER & letter dated 3/8/85 Operator and Senior Operator Training and Qualifications; SRO Experience I.A.2.1.2 Imediate Upgrading of SER & letter dated 3/8/85 Operator and Senior Operator Training and Qualifications; Training I.A.2.1.3 Imediate 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 Imediate 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
J TMI ACTION PLAN ISSUES (Continued)
SER/SSER resolved ITEM
[Verif. rpt. no.,
IfApplicable) l I.A.2.1.5 Immediate Upgrading of SER & letter dated 3/8/85 0 -- '-
and Senior Operator 7Training 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 Liter. sing Examination -
Increase Passing Grade l
i I.A.3.1.3.A Revised Scope and Criteria for SER Licensing Examination - With Simulators F
I.B.I.2 Evaluation of Organization and SSER 1 Management Improvements of Near-Team Operating License Applicants I.C.I.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.I.2.B Procedures for Transients and SSERs 6, 12, & 22 Accidents, Short-Ters;
[3/11/93; 50-446/92-60)
Inadequate Core Cooling; Revise Procedures j
Comanche Peak SSER 27 2
Appendix C 1
4 h
~.,._....
b. I'110h PLAN ISSUES (Continued)
SER/SSER resolved ITEM
[Verif. rpt. no.,
If Applicable]
1.C.I.3.A Procadores for Transients and SSERs 6, 12, 1 22 Acciaents, Short-Tem; Transients and Accidents; Reanalyze Guidelines l
l 1.C.1.3.B Procedures for Transients and SSERs 6, 12, & 22 Accidents; Short-Tem;
[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]
1.C.3 Shift Supervisor SSER 1 Responsibilities
[11/14/90; 50-446/90-40]
i 1
1.C.4 Control Room Access 14/90; 50-446/90-40]
I.C.5 Procedures for Feedback of SSERs 6, & 23 l
Operating Experience to Plant
[11/14/90; 50-446/90-40]
l Staff 1
I.C.6 Procedures for Verification of SSER 1 l
Current Performance of
[11/14/90; 50-446/90-40]
l 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 NTOL Applicants I.D.1 Control Room Design Reviews SSER 22 (Uriit 1)
SSER 26 (Unit 2) l Comanche Peak SSER 27 3
Appendix C l
t I
I TMI ACTION PLAN ISSUES (Continued) i SER/SSER resolved ITEM
[Verif. rpt. no.,
j If Applicable]
i 1
1.D.2.1 Plant Safety Parameter Display SSER 22 Console; Description 1
I.D.2.2 Plant Safety Parameter Display SSER 22 (Unit 1)
Console; Installed SSER 26 (Unit 2) i
.[3/11/93;50-446/92-60]
l 1.D.2.3 Plant Safety Parameter Display SSER 22 (Unit 1) 3 Console; Fully Implemented SSER 26 (Unit 2)
[3/11/93; 50-446/92-60, final verification j
tracked as insp. item]
I.G.I.1 Training During Low-Power S3ER 6 Testing; Proposed Tests I.G.I.2 Training During Low-Power SSER 6 l
Testing; Submit Analysis and 3
Procedures I.G.I.3 Training During Low-Power SSER 6 Testing; Training and Results i
II.B.I.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]
II.B.I.3 Reactor Coolant System Vents; SSER 6 & TS 3.6.1.7 Procedures
[3/11/93; 50-446/92-60]
t II.B.2.1 Plant Shielding to Provide SSERs 2 & 22 Access to Vital Areas and i
Protect Safety Equipment for Post-Accident Operation; Design II.B.2.2 Plant Shielding to Provide
.SSERs 2 & 22 Access to Vital Areas and
[Incorp.in11.B.2.3]
Protect Safety Equipment for Post-Accident Operation; Corrective Actions Comanche Peak SSER 27 4
Appendix C i
--.,vr ee
,e m--.-
~ - -,
e,
~,,--,,,,v-,y.,-,,,.v.m.,
w..e<w,...,-e-..-..m,
..www..&..--w-.,-e.,-+-,
mu-m
---Y 6
l.
TP! t:T:T: PLAN ISSL'ES (Continued) i SER/SSER resolved l
ITEP.
[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]
i Pentert SPfety Equipment for
[1/27/93; 50 446/93-05]
i Post-Accident Operation; Modifications II.B.2.4 Superseded by 10 CFR 50.49 II.B.3.1 Post-Accident Sampling; SSERs 6, 22, & 23 Interim System
[11/14/90; 50-446/90-40]
II.B.3.2 Post-Accident Sampling; SSERs 6 & 22 Ccrrective 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]
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 j
Damage; Develop Training l
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 Damage; Complete
[11/14/90; 50-446/90-40]
II.D.1.1 Relief and Safety Valve SSER 21 (Unit 1) l Testing Requirements; Submit SSER 26 (Unit 2)
II.D.I.2.B Relief and Safety Valve SSER 21 (Unit 1) l Testing Requirements; Plant-SSER 26 (Unit 2)
Specific Report II.D.I.3 Relief and Safety falve 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
T:.:,C.I A FLAN ISSUES (Continued)
SER/SSER resolved 1
ITEM
[Verif. rpt. no.,
If Applicable]
II.D.3.2 Valve Position Indication; SER & TS 3.4.4 & 4.0.5 lecnnical Specifications 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 1
Evaluatir n; Long-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 d
Initiation and Flow; Control
[4/24/89; 50-446/89-17]
Grade II.E.1.2.1.B 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 l
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
e Tm FTION PLAN ISSUES (Continued)
SER/SSER resolved ITEM
[Verif. rpt, no.,
If Applicable]
II.E.4.2.5.A Containment Isolation SSER 22 Ct;.c:c:.Lility; Containment 2
Pressure Setpoint; Specify Pressure II.E.4.2.5.B Containment Isolation SSER 22 i
l 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)
Purce Valves T
II.E.4.2.7 Containment Isolation SSER 23 & TS 3.6.1.7 Dependability; Radiation (4/24/89; 50-446/89-17)
Signal on Purge Valves II.E.4.2.8 Containment Isolation SSER 23 & TS T-3.3-4.2 Dependability; Technical Specifications II.F.1.1 Accident Monitoring; SSER 3 Procedures
[ Refer to II.F.1 items l
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 ters); 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
TM1 ACTION PLAN ISSUES (Continued)
SER/SSER resolved i
ITE"
[Verif. rpt. no.,
If Applicable) i i
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]
II.F.1.2.f Accident Monitoring; SER & TS T-3.3-7 &
Containment Hydrogen 3.6.4.1
[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 II.F.2.2 Instrumentation for Detection SSERs 6, 21, 23 of Inadequate Core Cooling;
[3/11/93; 50-446/92-60]
Subcool Meter; Install II.F.2.4 Instrumentation for Detection SSER 6, 21, & 23 & TS T-of Inadequate Core Cooling; 3.3-6 Additional Instruments
[3/11/93; 50-446/92-60]
l 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 i
Relief Block Valves and Level Indication; Technical Specifications II.K.1.5 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 i
Break LOCA and Loss-of-Feedwater Accidents; IE Bulletins; operability Status II.K.I.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 m
,-r w
r
TMI ACTION PLAN ISSUES (Continued)
SER/SSER rescived ITEM
[Verif. rpt. no.,
^
If Applicable]
II.K.2.13 Themal-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 Transients II.K.3.1.A Automatic PORY Isolation SSER 6 System; Design i
II.K.3.1.B Automatic PORY Isolation SSER 6 System; Test / Install
[5/18/89; 50-446/89-24]
II.K.3.10 Anticipatory Trip H:
Power SSER 25
[5/18/89; 50-446/89-24]
II.K.3.ll Justification for Use of SSER 6 i
l Certain PORVs II.K.3.12.A Confirm Existence of SER & SSER 22 l
Anticipatory Trip Upon Turbine (5/18/89; 50-446/89-24]
Trip; Proposed Modifications II.K.3.12.B Confirm Existence of SER & TS T-3.3.1-16 Anticipatory Trip Upon Turbine
[5/18/89; 50-446/89-24) l Trip; Modify II.K.3.17 Report on Outage of ECCS SER II.K.3.2 Report on Overall Safety PORY SER Isolation System II.K.3.25.A Effect of Loss of AC Power on SER Pump Seals; Proposed Modifications I
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
~..._,--.T..;
~-
i I
I i
l TFT ECTION PLAN ISSUES (Continued) l SER/SSER resolved ITEM
[Verif. rpt. no.,
If Applicable]
II.K.3.3 Report Safety and Relief Valve SER & TS 6.9.1.2 &
F;% es Promptly and 6.9.1.5 Challenges Annually II.K.3.30.A Schtdic for Outline of Small-SSERs 6 & 12 Break LOCA Model 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 Shov Co pliance with 10 CFR 50.46 II.K.3.5.A Automatic Trips of Reactor SSER 22 Coolant Pumps; Proposed Modifications II.K.3.5.5 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 Contrei'er
[12/31/92; 50-446/92-51]
Modification III.A.I.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 EOF 1
III.A.I.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 Facilitier; Modifications
[11/14/90; 50-446/90-40)
(Superseded by MPAs F063, F064 and F065)
Comanche Peak SSER 27 10 Appendix C l
-w-
-,a
---a w
-wwg-r.w-
--w q-
THI ACTION PLAN ISSUES (Continued)
SER/SSER resolved j
[Verif. rpt. no.,
If Applicable]
III.A.2.1 Upgrade Preparedness; SSERs 3 & 6 (App. G.,
[
E---rr- : Plans Sec. 4) l 111.A.2.2 Up;-ade Preparedness; SSERs 3 & 6 (App. G.,
l Meter-clogical Data Sec. 4)
III.A.2.3 Upgrade Preparedness; SSERs 3 & 6 ( App. G.,
Implement Plans Sec. 4) 111.D.1.1.1 Integrity of Systems Outside SSERs 4, 22, & 23
[
Containment; Leak Reduction
[10/21/91; 50-446/91-46]
j III.D.I.l.2 Integrity of Systems Outside SSERs 4 & 23 & TS 6.8.3 i
Containment; Technical Srt:ificatierr i
j III.D.3.3.1 Improved Plant Iodine SER & SSERs 6 & 22 Instrumentation Under Accident
[4/1/91; 50-446/91-07]
i Conditions; Determine Presence j
of Radioicdine III.D.3.3.2 Improved Plant Iodine SER & SSERs 6 & 22 l
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]
i 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 e..,r,,
-_,..r
,_ +-
.,-,-.--,...,,-m
,_e,.,m,,.,-.,w..,,w.,,,--y
.i-,....,
l l
)
i i
i APPENDIX D l
1 LIST OF PRINCIPAL CONTRIBUTORS l
i Contributor Gr o6riiz at ion E. Baker Office of Nuclear Reactor Regulation Froject Directorate IV-2 j
D. Graves Senior Resident Inspector Region IV I
B. Holian Office of Nuclear Reactor Regulai. ion Project Directorate IV-2 R. Jenkins Office of Nuclear Reactor Regulation Electrical Engineering Branch i
P. Madden Office of Nuclear Reactor Regulation Plant Systems Branch I. Miller Office of Nuclear Reactor Regulation j
Plant Systems Branch E. Peyton Office of Nuclear Reactor Regulation.
Project Directorate IV-2 l
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 Comanche Peak SSER 27 1
Appendix D i
l
y nc og[g UNITED STATES
/
NUCLEAR REGULATORY COMMISSION Aj) f j.i g
WASHINGTON, D.C. 20665 0001
- ,E
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MEMORANDUM FOR: The Chairman Comissioner Rogers Comissioner Curtiss Commissioner Remick Comissioner de Planque FROM:
James M. Taylor Executive Director for Operations
SUBJECT:
COMANCHE PEAK UNIT 2 FULL-POWER LICENSING (REFERENCE THE MARCH 30 MEMORANDUM FROM THE EDO TO THE COMMISSION) to the subject memorandum discussed information regarding thermo-lag testing primarily from a generic perspective.
The staff believes it also important for the Comission to understand the staff's rationale for accepting the thermo-lag installation'at Comanche Peak.
Enclosed is a discussion of the staff's rationale for finding the thermo-lag installation at Comanche Peak acceptable.
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[ ames M. Taylor xecutive Director for Operations
Enclosure:
As stated cc: SECY OGC OCA OPA PDR n,n
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l ADDITIONAL INFORMATION ON THERMO-LAG HOSE STREAM TESTING COMANCHE PEAK In May of 1992, TV Electric (TV) advised the NRC staff that they were going to conduct a comprehensive fire endurance testing program to qualify their Comanche Peak, Unit 2, Thermo-Lag fire barrier designs. The initial testing methodology was based on the acceptance criteria established by American Nuclear Insurers (ANI) Information Bulletin #5 (79), "ANI/MAERP Standard Fire Endurance Test Method to Qualify a Protective Envelope for Class IE electrical Circuits," July 1979.
As a result of the initial tests conducted in June and August 1992, the staff had concerns regarding the use of these criteria in that the criteria allowed the fire barrier test specimen to exhibit signs of barrier burn through, cable damage to occur, and allowed the barrier to be breached during the standard hose (solid) stream test.
Based on the staff's concerns, TU revised their fire testing methodology to eliminate the above concerns.
In a letter dated October 29, 1992, the staff concurred that TU's revised methodology was acceptable.
TU's revised hose stream test applies water through a 1%-inch fog nozzle set at a discharge angle of 30 degrees with a nozzle pressure of 75 psi and minimum discharge flow rate of 75 gpm.
The fog stream is applied to the specimen with the nozzle positioned 5 feet away from the center of the specimen at a pressure of 75 psi for 5-minutes.
NUREG-0800, Standard Review Plan (SRP) 9.5.1, " Fire Protection for Nuclear Power Plants," provides fire protection guidance which the staff finds acceptable with regard to meeting NRC fire protection regulations. An applicant can either follow this guidance or can propose to the staff an alternative method for achieving the level of fire safety required by the regulations. TU proposed an alternative method for hose stream testing. This method and the fire endurance testing acceptance criteria and methodology was found technically correct. The following presents the basis for the staff's acceptance of the TV alternative hose stream testing method:
1.
TU's revised fire endurance testing methodology and acceptance criteria adopted the fog hose stream testing criteria established by i
SRP 9.5.1 for penetration seals.
2.
The Comanche Peak, Unit 2, fire protection program is based on a
" defense-in-depth concept." The implementation of this program required the establishment of a fire prevention program; controls on ignition sources; fire protection features which provide fire barrier separation between safe shutdown trains; rapid detection of a fire and smoke condition; automatic and manual fire suppression and control methods; and, fire resistive structural building features.
3.
Comanche Peak, Unit 2, used fire resistive building construction techniques for their safety-related buildings.
In addition, the combustible fire loads associated with safety-related and safe shutdown plant areas at Comanche Peak are generally low.
f.
4.
If a fire were to occur in a safe shutdown area at Comanche Peak, structural building collapse is unlikely due to the fire resistive construction techniques used in the design of these structures.
Therefore, directional loads imposed on these barriers by falling structural objects during a fire is not expected.
Further, these areas are generally protected by automatic sprinkler protection and j
would actuate in a timely manner to control the fire.
5.
The fire brigade at Comanche Peak is trained in the application of water through fog streams for controlling energized high voltage equipment fires.
In addition, it should be noted that standpipe systems in nuclear power plants follow the design guidance of NFPA 14. This standard requires that the flow pressures at the hose valve where the hose station's manual fire fighting hose 1
(100 feet) connects to the standpipe be restricted to 100 psi.
Note that the friction loss in 100 feet of l\\-inch quality fire hose is approximately 25 psi i
at a nozzle flow rate of 86 gpm. This would yield a nozzle pressure at the i
hose stations of 75 psi..
During the November 16-20, 1992, fire protection inspection of CPSES Unit 2, representative interior hose stations were inspected. No playpipe type solid stream nozzles were found. The type of l
nozzles used were the electrically-safe fog type, i
It should be noted, the fog stream application (hose stream test) at 75 psi with a nozzle flow of a minimum of 75 gpm, when compared to the actual in plant hose station pressure and nozzle flow conditions, yields a ll~ gpm i
variance and no variance in pressure.
Since there was no variance in pressure, the variance in flow is considered to have a minimal affect on the I
eroding and cooling affects of water on the test assembly. The staff r
concluded that the hose stream testing performed by CPSES in their fire barrier testing program was acceptable.
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i MEMOPANDUM F0D:
The Chairman Comissioner Rogers Commissioner Curtiss Comissioner Remick Comissioner de Planque FROM:
James M. Taylor Executive Director for Operations
SUBJECT:
COMANCHE PEAK UNIT 2 FULL-POWER LICENSING (SRM M930316)
On March 16, 1993, the staff briefed the Comission on the status of the Comanche Peak Unit 2 full-power licensing review. On March 24, 1993, the plant achieved initial criticality and all low-power testing was successfully completed on March 28, 1993. An NRC inspection team observed initial criti-cality and low-power testing.
In a letter of March 28,1993 (Enclosure 1)
TU Electric notified the NRC that it is ready ~for operation above 5% power.
Enclosures 2 and 3 provide additional information regarding current 10 CFR 2.206 petitions related to Comanche Poak and technical information regarding
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Thermo-Lag. The staff is also responding on March 30, 1993, to the i
Comission's Order of March 26, 1993, in the Construction Permit Extension i
Proceeding.
The staff has assessed the status of the issues that the Office of Investigations (01) is reviewing pertaining to the Comanche Peak facility.
There is no change to 01's conclusion, stated in a memorandum of February 23, 1993, that the subject issues "would not preclude the Comission's consideration for a Full Power License."
On March 29, 1993, the Regional Administrator, Region IV, recomended to the Director, Office of Nuclear Reactor Regulation, that a full-power license be issued to Comanche Peak Unit 2.
This recomendation is enclosed for your information (Enclosure 4). On the basis of this recommendation, as well as on additional staff review, the Director of the Office of Nuclear Reactor Regulation has determined that the plant meets the Comission's regulations,
Contact:
Brian Holian, NRR 504-1334 r,
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t Multiple Addresses and the activities authorized by a full-power license can be conducted without endangering the public health and safety. Accordingly, I recommend that the Commission vote to authorize the Director of the Office of Nuclear Reactor Regulation to issue a full-power license for Comanche Peak Unit 2.
Original signea by J8mt3&M James M. Taylor Executive Director for Operations
Enclosures:
1.
Letter from TU Electric to NRC dtd. 3/28/93 2.
10 CFR 2.206 Petitions 3.
Additional Information on Thermo-Lag 4.
Memorandum from Region IV to NRR dtd. 3/29/93 cc w/ enclosures:
- See Previous Sheet for Concurrence effice PDIV-2/LA PDIV-2/PM,
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DISTRIBUTION Docket File PDIV-2 R/F j
TMurley JPartlow JRoe MVirgilio 5 Black BHolian EPeyton NRR Mail Room (WITS 9300051) 12/G/18 MBridgers (WITS 9300051)
JTaylor J5niezek HThompson JBlaha JMilhoan, Region IV WRussell FGillespie JKnubel, EDO
Enclosure i M
Log # TXX-93155 M=
File # 10010 0-.
j Ref. # 10CFR.50.57
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TUELECTRIC narch 28, 1993
% illiam J. CaMIL jr.
Go.a b ee Presoorm U. S. Nuclear Regulatory Commission Attn: Document Control Desk Wasnington, DC 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) - UNIT 2 DOCKET NO. 50-446 READINESS FOR ISSUANCE OF THE UNIT 2 FULL POWER OPERATING LICENSING REF:
1)
TU Electric letter logged TXX-93093 from William J. Cahill, Jr. to the NRC dated February 19, 1993 2)
TU Electric letter logged TXX-93011 from William J. Cahill, Jr. to the NRC dated January 8, 1993 3)
TV Electric letter logged TXX-93051 i
from William J. Cahill, Jr. to the NRC dated January 25, 1993 4)
TV Electric letter logged TXX-93140 from William J. Cahill, Jr. to the NRC dated March 22, 1993 Gentlemen:
TU Electric has ccmpleted and evaluated the low power physics testing and the accitional testing that can be completed prior to proceeding above 5% reactor power. Enclosure 1 provides a listing of the testing that is described in Chapter 14 of the Final Safety Analysis Report (FSAR) which was conducted during Mode 6 through Mode 2 since the issuance of the CPSES Unit 2 low power operating license.
Additionally, TU Electric has performed a self-assessment of the readiness of CPSES Unit 2 for proceeding above 5% power in accordance with the description provided in Reference 1.
This self-assessment has been reviewed and evaluated by the Station Operations Review Committee (50RC). The 50RC has concluded that CPSES Unit 2 is ready for operation above 5% power.
- 0 N. Ohve Stree LB. 81 Dallas.Tcass 73203
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i TXX-93158 1
Page 2 of 2 1
i TU Electric is at this time ready to receive an operating license for CPSES Unit 2 which authorizes operation up to 100% reactor power.
I Sincerely, o
William J. Cahill, Jr.
i Enclosure l
c - Mr. J. L. Milhoan, t egion IV Resident Inspectors, CPSES (2)
Mr. T. A. Bergman, NRR Mr. B. E. Holian, NRR Mr. L. A. Vandell, Region IV l
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.o Enclosure to TXX-93158 Page 1 of 1 A.
Deferred Preoperational Tests and Retests conducted during Mode 6 through Mode 2.
(Reference 2 Reference 3, and Reference 4).
- 7 1.
Pressurizer Spray Valve Leak Tightness 2.
Power Operated Relief Valve (PORV) Leak Tightness 3.
Reactor Cavity Humidity Detectors 4
Steam Dump Valve Stroke Verification 5.
Public Address and Emergency Evacuation Alarm System 6.
Main Stream Isolation Valve (MSIV) Stroke Timing 7.
Plant Computer Flux Mapping Module 8.
Plant Computer Data Archive Capability 9.
Plant Computer Delta I Module 10.
Heat Ventilating and Air Conditioning (HVAC) System Flow Balance Test results for the tests listed above bave been evaluated by the Test Review Grcup as acceptable for proceeding to full power.
B.
Initial Startup Tests conducted during Mode 6 through Mode 2 (Reference FSAR Chapter 14. Table 14.2-3 and Figure 14.2-48).
1.
Reactor Trip System 2.
Boron Reactivity Worths 3.
Rod Drop Tests
- 4 Reactor Coolant Flow Test 5.
Reactor Coolant Flow Coastdown
- 6.
Control Rod Drive Tests 7.
Rod Position Indicators S.
Moderator Temperature Reactivity Coefficient 9.
Control Rods Reactivity Worths 10.
Auxiliary Startup Instrumentation
- 11.
Chemical Tests
- 12.
Core Performance Evaluations 4
- 13.
Calibration of Nuclear Instrumentation
- 14 Radiation Survey
- 15.
Core Reactivity Balance
- 16.
Incore Nuclear Instrumentation 17.
Reactor Coolant Leak Test 18.
Rod Control System Test j
Test results for the testing listed above have been evalurted by the Test Review Group as acceptable to proceed above 5% power.
- Those items identified with an asterisk are complete for Mode 6 through Moce 2 but additional testing will continue as a nornal part of power ascension testing. Those items not asterisked _are complete.
10 CFR 2.206 PETITIONS The purpose of this enclosure is to provide the basis for the staff's recommendation that the full power license be issued for TU Electric's Comanche Peak Steam Electric Station (CPSES) Unit 2 with 10 CFR 2.206 petitions not finalized.
The two unresolved Comanche Peak 10 CFR 2.206 petitions are discussed below.
Comanche Peak Soecific Michael Kohn, on behalf of Messrs. Macktal and Hasan, submitted a 10 CFR 2.206 petition on June 11, 1992. The petitioner alleges that the purchase agreement for Tex-La's minority interest in CPSES by TU Electric violates NRC regulations on restrictive settlement agreements. The NRC acknowledgement letter, sent to the petitioner on August 12, 1992, stated that staff review has determined that the agreements do not appear to violate the provisions of the Energy Reorganization Act or 10 CFR 50.7.
Notwithstanding, letters to TU Electric and the former co-owners of CPSES were issued on January 12, 1993, that requested information pertaining to the settlement agreements. The responses to these letters have been received and are being evaluated by the staff. The Director's Decision is expected to be issued in April 1993. The staff does not believe these issues affect issuance of a full power license for Unit 2.
Although there may have been the potential for safety information to have been withheld, the petitioner did not identify any issues with respect to which he believed information had, in fact, been withheld.
On the basis of the following, the staff concludes that there is no safety significance associated with the issues currently identified in this petition:
(1) No violation of the regulations has been identified with respect to the settlement agreements, nor were any safety issues identified, (2) About 12,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of direct inspection (since the resumption of Unit 2 construction) has been conducted at CPSES Unit 2 by NRC personnel, and j
(3) A recent NRC review of the licensee's SAFETEAM program (Inspection Report 50-446/92-60) concluded that the program provides both:
(a) a means for employees to bring concerns to management, and (b) plant management with a mechanism for the early identification of issues that could impact the safety of the plant.
The Commission requested additional information regarding this issue by an Order issued in the context of the construction permit amendment proceeding, dated March 26, 1993. The staff response is being filed on March 30, 1993.
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Generic Thermo-Lao The Nuclear Information and Resource Service (NIRS) submitted a 10 CFR 2.206 petition on July 21, 1992, as supplemented by addendum of August 12, 1992. On February 1,1993, a Partial Director's Decision (DD-93-03) was issued regarding this petition.
On December 15, 1992, NIRS filed another petition pursuant to 10 CFR 2.206 regarding Thermo-Lag. This petition, which addresses numerous plants, including CPSES, Units 1 and 2, is being considered by the staff as a supplement to the petition filed on July 21, 1992. The acknowledgement letter, issued on February 4, 1993, denied the requested action to shut down all plants with Thermo-Lag, stating that no immediate safety concerns were raised. The review of the petition, which reiterates items from a previous petition, and includes issues regarding Thermo-Lag voiding and stapling is expected to be complete by May 1993. The staff's evaluation of CPSES Unit 2 fire barrier acceptability is presented in SSER 26, in which the staff concluded that with several commitments, TU Electric's fire barrier program is acceptable. The staff specifically requested information on the stapling issue and although not explicitly discussed in SSER 26, was considered in the discussion on page 9-20 in support of this conclusion.
i ADDITIONAL INFOPJiATION ON THEPJ'0-LAG This enclosure provides additional information regarding two issues relating to Thermo-Lag:
hose stream testing and seismic concerns.
HOSE STREAM TEST The American Society for Testing and Materials (ASTM), as part of their testing methods for determining the fire endurance for various types of building construction, adopted the current method for hose stream testing in 1933. Currently, fire testing standards are focused on building columns, walls and partitions, and floor construction.
The intent of the hose stream test is to impose a cooling, impact, and erosion effect on the building construction being evaluated. The weaknesses of a structural building system after being subjected to the hose stream test, fall into the following categories: structural failures, thermal (brittle) failures and erosion failures.
Focusing on raceway fire barrier systems, the staff, as part of their acceptance criteria development, examined the applicability of the ASTM standard hose stream test to these barrier systems.
Staff consensus, during this examination, identified the need for some form of hose stream test.
The staff elected to adopt approved hose stream testing methods identified by Position 5.a to Standard Review Plan (SRP) 9.5.1, " Guidelines for Fire Protection for Nuclear Power Plants." This SRP position established the fire endurance and hose stream testing acceptance criteria for other non-structural fire resistive barrier components (penetration seals) and allowed the use of i
the fog nozzle method for hose stream testing.
The staff's rationale for allowing the use of the fog method is based on the following:
Structural Walls, fire doors and dampers are structural building components, whose failure could either contribute to the structural failure of the building or fire growth within the building.
Fire barrier systems used to separate safe shutdown functions within the same fire area are not considered to aid in the prevention of structural building collapse.
Fire resistive construction techniques are used in the design of nuclear power facilities to prevent such structural failures.
In addition, the combustible fire loads are generally low.
Under actual in-plant fire conditions, structural collapse is unlikely.
Therefore, directional loads (simulated by the standard hose stream test) imposed on these barriers by falling structural objects during a fire is not expected.
Fire fighting activities, resulting from manual suppression can cause some level of barrier impact.
Manual fire fighting suppression operations in the areas of energized electrical equipment require the use of fog streams.
The fog nozzle test method (Pressure and Flow) simulates in-plant manual fire suppression techniques which would be employed by the fire brigade.
Cooling The fog strean method applies more water to the test specimen over a greater dration of time.
(Fog - 375 gallons vs. standard - 210 gallons) (Duration of application:
standard - 1 minute vs. Fog - 5 minutes).
The amount of water applied and the duration of application is sufficient to determine the thermal fragility of the barrier system under simulated i
fire fightir; hose stream applications.
Erosion The Fog method is sufficient to demonstrate erosion conditions which
]
would be encountered by the implementation of in-plant fire fighting techniques.
Experience (TV Electric tests) has demonstrated the ability of this method to impact the fire barrier material char layer, seams and joints.
j This method is capable of eroding the char layer and has made openings in the areas of joints and seams.
(Scheme 12-2, 24-inch wide tray with T-Section, hose stream test damage)
Additionally, the staff considered the fact that nuclear power plant fire protection programs are based on a " defense-in-depth concept," (e.g.,
prevention of fires; control of ignition sources; fire protection features which provide fire barrier separation between safe shutdown trains; rapid detection of a fire and smoke condition; automatic and manual fire suppression and control methods) in support of using the fog stream test method.
. SEISMIC ISSUES The NRC does not require that fire protection systems, including features such as fire barriers, be formally qualified for seismic events. Such qualification is required for safety-related systems that are used to mitigate design basis events such as large pipe breaks. The NRC is requiring licensees to address the potential consequences of events beyond the design basis as part of a systematic review of plant vulnerabilities (Individual Plant i
Examinations for External Events). One area specifically to be examined is fires caused by earthquakes.
The staff has, however, specifically examined the potential for Thermo-lag panels to break up during a seismic event, thereby creating a threat to nearby safety-related equipment. A 10 CFR 2.206 petition from Nuclear Information Research Service postulated such a breakup, with the panels acting as a shear (severing cables and shattering cable trays), thus jeopardizing safe shutdown.
l To the NRC staff's knowledge, the Thermo-Lag vendor has not performed seismic tests of prefabricated panels. The staff has reviewed a seismic analysis (performed by a consultant to the Thermo-Lag vendor) of such panels attached to cable trays and conduit sections.
In addition, the staff visited a plant to understand more about Thermo-Lag usage, installation details, and material properties.
It is the staff's judgement after this review, that Thermo-Lag i
panels are not likely to get detached from raceways during a safe shutdown earthquake. Although the material may crumble and crack, the staff concluded that considering the material properties of Thermo-Lag and the design of raceways, shattering of raceways and severing of cables are not credible scenarios, and that the safe shutdown capability would be maintained. As discussed above, the beyond design basis accident scenarios of earthquake induced fires will be considered under the Individual Plant Examinations for External Events.
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EW 2943 14:41 R !U US E FTS 'T29-E270 001 Las7a3sTATsa
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AR LtNOTON, Tt KAa 79011 4008 g3 Docket No. 50-446 MEMORANDUM FOR: Thomas E. Murley, Director t
Office of Nuclear Reactor Regulation FROM:
James L. M11hoan, Regional Adminustrator
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION UNIT 2 FULL-POWER LICENSE RECOMMENDATION The Region IV staff and I recommend the issuance f a full-power operating license for Comanche Peak Steam Electric Station (CPSES) Unit 2.
The bases for this recommendation are contained in a readiness assessment which was provided to you on February 1,1993, prior to the' issuance of the low-power license and the additional information provided bhlow.
Theconstructionandpreoperationaltestinlpectionprogramshavebeen s
I.
completed.
2.
The start-up test and operations inspection, programs have been initiated and an NRC master inspection plan for Unit $ has been developed and approved. The NRC resident inspectors, augmented by Regional staff, provided around-the-clock coverage of fuel loading, initial approach to criticality, and power ascension below S pe rcent of rated power. NRC inspectors also observed mode changes, zero power start-up tests, normal day-to-dayoperations,andmaintenanceactihities. Two issues were identified during initial licensed operations as discussed in Attachment I to this memorandum.
Those issues have been satisfactorily resolved.
In addition, a regional readines'$ assessment team inspection was conducted March 25-28, 1993, to evaluat's the licensee's performance during initial critical operations. The te'am concluded that TU Electric has implemented appropriate measurus including management oversight, corrective actions, and self-assessments tol support plant operations.
The operations, engineering and maintenance, organizations demonstrated a common resolve for the safe conduct of plant operations. Based on direct observation and evaluation of licensee performance, we believe that TU Electric's oarformance to date demonstrates their readiness to operate the plant aiove 5 percent power.
i 3.
The CPSES Oversight Panel met on March 11 and 28,1993, to perform an evaluation of licensee performance, and in ! review the results of the readiness assessment team inspection and th'e around-the-clock inspection coverage of the low power testing. The pan'el concluded that TU Electric is ready to conduct dual unit operations, including power operation of Unit 2.
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23/2943 84:41 R IU US RO FTS 72EHE278 Oc2 l.>
Thomas E. Murley
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Region IV staff were requested by memorandym dated December 3,1992, to identify any issues that could impact the issuance of a license for CPSES Unit 2.
There were no concerns expr ssed by any Region IV personnel relative to the issuance of a lo -power license for Unit 2.
Since the hun of the low-power licens, no concerns have been expressed to Region IV management.
5.
Acceptable prograss has been made in addressing those items identified l
in my February 1,1993, memorandum as being incomplete at the time of low power licensing. Resolution of the remaining open itee.s is not deemed necessary to support issuance of a full-power license. provides the status of those items.
NRC inspectors will continue to im lament an augmented inspection and evaluation program to monitor the icensee's performance during the period of power escalation to full-power. This will include:
Around-the-clockinspectioncoveragetomoqlitorthelicenste's l
performance during important tests and during parts of power escalation.
l Witnessing selected start-up tests and reviewing the start-up testing results.
Reviewing the licensee's sel'f-assessment to be performed at the 50 percent power plateau.
ir With respect to the staff requirements memorandud concerning the March 16, 1993, discussion on full-power operating licensirjg for Comanche Peak (Unit 2),
it is our understanding that your staff 1s responding to the Comission's request for additional information.
Region IV has reviewed and concurred in the response.
l In conclusion, we ' find that CPSES, Unit 2, constduction has been substantially completed in accordance with Construction Permit {CPPR-127, the FSAR, and NRC regulations; that fuel loading, initial start-up, and operations to date have been conducted safely; and that TU Electric is ready to safely operate Unit 2 above 5 percent power.
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. Ames L. Filhoan Vlegtonal Administrator Attachments: As stated cc w/ attachments:
(see next page) i
ELV2943 84:42 R IU US tft0 FTS 72EHE79 ag3 f
{
Thomas E. Murley -
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cc w/ attachments:
i J. Taylor,(MS17G21) t J. Sniezek, (MS 17G21) i RIY Coordinator, OEDO (MS 17621)
R. Bernaro, (MS EEE)
J. Partlow, (MS 12G18) l W. Russell, (MS 12G18)
J. Roe, (MS 13E4)
M. Virgilio, (MS 13A2)
- 5. Black, (MS 13H15) j B. Holian, (MS 13H15)
+
C. Rossi, (MS 9A2)
C. Grimes, (MS 11E22)
J. Montgomery L. Callan A. Beach S. Collins L. Yandell D. Graves i
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@s ATTACWENT 1 155UE5 IDEWT!FIED SINCE LOW-POER LICENSING I
Two issues were identified after the issuance of the low-power operating license that were of some potential significance and which were discu: sed in a staff note for the Commission on March 10, 1993.
The first issue involved the identification of a Borg-Warner pressure seal check valve in the auxiliary feedwater system (Valve AF-106), that failed a strve111ance test for back flow. The second issue involved the identification of noise on the reactor coolant system loose parts monitor that indicated the potential presence of a loose part in the rereter pressure vessel. The status of those issuer is described below:
l BORG-WARNER CkECK VAtVF AF-106 $TATUS On March 8,1993, auxiliary feedwater check valve AF-106 failed a back flow test. The valve was disassembled for raintenance and found to be rotated 10 degrees with respect to the disk seating surface., This resulted in the disk hanging approximately 1/2" off its seat at the lower surface. The valve was reworked and reassembled, and satisfactorily back flow tested on March 11, 1993. TU's investigation indicated that a satisfactory back flow test was completed last fall. Af ter the test was completdd, work instructions called for the installation of a block and key device that was intended to aid maintenance personnel in properly aligning the valve during future maintenance activities. However, it appears that some actio during 11e installation of the block and key device caused the misalignment of the valve. At that time, no post work testing was required. As an immedt te corrective action, the licensee determined that 25 of these check valves had the block and key device installed.
Fourteen of this number are safety-rflated, and all were back flow tested satisfactorily after installation of the device. The remaining Il check valves were evaluated through the use of tsmperature monitors, radiographs, or back flow testing and detemined to be satisfactory. All other Borg-Warner check valves installed in Unit 2 are bolted bonnet valves that are not susceptible to misalignment during yeassembly. The licensee completed a root cause analysis of this failure on March 23, 1993.
It was concluded that the cause for the event was uncertlduring the process to install ain, however, a contributing factor was the lack of verification of alignment j the block and key devices. The NRC readiness assessment team reviewed the issue and concluded that TU Electric had conductsd an appropriate root cause analysis. The corrective action:: implemented were found to address the condition in which the valve was found and should prevent recurrence. This same corrective action should also be effective in preventing stallar problems on the Borg-Warner pressure seal check valves wha re the block and key devices are not installed.
LOOSE PARTS MONITOR ALARM On March 7, 1993, with Unit 2 in Mode 4, operations personnel identified a high noise alarm on the reactor coolant system 160se parts monitor (LPM).
Subsequent evaluation of this condition by the 1(consee's plant engineering organization, in consultation with the LPM manuf3cturer (Babcock and Wilcox),
and the Nuclear Steam Supply System vendor (Westi,nghouse), concluded that the LPM high noise alarm was attributable to flux thtmble tube vibrations. Based on the technical analysis provided by Westinghouse, the licensee concluded I
03/29/93 14:43 R IU US ARC FT5 720-8278 825
-t-that there was no loose part present, that there were no short tern safety concerns related to the flux thimble tube vibratihns, and that the LPM system was acceptable for continued operation.
NRRandfegion!Ystaffagreedwith the licensee's conc 1L:,1ons. The licensee has an ' inspection and maintenance program, as required by NRC Bulletin 88-09, that will address vibration of the bottom mounted incore flux thimbles.
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ATTACHMENT 2 STATUS OF OPEN ITEMS FOR COMANCHE PEAK UNIT 2 The low-power license recommendation letter dated February 1,1993, provided the status of items that were not required to be icompleted prior to the l
1ssuance of the low-power operating license. Ths' current status of those items is discussed below, and none of the issues remaining open needs to be resolved prior to full-power licensing.
The Operational Readiness Assessment Team (ORAT) identified weaknesses
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~
in the areas of configuration control (system status control),
l procedures, and corrective actions which th's licensee indicated would receive necessary attention prior to fuel l'ad.
In the area of system e
toMode6andrevisedthecontrollingprocelduretoensurepositivestatus con i
system status control. RegionIVinspector;sconfirmedbyprocedure reviews and system walkdowns that the licertsee had taken appropriate corrective actions.
Intheprocedurearea,)thelicenseerevisedthe procedures in question and comitted to perform a complete procedure review (approximately 700 procedures) over %he next two years. Region IV inspectors confirmed that the specific procedure deficiencies identified were corrected by the licensee.
The contract auxiliary operators who nad not received training required by procedures were removed from plant duties. Also, the inspectors verified that the work performed by the contract auxiliary operato;rs was reviewed by the licensee, and that the operators were reass,igned to tasks authorized by management and commensurate with their current training. With regard to corrective actions, the licensee completed evaluation of the four items
^
identified by the ORAT. Region IV inspectors determined that post-test thatanextensivefieldverificationofabnleappropriatelyappliedand requirements on a feedwater check valve wer ormal procedures was being completed by the licensee on the committed schedule. Overall, Region IV is satisfied that the ORAT concerns have been satisfactorily addressed.
All TMI items, except the Safety Parameter Display System (SPDS),
were verified complete prior to low-power 1,1 censing. The SPDS was operational, but a rer,uired assessment after 30 days of operation had not been completed. ByletterdatedMalrch 22, 1993, the licensee indicated that no specific date fe,r the start of 30-day essessment had been established, but that the assessment would startwithin60daysafterfuelload(April;7,1993). This is the same approach used for Unit 1, and the Region finds it acceptable.
Thermo-Lag fire barrier material was verified as installed prior to low-power Itcensing; however, not all Thorso-Lag installations ind completed a 30-day cure time.
In addition,13 (box enclosure) installations that were configured differently from tested cor(figurations had not been adequately justified by analysis or testing. These installations resulted in the implementation of fire watches as a crampensatory measure pending completion of the cure time and co ' figuration upgrades.
Currently, the 30-day cure time has passed for all of the installations completed prior to the issuance of the 1cwypower license. Also, as discussed in draft Supplemental Safety Evaluation Report 27 (to be published concurrent with full-power license), testing performed on March 4, 1993, established a seven-day cure' time as acceptable for I
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all future installations. With regard to the box enclosure configurations, all were subsequently found acceptable after the staff's review, and these upgrades were certified domplete by TU Electric on March 10, 1993. Fire watches are no longer required with regard to these two specific issues.
Identified weaknesses regarding the control of temporary modifications as discussed in Inspection Re, port 50-445/97-62; 50-446/92-62, have been effectively addressed by interim actions taken by the licensee.
At the time of low-power licensing, the Region IV staff perceived that was being rapidly dispelled as the two,t 2 operations staff j
the lack of a sense of ownership by the Uni units were brought together under a single operating organizat, ion.
Since issuance of the low-power license, two special inspectilcns were conducted that looked In part, at this issue. The' inspection of the licensee's personnel error reduction program in Februa,ry 1993, and the augmented inspection of licenses performance during initial low-power operations during March 25-28, 19,93, determined that the plant staff has exhibited an increased sense of ownership of Unit 2commensuratewiththatseenonUnit11.
Test and retest deferrals discussed in Att chment 13.1 of the February 1,1993, low-power license reconsundation letter have i
been successfully completed in accordance fith the licensee's completed consistent with plant operational,, scheduled to be schedule.
The remaining deferred tests are requirements. Region IV continues to track the completion of these deferred tasks.
All NRC items required to be completed )riolr to exceeding l
l 5 percent power have been completed. Tie rpmaining open items Identified in the Inspection followup System are being tracked for corpletion at the appropriate time.
Atthistime,therearefouroutstandingalllegationsportainingtothe construction and operation of the CPSES facility. The Region IV Allegations Review Panel convened on March l11,1993, and determined that the resolution of the pending allegations was not necessary to support the issuance of a full-power license.
No additional allegations have been received for Comanche Peak since March' 11, 1993.
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