ML20058P393

NRC Staff Findings of Fact & Conclusions of Law in Form of Initial Decision.* Certificate of Svc

ML20058P393
Person / Time
Site:	Diablo Canyon
Issue date:	12/22/1993
From:	Hodgdon A, Jorgensen A NRC OFFICE OF THE GENERAL COUNSEL (OGC)
To:
References
CON-#493-14545 OLA-2, NUDOCS 9312270112
Download: ML20058P393 (150)

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{{#Wiki_filter:___ v /95 0 .r i i ):. h t. [ '93 DE 23 ft 8 :06 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMhBSSION ~iy_ a ; ;.. BEFORE THE ATOMIC SAFETY AND LICENSING BOAED "" "'id. ' ' In the Matter of ) ) Docket Nos. 50-275 OLA-2 PACIFIC GAS & ELECTRIC CO. ) 50-323 OLA-2 ) (Diablo Canyon Nuclear Power Plant, ) (Construction Period Recovery) Units 1 & 2) ) NRC STAFF'S FINDINGS OF FACT AND CONCLUSIONS OF LAW IN THE FORM OF AN INITIAL DECISION i l Ann P. Hodgdon Arlene A. Jorgensen Counsel for NRC Staff December 22,1993 9312270112 931222 1 PDR ADOCK 05000275 g PDR g i )

o -i-TABLE OF CONTENTS PAGE 1 I. INTRODUCTION................................... 1 II. BACKGROUND.................................... 1 1 III. STANDARD OF REVIEW AND GENERAL FINDINGS........... 4 IV. FINDINGS OF FACT................................ 10 Contention I: Maintenance & Surveillance........................ 10 ' Definition of Maintenance and Surveillance....................... 12 The Scope of PG&E's Maintenance and Surveillance Program............ 14 Management and Maintenance of Equipment Aging.................. 18 Conclusion on the Scope of PG&E's Maintenance & Surveillance Program.... 21 Implementation of PG&E's Maintenance & Surveillance Program....._..... 21 The NRC Inspection Program................................ 24 Issues Raised by MFP.................................... 28 A. Maintenance of Electrical Equipment that is Environmentally _ Qualified (10 C.F.R. I 50.49) ......................30 B. Check Valves / Inservice Testing...................... 35 C. Underground Cable Failures........................ 38 D. Wrong Motor Installed on MOV Actuator............... 41 - E. Storage and Handling of Lubricants................... 43

e: o -ii-EMiB F. Fuel Handling Building........................... 44 - 4 G. Tats of Containment Personnel Airlock................ 46 H. Component Cooling Water (CCW) Heat Exchanger....... . 49 I. Auxiliary Building Ventilation System.......:........... 50 - 1 J. Electrical Panel Covers........................... 52 K. Containment Equipment Hatch Gap.................... 54 L. Manual Reactor Trip Caused by Fuse Failure............. 55 - M. Limitorque 2-FCV-37. Test Failure.................... 56 - a N. Emergency Core Cooling System Accumulator Tanks........ 58 O. Corrosion of Underground Piping.................... 59 1 P. Control of Measuring and Test Equipment (M&TE)'......... 64 Q. Degraded Coupling on Centrifugal Charging Pump (CCP) . 66 i R. Inoperable High Pressure Turbine Stop Valve............. 67 S. Diesel Generator Failure to Achieve Rated Voltage.......... 68 T. Missed Surveillance Tests......................... 70 t U. Auxiliary Feedwater (AFW) Pump Test Procedure.......... 72 P V. Hold Down Motor Bolts on Centrifugal Charging Pumps.......- 73 f W. Reactor Coolant System (RCS) Imakage................ 74 X. Inoperable Reactor Cavity Sump Wide Range Ixvel Channel.... 76. f Y. Design Criterion Memorandum (DCM) Requirements........ 79 2. Isolated Pipe Support Snubber Damage................. 80 l AA. Gas Decay Tank Missed Surveillance.................. 82 t i i k r ~ E

l'l - iii - i PAGE l c i BB. Seismic Clips................................. 83 j i CC. Containment Fan Cooler Units...................... 84 DD. Debris (Housekeeping) Debris Issues.,................ 89 j EE. Steam Ccrem Feedwater Nozzle Cracking.............. '93 j i FF. Procedural Controls During Shot Peening............... 95 i t GG. Unplanned Engineered Safety Features (ESP) Actuations....... % HH. Limitorque Valve Failure ...........................'98 q i IL Motor Pinion Keys in Limitorque Motor Operators.......... 99 l JJ. Control of Lifting and Rigging Devices................. 101 KK. Main Feedwater (MFW) Pump Speed Probes............. 104 LL. Containment Ventilation Isolation (CVI) Signals........... 106 MM. Reactor Trip on Steam Generator I.ow level -............ 109 NN. Auxiliary Saltwater (ASW) Pump Crosstie Valve.......... 110 i 00. Testcock Valve on Diesel Generator.................. 113 ' PP. Main Feedwater Check Valve...................... 114 QQ. Auxiliary Saltwater (ASW) Pump Vault Drain Check Valves... 116 j i RR. SI-1-8805A Failed to Cycle on Actuation Signal.......... 117 SS. Fire in Electrical Panel.......................... 118 'IT. Chemical Volume Control System (CVCS) leakage........ 120 l Contention V: Thermo-Lag Compensatory Measures................ 122 l V. CONCwSION................................... 139 l 1 1 e 1

o e UNITED STArcS OF AMERICA l NUCLEAR REGULATORY COMMISSION-BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) ) ) Docket Nos. 50-275 OLA-2 PACIFIC GAS & ELECTRIC CO. ) 50-323 OLA-2 ) t (Diablo Canyon Nuclear Power Plant, ) (Construction Period Recovery) Umu 1 & 2) ) NRC STAFF'S FINDINGS OF FACT AND CONCLUSIONS OF LAW IN THE FORM OF AN INITIAL DECISION I. INTRODUCTION This is an initial decision on Pacific Gas and Electric Company's (PG&E) Application to amend the operating licenses for its Diablo Canyon Nuclear Power Plants Units 1 and 2 (DCPP or Diablo Canyon) to allow for 40 years of operation dated from i the ihuance of its operating licenses. j For the reasons discussed below, we find that there is reasonable assurance that PG&E's operation of DCPP in the recapture period will not adversely impact the public health and safety. We, thus, authod. c me Director of Nuclear Reactor Regulation to { 4 issue the license amendments. II. BACKGROUND On July 9,1992, PG&E submitted a license amendment request by which it j sought to extend the life of its operating licenses by more than 13 years (for Unit 1) and

o almost 15 years (for Unit 2) by " recapturing" the period spent in constructing the plants. l l The licenses, which are limited to a term of 40 years by section 103c of the Atomic Energy Act, 42 U.S.C. 5 2133(c), were issued consistent with a Commission policy under which that 40-year life extended from the date of issuance of the construction permit for a particular unit - for Unit 1, a term running from April 23,1968, to April 23,2008, and for Unit 2, a term running from December 9,1970, to December 9, l 2010. l In 1982, the Commission began issuing the 40-year operating licenses measured I l from the date of issuance of the operating license. It has also approved license amendments for many reactors conforming the earlier licenses to this new poFey. The Licensee is here seeking to amend its operating licenses to take advan%ge of the newer practice. As proposed, the extended expiration dates for Diablo Cuyon would be September 22, 2021, for Unit I and April 26,2025, for Unit 2. In response to a notice of opponunity for hearing on the proposed amendments (57 Fed. Reg. 32,575 (July 22,1992)), San Luis Obispo Mothers for Peace (MFP) timely filed a request for a hearing / petition for leave to inten'ene. This Licensing Board was established to rule on the request / petition and to preside over the proceeding in the event that a hearing was ordered. 57 Fed. Reg. 43,035 (Sept.17,1992). After a prehearing conference held in San Luis Obispo, California, on December 10, 1992, in which we heard argument concerning MFP's petition and the Supplement, in which MFP set fonh its proposed contentions, together with PG&E and i

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i' i l. the NRC Staff's responses, we granted MFP's petition for leave to intervene and request i l for a hearing. 37 NRC 5 (1993); see also LBP-92-27, 36 NRC 1% (1992). - -l 4 ? l , We admitted two contentions, MFP's Contention I and ' Contention V.6 37 NRC 5. - l 14-21, 26-28. MFP submitted nine other contentions concerning personnel errors, .i { counterfeit parts, handling. of hazardous materials, _ spent fuel" storage, emergency. I preparedness, the Staff's proposed no significant hazards consideration" finding, and the i j need to meet the requirements of the National Environmental Policy Act of 1969: (NEPA). Thew contentions were rejected as lacking in basis, as precluded bp law or as premature. 37 NRC 5 at 21-25,28-36. - L i l Subsequent to the issuance of our prehearing conference order, MFP filed three i late-filed contentions. We rejected 'all of them as' lacking in basis. LBP-93-9, - .l 37 NRC 433 (1993). We held hearings for seven days, August 17-24, 1993, in San Luis Obispo, California. Both PG&E and the NRC Staff prefiled direct testimony according to the. j Commission's regulations in 10 C.F.R. I 2.743 and according to the schedule established by this Licensing Board. MFP did not prefile direct testimony bot introduced some two - hundred documents at the hearing, whie.h MFP has used as the' basis of its proposed findings in which MFP urges us to find in its favor and deny the amendments. For the. reasons discussed below, we decline to do so. Prior to MFP's challenge to PG&E's application, only one such amendment. request had ever been challenged. See Vennont Yankee Nuclear Pbwer Corp. (Vermont Yankee Nuclear Power Station), LBP-90-6,'31 NRC 558,564 (1990). 'Ihat challenge i

o o was, however, withdrawn prior to a hearing as a result of a settlement. Thus, this was the first such hearing on a construction period recovery. As we stated in LBP-93-1, operating license hearings are conducted as paper programs. 37 NRC at 19. In contrast, the hearing held in August concerned the implementation of those programs. PG&E and the NRC Staff introduced the testimony of expert witnesses laoking at the surveillance and maintenance programs as a whole. These witnesses testif2ed that those programs are well conducted and support the license amendments sought. MFP relied on documents, largely identified on the eve of hearing or during the hearing, to demonstrate through cross-examination that these programs are deficient and do not support the license amendments. In preparing these findings, we have closely examined the documents introduced by MF? as exhibits in this proceeding, as well as the expert testimony introduced, and find that cumulatively the documents evidence a well conducted surveillance and maintenance program and support the testimony of PG&E and the Staff's witnesses. III. STANDARD OF REVIEW AND GENERAL FINDINGS i 1. While the licensee or applicant in an NRC proceeding has the ultimate burden of persuasion, an intervenor has the burden of going forward on the issues raised by its contention. See Philadelphia Electric Co. (Limerick Generating Station, Units 1&2), ALAB-262,1 NRC 163,191 (1975); see also Vermont Yankee Nuclear Power Corp. v. N.R.D.C.,435 U.S. 519,553-54 (1978). This burden can be sustained either by direct evidence or by means of cross-examination. Tennessee Valley Authority i l (Hartsville Nuclear Plant, Units 1 A,2A,1B & 2B), ALAB-463,7 NRC 341,356 (1978). 4

~_ 4 l The mere admission of a contention into a proceeding does not shift that burden of going i 1 forward, and the intervenor must still come forward with sufficient evidence to warrant ' j further consideration of the matter it has raised. Metropolitan F4ison Co. (Ihree Mile 1 i l Island Nuclear Station, Unit 1), ALAB-722,19 NRC 1193,1145 (1984), irwised inpair l on other grounds, CL1-85-2, 21 NRC 282 (l985); Public Service Co. ofN:.e llanpshire. .j ~ (Seabrook Station, Units 1 & 2), LBP-83-20A,17 NRC 586,589 (1983). Ql MFP Proposed Finding 13. l 2. MFP has set forth many technical analyses, opinions and conclusions in its proposed findings. See, e.g., MFP Proposed Findings 19-61. However these findings ] 1 may not be adopted, for technical analyses,' opinions, and conclusions in NRC i proceedings must be sponsored by experts.who can testify to the soundness of the-i 4 conclusions and opinions set forth. Duke Power Co. (William B. McGuire Nuclear l 1 3 Station, Units 1&2), ALAB-669,15 NRC 453,477 (1982); Southern Cahfornia Edison } j Co. (San Onofre Nuclear Generating Station, Units 2&3), ALAB-717,17 NRC 346,367 A (1983). Where experts do not appear for cross-examination to support conclusions, these 1 1 j conclusions may not be accepted. Louisiana Power and Light Co.- (Waterford Steam 4 Electric Station, Unit 3), ALAB-732,17 NRC 1076,1088 n.13 (1983). Thus, where i 1 MFP has drawn conclusions in its proposed findings on the basis ofits interpretation of the documents or generalizatic fmm several documents, these findings may not be adopted by the Board in that they were not sponsored by experts who could be cross-examined as to the soundness of their conclusions.

. ') 3. MFP has highlighted virtually every statement in PG&E's self examination 'l of its operations in nonconformance reports (NCRs) and Licensee Event Reports (LERs) i and in the Staff's inspection' reports capable of a: negative interpretration. Having - l a examined these reports we are convinced that it is better to overreport and raise questions. l than to underreport and nct manifest an inquiring attitude. We think it is important that j i we not create or even condone a process that penalizes a licensee for trying to improve j 1 its performance where that performance is already good. De documents, as a whole,- and the testimony show strong and well functioning. surveillance and maintenance programs at Diablo Canyon where the put,Isiis are squarely faced and solutions sought. ] The documents on which MFP has sought to rely, as the experts testified, support the l grant of the license amendments sought. i 4. Further, MFP's generalizations are often without any factual predicate in - q the record. For example, the first specific' matter MFP addresses in regard to Contention - i I on surveillance and maintenance programs is an alleged " Reduction in Safety Margins." \\ MFP Proposed Findings at 13. To support this generalization, MFP points primarily to'. 1 l l Limitorque valve 2-FCV-37, which it states was inoperable for some period between l 1990 and 1993, and three Containment Fan Cooling Units (CFCUs), which it states were inoperable for almost a year. MFP Proposed Finding 27. However, when the MFP proposed finddg on the Limitorque valves (MFP Proposed Finding 261) is examined, there is no citation to show the hoperability of the valve for that period, and the record establishes timi the CFCUs were never inoperable. - NRC Staff Testimony at 8 (Narbut); NRC Staff Exhibit 2. Dus the conclusions and generalizations in MFP's " General l

. i Findings Regarding Contention I" (MFP Proposed Findings 20-61) not only lack any evidentiary support for the generalizations that are made, but also lack a factual basis.

.5. In this pr -x+*.g, as in all other NRC proceedings, written testimony-must be served on all other parties fifteen (15) days prior to the commencement of the hearing. See 10 C.F.R 6 2.743(b). Exceptions may be made, in' the' discretion of the presiding officer, only where all parties consent to its introduction or those parties have had a reasonable opportunity to examine the evidence. :14. Although' the regulation' speaks in terms of written testimony, it applies as well to exhibits, and it would be an-error to premise conclusions on documents that were not given'to other parties'in l sufficient time for them to digest their meaning, to ascertain whether rebuttal evidence l was needed and to prepare examination on the document. _ Tennessee Valley Authodty I 1 (Hartsville Naclear Plant, Unit IA, IB, 2A and 2B), ALAB-367, 5 NRC 92,117-8 (1977). The Board provided that material on which a party was to base its case'was to be served sufficiently in advance of the hearing to' allow other parties to prepare. LBP-93-1, 37 NRC 5, 21 (1993); see also Memorandum and Order (Notice of Prehearing Conference and Evidentiary Hearing), July 8,1993. This was particularly important with regard to documents on which MFP would rely, as MFP did not introduce any direct testimony, but pr-W only by means of cross-examination. See Public Service Co. ofNew Hampshire (Seabrook Station, Units 1 and 2), LBP-83-20A,17 NRC 586,589 (1989) (duty to reveal documents relied upon before hearing when proceeding only by j cross-examination). Where other parties do not receive those documents in ample time for them to prepare, they are prejudiced. See Penarylwmia Power and Light Co.. .l I

. 1 (Susquehanna Steam Electric Station, Units 1 and 2), ALAB-613,12 NRC 317,338 (1980), quoting Nonhem States Power Co. (Tyrone Energy Park, Unit 1), LBP-77-37,= 'l a 5 NRC 1298,1300-01 (1977); see also Metropolitan F4fson Co. (nree Mile Island l Station, Unit 1), ALAB-772,19 NRC 1193,1245 (1984), rev'd inpart on other grounds, _j i CLI-85-2, 21 NRC 282 (1985) (necessity of intervenor revealing testimony. and i documents supporting its case before applicant presents its case). j i 6. In this proceeding MFP, to a very large extent, attempted to make its case - on the basis of documents it did not supply in advance of hearing as required by the ' -i orders of the Board and the Commission's regulations. Rather MFP identified these i documents on the eve of hearing or during the course of the hearing. Such documents-l and cross-examination based on those documents may not be the basis of findmgs herein. l Hartsville, supra; Susquehanna, spra; Seabrook, spra. -l 7. Although these actions >um MFP's case, the situation surrounding the issue of " Maintenance of Environmental Qualification of Electrical Equipment" was particularly egregious. See MFP Proposed Findings 62-94. There MFP cross-*==iw ] and relied on an exhibit composed of Telatemp stickers (MFP Exhibit T-4), but did not identify that it intended to rely on that mauial until the very moment of its introduction.' .i Under Commission case law, no relimce may be placed on MFP's reading of these i documents or conclusions it draws from the documents. See MFP Proposed Findings 77-94. 8. Similarly, in regard to most of the other equipment on-which MFP proposes findings, the documents on which MFP relies were not identified by MFP in ' l

i i -9' j i i i i sufficient time for other parties to prepare testimony on the documents, and no reliance j may be had on those documents or on the conclusions MFP would seek to draw from i those documents in this proceeding. See Hartsville, sqprn; Seabrook, supra.x ) i 9. MFP claims that PG&E and its witnesses cannot be credited because they represent an industry point-of-view. See MFP Proposed Finding 24. However, the value ' l of the testimony of a witness is not undermined by his being a hired consultant or i employed by a party. ' Metropolitan FAson Co. (Ihree Mile Island Nuclear Plant, b Unit 1), ALAB-772,19 NRC '1193,1211 (1984), rev'd in part on asher grounds, - CLI-85-2,21 NRC 282 (1985); Louisiana Power & Light Co. (Waterford Steam Electric Station, Unit 3), ALAB-732,.17 NRC 1071,1091 (1983). Moreover, these witnesses were subject to cross-examination at the hearing, as were the Staff witnesses. The' 1 conclusions in their testimony on the adequacy of PG&E's maintenance and surveillance { programs was not wanaamd in cross-examination. 10. On the other hand, MFP, which represents a different point-of-view m. regard to the continued operation of Diablo Canyon, did not present any witnesses for j) cross-examination. It seeks to make its case in its proposed findings. through I generalization of what it concludes selected. documents show..However, these 1 ( interpretations and generalizations by an organization that is also interested in a particular 1 result in this proceeding were not subject to testing at hearing, and could not be adopted by the Board in view of the testimony by PG&E's experts and the NRC Staff that - PG&E's surveillance and maintenance programs were adequate. I

m i I1. MFP proposes. findings concluding that "PG&E. has demonstrated a repetitive pattern of personnel errors which jeopantize the safety of the plant." MFP Proposed Finding 53. These findings may not be adopted. : At the pid.sring conference in this proceeding a contention on " personnel errors" was rejected. See 37 NRC at 22-

23. Further, these findings are based upon conclusions, technical judgments,' op' ions m

and generalizations that have no support and were not subjected to cross-examination. ~ l IV. FINDINGS OF FACT Contention I Maintenance & Surveillance .j I-1. . Contention I, as admitted by the Board 8 reads as follows: l I - The San Luis Obispo Mothers for Peace contends that Pacific Gas and { Electric Company's proposal to extend the life of the Diablo Canyon l Nuclear Power Plant for more than 13 years (Unit 1) and almost 15 years. l (unit 2) should be denied because PG&E lacks a sufficiently effective and ' l comprehensive surveillance and maintenance program. l I-2. Testimony in this pranmiing was presented by a panel of witness from~ j PG&E consisting of: Bryant W. Giffin, Manager of Maintenance Services (DCPP); William G. Crockett, Manager of Technical and Support Services (DCPP); David A. Vosburg, Director of the Work Planning Section, Maintenance Services Dv, parts, cat (DCPP); Steven R. Ortore, Director of the Electrical Maintenance Mian, Maintenance l i Services Department (DCPP); Tedd Dillard, Supervisor of Component Programs for the j l l l 8 Prehearing Conference Order (Ruling Upon Intervention Petition and Authorizing' Hearing), LBP-93-1,37 NRC 5,14 (1993). l \\ 1 .1

I I f - II - l 4-l [ Nuclear Division of Florida Power & Light Company; arJ David B. Miklush, Manager i of Operational Services (DCPP).2 4 I-3. Testimony was also presented by a' panel from the NRC Staff consisting: a ] j of: Paul P. Narbut, Regional Team Imder, Region V, Division ~of Reactor Safety and i Projects; Mary H. Miller, Senior Resident Inspector (DCPP), Region V; and Sheri R. Peterson, Senior Project Manager '(DCPP), Office of Nuclear Reactor Regulation.' l -I-4. Mothers For Peace (MFP) filed no testimony and presented no witnesses 1 in the pK-: Mag, but instead chose to rely on the introduction and cross-examination of ) numerous exhibits consisting of PG&E's internal Nonconformance Reports (NCRs), Licensee Event Reports (LERs) filed with the NRC, PG&E wri++;-:- '- -re w'8 $e 3 ) 1 i NRC, and NRC StaffInspection Reports and Notices of Violation.: See Tr.- 576-79. Tae e majority of these exhibits were not linked to the prefiled ' testimony of'either PG&E' or ~ i the NRC Staff on Diablo Canyon's maintenance _and surveillance program, but were l documents alleged by MFP nevertheless to' be relevant to PG&E's maintenance and surveillance program. MFP failed to' produce any' competent technical experts to - establish that these documents evidenced any breakdown in PG&E's maintenance and ~ surveillance program or that the documents could form a basis for a showing that these ~ programs were inadequate. 2 Testimony of Pacific Gas and Electric Company Addressing Contention 'I: Maintenance and Surveillance, admitted but not bound in, Tr. 590. 3 NRC Staff Testimony of Paul P. Narbut, Mary H. Miller land Sheri R. Peterson Regarding Contention 1: The Surveillance and Maintenance Program at Diablo Canyon, : ff. Tr. 2159. 4 ,. n.a -.4, n

I 1 I i i I 4 Definition of Maintenance and Surveillance - 'I I-5. Maichiice is defined as the work of keeping something in suitable'- i condition. Tnere are two kinds of maintenance: preventive and corrective. Preventive l mamtenance is regularly scheduled work puformed on structures, systems or components j a l (SSCs) that keeps failures from occurring due to predicted component degradation. i Industry-wide operating experience is often taken into account in determining what type j preventive maintenance is -=y and how often it should be performed. Corrective maintenance is performed after a failure occurs, or a component exhibits degraded l t capability. Surveillance tests are conducted to identify failures or degraded performance that needs to be w,nwic4 prior to a system being called upon to perform a safety i function. NRC Staff Testimony, ff. Tr. 2159 at 2. I-6. The NRC's maintenance rule,10 C.F.R. I 50.65,' does not become effective until 1996 and, therefore, does not yet apply to nuclear plant licensees. However, the legal requirements governing the implementation of maintenance and i surveillance programs at Diablo Canyon in the interim are found or referenced in the Technical Specifications for Diablo Canyon and 10 C.F.R. Part 50, Appendix B.' PG&E i Direct Testimony at 21; Tr. 2280 (Narbut). PG&E has also' committed in.the Final i Safety Analysis Report Update (FSAR) for Diablo Canyon to follow the standards of ANSI 18.7-1976/ANS-3.2, " Administrative Controls and Quality' Assurance for the - Operational Phase of Nuclear Power Plants," as endorsed and modified by the NRC in I Regulatory Guide 1.33, " Quality Assurance Program Requirements (Operation)." PG&E Direct Testimony at 21. i i I 4 h l l

i i a I-7. The NRC Staff uses the 18 quality assurance criteria in 10 C.F.R. Part 50, Appendix B as the foundation of the inspections of the plant.L Tr. 2280 (Narbut).' To 1' determine the adequacy of surveillance and maintenance at various plants, the Stafflooks at performance based criteria, such as work 'in progress, the procedures, and j j qualifications of personnel. Id. I-8. 'Ihe Commission has specifically declined to formally adopt the standards' l set forth by the Institute of Nuclear Power Operations (INPO) in its docunr.i 90-008, entitled " Maintenance Programs in the Nuclear Power Industry" (Reviraon 1, March i 1990);* however, this document was introduced by MFP. (MFP Exhibit 4). This i document was also referenced by PG&E in its direct testimony and was extensively examined in the record. PG&E Direct Testimony at 23; Tr.1488 (Dillard). We decline to adopt the standards stated therein but we do find it useful in looking at the elements of a comprehensive maintenance program. I-9. Some of the INPO 90408 elements that 'are addressed in PG&E's maintenance and surveillance program include a competent and qualified organization; a mix of preventive and corrective maintenance to ensme equipment degradation is identified and corrected prior to failure; accurate procedures or work instructions; maintenance planning and scheduling; post-maintenance testing; a detailed root cause analysis program to understand the cause of equipment failures; and 'a maintenance history program to provide historical data for maintenance planning. PG&E Direct Testimony at 24; See also MFP Exhibit 4.~

• 54 Fed. Reg. 50,611 (December 8,1989).

=

. I-10. PG&E defines maintenance at Diablo Canyon as the aggregate of actions that (1) minimizes the degradation or failure of systems, structures, and components ("SSCs"), and (2) promptly restores the intended function of SSCs if they experience operability or functional problems. PG&E Direct Testimony at 4. " Surveillance" at DCPP is the aggregate of periodic tests and/or inspections which verifies that SSCs continue to function in accordance with predetermined specifications or are in a state of readiness to perform their particular safety functions. Id. at 5. The Scooe of PG&E's Maintenance and Surveillance Procram I-l 1. Maintenance and surveillance at DCPP encompasses aspects of several activities at the plant including: (1) surveillance required by Technical Specifications; (2) equipment surveillance not required by the operating license; (3) Inservice Inspection (ISI) and Inservice Testing (IST) Programs; (4) the Environmental Qualification (EQ) Program; and (5) the Maintenance Program. PG&E Direct Testimony at 10. 1-12. Surveillance testing required by DCPP Technical Specifications has been developed and implemented by PG&E in accordance with the industry standard ANSI N18.7-1976/ANS 3.2, " Administrative Controls and Quality Assurance for the Operation Phase of Nuclear Power Plants." PG&E Direct Testimony at 11. Such surveillance i testing is administratively controlled in accordance with PG&E procedure NPAP-C3, " Conduct of Plant and Equipment Tests." Id. Technical Specification surveillance testing at DCPP is done to assure that safety-related equipment failures or substandard i equipment performance will not remain undetected. Id. at 12. 4

-. - - _ ~ -- -. -. l ! j. l I-13. In addition, there are numerous plant activities besides1 Technical i- -Specification-required surveillance test procedures that continuously provide current-i i information about the condition of SSCs at DCPP. 'Ihese include routine plant operator 4 equipment in= pac +ians or walkdowns, predictive maintenance program testing, preventive f maintenance program inspections, pmcedure functional checks, performance tests of i l equipment not controlled-by Technical Specifications, erosion / corrosion monitoring, post-maintenance testing, and system engineer walkdowns. PG&E Direct Testimony at 12-14. Maintenance tasks are scheduled and performed on the' basis of information' derived from these activities in order to maintain equipment performance at the required ~ level for the life of the plant. Id. at 14. I-14. 'Ihe ISI and IST programs meet the requisements ' of J 10 C.F.R. Il 50.55a(b)(2) and 50.55a(g), as well as' plant Technical Specifications, and include inspection, testing and maintenance of pressure-retaining components as required by the American Society of Mechanical Engineers (ASME) and the Boiler and Pressure Vessel (B&PV) codes. PG&E Direct Testimony at 14.' Commission regulations also require revision of ISI and IST Programs as==y to comply (to the extent practical within the limitations of design, geometry, and materials for construction of components) with the edition of the ASME B&PV Code and Addenda in effect and' adopted by the NRC twelve months prior to the start of each ten year inspection interval. 10 C.F.R. I 50.55a(g); PG&E Direct Testimony at 15-16. These programs are in place to ensure that pressure-retaining components will be adequately inspected,- tested,' and maintained throughout the term of the 40 year epwding license. Id. at 16.'

. 1 I-15. Post-maintenance and post-modification testing (PMT) also is encompassed within the scope of PG&E's maintenance program. The primary objective of PMT is to ) ensure that all plant equipment that has undergone maintenance or modification has been demonstrated to be fully functional or operable prior to being returned to service. PG&E i Direct Testimony at 51. PMT at DCPP consists of two types of testing: (1) Maintenance. or Modification Verification Tests performed by the implementing organization without l actually operating the equipment; and (2) Operability Verification Tests to prove Technical Specification operability. Id. at 52. Documented completion of the required PMT is an essential element of the equipment control process used by the Operations Department when returning equipment to service. Id. at 53. I-16. The planning and scheduling of maintenance tasks at DCPP is part of a work control process which ensures that maintenance activities (including both preventive and corrective maintenance tasks) are planned and performed in a safe, timely, efficient j and controlled manner. PG&E Direct Testimony at 43. Administrative controls for the work control process at DCPP are integrated into the computerized Plant Information Management System (PIMS), thereby making the data available on a plant-wide basis and ensuring that maintenance tasks receive appropriate levels of review and are tracked through final resolution. Id. at 44-45. 1-17. PIMS is a computer-based system at DCPP which provides integrated access to up-to-date information on plant component history, maintenance task instructions and history, problem reports and status, inventory control, and radiation exposure tracking. PG&E Direct Testimony at 20-21. It also provides a means by which l l l

~. i j a .e 17 - all porwiusel working at the plant can document plant equipment problems and request interdepartmental support. Id. at 44. PG&E uses a. reporting' system to document failures or degradation of structures,. systems or components consisting of nonconformance reports'(NCRs), which are a mechanism for' personnel to document certain deficiencies at Diablo Canyon, initiate corrective action and establish a completion. schedule for resolution of a nonconformance. Staff Direct Testimony, ff. Tr. 2159 at 3. In Diablo Canyon procedure OM7.ID3, PG&E defines a nonconformance as: A quality _ problem that constitutes a significant condition adverse to' quality. Id. 1-18. To be classified as a nonconformance, the quality problem must satisfy one or more of eight criteria outlined in the procedure. The criteria include NRC violations, programmatic or implementation breakdowns, design deficiencies,' defects, frequently recurring and NRC reportable events. If the problem identified is determined to be a nonconformance, the nonconformance report (NCR) documents the event description, - root cause determination, safety analysis, and action taken to correct the nonconformance and prevent recurrence. Staff Direct Testimony at 3. 1 I-19. A detailed root cause analysis program provides for the systematic analysis of unplanned occurrences pertaining to maintenance. PG&E Direct Testimony at 59. i This root cause analysis program is controlled by Procedure NPAP C-26, " Root Cause Analysis," and provides guidance in several techniques including cause and effect analysis, change analysis, event and causal factors analysis, barrier analysis, and human j 1 factors surveys. Id. PG&E also tracks component histories as part of root cause analysis and component failure trending at DCPP. PG&E Direct Testimony at 60. There are l . -a e

I approximately 187,000 components in the DCPP Component Data Base, each with its own maintenance history available in PIMS. Id. at 60-61. Component experience also is available from an industry-wide database, the Nuclear Plant Reliability Data System (NPRDS), maintained by INPO. Id. at 60. I-20. Licensee event reports (LERs) are submitted to the NRC pursuant to 10 C.F.R. 6 50.73. Staff Direct Testimony, ff. Tr. 2159 at 3. The threshold for requiring an LER is higher than the threshold for an NCR. Typically, the Licensee issues anywhere from 60 to over 100 NCRs each year compared to 20 to 30 LERs each year. Id. While NCRs are used internally at Diablo Canyon, LERs are submitted on the Licensee's public docket and used by the NRC for trending purposes and identifying significant events. Both reports are a means for licensees to document self identified problems. Id. Management and Maintenance of Eauioment Aging I-21. Management of equipment aging is inherent in many of the maintenance and surveillance activities discussed above. However, PG&E has also initiated other l specific programs expressly directed at aging management issues. PG&E Direct Testimony at 62. Among these programs and activities that serve to mitigate the effects t of age-related degradation at DCPP are the following: the preventive, predictive and corrective maintenance programs; surveillance test programs; fatigue monitoring; the EQ l program; the Reactor Vessel Embrittlement Management Plan; the Motor Operated Valve (MOV) testing and evaluation program; the Steam Generator Strategic Management Plan; the erosion / corrosion program; and the structural monitoring program. Id. at 63-64.

n 19 - j 't 1 Each of these programs and activities produces. specific results or corrective actions to-W maintain and/or restore equipment to its required performance level sick.cr the aging scunwd prior to or during plant operation. Id. at 64. 1 I-22. Since beginning plant operation PG&E has also made a number of major-- plant modifications to improve reliability or upgrade safety-related equipment that also help minimize the effects of age-related degradation on the plant over its operating life, j i f.e, the 40 years contemplated in the FSAR. PG&E Direct Testimony, at 65-69. Some j of these modifications include the following: copper removal, iWing rd-t of j all feedwater heaters and retubing of all moisture separator reheaters;= addition of a. i Condensate Polisher System; steam generator blowdown rate increase;: fuel design -l l } improvements; removal of Boron Injection Tanks; reduction of the boron concentration in the Boric Acid System; installation of a digital Feedwater Control System; Chlorination 1 System modifications; and installation of an on-line fatigue monitoring' system. Id.- i I-23. PG&E also has established an aging management program pursuant to l .i Program Directive TS1, " Plant Aging Management," that addresses age-related l t degradation over the course of 'the plant's 40-year operatingLlife. - PG&E Direct l Testimony at 70. The aging management program collects data from'new research findings, industry operating experience, the NRC, the Electric Power Research Institute (EPRI), and vendors, for inclusion in appropriate programs. -14. I-24. PG&E's aging management activities also include several special maintenance pmgrams which have been established to monitor and manage certain critical l components subject to complex aging mechanisms, as1well as certain designated + -- +r,. -.<-c>.- .. ~ - -,, ,wst--e,...y.,5 ,., -v y, --r v e c he r t u v y-9 p - n -f-N.* y e r m e v y

. components with a limited life. PG&E Direct Testimony at 72. For example, steam generator tube degradation is monitored and managed by careful chemistry control during operation and by an extensive cleaning and inspection program during each refueling outage. Id. I-25. The reactor pressure vessels at Diablo Canyon are also addressed by special maintenance programs. The DCPP Reactor Vessel Radiation Smveillance Program is designed to monitor changes in material and mechanical properties of DCPP reactor pressure vessels (RPVs) in order to ensure their continued safe operation throughout the 40-year operating life of the plant. PG&E Direct Testimony, at 75. Compliance with all NRC regulations governing RPV integrity has been documented in PG&E's response to Generic Letter 92-01. Id. at 76. I-26. Erosion / Corrosion ("F1C"), which refers to the process of wall thinning in susceptible piping or other pressure boundary components caused by the flow of water or wet steam, is a normal part of the nuclear power plant aging process. PG&E Direct Testimony at 77. The management of E/C is an integral part of maintenance at DCPP. Measures to control E/C at DCPP include the replacement of certain piping with E/C resistant material such as stainless or chrome-moly steel. Id. I-27. The record also indicates that PG&E is managing the aging of passive, long-lived structural concrete and steel at DCPP. Conditions such as spalling or cracking of concrete, corrosive or caustic attacks from leaks, spills, or exposure to the environment, mechanical damage, and rust are routinely identified and reported by plant personnel. PG&E Direct Testimony at 81-82. PG&E's maintenance work at the intake

j l structure is an example of this process. Tr. 1737-38 (Giffin).. This type of maintenance i will assure that these structures will perform their intended functions for the life'of the plant. Tr. at 1741-42 (Giffin). For safety-related structures, functional surveillance j requirements are specified in Technical Specifications. PG&E Direct Testimony at 82. Periodic surveillance testing verifies the' operability of these structures. Id. Conclumian on the kana of PGAF's Maintenance & Surveillance Pmeram. l I-28. Based on the foregoing, it would be difficult for the Board to find that the 1 l scope of PG&E's maintenance and surveillance program is not sufficiently comprehensive j to support the expected 40 year life of the plant. Indeed, we find, rather that PG&E has - j addressed the areas set forth in the Commission regulations at-10'C.F.R. Part 50,. Appendix B, and in INPO 90408,~and as required by. the Technical' Specifications. Nothing in the record before us seriously calls into question the scope of.PG&E's a program. Most of the exhibits introduced by MFP run, rather, to the implamantation of the maintenance and surveillance program at Diablo Canyon', an issue we address below. l 1 Implementation of PG&E's Maintenance & Surveillance Prorrmm -

I I-29. The DCPP Maintenance ' Program has. been implemented through procedures which incorporate information from the Technical Specifications, design basis criteria, industry standards, equipment vendor and manufacturer recommendations, NRC.

Safety Evaluation Reports, and PG&E maintenance experience at both nuclear and fossil plants. PG&E Direct Testimony 'at 18-19.~ 'Ihe Maintenance Program includes' maintenance tasks, categorized as either preventive or corrective in nature, on both: safety-and non-safety-related SSCs. Id. at 19. i l i ..n

. I-30. The effectiveness of maintenance of a nuclear power plant is demonstrated 4 partly by the overall safe and reliable operation of the plant over time. Nuclear power i plants with consistently high capacity factors, long continuous operation, and short refueling outages are usually the best maintsined plants. PG&E Direct Testimony at 156, 163. This is because reliable and continuous operation at high capacity factors between scheduled refueling outages and across many operating cycles cannot be sustained unless plant equipment and facilities are effectively maintained. Tr.1493-94 (Dillard); Tr. 2091-92, 2111-12, (Giffin); Tr. 2094-95, 2112 (Miklush). 1-31. Moreover, effective maintenance is a necessary element in the achievement of high operating performance by a nuclear power plant. This is because a well-maintained plant will rur. well, and a high capacity factor is indicative of a plant that is running well. Tr. 2112 (Giffin). Operating performance cannot be achieved over long periods of time without effective maintenance af important equipment. Tr. 2112 (Miklush); Tr. 2266 (Miller). I-32. Excellent plant operating performance is also an indicator of an effective surveillance program. This is so because plant Technical Specifications require licensees periodically to perform thousands of surveillance tests on equipment important to safety. Many of these Technical Specification-mandated surveillance tests require the plant to commence shutdown within a relatively short period of time if the test does not demonstrate the equipment to be fully operable. Thus, the plant's surveillance program must maintain and test safety-related equipment in a well-coordinated manner at a very high level of performance to assure that overall plant operating performance is not

i-i ; adversely affected due to a Technical Specification-mandated; power reduction or shutdown. PG&E Direct Testimony at 165-66. I-33. The effectiveness of maintenance is also demonstrated by the short duration ~ of planned refueling outages, coupled with event-free operation following such outages. This is because the large scope of maintenance activities, performed'during planned outages requires comprehensive training of maintenance pernnnel and thorough planning j and execution of maintenance tasks during the outage. PG&E Direct Testimony at-166-67. I-34. The length of time required to demonstrate that high operating performance has been sustained varies from plant to plant. However, high operating performance for Li more than one or two fuel cycles for each unit of a plant after the first five years of - startup operation is strong evidence that the licensee's maintenance program is effectively. I i assuring that equipment-is maintained at a high. level of safety 'and ' operability. i ? Tr. 2091-92 (Giffin); Tr. 2092, 2094-95 (Miklush). l I-35. Evidence in the record demonstrates that DCPP has sustained high ~ operating performance, as indicated by operating capacity factors and short refueling outage durations, over its eight year operating history. PG&E Exhibits 14, 15, 16;- Tr.1493-94 (Dillard). In addition, the record demonstrates that this excellent overall l l operating performance has increased, rather than decreased, in recent years as the plant I has evolved into a mature operating plant. PG&E Exhibits 14, 15, 16. i i I l ~ l l

O e The NRC Inspection Program I-36. Another measure of the effectiveness of the maintenance and surveillance program at Diablo Canyon is to look at the NRC inspection program for maintenance and surveillance at Diablo Canyon, which consists of several elements. First, there are the monthly inspections performed by the resident inspectors. Procedures require the resident inspectors to observe surveillance testing in a unit each month and to observe a surveillance test in detail every other month. Hence, resident inspectors observe about 48 maintenance activities a year and about 18 surveillance tests. Staff Testimony, ff. Tr. 2159 at 4. I-37. Second, the inspection program involves regional specialists performing required (core) inspections in such areas as inservice examinations (examining a portion of the important welds periodically) and inservice testing (examining the testing as required by the American Society Mechanical Engineer (ASME) codes adopted by the NRC at 50.55(a)). Also, the NRC has a van which performs independent nondestructive examinations, such as X-ray of welds. This van and the NRC's chief nondestructive examiner examined Diablo Canyon's work in 1987 and found their nondestructive i examination programs sound. Id. I-38. Third, regional and NRC headquarters inspectors performed special i inspections in specific areas of interest such as erosion / corrosion of piping, motor operated valves, procurement, environmental qualification, and the 6th diesel generator installation. Id. at 5.

L w I j I-39. Fourth, NRC management has periodic meetings with PG&E management. l These meetings often deal with maintenance and surveillance issues'such as a meeting j held in July 1993 on steam generator tube inspection methods and plans. Id. I-40. Finally, periodically there have been team' inspections. These are multi-i discipline teams of inspecivis looking at a specific topic in depth. There was a team - inspection that===minad maintenance at Diablo Canyon in 1988. The last team at Diablo i evaminaA shutdown risk and was led by an NRC headquarters specialist. The team found ' l PG&E's activities to be sound.14.= j i I-41. The performance of maintenance and surveillance at Diablo is considered to be superior and clearly supportive of safe facility operation. Staff Direct Testimony: i ff. Tr. 2159 at 5. PG&E's performance has been, at worst, good and has improved over the years. Gradual trends over the past seven years show'a reduction in the number of I equipment failures, reduction of safety significant problems, increased management involvement in maintenance issues, and more timely identification and resolution of l problems. Id. I-42. While there have been problems associated with Diablo Canyon mainte-nance and surveillance in the past, these problems do not show any breakdown in Diablo j Canyon's surveillance and maintenance program. Staff Direct Testimony, ff. Tr. 2159 at 6. Problems have been identified in NRC inspection reports, licensee event reports,- and management meeting reports. For exaraple, improper maintenance of the j i containment fan cooler units (CFCUs) and the corrosion of the diesel fuel oil piping are two instances. Id. NRC follow-up of these problems nee <wi whether the Licensee had m..

. done a thorough job of understanding the cause and breadth of the problem and whether the Licensee had taken prompt and effective corrective action. Id. l I-43. MFP in its proposed findings on Maintenance and Surveillance has sought to generalize from individual deficiencies to concluc'e that PG&E's maintenance and surveillance programs are inadequate. See, e.g., MFP Proposed Findings 19-22. However, these are the conclusion of the MFP's lay persons and attorney, who prepared MFP's proposed findings, and were not supported by any witness, technical or not, at the hearing. There was thus no opportunity for the other parties to cross-examine on these conclusions. Because MFP's general conclusions on whether PG&E's maintenance and surveillance program were adequate lacked the support of any witness, and because these conclusions were not subject to testing at a hearing, they cannot be adopted by the Board. See McGuire,15 NRC at 477; San Onofre,17 NRC at 367; Waterford,17 NRC at 1088

n. 13.

1-44. Similarly, MFP's proposed general findings on " Failure or Unreliability of Important Safety Systems" (MFP Proposed Findings 25-30), " Untimely or Ineffective Response to Maintenance Problems" (MFP Proposed Findings 31-37), " Breakdown of Multiple Barriers" (MFP Proposed Findings 38-42), and " Repetitive Patterns of Failure" (MFP Proposed Findings 44-61), must be rejected. Leaving aside the paucity of citations and that most of the examples MFP cited do not show the failures claimed, these are again the generalizations of MFP's laypersons and attorney, without any competent technical expert to support these conclusions. Thus, in the face of contrary direct

e evidence in the record from PG&E's and the NRC Staff's expert witnesses, these findings cannot be adopted. I-45. PG&E has dealt with problems in the maintenance 'and surveillance areas effectively. Staff Direct Testimony, ff. Tr. 2159 at 6.. PG&E has corrected a great majority of the problems promptly. Their corrective actions generally addressed the specific problems as well as more generic concerns, if applicable. Many of the problems noted were initially identified by PG&E through their maintenance and surveillance program. The scope of corrective actions typically included changes to equipment, procedures, drawings, training, maintenance practices, or design documents. M. at 7. I-46. In cases where the Licensee's actions were untimely or inadequate,' the NRC has issued violations. In all cases, the NRC performs follow-up of the significant problems to ensure effective corrective actions were taken. - Progress continues in the maintenance and surveillance area as a result of PG&E's effectively resolving the problems noted. Staff Direct Testimony, ff. Tr. 2159'at 7. I-47. Additionally, the vast majority of corrective actions taken by DCPP are effective in dealing with maintenance problems. Tr. 2203 (Miller). Further, the Staff resident inspector pointed out that in many plants the NRC has a problem with lack of identification of problems, whereas at Diablo Canyon they identify things "to gruesome detail" and fix them. Tr. 2216 (Miller). All the NRC witnesses testified that DCPP's - maintenance and surveillance programs had elements of " openness" and " aggressive self-identification" which indicated that the programs were extremely healthy and effective. Tr. 2271-2 (Narbut); Tr. 2270-1 (Miller); Tr. 2274 (rmwn).

. Issnac Raimart by MFP I-48. MFP did not seriously challenge any of the prefiled testimony in this case, except in a few discrete areas, such as cable failures, control of measuring and test equipment (M&TE), and the containment fan cooling units (CFCUs). Rather, MFP chose to rely on the introduction of numerous documents including NCRs, LERs, NRC Staff Inspection Reports (irs), NRC Notices of Violation (NOVs) and corr =w= dance. See NRC Staff Proposed Finding I-4 above. I-49. According to MFP, these documents taken together signify that the1 maintenance and surveillance program at Diablo Canyon is deficient in that: inadequate maintenance has resulted in the failure or unreliability ofimportant safety systems; PG&E has shown a pattern of untimely or ineffective response to maintenance problems; these problems are the result of more than one error, resulting in the breakdown of the multiple barriers that should have prevented the problem; and PG&E has demonstrated a repetitive pattern of failures, thus indicating programmatic deficiencies. MFP Finding 21. However, such technical analyses must be sponsored by an expert whose. opinion is subject to examination on the reliability of factual assertions and the soundness of the opinion expressed. See Southern California Edison Co. (San Onofre Nuclear Generating. Station, Units 2 and 4), ALAB-717,17 NRC 346, 367 (1983). 'Ihus, these generalizations and conclusions cannot be adopted. - I-50. Moreover, we agreed at the outset in this proceeding that we would eschew playing a " numbers game" with regard to reported incidents represented by these documents. Tr. 599 0%chhoefer). Yet, that is ' precisely what MFP would have us'do,

i .\\ j l l i I i l by here addressing document after document some of which address the maintenance and surveillance program at Diablo Canyon and some of which do not. A large number of the documents address isolated and widely disparate component failures or penonnel errors that have been identified and curistod. Others evidence a maintenance and. surveilkm program that is, indeed, working, and working well. I-51. To put the matter in context, the record demonstrates that there are some 20,000 maintenance and surveillance work orders performed each year, and of these some 7,000 are corrective maintenance task at Diablo Canyon. PG&E Direct Testimony. l l at 40,43. Error-free maintenance is not a standard. Tr. 2275 (Peterson). Rather, the I NCRs and LERs in the record demonstrate PG&E's system for issue identification, root cause evaluation, issue resolution and application oflessons learned and that PG&E has a good tracking system with a good database. Tr. 2273-74 (Narbut). 1 I-52. The events described in the documents MFP sought to have. admitted-i apresent relatively few random and isolated occurrences that are not concentrated in any one aspect of the program. Rather, they have involved different equipment, systems, procedures, locations or operating conditions, and have occurred at random times in a. random manner. PG&E Direct Testimony at 113-14. I-53. Nevetheless, in spite of the fact that we see no pattern or pervasive' i breakdown in the Diablo Canyon maintenance and surveillance program, we address each i 1 l l l~ l l ~

) 1 1 : i of the exhibits relied upon by MFP seriatim to ensure completeness of the record before l J us.5 -i A. Maintenance of Fletrical Equipment that is Envimnmenta11v Onnlified I (10 C.F.R. 8 50.49) I-54. DCPP's Equipment Qualification -(EQ) Program complies with' the.., j requirements of 10 C.F.R. 6 50.49, which requires that electric equipment important to I safety located in a harsh environment be environmentally qualified. PG&E Direct l 1 Testimony at 17. DCPP's EQ program was evaluated and found by the NRC to be in ] conformance with applicable requirements in 1981 and 1985. Id. I I-55. Maintenance of EQ equipment is a formal program at DCPP. Id. Each-i i i item of EQ equipment has a calculated qualified life curispording to the time equipment-l l can operate under its normal, installed operating conditions and still be qualified for the j postulated post-accident environment, Id. at 79. EQ equipment is specified, designed and ] i fabricated for the anticipated service conditions. Id. To the extent EQ requirements I dictate particular maintenance requirements, these requirements are incorporated into the 1 maintenance practices for the specific equipment. Id. at 79-80. I-56. PG&E has established a program for temperature monitoring in connection i i with maintenance of EQ equipment.14. at 80. Maintenance Procedure (MP) E-57.8A ' provides guidance for the methods of specific device temperature monitoring to obtain ~ qualitative temperature information.14. In accordance'with this procedure, temperature i I 8 i We note that MFP did not make findings on all the documents that it introduced at the hearing. As MFP has abandoned these issues, we do not make findings on them. I -I I

1 i 1 indicating stickers are placed on various EQ devices to identify any equipment that may ] be exposed to temperature extremes higher than previously. considered for qualified life j purposes. These stickers identify momentary peaks and are'sometimes augmented by ' .j J continuous temperature iewiding devices. Id. 'I l I-57. After finding " hot spots" at which. temperatures were higher than previously anticipated, PG&E inaugurated its temperature monitoring program involving the use of "Telatemps," stickers that record a peak temperature by incw n.deg a r I chemical that changes color when that temperature is reached. Tr.1843,1845-462,2043 ' (Ortore). Iater, PG&E brought in Sargent & Lundy as consultants to assist in analyzing temperature data and in recalculating qualified life so that any =*-ry adjustments to the preventive maintenance program based on recalculated qualified life' could be made. Tr.1845 (Ortore). I-58. During refueling outages, maintenance personnel read and record on data sheets the data provided by the stickers, remove the stickers and affix them to the data sheets and attach new sticker., to replace those removed. Tr.1847 (Ortore). 1-59. MFP introduced into the record three documents related to PG&E's 1 maintenance of EQ equipment: MFP Exhibit T-2, PG&E Temperature Monitoring 'j Pmcedure (MP E-57.8A); MFP Exhibit T-3'(Sargent &~Lundy Engineering Report 1

1. 8664-03, "Effect ofIAx:alized High Temperatures Upon EQ Components," February 27, 1990); and MFP T-4 (Teletemp [ sic] Temperature Sticker Data, Units 1 and 2). To

• MFP T-4 consisted of Xerox copies of data sheets to which tb
• tickers that had; been removed from EQ equipment had been affixed together with the ar data as read i

and recorded by maintenance personnel. ..~a

E 4 ! l 1 i j show that the maintenance program for environmentally qualified electrical equipment' t t was not adequate. MFP had given no advance notice of its intention ~ to introduce these documents and any value they might have had as evidence was substantially reduced by j r the element of surprise that was associated with their use by MFP.- 3 1 4 i i I-60. The Licensing Board does not adopt MFP's findings -ning l maintenance of EQ equipment for both procedural and substantive reasons. MFP did not. i i -i timely file any EQ contentions although it had every opportunity to do so. The Licensing i t 4 Board specifically rejected MFP's third late-filed contention,' which alleged that: 4 i [D]eficiencies exist at the DCNPP with the environmental qualification of - ] safety-related and non-safety-related electrical cables (Okonite cables or : other cables with bonded jackets). Furthermore, deficiencies exist in the ~ j adequacy of maintenance and surveillance practices at DCNPP to verify-l 1 that the actual operating environment of these cables are [ sic] bound by the i environmental parameters used to qualify the equipment. 'Because these j deficiencies make the plant more vulnerable to a severe accident,~ Pacific j Gas and Electric Company's ("PG&E") license amendment request must' l be denied. 37 NRC at 449. In rejecting the contention, the Licensing Board stated that " insofar as i environmental qualification is concerned, this contention lacks any basis.... We are 3 thus denying the contention but permitting the failed-cable question at Diablo Canyon to i i j be litigated under Contention 1."7 j f I-61.. Moreover, the Licensing Board indicated in its first prehearing conference .F y i order in admitting Contention I that "to the extent that MFP is asked to do so, it must 1 1 identify prior to hearing all of the incidents on which it intends to rely in advancing and. j-p 1 1 i l The " failed-cable question" did not concern EQ cable. ' See 37 NRC at 448, ciring 7 i Affidavit of Ann M. Dummer. i .i 1: -}

e l l : ) { going forward with its contention." 37 NRC at 21. On June 21,1993, MFP filed e li*i l of documents, mostly NCRs, LERs and irs, on which it indicated it would rely. On i s August 13,1993, MFP provided for authentiation a second list of documents, in part j a sub-set of the previous list but also incorporating new documents.. At the hearing, MFP F produced a document it called a " road map" on which it li:'ed some 200 documents that - 2 it would seek to introduce. None of these lists indicated tb3. MFP would seek to j introduce the ".elatemp records and other documents concerning EQ. Therefore, this~ matter may rat be considered. See Hartsville, sapm. i I-62. MFP bases its conclusion that this matter presents an important safety j concern on NRC Information Notice 89-30i High Temperature Environments at Nuclear i . See MFP Proposed Findings 67-76. However,'this-I Power Plants (March 15, 1989). Information Notice was not introduced in evidence and notice ofit cannot be taken under i ) l' 10 C.F.R. I 2743(i) as it does not constitute a technical or scientific act within the' knowledge of the Commission as an expert body. Moreover, PG&E's program predates 's i a i i NRC Information Notice 89-30 and thus was not the occasion for PG&E's initiating its } temperature monitoring program. Tr.1844 (Ortore). MFP would have the Board adopt ] findings regarding the range of temperatures to which certain EQ equipment was exposed. ] No record evidence substantiates this use of Telatemp data, as the testimony was that the i I readings were " qualitative." PG&E Direct Testimony at 80 (Ortore). MFP would have ) the Board adopt a finding that "the temperature measurements taken in 1988 were I relatively high for the 133' elevation - 140 and 160 degrees - and that PG&E testified that ; 160 degrees is "quite wana." However, in answer to MFP counsel Ms. Curran's l i i 4

I l^ 1 : question as to whether he considered 160 degrees Fahrenheit mlatively hot, the witness. ~ Ortore, asked, " Relative to what?" to which Ms. Curran mplied, " Relative to what-equipment is generally qualified to see." Whereupon, the witness Ortore said, "No." i Tr.1887 (Ortore). 'Ihus, the record does not support a finding that 140' and 160' were relatively high for the 133' elevation and no other reference to temperatures at that i elevation was offered. I-63. MFP would have us find that absolute temperatures and absolute temperature ranges are measurable by the Telatemp sticker program. MFP Proposed- [ Findings 76-77. The testimony was to the contrary. PG&E Direct Testimony at 80 l l (Ortore). Also, them was no evidence to lead to the conclusion that there was any ~ failure involved when fewer than two stickers were found for a given component. There was no testimony to support MFP's view of the Telatemp program. :'Ihe testimony was i that the program was used as intended, as an indicator. There was nothing to support 1 l MFP's conclusions that PG&E did not follow-up the Telatemp information as appropriate, -l in reevaluating the qualified life of a particular component based on the history of the temperatures it had experienced. Rather, the evidence was to the contrary. See Tr.1845 (Ortore). I-64. MFP would have us find that PG&E failed to install a second set of l l stickers on FCV-440 in IRS. MFP Proposed Finding 90. MFP offers no basis for this i finding and we fail to understand how MFP could have such information, as IR6, the outage at which the stickers installed at IR5 will be removed, has not yet occurred. PG&E's witnesses testified that PG&E includes Telatemp stickers on every piece of EQ l l l 1

1 , equipment in containment. Tr.1873 (Ortore). 'Ihe inference that MFP would draw, that stickers not found were never affixed, is not permissible, as it has no record support. I-65. MFP proposed a finding that PG&E's program for monitoring l'edimt high temperatures is deficient, inconsistent and inaccurate, yet MFP has provided no l 1 record basis for such a finding. There was testimony that the temperature monitoring ) program was but one of the tools PG&E employed in monitoring EQ equipment. See - j I i Tr.1851,1865 (Ortore).. Even if that program was less than perfect in that stickers may i have been dislodged or failed for other reasons, the testimony was that the program j performed as intended. And that was all it was designed to do. 'Ihe only applicable l regulation,10 C.F.R. I 50.49, does not prescribe a specific monitoring program. .] l I-66. The Board concludes that even if as a procedural' matter this sub-issue could be considered, PG&E's monitoring for high temperature of electrical equipment j i was appropriate and Telatemps were properly used as a part of that monitoring program. f B. Check Valves / Inservice Tactine -) I-67. MFP Exhibits 6-11 and 13 were admitted by the Boani regarding' the issue of Inservice Testing of Check Valves. Tr. 622. For many years,~ the ASME code did not require testing of leak tightness of these valves if their position was normally closed. Tr. 602 (Crockett). However, in 1988 the NRC issued an Information Notice'that some valves should be tested even in a closed position.= Id. 1-68. In 1989 the NRC issued a generic letter, which was a follow up to the 1988 Information Notice, that provided the industry with more specific requirements on - testing these types of valves, addressing criteria and alternate methods of testing the

j o, valves. Tr. 603 (Crockett). ne notice contained a list showing which plants the NRC ' knew would have potential problems in this area, and a list of plants the NRC felt had - adequate IST programs and did not require a followup. Diablo Canyon was on that { second list. Id. The NRC at that time thought that Diablo Canyon had met all the [ requirements. PG&E had a review ongoing before'the generic letter was issued. Id. I-69. This issue is a generic problem. Tr. 604 (Crockett). PG&E identified i some check valves that were not being tested in their closed position, and added these to the program and wrote an NCR.- Id. Since then, other valves have been found,_ sometimes through Westinghouse letters, and it isn't obvious that they have a safety I function. Id. PG&E has continued to identify other check valves that have been added to the program. Id. In every one of these cases of newly identified valves, when PG&E went back and added it to the program and tested the leak tightness'of the. valves, the- [ valves were tight. Id. It was never a problem that the valves were indeed leaking. Tr. 608 (Crockett). 1 i I-70. MFP Exhibit 6 is an NCR and MFP Exhibit 7 is an LER regarding the { same issue. Tr. at 610 (Crockett). nese documents relate to PG&E's identification of I certain safety injection pump discharge check valves that were being tested by a normal, ] accepted method in the industry, but for which further testing'was warranted to - l demonstrate that there was no leakage. De valves were added to the IST program. ' Id.. i at 606-08; MFP Exhibit 7 at 8. I h '{ 8 1 i I

i '! ~ 'i I-71. MFP Exhibit 8 is an NCR that does not'show any deficiency. While an l NCR was initiated, the evaluation showed that the smvain-being performed on the i valves were sufficient to meet the ASME Code. Tr. 611-12 (Crockett, Vosbmg); MFP Exhibit 8 at 1. t I-72. MFP Exhibits 9,10, and 11 are all PG&E NCRs related to check valve f i testing initiated as part of PG&E's continuing review of the scope of the IST program. In each case, the IST program has either been revised to inw@urs the valves or tests 1 at issue, or commitments to do so have been prW. MFP Exhibit 9 at 8; MFP j t Exhibit 10 at 7; MFP Exhibit 11 at 12-15. 'Ihese nonconformances were also determined ~ to be the result of personnel errors, wherein for example, a commitment to do a certain i full flow stroke of a check valve was omitted from.the program. MFP Exhibit 9; i Tr. 612 (Crockett). The corrective action for this was to include the requirement to place NRC approved relief requests in the Commitment Management Database. Tr. 613 - (Crockett). I-73. MFP Exhibit 13 is another LER related to the generic inquiry described ~ above. The particular check valve in question, CVCS-8440, is intended 'to ' lose and c isolate the volume control tank (VCT). PG&E identified a potential safety function whereby leakage from valve 8440 could find a path back to the VCT. This was not an obvious path, but PG&E identified the issue and 'added the valve to the test program. Tr. 604-06 (Crockett). I-74. The Board declines to view these NCRs and LERs as indicative;of any breakdown in PG&E's maintenance and surveillance program. Clearly, this is a generic 3 ~

- ~.. ~.- t O. l ! problem that is being addressed industry wide, and the fact that PG&E is continuing the j i program to identify these valves gives us reasonable assurance the program is, in fact, working. Had PG&E failed to respond to the NRC Information Notice and not initiated any of thest NCRs and LERs, we would be inclined to agree with MFP that the IST. i program was deficieni. But these documents compel precisely the opposite conclusion. C. Undergr sund Cable Failures I-75. MFP introduced Exhibits 14-21 regarding this issue which was addressed in both the NRC Staff and PG&E's direct testimony. NRC Staff Direct Testimony, ff. Tr. 2159 at 9-10; PG&E Direct Testimony at 108-110. Additionally, there was extensive i discussion of this issue during the hearing. Tr. 623-74. l j I-76. In Licensee Event Report 93-05, PG&E reported five instances between l October 1989 and March 1993 of degradation in the cable insulation for.12kV and 4kV 3 i cables that supply power to plant components. MFP Exhibit 15. ' Both types of cable are j i contained in two separate underground duct bank conduits, one for each unit, routed between the turbine building and the intake structure. PG&E Direct Testimony at 109.' i t Two of the 4kV failures occurred on cables that are safety-related and supply electrical i i power to safety related cooling water pumps, and one 4kV failure occurred on a non-l safety related cable that supplies a load center transformer. - Id. The hon-safety related 12kV cables supply electrical power to the main circulating water pumps. Id. 1 I-77. PG&E has preliminarily concluded that the 12kV cable failures occurred 1 J due to a combination of submergence of the cables and chemical attack accelerated by the submergence. NRC Staff Direct Testimony, ff. Tr. 2159 at 9. 7he submergence was

0 0 i I caused by insufficient preventive maintenance of the sump pumps which were designed to remove water from the cable conduits. Id. The NRC inspection verified that the -Licensee has repaired the inoperable pumps and initiated a preventive maintenance 5 program for them. Id. Further, all of the failed cables have been replaced. Id. I-78. PG&E is continuing work to identify the cause of the 4kV cable failure. 1 i NRC inspections concluded that the Licensee's follow-up actions in response to those equipment and cable failures have been adequate and reasonably thoinugh. NRC Staff ' Direct Testin:ony, ff. Tr. 2159 at 10. The NRC concluded that plant safety had not been' significantly reduced by these cable failures, due to' the presence' of other, unaffected i cables for redundant safety-related pumps. Id. t I-79. Prior to this event, there was no maintenance for the sump pumps and 1 drains. The preventive maintenance program, whose scope involves recommendations I from the vendor, recommendations from the staff and recommendations from industry based on experience did not include these small sump pumps. _ Tr. 639 (Ortore). I-80. While damage to the non-safety' related 12kV cable was due to'being - submerged, the safety-related 4kV cable had no evidence of water damage. Tr. 644-45 (Giffin). That is, even though the 4kV cable may have been submerged, there was no indication of water intrusion into the installation. Tr. 646 (Giffin). I-81. These cables are periodically tested and two of the failures (one 4kV and. one 12kV) occurred in testing, not while the equipment was in service. Tr. 652 (Ortore). ~ There is a ground system installed on the 4kV cable and an indication showed up in the control room that there was a momentary ground on the cables, so the pump was taken ~

i l i. out of service and hi-pot testing was done. Tr. 652 (Giffin). If the cable failed due to a ground, there would be a ground alarm, and that system is designed to operate for a minimum of one hour after the ground fault is alarmed. Tr. 653-54 (Ortore). - There is no difference between normal operation of the pump and the pump running in a safety- ) related mode, since the current through the cable is the same. One of the pumps is operating continuously. Tr. 655 (Vosburg). I-82. The first indication of the 12kV cable failure was a ground alarm in the control room. Resisters heated up and dust collected on them, producing smoke, which set off the smoke detector. Tr. 658-59 (Vosburg). Regarding the 4kV cable failures, one i l was during high potential testing; the other two occurred while the pump was operating.' i When the ground fault came in, the redundant pump came on line. Tr. 659-660 (Ortore). I-83. The inoperability of the sump pump is related only to the 12kV cable j failures, although the 4kV cable runs in the same duct work. Tr. 662 (Giffin). On each unit, there are two trains of auxiliary salt water, two complete trains. Each of these trains can be cross-tied to the other unit. Tr. 664 (Vosburg). While the same cable is used in all four trains, it is extremely unlikely to have a common mode failure, as the same failure on four cables would have to happen randomly at the same point in time to t have an effect. Tr. 665 (Vosburg). And over the course of years there have been but two indications of degradation of the cable, which occurred years apart. Id. While MFP i would have us find the potential to fail for all 4kV cable in the plant, the witnesses testified that no other 4kV cable used in the plant is subjected to the same submergence. l 1 r

Tr. 663 (Ortore). Further, during testing, the vendor had to take the cable in excess of 60,000 volts to have it fail. Tr. 657 (Giffin). l I-84. Thus, while PG&E is still investigating the root cause for the 4kV cable. t failure, the Board is satisfied that a common mode failure is unrealistic at best, and we accept the Staff's evaluation that plant safety has not been reduced due to the existence of other unaffected cable for redundant safety'related pumps. See NRC Staff Proposed r Finding I-77 above. Further, we find that'with-the sump now added to PG&E's, l t maintenance program and with the cable replaced, it is unlikely that such a cable failure i will recur. NRC Staff Proposed Finding I-82, above. l D. Wrong Motor In=talled on MOV Actuntar I-85. MFP Exhibit 24 is a PG&E NCR which addresses an error made during l maintenance to replace the motor on an actuator for a motor operated valve. An j individual preparing the work order made an error in reading the motor size from a table and specified on the work order a motor with an incorrect torque value. -Tr. 689-90 (Giffin). The wrong mora was then installed on the valve. M. I-86. However, even with the incorrect motor installed, the MOV still was able to shut, which was its intended safety function. 'Ihus, there was no actual or potential safety significance attributed to this event. Tr. 690-91 (Giffin). This valve is normally open but is required to be closed during the switchover sequence from injection phase to cold leg recirculation phase post LOCA and to remain closed. MFP Exhibit 24 at 4. I-87. MFP argues that there would be insufficient time for manual closure of the valve, and the ability of the motor to close the valve is not sufficiently assured. MFP

i , Findings at 150-151. However, MFP misreads the NCR which specifically states that i the capability of the motor is more than adequate to shut the valve and once it goes closed, the motor may burn out, but not until after the valve has performed its safety function. Id. at 5.8 I-88. At this point the Board must make clear that MFP's stratagem ofinvoking " judicial notice" to introduce new and unsupported documents into the record through its l findings is not acceptable. A judicially noticed fact must be one not subject to reasonable dispute in that it is either (1) generally known within the territorial jurisdiction of the court or (2) capable of accurate and ready determination by resort to sources whose accuracy cannot reasonably be questioned.' But it cannot be employed to introduce into 1 the record documents that are not available to the parties during the evidentiary hearing. Nor can such documents be relied upon for findings absent proper admission into the record. We reject these documents and decline to adopt any finding of MFP that relies upon " judicial notice" of documents that were unavailable to the parties at the hearing.20 I-89. MFP further argues that the error that resulted in the installation of the wrong motor was a breakdown in the program because this error was carried through from the work order to installation and inspection. MFP Proposed Finding at 147. Yet- ' We would also point out here that the argum nt by MFP regarding the margin of error in the calculated thrust is nowhere to be found in the document itself and was not addressed on the record at the evidentiary hearing. Therefom, we cannot adopt this finding which is based on sheer unsupported speculation. ' Rule 201, Federal Rules of Evidence; see also 10 C.F.R. 6 2.743(i). ' See Long Island Lighting Co. (Shoreham Nuclear Power Station, Unit 1), LBP-30-88-13,27 NRC 509,565-66 (1988).

. this error was found by routine verification processes and the corrective action initiated applied to all the individuals involved in the incident. MFP Exhibit 24 at 2,'6-7. I-90. We decline to adopt MFP's view that the involvement of more than one - person in the same incident somehow equates to a breakdown in the' system. This was an isolated event, as walkdown of all other MOVs for which similar work had been performed showed that this was the only case in which the error had been made. Tr. 691 (Vosburg). E. Storare and Handline of Lubricants I-91. MFP Exhibits 27 and 28 are two PG&E NCRs,- related to control of lubricants and greases. MFP Exhibit 27 relates to an incident in 1993 where improper oil for the ASW pump motor was used to top off the oil after an oil change, but Chevron assured PG&E there was no problem with incompatibility. Tr. 728 (Giffin). While about 10 percent of the oil in the motor was not proper, it did not affect operability.. Tr. 730 (Giffin). I-92. Exhibit 28 relates to instances in 1991 where the wrong cartridge of grease was used in grease guns, but again, there was no safety significance regarding the mixing of these greases. Tr. 732 (Giffin); MFP Exhibit 28 at 5.- Corrective actions were taken at the time. MFP Exhibit 28 at 7-8. There were no similar incidents until the 1993 incident concerning lubricants for the-ASW pump motors. MFP Exhibit 27,- NCRi DCO-93-MF-N039 (July 27,1993).-- I-93. The lubricants used in the maintenance of equipment at DCPP_ are kept at a central location in the turbine building (elevation 85). 'Ihey are controlled by'a.

i , procedure which specifies the appropriate oils for each component and by a check-out log which documents the use of the oils. Tr. 730-31 (Giffin). The location of the oils and the control procedures for their use represent a balancing of the needs of various departments as well as a balancing between the need for control versus the desirability of efficient and timely access to the oils. Tr. 737 (Vosburg). I-94. The Board finds the corrective actions taken by PG&E are not wanting, particularly given the dissimilarity between the events described in the two NCRs. Clearly the program needs constant management vigilance, but we find existing procedures are adequate. We further note that this problem is being addressed by the current preventive maintenance programs to change oils periodically, to monitor equipment for excessive wear, and to sample oils as part of the predictive maintenance program. Tr. 740, 744 (Vosburg). F. Fuel Handline Buildine I-95. MFP Exhibits 38, 39, and 41, were admitted (Tr. 828) regarding the fuel handling building. MFP Exhibit 38 is a PG&E NCR describing two 1991 events in which the fuel handling building (FHB) failed to pass the required surveillance test acceptance criteria for building negative pressure. While in both cases the pressure was not as negative as required, sufficient negative pressure was maintained at the surface of the spent fuel pool and there was no safety significance of the surveillance test failures, i PG&E Direct Testimony at 105, MFP Exhibit 38 at 2-3,13-14. I-96. The first of the two test failures (event 1) addressed in MFP Exhibit 38 and PG&E's Direct Testimony, occurred on January 18,1991. MFP Exhibit 39 is the LER

1 .) 0 8 l addressing this same incident. Tr. 784 (Crockett). Prompt action was initiated to investigate the situation, determine the root cause, and implement cornetive actions. 'I PG&E Direct Testimony at 104-05 (Crockett). PG&E determined that the cause of the reduced negative pressure was the existence of small leakage paths into the building due to building siding and seal degradation. Cormctive actions included sealing the leaks and residing both FHB buildings. Id. at 105. This is an example of stmetural degradation - that was identified by PG&E through its surveillance program and corrected. Tr. 807-8 ~ (Giffin). It does not suggest a maintenance progism breakdown. I-97. Event 2, discussed in MFP Exhibit 38, regards an incident involving a failure of the FHB to pass a surveillance test in August 1991. PG&E determined that the setpoint for the controller had drifted, increasing the supply flow into the FHB. MFP Exhibit 38 at 3. This setpoint drift was identified and corrected. Tr.'783 (Crockett). I-98. The degradation of the FHB was due to aging and in'the' process of testing-and monitoring the aging was detected. Tr. 807 (Giffin). Contributory causes were determined to be dirty exhaust fan ducts and blocking of an FHB exhaust duct. MFP ~ Exhibit 38 at 1. Additionally, a significant corrective action taken is that the surveillance j i is now performed within seven days of any movement of the fuel, rather than at the 18 month frequency specified in the Tech Specs. Tr.- 812 (Crockett). I-99. MFP Exhibit 41 is a November 1991 Monthly Summary of the DCPP : Onsite Safety Review Group (OSRG). It appears'that MFP offered this exhibit in an - attempt to show that there is some linkage beta.~., a problem with controlling doors in' the FHB and the events described in Exhibits 38 and 39. However, there is no =. -.. ,,,. + ~,

i i - 4 i connection between the event described in MFP Exhibit 41,. referencing NCR l t i l DC2-90-TN-N015, a PG&E NCR that was not admitted or addressed in the record, and i i MFP Exhibits 38 and 39. The NCR referenced in MFP Exhibit 41 relates to control of i i j personnel doors as part of the FHB boundary. The control of the doors'is not a - 1 maintenance issue. Tr. 820 (Vosburg). Therefore, we decline to give this document any .l evidentiary weight. I-100. We do not find that the incident with the fuel handling buildir.g is evidence j of a breakdown in the maintenance and surveillance program at Diablo Canyon Rather, l the degradation was found through performance of required surveillance. G. Tests of Containment Personne1' Airlock ' I I-101. MFP Exhibits 42, 43, and 44 were admitted (Tr. 854) regarding the i containment personnel airlock. Exhibit 42 is a PG&E NCR related to a failure in several 'i cases to specify a post-maintenance test (PMT) for work involving removal of a pressure I gauge from the personnel airlock. MFP Exhibit 43 is the LER addressing the same issue. l MFP Exhibit 44 is a separate LER addressing a completely unrelated missed surveillance f i regarding the automatic leak rate monitor. J I-102. The issue addressed in MFP Exhibits 42 and 43 was identified by PG&E.' ) I On April 25,1993, the Unit 2 personnel airlock containment differential pressure gauge was removed for maintenance and reinstalled. The work order failed to specify a-post-maintenance local leak rate test of the instrument fittings as required by Technical Specifications. MFP Exhibit 42 at 2. ' Apparently, in speciffng the &=y PMT, the focus was only on fixing and testing the gauge. The effect on the pressure boundary of __._.,.i

C a removing the gauge, and the related need for a local leak rate test, was not identified. Tr. 830 (Vosburg). The deficiency hem was in the documentation to the work planner .l t who wasn't told this work would need a PMT, not a personnel error. Id. 1-103. As reported in MFP Exhibits 42 and 43, in investigating this issue, PGAE identified two similar instances in September 1990 when gauges were removed 'from service for calibration and returned to service without the -=ry post-maintenance i local leak rate test. However, in the earlier incidents prior to entering a mode where it~ i was required to be tested, an integrated leak rate test on the personnel hatch, which tests ' i the entire boundary, was done. Tr. 832 (Vosburg). I-104. PG&E has instituted corrective actions to address the issue identified in. MFP Exhibits 42 and 43. PG&E has updated the component database for the giw.cel - airlock gauges to include a waming to alert the work planner that a post-maintenance leak rate test is e=ry. MFP Exhibit 42 at 9. j I-105. MFP Exhibit 44 is an LER addressing an unrelated incident of a missed conditional surveillance test. STP I-1B, " Routine Daily Checks," requires Operations to verify on a daily basis that the leak rate monitor on the personnel airlock door seals is energized and in the automatic mode. If the leak rate monitor.is in the ' manual mode, then STP I-1B requires that Operations verify that conditional surveillance STP M-8F has ' been performed if containment entries have been made. MFP Exhibit 44 at' 2. A PG&E system engineer determined on September 27, 1991, that the leak rate monitor was. inoperable when in the automatic mode. Subsequent review determined that the monitor ~ had been inoperable since June 5,1991, but Operations assumed that subsequent. .l i

i A P f operating shifts would perform a manual leak-rate test following receipt of an automatic - l j leak-rate test failure alarm. Id. at 3. PG&E found that 17 containment entries were made during the period ofinoperability.' Id. The conditional surveillana STP M-8F was missed on those occasions because the majority of plant upr.r. tors did not recognize that I the leak-rate monitor was unable to perform its automatic function, and they reset the failure alarms without manual retesting following door entry. There was no indication l that the personnel airlock doors were not capable of performing their intended function f during this period. Id. at 6; Tr. 837. (Crockett). I-106. For NRC reporting purposes, and in accordance with normal accounting methodology throughout the industry (Tr.1149, Crockett), this event constituted one i missed conditional surveillance test incident. Tr. 845 (Crockett). PG&E testified that there have been 65 instances of missed Technical Specification surveillance tests over the operating life of the plant. Tr. 836 (Crockett). 'Ihis is in the context of approximately l 10,000 Technical Specification surveillance tests perfoi...ed each year. Tr. 834 i (Crockett). In the past ten years only 65 have been missed; progress has been' made in that only three have been missed in the past two years: Tr. 836 (Crockett). l I-107. There was extensive discussion in the record of how missed surveillawa= are counted and tracked, and why, if the root cause is the same, multiple occasions are counted as one missed surveillance. Tr. 844 (Crockett). The-vast majority of surveillances are scheduled, rather than conditional (as is the one under discussion referenced in MFP Exhibit 44). Tr. 846-849 (Vosburg). No matter how many times it occurs, if the root cause is the same, a missed conditional surveillance is one missed

. surveillance. Tr. 848 (Crockett). Counting each occurrence would not produce meaningful data; rather, the root cause is what matters. Tr. 852 (Giffin). f I-108. The Board is satisfied that the incidents represented by these exhibits do t not speak to the adequacy of the maintenance and surveillance program at Diablo Canyon. We are also satisfied that counting missed surveillance by root cause rather than i by occurrence is a more useful way to approach improvement in the program and we recognize that the industry and the NRC accept this practice as appropriate methodology i for documenting the event. Tr.1149 (Crockett). H. Comnonent Cnaline Wat- (CCW) Heat Fehaneer I-109. MFP Exhibit 47 admitted by the Board (Tr. 868) is a PG&E NCR related - to eddy current testing conducted by PG&E during a Unit 2 refueling outage (2R5) on the component cooling water (CCW) heat exchanger tubes. In March 1993, PGAE i conducted the eddy current testing on one CCW heat exchanger.. 'Ihis testing was part l l l of the ISI program that looks at performance to identify and predict early if there is any t degradation of equipment. Tr. 856 (Crockett). The testing indicated fretting on the l I outside diameter of the tubes at the baffle plate. MFP Exhibit 47 at 3. PG&E then j i expanded the scope of the eddy current testing campaign to include 100 percent of the i l tubes for both CCW heat exchangers. Id. at 3-4. The results of the testing revealed j several tubes that needed to be plugged on both heat exchangers. 14. 'Ihis was - accomplished. Tr. 856 (Crockett). I-110. PG&E determined the root cause of the fretting to be a flow induced l vibration on these tubes. Tr. 857 (Crockett). Accordingly, operating procedures are l

_ ~ _ - 1 4 C. i i 'I 4 j being revised to address maximum flow limits on the heat exchanger. MFP Exhibit 47 ) at 8. Surveillance or maintenance of these heat exchangers was scheduled based on the ' ) expected wear and service life of the heat exchanger. Tr. 858 (Crockett). PG&E l l I { identified the issue and took appropriate maintenance actions. Tr. 858 (Crockett). j i j I-lli. MFP notes a concem that if PG&E does not conduct eddy tests lmore i j frequently the margin of operability before degradation will be too small. MFP Proposed j ~ s y Finding 210. However, PG&E is performing an engineering analysis to determine the 2 i precise interval for testing and eddy current testing will again be conducted at the next t i i ] refueling outage or sooner. Tr. 866 (Giffin). Rather-than a deficiency in the l i maintenance and surveillance program, this demonstrates that in-service inspections are 1 l being carried out in an appropriate manner. i i I. Auxiliary Buildine Ventilatian System l l I-112. MFP Exhibits 49 and 50 admitted by the Board (Tr. 886) relate to an j instance in which the Unit 1 Auxiliary Building Ventilation System _ (ABVS) was-improperly rendered inoperable while a preventive maintenance task was being-t j performed. MFP Exhibit 49 is the PG&E NCR addressing this event. The incident was i reportable and MFP Exhibit 50 is the associated LER. ) i 1 I-113. More specifically, on March' 2,1993, maintenance personnel were l preparing to perform a preventive maintenance task to clean a flow element associated with the Unit 2 ABVS. A clearance was placed on the system. However, in. subsequently revising the clearance and implementing the work order, gim d improperly closed a damper. This activated the ABVS logic to shut down the only, i I i . s-m .y y 7- '3,,,, -, _ _.

O O I operable ABVS fan. 'Ihe redundant fan was already out of service for the maintenance. As a result, the Technical Specification requirement was not met for.15 minutes.- MFP Exhibit 50 at 2-4; Tr. 881-82 (Giffin). The event was not safety significant in that failure of the ABVS does not immediately jeopaniize.the operability of any safety-rdatevi equipment, since 24 hours is available prior to adversely affecting equipment. MFP Exhibit 49 at 8; MFP Exhibit 50 at 5-6. I-114. As referenced in the NCR, non-cognitive personnel error v.as ultimately identified as the root cause of the event. MFP Exhibit 49 at 7. One cause was error on the part of the system engineer in that he was unaware of the potential for error. that: might result from a change to a clearance that supported a preventive ~ maintenance activity. Id. 'Ihe other was error by the mainterm foreman, who misinterpreted unclear work order steps and mistakenly authorized closing the wrong damper. The work ' instructions were determined to be the barrier that could have prevented the human error from occurring. Tr. 883 (Giffin). However, it should be noted that the circumstances of this incident were unusual and, when this recurring maintenance task had been performed previously under different alignment conditions,' the damper involved had been' closed. MFP Exhibit 50 at 3. In any event, one of many corrective actions identified was to revise the recurring task work order for this preventive maintenance work to i specifically identify the dampers to be closed. Tr. 886 (Vosburg); MFP Exhibit 49 at 11. I-115. The Board agrees with MFP that this is, indeed, an example of personnel error and lack of adequate communications. But this incident in and of itself appears isolated and does not wrwit a breakdown in the overall maintenance and surveillana d _,,.v.. e..q+ y 9v--,rsqy

1 ! program. We are also mindful that this event had little safety significance given the design basis of the ABVS. See NRC Proposed Finding I-113. J. Electrical Panel Covers I-116. MFP Exhibits 51 and 52 are two PG&E NCRs admitted by the Board (Tr. 900) addressing failures to re-install or secure equipment covers or a divider plate in an electrical panel following maintenance work within the electrical panels. Given the ) similarity between these two documents, PG&E's witnesses stated that the first NCR is being canceled and all actions taken will be done under the NCR that is MFP Exhibit 52. 1 Tr. 887 (Giffin). 1-117. MFP Exhibit 51, dated June 7,1993, relates to covers not installed in the hot shutdown panel for both Unit I and Unit 2. The covers would complete the enclosure that surrounds the switches inside the panel. The covers were found lying in the bottom of the back of the hot shutdown panel, and mounting screws were nowhere to be found. MFP Exhibit 51 at 1. PG&E determined that the as-found condition did not impact safety. Id. at 3-4. I-118. MFP Exhibit 52 documents one instance in April 1993 in which the rear hinged panel of the Unit 1 RHF panel was found with no fasteners installed to secure the hinged panel to the main panel. The fasteners were in a plastic bag in the bottom of the 1 panel. MFP Exhibit 52 at 1. PG&E's witness stated PG&E has determined that the ) preliminary root cause for this event is that both construction and maintenance personnel worked on the panel, and, essentially, construction thought maintenance should do it and maintenance thought construction would do it. Tr. 888-89 (Giffin). PG&E specifically

~ 1 e I evaluated this condition for a potential loss of seismic qualification impacting operability of a vital 4kV bus, and concluded that the bus and its associated diesel generator would have been operable before and after a postulated seismic event. MFP Exhibit 52 at 5. Furthermore, system mdundancy in this case precluded potential safety impact. Id. at 6. I-119. Further, regartling this NCR, MFP takes one statement out of context from i

p. 3 of MFP Exhibit 52, and ignpres statements on pages 2 and 5-6 of that exhibit which I

analyze the event. MFP's onnions and conclusions on this matter are unsupported. MFP Findings 233-34. MFP has no basis to disagree with PG&E's safety analysis. ' As i we have stated previously, MFP provided no expert testimony and presented no witnesses for cross-examination of the hearing. As this Board pointed out to MFP during the hearings, "in developing a record that contains a let of raw data in it someplace and l leaving it to the Board to reconstruct what your view may well be... you really do have some burden of showing us what your view is as you demlop she record, [ emphasis supplied] and you just can't rely on us to comb through'this record and reconstruct a theory for you out of a voluminous, sort of, raw data... and then later having findings selected out of that all from the paragraph that outlines the error and not taking account of any of the paragraphs that outline the remedies." Tr. 697-98 (Kline). The Board cannot adopt speculative theories put forth extra-record by MFP's lay' persons 'and attorney which have no foundation on the record before us. I-120. PG&E's evaluation of root causes and corrective actions 'related to these incidents is ongoing. Tr. 898 (Giffin). All panels involved in these two NCRs have been restored to the correct configuration. Tr. 891 (Giffin). ~

> I-121. The Board does not find the occasion of these incidents rising to the level j of demonstrating a fundamental flaw in PG&E's maintenance and surveillance program. A part was found by PG&E not to be in perfect order, it was repaired by PG&E and the j I event lacked any safety significance. This is not an example of a dysfunctional l maintenance program. K. Containment Eauinment Hatch Gan ( I-122. MFP Exhibits 53 (a PG&E NCR) and 54 (an LER) were admitted by the t Board (Tr. 912) and both address the same incident at DCPP Unit 2 in March 1993. As. l l documented therein, the containment equipment hatch should have been closed while fuel was being offloaded. A maintenana crew assigned to the task closed the hatch, installed l the four bolts mquired by Technical Specifications and procedure,- but did not notice a [ gap at the upper portion of the hatch. MFP Exhibit 53 at 2-3. The gap in the hatch seal constituted a violation of the Technical Specifications. MFP Exhibit 54 at 1. j I-123. PG&E attributes this incident to personnel error. Tr. 901 (Giffin). The worker failed to follow procedures specifying that he verify that the hatch was closed. l Tr. 907 (Giffin). The condition was subsequently identified by a shift foreman responsible for the fuel movement inside containment. Tr. 901 (Giffin). While this same .f I problem had happened once before, the incidents were ten years apart and could hardly~ be said to have any real connection. Tr. 903 (Giffin). l } I-124. As PG&E testified at the hearing, it is really not "possible to completely eliminate personnel errors in any endeavor." Tr. 903 (Giffin). With the number of man l \\ hours of work performed and the number of procedures involved, a tremendous amount - j i -i i ~ _ ~

l 4 s ) i l J of work on a tremendous number of different systems is done. Id. PG&E testified that ) J PG&E tries to give the people performing the work the tools that are -=y to do that. - i ) job; whether that be training or equipment, so they won't make a mistake. Id. ~ But mistakes happen. That is why the design and the redundancy is important, to minimize any potential for mistakes causing problems. Id. 1-125. The Board finds that, though this is clearly an incident of giw.d error, l it is isolated and does not reflect any systemic problem. The record reflects that the i procedure was clear (Tr. 904, Giffin) and that the worker was experienced. Tr. 905 (Giffin). PG&E found the condition, took corrective actions, and counselled the. maintenance grwn.cl involved. MFP Exhibit 53 at 10. In addition, PG&E is also revising the procedure to include in the work order a hold point to allow Quality Control to verify the hatch closure. Tr. 908-9 (Vosburg). No matter relevant to PG&E's overall . i surveillance program 'vas identified. j L. Manual Pe r Trio enn=1 by Fuse Failure o q I-126. MFP Exhibits 55 (an LER) and 56 (a PG&E NCR) were admitted by the i l Board (Tr. 912) with virtually no cross-examination or development of any kind in the record. MFP Exhibit 55 is an LER reporting a manual reactor trip caused by a rod i control power supply fuse failure due to a personnel error. - MFP Exhibit 56 is the PG&E NCR initiated to address the incident. Both the LER and.NCR' address the event, its causes, and the corrective actions. There is no indication in either document that this isolated event reflects a common problem or a systemic maintenance deficiency.- l v -- v --r ~ e,- .n, -a,>e,,,r---< e-

> I-127. The trip resulted from operator action (to trip the plant) when manual control rod movement Wma inoperable. This inoperability was caused by the failure of a fuse in the bus duct disconnect to the rod control power supply. MFP Exhibit 56 at 1. Upon investigation, PG&E determined that the fuse that failed was of an old style with known reliability problems that was to have been replaced in 1989. Id, PG&E also found other examples of these old fuses in similar locations. Id. PG&E determined the most likely root cause to be that, in October 1991, the contract electrician that was to have replaced the old fuses in the bus duct disconnect actually replaced fuses in the rod control power cabinet. Id. at 4. Following the trip, PG&E took action to replace all of: the bus duct fuses for the Unit I rod drive control cabinets. Id. at 7. The corrective actions also include other measures such as providing unique panel identification for each panel location in the rod control system and. worker-tailboards to communicate expectations. Id. at 10. No matter generally relevant to PG&E's' surveillance and maintenance program was identified. M. Limitomue 2-FCV-37 Test Failure I-128. MFP Exhibit 57 is a PG&E NCR describing the failure of one Limitorque motor operated valve, 2-FCV-37, to close on demand from the control room during the i i performance of a routine surveillance procedure in January 1993 (admitted, Tr. 917). l 'Ihe cause of the test failure was determined to be an incondly installed quad ring. Tr. 913 (Giffin). 'Ihis is an example of the IST/ISI test program working as intended. I-129. 'Ihe personnel error in the quad-ring installation occurred during' a maintenance overhaul of the operator in 1990. In the NCR process, PG&E determined i l j

~ 57 - 1 that the root cause of the error was that the work procedure did not give sufficient I detailed guidance to ensure that the quad rings were properly installed. MFP Exhibit 57 i at7. I-130. PG&E has taken corrective actions to address this exurrence. The component was returned to operable status and the maintenance procedare was revised j l to provide the necessary guidance on re-assembly of Limitorque operators. MFP Exhibit l 57 at 14,16. There is no indication in the record to show that this issue is more than a single isolated ewat. There were no other cases cited in which the quad rings in Limitorque operators were improperly installed, either before or after the procedure revision. While a previous occurrence was referenced in the NCR (MFP Exhibit 57 at 17-19), PG&E's witness distinguished this occurrence from the earlier one. Tr.~ 916 (Vosburg). That incident was not related to maintenance, and did not share a root cause with the primary incident addressed in the MFP Exhibit 57. 1-131. It is clear from a reading of the entire document that the valve was not, in fact, inoperable from a safety standpoint See MFP Exhibit 57 at 9-12. Q: MFP l Proposed Finding 250L.63, which ignore the safety analysis in the Exhibit. As we stated on the record when these documents were admitted, the whole document must be considered, notjust selected portions. Tr. 696 (Kline). This document does not evidence-l a flaw in the PG&E surveillance and maintenance program. I l l 1 l i

s o - 58 1 N. Emercency Core Cooline System Accumulator Tanks I-132. MFP Exhibits 59 (admitted Tr. 2178), 60 (admitted Tr. 945-46), and 61 l (admitted Tr. 945-46), all relate to the issue involving intragranular stress corrosion cracking (IGSCC) identified by PG&E in the emergency core cooling system accumulator tank sample and fill line nozzles. I-133. Small leaks in these nozzles in Unit 2 were first identified by PG&E in 1985-87. At that time, all leaking nozzles were identified by PG&E and successfully repaired or replaced. The surveillance inspection frequency was also increased. MFP Exhibit 60 at 3. In addition, PG&E cut out several nozzles for metallurgical analysis. Tr. 934 (Crockett). Where cracking was observed, crack propagation was from the inside diameter to the outside (MFP Exhibit 60 at 3); this condition was not, therefore, the result of any exterior corrosion. 1-134. In 1991 the NRC issued an I&E Notice about the possibility ofIGSCC in accumulator tanks and during the Unit 2 outage further inspection of Unit 2 was done. PG&E found funher indication of cracking on certain nozzles and all nozzles with these indications were repaired. Tr. 934-35 (Crockett); MFP Exhibit 60 at 3-4. I.a 1992, during planned surveillance on Unit I nozzles, some indication of cracks was again identified and repairs were made. Id. at 4. PG&E determined that certain of these l nozzles in both units were manufactured of a material with high carbon content, more susceptible to IGSCC. Tr. 936, 941 (Crockett). I-135. PG&E has been investigating this issue since its initial identification and has implemented appropriate corrective maintenance and increased smveillance.

s

• Tr. 939-41 (Crockett). PG&E's voluntary LER also committed to an enhanced periodic inspection program for accumulator nozzles. MFP Exhibit 60 at 8. The April 1993 NRC inspection report, IR 93-08, concluded that PG&E was implementing an acceptable examination program for the accumulator tanks. Tr.' 2178 (hies); MFP Exhibit 59 at 5.

The NRC Staff witnesses also testified that not only was PG&E's program l 3 acceptable, it was considered "very good work to be looking this hard and finding these things and fixing them." Tr. 2178 (Miller). We do not see the issue presented by these documents as evidence of any deficiency in PG&E's' maintenance and surveill=*. program." O. Cormsion of Undareround Pinine I-136. MFP Exhibits 62, 63, 64 and 64A, admitted by the Board (Tr.1096), are .i all PG&E documents involving the identification, investigation, and resolution of < l t corrosion problems on certain underground piping, specifically the diesel fuel oil line for Unit 2; fire protection carbon dioxide (cardox) piping, which supplies the Unit 2 diesel generator fire suppression capability; and the ASW annularpiping. All of this equipment is located below ground in a concrete trench on the seaward side of the turbine building. PG&E Direct Testimony at 99; Tr.1059 (Crockett). I-137. Corrosion identified on the cardox piping is addressed in MFP Exhibit 62, an PG&E NCR. This cardox piping is located in the same trench as the diesel fuel oil u line piping that is addressed in MFP Exhibit 64.- Tr.1065 (Crockett).' In January 'of - i " MFP also tries to raise in its findings a financial concern regeidng replacement of these nozzles, an issue outside the scope of the admitted contention in this prMing. See Prehearing Conference, December 10,1992 at 56; Tr. 815.

. i 1993 corrosion was found on a stainless steel bellows used within the cardox piping that supplies the diesel generator. Id. The piping was repaired and returned to service. MFP Exhibit 62 at 1. This piping serves a fire protection function, and an acceptable compensatory measure is a continuous fire watch in areas where redundant systems or components could be damaged and a fire watch patrol in other areas, which was already established. MFP Exhibit 62 at 1, 3; Tr.1066 (Crockett). PG&E also plans to replace the cardox piping in 1993, moving it out of the trench and inside the turbine building to eliminate this source of corrosion. MFP Exhibit 62 at 12; Tr.1066 (Crockett). l I-138. MFP Exhibit 63 is a PG&E NCR addressing PG&E's identification in June 1992 of a hole in the ASW annubar piping due to corrosion. The annubar is an instrumentation housing that protrudes from the pipe in which a flow instmment to measure the flow can be inserted. Tr.1061 (Crockett). The leakage did not affect operability of the ASW pump. Tr.1062 (Crockett). The root cause for the corrosion was determined to be degradation or breakdown in the coal tar coating exposing the pipe to standing water and the saltwater air environment. MFP Exhibit 63 at 1, 9. Standing water seeping through the coating would corrode the pipe, ecpadMly at the water / air i interface. Id. at 9. The standing water was due to inadequate drainage caused by flow ) blockage by pipe supports and external debris. Id. However, the ASW system was determined to have been operable, even with the as-found through-wall hole in the annubar piping. The ASW system would have been able to provide adequate cooling water. Id.; Tr.1085 (Vosburg, Giffin). 1

'l ( l I-139. PG&E replaced all ASW annubar piping in both units (MFP Exhibit 63 d at 1) and set up a multi-discipline corrosion task force to identify galvanic corrosion on ' j i exterior surfaces of pipes. Tr.1062 (Crockett). This effort encompasses both the'cardox: and diesel fuel oil piping corrosion. Comprehensive actions to prevent recurrence have .l i been identified and are being taken, including inspection procedures to emphasize visual l q inspection for corrosion as well as design changes.- MFP Exhibit 63'at 1,12-13,29-31. { .i The corrosion task force will specifically look at corrosion. issues and make recommendations from a global perspective. Tr.1088 (Crockett). l I-140. MFP Exhibit 64 is a voluntary LER, filed in October 1992, related to the 1 degradation in the diesel fuel oil line piping. 'Ihe diesel generator fuel oil system is used i l i to transfer fuel oil from underground storage tanks to the emergency diesel generators. !i PG&E Direct Testimony at 99. This small diameter pipe runs below ground in the same i trenches as the ASW and Cardox piping.14. ' Exterior corrosion on this piping was first j observed in 1990.. This exterior corrosion was due to deterioration of the external 1 coating on the piping in areas where the coating had not been fully effective..The. deterioration had allowed exposure of the pipe's exterior surface to the ' salt air. environment. PG&E Direct Testimony at 100; Tr.1060 (Crockett). The corroded area was cleaned, inspected, and recoated, but there was no damage to the pipe due to corrosion. Id. The inspection frequency was increased from ten years to five years. MFP Exhibit 64 at 3; Tr.1061 (Crockett). I-141. Subsequently, in 1992 following a caustic spill, PG&E again inspected the i piping in this trench and identified more corrosion on the diesel fuel oil line. Id. l j . a

i 1 1 62 - Ultrasonic testing identified 'one location on the DFO piping that was below minimum d i 1 wall thickness requirement and this piping was subsequently replaced. MFP Exhibit 64 at 1, 5. PG&E identified corrective actions to address the corrosion and actions to 1 prevent recurrence. Tr.1064 (Crockett); MFP Exhibit 64 at 8-9. 'Ihis includes raising ) i the piping within the trench to minimize any potential exposure to shading water. Tr. i i 1084 (Giffin). Neither the corrosion identified in 1990 nor that identified in 1992' ) \\ affected operability of the piping. MFP Exhibit 64 at 8; Tr.- 1063-64 (Crockett). i 1-142. The corrosion issue was identified by the maintenance and surveillance a program, and PG&E's actions subsequent to the discovery of the corrosion were thorough -i in identifying the amount and extent of corrosion and correcting the problem. ' NRC Direct Testimony, ff. Tr. 2159 at 12. PG&E's long term actions, which included looking for similar situations in other systems, were identified, appeared to be adequate, and will be followed up. Id. I-143. Finally, MFP Exhibit 64A is a December 1992 Monthly. Rgort of the OSRG that discusses the NCR (MFP Exhibit 63) on the diesel fuel oil piping corrosion I and the ASW annular leakage. The OSRG is by definition intended to provide critical - input into PG&E's nonconformance evaluation process. Tr.1068 (Giffin). The OSRG is only one part of the issue evaluation process. Tr.1069 (Giffin). In 'this case, the OSRG merely expressed the concern that the 1990 disposition of the diesel fuel oil piping issue may not have been adequate. 1-144. The NRC echoed the OSRG's concerns that the actions taken at that time were not sufficient. The problem had been first identified in February of 1990 during ll l i

4 I i - 4 i the performance of the ten year pressure test of the system. _ 'Jhis test requires a visual j check for leaks, but not for corrosion. NRC Staff Direct Testimony, ff. Tr. 2159 at 12. 1 Action taken by PG&E included: 1) correction of the problem found and 2) shortening i the test interval to five years. These actions at that time were not -sufficiently ~ -l comprehensive or conservative to prevent recurrence. Id. An additional NRC concern involved the coordination bemaa the maintenance and operations orgmai=* ions, in that the other fuel oil system had been taken out of service leaving the corroded system in service. However, later PG&E determined, by analysis, that the fuel oil system had 1 been opemble despite the corrosion. Id. In sumniary, the surveillance test designad to j catch this sort of problem did, in fact, catch it. 'Id. at 13. i I-145. Thus, while further actions arguably could have been taken in 1990, in the. j 1 vast majority of cases in which a problem has been identified, PG&E has taken prompt 1 action commensurate with safety significance. NRC Staff Direct Testimony, ff. Tr. 2159 at 13. 1-146. MFP expresses concern that PG&E will not respond appropriately to this i issue. See MFP Proposed Findings 300-303. MFP also postulates that, since there were no maintenance procedures in place prior to these occurrences, the program is deficient in its entirety. See MFP Proposed Finding 310.- MFP ignores the record. '1here is i ample evidence that PG&E is now addressing the issues and there is nothing in the record l t to contradict PG&E's assertions that the problems have been addressed and that programs are underway. Rather, the volume of NCRs in the record demonstrate that PG&E has addressed and resolved issue after issue. As the Staff pointed out, "in the vast majority

e j .g. i i i l of cases in which a problem has been identified concerning equipment, by either the NRC .l or the Licensee's staff, the Licensee has promptly started corrective actions. The time l for the Licensee to fix equipment problems has been appropriate to'.the safety significance. The Licensee has typically returned equipment to service well before. expiration of the time allowed by the Technical Specification, and has had excellent per. l formance and management involvement in maximizing equipment availability." ' NRC f r r Staff Direct Testimony, ff. Tr. 2159 at 14.~ l 1-147. We decline to find deficiencies in a program based upon a skewed view ~ .1 that does not reflect the on-going overall program in so complex an endeavor as operating l l a nuclear power plant. None of the documents show that PG&E was in violation of Tech - Specs or outside its design basis, or that the public health and safety was affected. P. Control of Maenrine and Test Eauipment (M&TE) = l I-148. MFP Exhibits 65, 66, 67, (admitted Tr. Ill2) and 69,- 70, and 71' { (admitted Tr. 2197), all relate to PG&E's program to control Measuring and Test Equipment (M&TE). These exhibits refer to an NRC inspection in February 1991 (MFP \\ Exhibit 69) that found deficiencies in the M&TE programs administered by 'the Mechanical Maintenance Section at Diablo Canyon. The NRC also found (MFP Exhibit : 4

70) that similar deficiencies had been identified previously by PG&E's own' Quality l

t Control (QC) and Quality Assurance (QA) Departments which had not been aggressively : l corrected at the time. PG&E Direct Testimony at 102. I-149. These deficiencies resulted in a non-escalated NRC enforcement action (one - Severity Level IV violation). MFP Exhibit 71. However, this historical enforcement l , -,. +

action, dating from 1991, does not indicate any current MATE problem, nor does it controvert other testimony specifically establishing the current effectiveness of that program. L See, e.g. Tr. 2192-% (Miller). 1-150. MFP Exhibits 66 and 67 specifically document PG&E's corrective actions to address the 1991 violation. MFP Exhibit 66 is the PG&E response to the Notice of .. Violation filed'with the NRC,' documenting actions to be taken not only to address M&TE control, but also to improve PG&E's timeliness of problem resolution. MFP Exhibit 66 at 3-4. Corrective actions are also documented in PG&E's NCR (MFP l [ Exhibit 67) on the subject initiated in the same time period. MFP Exhibit 67 at 19-28; l Tr.1103 ('Giffin). l-I-151. PG&E provided uncontroverted testimony on the scope of the program to control M&TE. - PG&E Direct Testimony at 58. The current NRC Staff senior resident ing+2ei also testified that she is satisfied with the current M&TE program based on her own in depth follow-up inspections. Tr. 2192-94 (Miller). I-152. - MFP Exhibit 65 is a PG&E NCR noting that PG&E QA identified several i more recent instances 'of M&TE documentation deficiencies. These minor deficiencies i j ~ involved 'a failure to document use of MATE in either the work order or the PIMS l program. The earlier deficiencies cited by the NRC involved multiple failures to document the use of M&TE in either place; thus, the current NCR is distinguishable. Tr.1099-1101 (Giffin). Corrective actions have been taken for the recent specific issue and QA conducted a review during the most recent outage and found no similar - problems. Tr.1100 (Giffin); Tr.1103-4 (Vosburg). The NRC Staff inspector also I i I -o. ..-,.. ~ .n.

l . concurred that this 1993 NCR does not in any wsy disturb her overall conclusion regarding present M&TE control. Tr. 2194-% (Miller). I-153. The Board finds that these exhibits undoubtedly document a problem in the M&TE program in the 1989-91 time frame. However, we cannot ignore uncontroverted testimony by both PG&E and the NRC Staff that this problem has been addressed to the satisfaction of the NRC and no longer represents a matter of concern. Tr.1101-03 (Giffin); Tr. 2195 (Miller). On the record before us, we find that PG&E's corrective actions have been satisfactory in this instance. Q. Degraded Coucline on Centrifugal Chareine Pumo (CCP) I-154. MFP Exhibit 73 (admitted Tr.1125) is a PG&E NCR documenting the identification of an increase in vibration on Centrifugal Charging Pump (CCP) 2-1 by Predictive Maintenance. MFP Exhibit 73 at 1. The scope of the predictive maintenance programs includes vibration monitoring and diagnostics. PG&E Direct Testimony at 39-40. I-155. This MFP exhibit documents PG&E follow-up on vibration indications in a CCP motor. PG&E found a coupling sleeve in the speed increaser side of the pump to be stiff due to hardened lubricant. MFP Exhibit 73 at 2; Tr.1121 (Ortore). The condition was corrected and the NCR documents PG&E's investigations into the cause of the hardened lubricant. Tr. 1121-27 (Ortore); MFP Exhibit 73, Attachment. l

l .i 1 l. l .i i-I-156. There had been at least one previous occurrence of vibration problems with - 1 l this CCP 2 and apparently the corrective actions in the previous event had not been= j l properly implemented. MFP Exhibit 73 at 9-10. However, we agree with PG&E that - -l '1 this is an example of precisely what the predictive maintenance program is designed to ~ l do: detect potential problems while the equipment is stih able to perform'its safety. f function. In this case, the defective coupling was rMeW prior to any failure. M. at 4.- R. Inoperable High Pressure Turbine Stop Valve l I-157. MFP Exhibit 74 admitted by the Board (Tr.1138) is an LER documenting =! a unit shutdown that commenced in March 1992, in accordance,with Technical 1 i Specifications, when PG&E determined that one'high pressure turbine stop valve was l i L inoperable. The immediate cause of the inoperable valve was attributed to. equipment l failure. MFP Exhibit 74 at 3-4. PGAE testified that this' equipment problem is not-i l related to a maintenance deficiency. Tr.1126-27 (Vosburg). PG&E documented in the l LER its investigation into the root cause of the component failure. MFP Exhibit.74 -] at 4-5; Tr.1127 (Vosburg). Installation and maintenance on these valves is contracted l to Westinghouse, the manufacturer, which did not definitively establish the root cause. - Tr.1128,1135 (Giffin). The root cause was narrowed down to two possibilities: one is that the pins used to stake the nut that attaches the disk to the swing arm of the check i l valve had come out, allowing the nut to back off and the check valve disk to fall off. q

We note that on p.1 of this NCR,.it is stated that _"this is the third occurrence involving CCP 2-1", but the only other occurrence, referenced in the document is one q

from April 1989; thus, it is unclear on the face of the document as to what the reference on page 1 pertains.

-G-i Tr.1127 (Vosburg). The other possibility was that the nut had been pulled off the threads on the stem. Id. No matter which of the two root causes one selects, the t corrective action to prevent reoccurrence would be the same. Tr.1128 (Giffin). I-158. Other similar valves were inspected and verified to be satisfactory. Tr.1128-29 (Vosburg). Corrective actions being taken pending the final root cause determination also include acoustic monitoring of Unit 1 and Unit 2 stop valves to demonstrate operability. MFP Exhibit 74 at 6. Additionally, preventive maintenance has been enhanced to specifically address both of the two possible root causes of this equipment problem. Tr. 1134-35 (Giffin). I-159. The Board does not find that any deficiency in PG&E's maintenance and surveillance program is evidenced by this random and isolated occurrence involving a single valve. S. Diesel. Generator Failure to Achieve R*d Voltare I-160. MFP Exhibits 75 and 76 (admitted Tr.1145) document an instance of a personnel error in a maintenance activity. MFP Exhibit 75 is an NCR and MFP Exhibit 76 is a special report to the NRC, both addressing the same incident. On December 29, 1992, Diesel Generator (DG) 2-2 was being surveillance-tested in accordance with procedure as a post-maintenance test. The DG started and accelerated, but did not load because the generator did not achieve the rated voltage. MFP Exhibit 75 at 1; MFP Exhibit 76 at 1. i i I-161. Upon investigation, PG&E determined that all four generator slip ring i brushes were out of position. MFP Exhibit 76, Enclosure at 1. 'Ihis'mispositionmg was i j

attributed to inadvertent and undetected movement'of the slip ring. brushes due to j i loosening of brush rigging mounting bolts during inspection of the shaft end for thrust load axial movement followed by retightening of the mounting bolting without slip ring. .j brush position inspection. Id. 'Ihis was deemed a psiegel error by a mechanical maintenance worker. Tr.1140 (Giffin). I-162. In its root cause analysis, PG&E identified the need for advisory or caution l i l statements in the maintenance instruction regarding the inspection of slip ring brushes. j ~ i MFP Exhibit 76, Enclosure at 2. 'Ihe maintenance manual was also to be revised to I address this point to minimize the potential for future similar errors. Id. at 3.1 PG&E -j will also ensure that electrical personnel are involved when ' work involving the generators is performed. Tr. I144 (Giffin). l I-163. MFP would have the Board find that the program is deficient in its entirety because the procedures in this instance were not satisfactory. However, MFP chooses i .I to ignore the corrective actions contained in these documents. - As we have stated before, the entire document comes in for consideration, notjust selected portions thereof. Here maintenance was being conducted, a follow-up surveillance test was performed to verify that the maintenance had been properly completed, and the problem was identified and eviisted. PG&E also took actions to improve its procedures to prevent future similar errors. Tr. I140-41 (Giffin). i I-164. The Board declines to view this unsatisfactory surveillance as indicative of a defective surveillance and maintenance program. 1 i

, T. Missed Surveillance Tests j I-165. MFP Exhibits 77,78, and 79,- (admitted Tr.1153) address two separate issues of missed surveillance tests. The first issue, addressed in' MFP Exhibit 77 (a PG&E NCR), concerns surveillance of a Component Cooling Water (CCW) valve. 7he valve was stroke tested in October 1992 'and based on the stroke time results, which were in an alert range, the surveillance frequency should have been changed from every 92 days to every 31 days. The frequency change-was not correctly entered into the computerized schedule due to a personnel error. Tr. 1147-48 (Crockett). As a result, the surveillance schedule was not changed and the stroke test was not performed again until its normal 92 day frequency, at which time the error was identified. Tr. 1147-48 (Crockett); MFP Exhibit 77 at 3-5. Thus, the surveillance was missed twice in the - l interval. These two misses are included as part of the previously mentioned 65 missed surveillances. Tr. I149 (Crockett); see also discussion at finding I-106 above. 'Ihe' I .i second issue addressed in MFP Exhibits 78 (a PG&E NCR) and MFP Exhibit 79 (the a corresponding LER), involve two occurrences of personnel errors during surveillance testing of the ASW pumps. Tr. 1149-50 (Crockett). The first instance, on August 21, l 1 6 1991, involved a personnel error in the conduct of the surveillance test. ; The reviewer used an incorrect pump curve to determine the required differential pressure and thus-l failed to recognize that the ASW pump should have.been declared inoperable. MFP Exhibit 79 at 3. In this instance no surveillance was actually missed, and all schsequent tests actually showed the pump was indeed operable. Tr.1150-51 (Crockett). As a' l . J

c . corrective action, there is now an independent verification of the data sheet. Tr.1152 (Crockett). I-166. The second instanca related to the ASW pump occurred on November 14, 1991, when the surveillance was again performed and the reviewer failed to correctly evaluate the differential pressure to require that the testing frequency be doubled. Hence, a surveillance that should have been conducted in January 1992-was missed. MFP ) i Exhibit 79 at 3. This also constitutes one of 65 missed surveillance mentioned previously. Tr.1151 (Crockett). PG&E's corrective actions to this example are i documented in the exhibits and included a training session covering alert test frequency tracking. MFP Exhibit 79 at 6. I-167. Taken together, these documents do not in themselves suggest any programmatic breakdown in the DCPP surveillance testing program. The documents reflect personnel errors that do occur occasionally and PG&E has made procedural and program changes to help alleviate the potential for the particular errors recurring. Tr.1152 (Giffin). I-168. The Board takes note of the fact that the NRC Staff is satisfied that " performance of maintenance and surveillance at Diablo is considered to be superior and - clearly supportive of safe facility operation." NRC Staff Direct Testimony at 5. 'While MFP would require totally error-free performance, personnel errors do occur from time to time in any human endeavor. The record supports the fact that these errors were addressed and corrected and the equipment was not affected adversely.

i i i i I I ! .i U.. Anviliary Feedwnt- (AFW) Pumn Test Pro :edure t I-169. MFP Exhibits 81 and 82 concern an error ie the surveillance test procedure.- } for the ASW pump. MFP Exhibit 81 is a PG&E NCR that states that test procedure STP P-6B was used with an incorrect diagram on a nurr'oer of occasions for: routine r surveillance of the AFW pump.' This error was cited by the NRC as a Severity IevelIV. a i violation in October 1992 for failure to initiate an Action Request (AR) to identify the problem of test procedure STP-P6B and to evaluate the-potential impact of'carlier f r surveillance tests performed with the incorrect instructions on the operability of the pump. MFP Exhibit 82 is PG&E's response to that Notice of Violation. PG&E agreed -l with the violation. MFP Exhibit 82 at 3. " I I-170. The error was that a diagram used in a surveillawe procedure showed an L arrow pointing to the general location of certain bearing caps. Dunng a procedure _{ i revision process, the arrow was inadvertently drawn pointing to the housing of the pump - instead of the bearing. Tr. I156 (Crockett). A PG&E system engineer reviewed the procedure several months later and identified the problem as a typographical error and corrected the diagram. Id. at 1155. Although the procedure had been used in the interim, the diagram is only an aid and is not normally needed to assist an operator in : finding the bearing caps. Id. at 1156. There is no indication that the tests were not appropriately performed, while this error mmained in the procedure. Id. at 1157. i " We note that this exhibit also includes PG&E's response to a separate violation, but that matter was not addressed in the record and we do not address it here.

v I-171. He NRC's NOV regarding this incident was issued because PG&E's ~ q I system engineer, in correcting the procedure diagram error, did not initiate an Action l Request (AR) in accordance with PG&E's own procedures to track the matter and to ensure that tests conducted while the error was in the procedure were conducted correctly. Id. at 1156; MFP Exhibit 82 at 3. PG&E addressed this issue in its response; f to the violation. MFP Exhibit 82 at 4. l I-172. This minor error in a procedure had no effect on the actual test data and - is not so significant as to cause us concem about PG&E's overall surveillance and maintenance programs. This isolated occurrence bears no relation to any of the other . i issues put forth in the record. Standing alone, it cannot be representative of anything definitive about PG&E's maintenance and surveillance program. 1 V. Hold Down Motor Bolts on Centrifneni Chereire Pumns? t i I-173. MFP Exhibit 83 is a PG&E NCR (admitted Tr.1173) that reports several discrepancies in centrifugal charging pump 2-1 hold down motor bolts found by PG&E during preventive maintenance in July 1992. -De problems included unmarked bolts, bolts machined down to their root diameter, washers that had elongated holes and stacked - washers. In addition, other bolts were overtorqued. MFP Exhibit 83 at 2-3.' PG&E - determined that during the original plant equipment-procurement the procurement - l 1 specification and vendor supplied information did not adequately address the configuration of motor / pump hold down bolts for skid mounted equipment. MFP Exhibit 83_ at 1,4. l ~ i I-174. This procurement was completed during plant construction in the 1970's. Tr.1162 (Giffin). MFP focused'on earlier similar events concerning mounting bolts, j l i ,-.3

i .\\

I most of which concern 1983 and 1986. However, as the NCR itself states regarding t

I these earlier events, "In 1986 NECS GC operated under different requirements than. l DCPP." The current procurement program is a different program from what it was during the construction of the plant. Tr. I167,1173 (Giffin). MFP Exhibit 83 at 11-12. j See also Tr. 1165-67,1173 (Giffin). PG&E's actions to address the specific NCR. i introduced into evidence are clearly stated in the document. MFP Exhibit 83 at 6-9. l I-175. This single example does not in itself demonstrate any current problem with procurement of equipment. MFP has failed to show that the surveillance and l maintenance program is inadequate in failing to identify material' deficiencies. See Tr.1170 (Bechhoefer). As we pointed out in rejecting a contention on the subject of PG&E's procurement program, an isolated incident or two simply "does not raise a sufficient question to constitute an adequate challenge" to PG&E's procurement program; f LBP-93-1, 37 NRC 5, 23-24. W. Reactor Coolant System (RCS) Imhoe I-176. MFP Exhibits 84 and 85 (admitted Tr.1189) are a PG&E NCR and the corresponding LER related to an incident in August 1991. These documents involve an unusual event declared at the plant when a calculated RCS leak rate was determined to be above the.1.0 gpm rate allowed by PG&E's Technical Specifications. - Tr.1177 (Vosburg). The condition was+ identified during surveillance testing by PG&E's Operations Department and appropriate actions were taken to reduce the leak rate. Id. at 1177-78. i e i

4 .4 1 1 i 4 i' { I-177. A review of the previous leak check surveillance identified a calculation - error in that individuals performing the test made a mathematical error. and did not. l identify the excess leak rate at that time. For this reason, no action was taken within the ' j four hours required by Technical Specifications measured from the time of the earlier j test. Once the excess leak rate was identified, however, appropriate actions were taken i 1 [ within the Technical Specification time allowed. Tr.1179_ (Vosburg); Tr. I180 (Giffin); j 1 MFP Exhibit 85 at 5. 'Ihe personnel involved in the mathematical error were counseled i ' i i on the need for attention to detail and self-verification. MFF Exhibit 85 at 6. l I-178. The leakage was identified as originating in the Charging Subsystem of the l Chemical Volume Control System (CVCS). The Charging Subsystem returns coolant to i 4 the RCS during normal plant operations following chemical treatment by the CVCS.. 1 f MFP Exhibit 85 at 3. PG&E reviewed similar valves on both units potentially subject - ) to leakage. Tr. I182 (Giffin). The corrective actions taken will minimize the potential i 3 j for bolting corrosion similar to that which led to the RCS leakage because the carbon steel bolts have been replaced with stainless steel bolts'that are not subject to boric' acid degradation. Tr. I184 (Giffin); MFP Exhibit 84 at 14. Existing surveillance is adequate to detect leakage. Tr. I185-87 (Vosburg). I-179. MFP argues that PG&E's corrective action regarding these valves is ' vague and indefinite." MFP Proposed Finding 420. MFP Exhibit 85 at 21-29. MFP si.lectively cites to the TRG meetings minutes, ignoring the many' actions taken, and the testimony that the problem has been resolved. Tr. 1184-87 (Griffin', Vosburg);' MFP Exhibit 84 at 14. However, as PG&E testified, the NCR yw is cumulative; NCRs w w+.w%. .-.a.- --..,w % me cw...e'cw-s-e,w,--%--

i l are continually updated until closed out and this one has been closed out. Tr. 1188-89 l (Giffin); MFP Exhibit 85 at 29. j I-180. Further, MFP argues that the maintenance and surveillance program was - i " negligent in its obligation to detect degradation of equipment before it became self-evident." MFP Proposed Finding 413. However, MFP ignores the safety analysis, { i which states: " Ample time is available to detect leakage, evaluate it, and take the proper i and prudent compensating action. The leakage rate was carefully monitored and the rate l of increase was evaluated by the trend of the results of STP R-10C. When the detected leak rate did exceed TS limits, appropriate actions were taken to mitigate the leak rate." MFP Exhibit 85 at 12. I-181. To the extent that monitoring was conducted, the surveillance program could not be deemed negligent. Corrective action to check the valves and replace the 'i bolts perhaps could have been taken somewhat sooner, but this event does not evidence a breakdown in the overall program. X. Inonerable Panetor Cavity Sumo Wide Range I2 vel Channal 1-182. MFP Exhibits 86, 87, and 89 (admitted Tr.1202) and Exhibit 88 (admitted Tr. 2201) address an issue that was addressed in the Direct Testimony of both PG&E and ' the NRC Staff. Containment reactor cavity sump wide range level channels are i post-accident instrumentation used to provide quantitative data to the Afety Parameters j Display System (SPDS) about the water level inside the containment structure. "These j i data are used to verify the occurrence of a loss-of-coolant accident (LOCA)'and to ' j i i

= - - i o j l 77-evaluate plant conditions to assure proper response to an accident." MFP Exhibit 87 j at 5. I 1-183. A normal channel indication of 0 is difficult to distinguish from a failed channel indication of slightly below 0. But the SPDS provides notice of a failed channel by displaying a question mark when there is a problem with input data. Id. at 2. A blue flashing path means that the channels supplying data to SPDS for that logic path are not operable. MED Exhibit 89, Enclosure at 1. i 1 1-184. The first inoperable level indication occurred in November 1990..The immediate :ause was a blown fuse. However, after extensive evaluation, PG&E could not identify the root cause because of the intermittent nature of the failure. PG&E Direct Testimony at 94; Tr.1194-5 (Crockett). Subsequently, in October 1991, another ) intermitteat failure occurred. While again the root cause could not be found,' PG&E replaced the instrun,ent and much of the interconnecting cable. Tr. 1194, 1198 - i (Crockett). This has resolved the problem, to the extent there has been no other intermit'ent failure since 1991. Tr.1199 (Crockett); MFP Exhibit 86; MFP Exhibit 87. 1-185. MFP Exhibit 88 is an NRC Inspection Report, 92-01, that enclosed a Notice of Violation. As the NRC Staff explained, in 1992 PG&E was cited for failing to maintain operable one channel of the Reactor Cavity Sump Wide Range Level Indication in October 1991. MFP Exh' bit 88. The NRC Staff was also concerned that a similar problem had occurred in November 1990. However, the NRC Staff's concern in both cases was not the equipment problem in itself, but the cperators' lack of-i

. awareness for several days of the problem. NRC Staff Direct Testimony at 11; Tr. 2199 (Nazbut). 1-186. The delay by the operators in ranagnidag the failed indicators was attributed in part to the fact that the indicator failed low,' thus giving the same indication as a normal indication of 0 on the instrument. Training was given to the operators following the November 1990 occurrence on interpreting information available on the. Safety Parameter Display System (SPDS) with respect to failed indications. PG&E Direct Testimony at 94. Following the October 1991. occurrence, additional corrective actions were taken, including procedural requiremems for the operators to check the SPDS, which plainly shows failed channels, each shift. Id. at 94-95.' Corrective actions are also documented in PG&E's response to the Notice of Violation. MFP Exhibit 89 at 2-3. I-187. MFP argues that PG&E does not maintain the SPDS in a reliable condition and there is uncertainty in the maintenance of the system, which threatens the public' health and safety. MFP Proposed Finding 430. MFP misunderstands the system: it is not the SPDS itself that is unreliable, it was the sump level indicator that was broken. Further, as PG&E explained, this particular indicator is only one of many parameters used to determine the occurrence of a loss ~of coolant accident (LOCA). Tr. 1200-01 (Vosburg). Also, there is no regulatory requirement for the system to be seismically qualified. Tr.1197 (Giffin). 1-188. As the Staff stated, "it is clear from other NRC inspection activities that - the operator knowledge concerns found with the level indication ' surveillance are.not

~_ - ~ c - D-widespread." NRC Staff Dimet Testimony, ff. Tr. 2159 at 11. The Staff also stated that - PG&E's " subsequent corrective actions appropriately addressed 'each of the concerns ~ i =@=wl with the level indication, as_well as generic actions to ensure that instruments ^ like the sump level indications are properly maintained and monitored." Id. This item has been closed by the Staff and is followed up by focused inspection. Id.; Tr. 2203 (Miller). This item does not indicate any. inadequacy in PG&E's surveillance and.. maintenance program. j l i Y. Decien Criterion Memorandum (DCM) Reauirements I-189. MFP Exhibit 90 (admitted Tr.1211) is a PG&E NCR that relates to a PGAE initiative to upgrade the design basis documentation for DCPP. 'Ihe NCR relates j to one specific issue that emerged from that initiative that is related to maintenance and. l surveillance requirements. PG&E Exhibit 23 (admitted Tr.1840) is' a later' draft of the l NCR addressing the same issue. I-190. PG&E's design documentation initiative will improve the design 2 documents, enhance the engineering understanding of these documents, and improve document accessibility. Tr.1202-3 (Crockett). It is an initiative specifically encouraged-by the NRC and is considered to be a program strength. Tr.1024-5 (Crockett). 1 1-191. The issue in the NCR relates to differences between the design documents and existing maintenance and surveillance. The discrepancies identified to date have not i i been significant. Id. at 1203. A typical example is that no survaill=r* test existed to i verify the function of the diesel fuel oil day tank low level switch transfer pump. MFP Exhibit 90 at 2. This specific discrepancy is not significant because the operating l g, --,.----.w.,.,,,-.a,-,w, e-,n..-,.,b.-,,---..,.,,,-- e-

l I L l i procedure calls for the diesel fuel oil transfer pump to be started manually when the j diesel generator starts. Tr.1203-4 (Cmckett). l 'I I-192. These NCRs readdress an ongoing PG&E evaluation. Although not completed, the initiative is a program enhancement and, based on current evidence, is not [ l necessary to achieve an adequate maintenance program. Tr.1206 (Giffin). We agree with MFP that consistency between design and maintenance is fundamental to an adequate f i maintenance program. But we note that there have been pilot programs before that have t looked at the consistency between the surveillance and maintenance program and the' design documents. Tr 1207 (Crockett). No significant problem has been identified in PG&E's surveillance and maintenance program in regard to design basis documentation. I Z. Isohtad Pipe Suonort Snubber Dammee i I-193. MFP Exhibits 91 and 92 (admitted Tr.1222) describe an event conceming { a damaged snubber at pipe support 1-171SL on a main feedwater flow control bypass line for Steam Generator 1-1 found by PG&E during a structural inspection walkdown. MFP [ Exhibit 91 at 1. MFP Exhibit 92 is a voluntary LER filed by PG&E for the NRC's information purposes. Tr.1212 (Giffm). The function of a snubber is to allow a line .l to move smoothly if there is any shaking, mainly in a seismic event. Tr.1213 (Giffin). This snubber does not serve a safety function. Tr. 1214-1215 (Giffin). Investigation by l l PG&E concluded that the snubber had locked in place during plant operation. Subsequent compressive thermal loading that ce.mW during system cooldown as a result of a plant I ~' trip caused the snubber to buckle. MFP Exhibit 91 at 1. 1

i I-194. PG&E determined that the failure of an internal part (verge wheel) had i caused the snubber to lock in place. Evaluation of the failed verge wheel conducted by i Technological & Environmental Services (a PG&E organization) determined that the failure was due to stress corrosion cracking. MFP Exhibit 91 at 1. 'Ihe snubber damage in this case was attributed to a combination of three factors: the verge wheel material l i (440C stainless steel, heat treated to high strength level) is susceptible to stress corrosion j cracking when exposed to a contaminated environment and high tensile stress; the outdoor l 1 salt air environment; and the pinion hole in the failed wheel was smaller than the j dimension specified by the manufacturer, which resulted in a large amount ofinterference I which created tensile stress in the wheel.. Id.. The plant trip in April 1992 had caused a full thermal cycle on the line. MFP Exhibit 91 at 3-4. 3 1-195. As corrective actions, during the IR5 outage PG&E removed and tested j Unit I snubbers of the same type located outdoors. All passed, but some ' snubbers were nonetheless replaced with a different type and others were overhauled. Id. at 5, 7,11. Unit 2 snubbers of this type located outdoors are to be replaced and caps or rubber boots ~ are to be installed to protect the snubber from water intrusion. Tr.1213 (Giffin). 1-196. This incident is related to a manufacturing defect. PG&E has a program to inspect and test snubbers at DCPP although it cannot detect internal stresses on the componentsinside. Tr. 1218-19 (Giffin). This particular condition would not have been identified absent disassembly of the snubber. Id.

a l t i i AA. Gas Decay Tank Miccad Surveillanm f 1 I-197. MFP Exhibits 95 and % were admitted by the Board (Tr.1227) and relate - to a missed surveillance test. MFP Exhibit 95 is the PGAE NCR and MFP Exhibit % is the corrsporgiing LER. Radiation monitors are installed on each' Gas Decay Tank. l (GDT) to provide alarms in the auxiliary building control room and the main control room in the event that the quantity of radioactivity in the GDTs approaches'the limit of. j t 10,000 curies of noble gas (as XENON-133 equivalent). MFP Exhibit 95 at 6. l t I-198. Technical Specification 4.11.2.6 requires that the quantity of' radioactive I material contained in each GDT shall be determined to be within the limits set forth in' ~ TS 4.11.2.6 at least once per 24 hours when radioactive materials are being added to the l tank. Id. at 2. However, on October 12,1992, the time limit for the surveillance was i exceeded when a technician failed to perform the ' required GDT within the 24 hour time { limit. Id. I-199. The root cause of this event was inadequate instructions. The instructions on the checklist referred to the test as " daily" rather than "24 hours." 'Ihe technician - could not get to the area to perform the test because the area was secured due'to wet l paint on the floor. Because the test was listed as a daily test, he assumed that he could perform the test later in the day. MFP Exhibit 95 at 3; Tr.1225 (Crockett).. Within the allowed tolerance for a surveillance, the test was in fact due at 2:20 p.m. However, it ' I i was not performed until 4:25 p.m. - approximately two hours late. Id. at 1226; MFP Exhibit 96 at 4.

_,,s -A E 4 -ae a-83 - t I-200. This missed surveillance was a 24-hour Chemistry Department surveillance and not a maintenance surveillance. Tr. 1224-25 (Crockett). PG&E has taken corrective actions to address this incident, including revising the shift turnover checklist to specify - the chemistry surveillance as a 24 hour surveillance. MFP Exhibit % at 5. I-201. The surveillance was missed only once. Tr.1224 (Crockett). This missed i surveillance is one of the 65 reportable instances of surveillance test misses discussed earlier. Id. This incident has nothing to do with the maintenance program at Diablo Canyon and does not represent a programmatic breakdown. As discussed earlier, the total number of missed surveillam at DCPP does not reflect, in context, a pervasive problem with the PG&E survtillance and maintenance program. i BB. Seismic Clips I-202. MFP Exhibit 98 was admitted by the Board (Tr.1249) and is an NCR l related to PG&E's discovery in October 1992 that the Unit I reactor trip and bypass l breakers did not have seismic clips installed as required by plant procedures. Immediate i actions were taken to install the clips. MFP Exhibit 98 at 9. The Unit 2 breakers were inspected to ensure that all clips were properly installed. No problems existed with those breakers. Id. at 9-10. I-203. PG&E analyzed the system performance without seismic clips and determined that the system would still perform its safety function of providing a P-4 signal and tripping the turbine. MFP Exhibit % at 8. Without the clips installed during a seismic event, the turbine might be tripped before the reactor trips, but this is-a previously analyzed event. Id. at 8-9. No inoperable conditions existed; thus, there was

I 4' D l l i no potential impact on public health and safety. Id. MFP is mistaken in asserting without any proof that these breakers would not have activated as required (MFP Proposed Finding 464), as the Westinghouse evaluation notes that missing seis:nic clips will not prevent the breakers from performing their safety function. Id. at 13; Tr.1242 l (Vosburg). t I-204. The identified situation resulted from confusion between Operations andI l I&C following testing of the breakers. Tr. 1244-45 (Vosburg). PG&E invaetiW the. l incident and identified a need for procedural or programmatic controls to ensure that reactor trip and bypass breakers are secured with seismic clips prior to entering a mode where they are needed. These actions are being taken (MFP Exhibit % at 10) and are. commensurate with the scope of the problem identified. Tr. 1247-49 (Vosburg, Giffin). l I-205. The record is clear that this isolated incident involving seismic clips standing by itself, as it does, is not symptomatic of any pervasive failure in PG&E's I maintenance and surveillance program. CC. Containment Fan Cooler Units I-206. The problems experienced with the Containment Fan CoolerL Units (CFCUs) was one of the bases for MFP's Contention 1. See 37 NRC at 14-21. Both the i i ) -licensee and the NRC Staff W the admission of the contention and of this basis. l The Staff argued that MFP's reliance on an Enforcement Conference held to' discuss the' l 1 1 possibility of eenhted enforcement on that matter was improper in that the Staff's j position was superseded by the Notice of Violation that subsequently issued. The Staff argued that to the extent MFP's reliance was on the CFCUs' inoperability, that reliance ) i

.1 l s i ; l 1 wts also improper, as the NOV concluded that the CFCUs were never inoperable. See .l NRC Staff Testimony at 8 (Narbut). However, the Board held that the Staff's objections did not indicate a lack of basis but rather constituted an evidentiary matter. 37 NRC q at 19-20. I-207. Both PG&E and the NRC Staff addressed the matter in their testimony. PG&E Direct Testimony at 88-89; NRC Staff Direct Testimony at 7-9. Containment Fan. 1 Cooler Units (CFCUs) are used to cool the containment etir@w and equipment' l . I located in the containment building in normal operation and, in the event of a LOCA, they serve to limit the pressure peak in containment in conjunction with the~ containment - j a i spray system. NRC Staff Direct Testimony at 7 (Narbut). There were two problems l 1 with the CFCUs at DCPP. The first problem, found by plant' personnel, surfaced in. l February 1992. It involved the failure of the backdraft dampers to close.. Id. Ihe second problem, discovered in October,1992, involved cracking of the backdraft damper -j i blades. Backdraft dampers prevent reverse flow through CFCUs not in operation. Id. l I-208. In January,- 1992, the Licensee identified that some of the counterweights 4 for the CFCU backdraft dampers had fallen off. The counterweights close the dampers. .{

14. Subsequently, the NRC and the licensee concluded that errors had occurred during the previous refueling outage in reassembling the-dampers in ' Unit 1, that post.

~ maintenance testing instructions were not sufficient,' that reinspections done in March I 1992 were done without appropriate procedures and that the licensee had not done an in-depth review when a bolt was found on adjacent deck grating and fan reverse rotation was noted a year earlier, in March 1991. NRC Staff Testimony at 8 (Narbut).- ,,...-_l _ _. _. ~, _ ,, ~,.

.) . I I-209. 'Ihe CFCUs were subsequently analyzed to have been operable during the ) entire period, but this was not evident at the time. Id. j I-210. In October 1992, cracking cNe to fatigue was discovered in some of the - 1 s backdraft damper blades in Units I and '2. "Ihe blades were replaced with blades of j higher strength material. Id. l I-211. The NRC identified several causes for the licensee's failure to discover [ these problems earlier: 1) the quality organization was not sufficiently involved; 2) the i maintenance organization tried to resolve problems such as broken bolts and reverse rotation without involvement of the engineering organization; and 3) there was a lack of _ l I attention to detail by maintenance personnel in performing the reassembly work and q communications regartling the condition of the dampers were not effectively relayed to j i i PG&E management. Id. I-212. The resolution of the CFCU problems took several months. Id. at 8-9. Subsequently, the Licensee put together a team to perform an integrated review of the problems. The maintenance and surveillance test procedures were improved. Id. In j l addition, as a result of the team's recommendations, the licensee provided guidance to j t engineering personnel to define their responsibilities related to maintenance and-l 1 component tending, - added additional personnel - to the system engineering staff, - reemphasized more direct communications beres those denhng with plant problems, ~ reevaluated inspection activities, re-emphnimi the development of a questioning attitude - i in the reviews of plant problem reports, and required additional training in the evaluation of degraded plant equipment. Id. . i 1

i l l j i i 1-213. As noted above, both PG&E and the NRC $taff addressed the events j leading up to the NOV issued with regard to the CFCUs in their direct testimony. PG&E l l Direct Testimony at 88-89; NRC Sta*f Direct Testimony at 7-9. MFP offered six ' l:! exhibits related to that subject: MFP Exhibit 100, NRC DC492-MM-N022 Ganuary 4, l 1993); MFP Exhibit 101, LER 1-92-023-00 (November 20,1992); MFP Exhibit 102,- i l NRC IR 92-17 (May 8,1992); MFP Exhibit 103, LER 1-91-019-01 (June 5,1992); MFP 4 i Exhibit 104, NRC DCO-92-MM-N007 (February 12,1992); and MFP Exhibit 140, NRC i l Management Meeting, Report 92-3 (April 16,1992).' We admitted two exhibits j t j introduced by the NRC Staff, NRC Exhibit 1, NRC Enforcement Conference 92-19 (June i 18, 1992) and NRC Exhibit 2, NRC NOV from IR 92-17-(June -19,1992) and one l introduced by PG&E, PG&E Reply to NOV in NRC IR 92-17.(July 20,1992). 1 i I-214. We do not adopt MFP's Proposed Findings concerning the' CFCUs for l l several reasons. See MFP Fwposed Findings 473-508. MFP's Proposed Findings i i incorrectly indicate that the CFCUs were inoperable. Although the CFCUs were j conservatively declared inoperable in February 1992, subsequent analysis determined that 1 j they were never inoperable. NRC Exhibit 2 at 1. Contrary to' MFP's characterization in Proposed Finding 480, the NRC did not " cite PG&E for four apparent Severity level l IV violations." (MFP's Proposed Finding 480 gives MFP Exhibit'102 at 2 as authority j for the finding). Rather, the exhibit idenn)fes three apparent' violations: "(1) operation j of Unit 1 in Modes 1, 2, and 3 with three inoperable CFCUs between March 27,1991 and February 22,1992, contrary to Technical Specification (TS) 3.6.2.3.a; (2) failure to i f take appropriate corrective actions after observing reverse rotation of Unit 1 CFCUs on 1 3

a y i -M-i I I j March 25,1991; and (3) failure to follow approved procedures (and Work Order instructions) while inspecting Unit 2 CFCU dampers in February 1992." The NRC subsequently cited PG&E for three Severity level IV violations, as found above, and r noted that the apparent violation regarding CFCU 'umperability had been eliminated, as. i l the NRC agreed that the requirements of CFCU Technical Specification 3.6.2.3 had not l been violated. NRC Staff Exhibit 2.8d I I-215. We found the NRC's testimony offered at the hearing to be persuasive on - 1 i the issue of the importance of the CFCU event. Ms. Miller, the ' senior resident i h inspector, and Mr. Narbut, Regional Team leader, Region V, were in agreement that - l the CFCU event was oflow safety significance. Tr. 2220 (Miller, Narbut). Mr. Narbut._ } noted that the NRC concentrates on negatives, but that the positive of the CFCU story-l j was that PG&E employees reacted when they found parts on the decking. Tr. 2215 { i (Narbut). i 1-216. We find Ms. Miller's assessment of the significance of the CFCU. event I revealing: Ms. Miller said that where events such as those surrounding the CFCUs, the main feedwater pump and containment sump were "your total population for a SALP. period, that speaks to a well-run plant." Tr. 2220 (Miller). 1-217. In sum, although PG&E's maintenance in regard to the CFCUs was "not good " Tr. 2214 (Miller), that incident does not of itself or even in relation to other i incidents indicate a programmatic problem with' maintenance at' DCPP.- .l t [ 24 We do not adopt MFP's Proposed Findings 481 and 483, as MFP's reliance on the - Enforcement Conference is misplaced to the extent it was superseded by the NOV of-- I June 19,1992. 1 4 i I '!l .] l

e 1 i 89 - DD. Debris Mousekeeping) Debris Issues ') J l I-218. MFP Exhibits 105,106,' 107 and 108 are NRC documents that were 1 l admitted by the Board (Tr. 2241) and MFP Exhibits 109,110,111 and 113 are PG&E j = l documents (admitted Tr.1519). 'Ihese are series of exhibits loosely connected in that - ] '1 they all relate to control of debris. PG&E also introduced two exhibits on this topic (PG&E Exhibits 25 'and 26, admitted Tr.1518), and addressed the issue in two areas in j its direct testimony. PG&E Direct Testimony at 97-98, and 106.' '1he exhibits actually' i j address three very separate debris control issues:. (1) foreign material within the Residual ; j Heat Removal (RHR) recirculation sump lameM inside containment; (2) control of L matenal inside containment generally, but outside the RHR recirculation sump; and (3) i j the Foreign Material Exclusion (FME) controls. Tr.1508 (Crockett); Tr. 2235-36 1 l (Miller). ii I-219. The first issue is addressed in MFP Exhibit 113, a March 1990 PG&E i response to a Notice of Violation, and is also addressed in PG&E's Direct Testimony. l PG&E Direct Testimony at 106. 'Ihe RHR recirculation sump'in each' containment j building provides a collection point for water during a postulated accident, so that the - 1 water can be recirculated, cooled and returned to the system for accident ' mitigation.. } } These RHR sumps are surrounded by debris screens. M. The Notice of Violation issued i i j by the NRC cited three violations mlated to the RHR recirculation sump screens., One -I ( j of these violations involved inadequate engineering and construction completion, which - i resulted in unacceptable gaps in the semen structure. M. Another involved operational ) j control that allowed an unattended open access hatch in the screens. Neither of these two i i j i

.=. i. i -M-l violations bear any relation to maintenance or surveillance. Id. The third violation i included a reference to poor performance of a visual surveillance of the sumps, which j resulted in unidentified debris remaining in the sumps. Id.; MFP Exhibit 113 at 11. 1-220. This violation occurred in 1989 during the Unit I third refueling outage. Tr.1508 (Crockett). 'Ihe deficiency regarding inadequate visual surveillance of the j i sumps was an isolated personnel error - an individual failed to implement an otherwise l l clear procedure. ~ PG&E Direct Testimony at 106 (Crockett). Corrective actions are documented in MFP Exhibit 113 at 12. Since those actions were taken, PG&E has not i experienced any recurring problems in this area. Tr. 1508-09 (Crockett). I-221. The second issue grouped in this category by MFP relates to miscellaneous. material - particularly tools - left inside containment (but outside the' sump), identified ' as the plant heated up following outages. Tr.1509 (Crockett). The instances occurred i following the Unit 2 fourth refueling outage (in October 1991) and following the Unit l' ~ fifth refueling outage (in October 1992). PG&E Direct Testimony at 97. They are- { documented in MFP Exhibits 105,109, and 111. i . I-222. MFP Exhibit 109 is the PG&E NCR that addresses this issue, and was i revised to incorporate the more recent 1992 Unit 1 event.~ It documents a substantial l number of enhancements. to prevent recurrence. MFP ' Exhibits '109-112. Most significantly, important work orders now include specific instructions'on control of i material inside containment. Tr. 1509-10 (Crockett); see also PG&E Direct Testimony at 98. These corrective actions appear to have been successful. During the most recent i .-- - -._ ~, - - -. _ .-....s.

. Unit 2 fifth refueling outage there were no simdar findings. Tr.1510 (Crockett); Tr. 2237 (Mdler). I-223. There is a clear distinction between these first two debris issues although they both involve the RHR sump environment.' The first, debris in the sump, could result ~ in blockage of tubes in the RHR heat exchanger, which is why the NRC considered it - significant enough to issue a civil penalty. MFP Exhibit 113. 'Ihe second, debns m l l containment generally, relates to the potential blxkage of the sump screens. MFP l Exhibits 105,109,111. However, because of the sim of the containment sump structure, I it would take a very large amount of debris outside of the sump to have a significant - effect on the RHR pumps. Tr. 1510-12 (Vosburg). I-224. The third issue grouped in 'this category by MFP relates to PG&E's program of foreign material exclusion (FME). This program:is intended to prevent objects from entering a system while maintenance work is being pfvis.e4 on specific-q equipment. Tr.1512 (Crockett). MFP Exhibits 107 and 108 are two NRC inspection reports addressing an issue with respect to this program that arose during the Unit 1-second refueling outage in 1988. Id. PG&E took action to address this issue at that time, as documented in PG&E Exhibits 25 and 26, which are PG&E's responses to the two NRC inspection reports. Tr.1513 (Crockett); see also, PG&E Exhibit 25 at 2; PG&E Exhibit 26 'at 2. PGAE's actions to address the specific 1988 FME issue enhaami an already successful program. Tr.1514 (Crockett). I-225. MFP Exhibit 100 is a 1992 PG&E NCR addressing a number of more recent violations of FME area boundaries. The document identifies further program I -m

i i 1 j -M-t, ) i 1 enhancements. MFP Exhibit 110 at 7-14. FME is an issue that requires constant i j vigilance, and MFP Exhibit 110 shows only that PG&E promptly identified new issues j in the area and took action to address those issues. Tr. 1515-16 (Crockett). i i 1-226. MFP Exhibit 106 is another NRC inspection report reflecting a focused j review of plant shutdown issues, including FME. ' The NRC~ Staff identified that the [ L DCPP FME process did not include a procedural requirement' regarding. capping-j three-eighths inch instrument tubing that. supplies signals to pressure transmitters.- 4 i Tr.1516 (Crockett); Tr. 2227-29; 2237-38 (Miller, Narbut)..-This issue had minimal. ) safety significance given the low likelihood of material actually getting into the instrument j i lines at issue and the lack of use of the instrument tubing in plant operation.~ Tr. 1516-17_ j (Crockett); 2228-29 (Miller, Narbut); Tr. 2237-38 (Miller). The Staff cimr idaed this as the only finding after looking "a great deal" at the FME issue. : Tr. 2237 (Miller). l i Thus, the very specific and minor finding in MFP Exhibit 106 can be accorded little or l no weight in assessing overall performance in the FME area. I-227. A fourth issue related to debris, and only peripherally addressed in the { i documents offered by MFP, is general plant housekeeping. This issue is raised by PG&E .{ I in its own 1993 draft self-evaluation. MFP Exhibit 35, at Finding MA.2-1. This critical-evaluation, intended to foster continuous improvement, includes specific recommended actions to foster good howk@g. Raseri on the entire record'of this proceeding, 'I however, PG&E's current housekeeping at DCPP is good. Tr. 2239 (Miller). MFP j Exhibit 106 identifies housekeeping throughout the plant, and the policies and procedures in place regarding this issue, to be a DCPP strength. MFP Exhibit 106 at 21; Tr.1517 - ) l ,,,. ~, .., -. _.. -.. _..,. _ -.. _ - ~. .. -., ~ _. J

e a (Crockett). The NRC Staff witness also described the plant as "very clean" and identified i various strengths in this area. Tr. 2239 (Miller). j I-228. 'Ihe Board does not find these few instances. of unconnected events concerning ' debris" as evidence of any breakdown in PG&E's surveillance,and' maintenance program. There have been instances in the past of less-than perfect" performance in the area of foreign materials exclusion, but nothing in the recorti shows. that this is either widespread or on-going. EE. Stemm Generator Feedwn'ar NaEle Crmekine I-229. MFP Exhibit 117 was admitted by the Board (Tr.~ 1556) and documents ~ j another emerging equipment issue identified and addressed by PG&E. Ihis issue was also addressed in PG&E's direct testimony. PG&E Direct Testimony at 91-93. MFP Exhibit 117 is a voluntary LER addressing PG&E's identification during surveiH== tests. ofindications in steam generator feedwater nozzles. The exhibit also documents specific corrective actions taken by PG&E. i 1-230. The steam generator feedwater nozzle 'is a. 20 inch: diameter piping -l connection through which feedwater flows into each steam generator. 'Ihese nozzles and the immediate upstream piping are susceptible to intenor surface cracking as found at : other nuclear power plants. PG&E Direct Testimony at 91. 'Ihis surface cracking results from nozzle metal temperature difference caused bylcertain relatively infrequent operating flow conditions during which cold water is flowing into the ' steam generator through a hot nozzle. Id. The issue was identified at Diablo Canyon by a PG&E - engineer, who observed the problem at another nuclear plant and recommended that ..=. -.a.=.-

l i 94 inspections be conducted at DCPP. PG&E Direct Testimony at 92. PG&E investigated. the issue and initiated appropriate inspections and repairs. To minimize future potential problems in this area, a design change is being developed. Id. at 92-93. PG&E has also t conservatively decided to perform surveillance on the nozzles and adjacent piping during { each refueling outage. MFP Exhibit 117 at 6. I-231. PG&E's surveillance that detected this issue was one cycle ahead of the normal inspection interval. Tr.1554 (Giffin). At the time of discovery, the crack indications did not exceed code allowables and would'not have done so for at least another full cycle of operation. PG&E Direct Testimony at 92. Therefore, even following the normal scheduled inspection interval, the cracking would have been identified before it exceeded the code allowables. Tr.1554 (Giffin). The NRC Staff, f following a briefing by PG&E, concluded that PG&E's analysis of this issue was reasonable. Tr.1556 (Crockett). I-232. PG&E's handling of this issue evidences an effective maintenance and i surveillance program. PG&E assimilated industry experience, conducted prudent l inspections, performed a thorough analysis to identify corrective actions, and kept the NRC well informed. l 1 I-233. MFP argues in its proposed findings that PG&E did not syyivyriately respond to the NRC Information Notice, IE Bulletin 79-13, Revision 2, ? Cracking in Feedwater System Piping." MFP Exhibit 117 at 2; see MFP FivyOW Findings 550-53. However, that is not evidenced in any measure by'this exhibit. We are satisfied with PG&E's analysis of the issue.

l 95 - FF. Prneedural Controls Durine Shot "=.-ir. I-234. MFP Exhibit 118 was admitted by the Board (Tr. 2208) concerning; procedural controls of airborne radiation'during shot peening. Steam Generator Tube Degradation was also addressed in PG&E's direct testimony. PG&E Direct Testimony at 73-74. ream generator shot peening is a one-time (per unit) evolution conducted at - DCPP as an initiative to forestall steam generator degradation. PG&E Direct Testimony. at 73; Tr.1560 (Giffin). MFP Exhibit 118 is an NRC radiological in@ report addressing a health physics problem experienced during the first (Unit 1) implementation of this activity. Tr. 1558-60 (Giffin).. l I-235. The inspection report does not reflect a maintenance pmblem. Rather, it reports one violation regarding inadequate radiological controls during the shot peening. l evolution. This is a health physics issue. Tr. 2207 (Miller). 'Ihe violation did not affect l k in any way the success of the shot peening opr: ration. Tr.1559 (Giffin). 'Ihe Staff also observed that the shot peening itself appeared to be " exemplary." Tr. 2207 (Lwn). j I-236. PG&E Exhibit 22 (admitted Tr.1565) is the licensee's response to the. . I ~ health physics violation. It includes corrective actions. PG&E Exhibit 27 at 3. After i these actions were put in place, the shot peening evolution for Unit 2 was successfully { completed with no radiological control problems. Tr.1560_ (Giffin). ] I-237. MFP argues that this is related to maintenance to the extent the shot peening itself was a maintenance activity. MFP Pro osed Findings 560,556.- However, i based on the record before us, we are inclined to agree that this is a health physics matter i and not a maintenance matter. Hence, MFP FAtibit 118 can be accuids! no weight. I o m.m..m --m a +-e +-- 'rY4

4 96 - l The document is only peripherally related to nt intenance. Even with respect to the q health physics matter, radiological controls, it appears to describe an isolated issue that was addressed successfully. It rehts to an evolution that will not be conducted again at DCPP. GG. Unplanned Enrineered Safeiv Features ESF) Ar*=tirwis I-238. MFP Exhibits 119,120,121,122,122A,123,124,126 and 127 (admitted Tr.1588) are eight different exhibits that document five separate unplanned En'gineered Safety Features (ESF) actuations. 1-239. Engineering safety features are systems that are part of the plant design to mitigate postulated accidents. An unplanned ESF actuation at DCPP generally has little I or no effect on the plant. Actuation of safety injections is an exception since it results in a plant trip. Unplanned ESF actuations are reportable, by LER, to the NRC. 1 Tr.1567-68 (Vosburg). 1-240. MFP Exhibits 119 and -120 document one ESF actuation caused by a maintenance person performing a surveillance test in September 1992. Id.' at 1569. The ' root cause was a personnel error. MFP Exhibit 120 at 3. Corrective actions are-documented in the LER. Id. at 4. I-241. MFP Exhibit 121 documents an ESF actuation in March 1991, due to a personnel error in the Operations Dep rhi, cat while performing a surveillance test.' The individualinvolved turned the wrong switch. Tr.1569 (Vosburg). Corrective actions are documented in the NCR. MFP Exhibit 121 at 6-7. No actions to enhance the surveillance program were necessuy. Tr.1570 (Vosburg).

l 97 - i 1-242. MFP Exhibit 122 documents an ESF in July-1991, resulting from a-1 personnel error in the Operations Department while phming a surveillance test. - 2 Following an interruption in the test, the operator actuated a wrong switch. Corrective actions are documented in the LER. MFP Exhibit 22 at' 5-6. No actions to' revise the surveillance program were -

y. Tr.1570 (Vosburg).

I-243. MFP Exhibits 123 and 124 document one ESF actuation in May 1991, i caused by an error by a maintenance technician. In performing a surveiH== test, the. l l technician pulled a fuse in the wrong <hnanel. Id. at 1571; MFP Exhibit 123 at 5.' i l Corrective actions, including a physical barrier to prevent recurrence, ' arc documented in the LER. MFP Exhibit 123 at 6. i I-244. MFP Exhibits 126 and 127 document the last of these inadvertent ESF. ~ i actuations. This occurred when two maintenance technicians working on the' solid state protection system in October 1991, failed to follow the procedures for reconfiguring the system. Tr. 1571-72 (Vosburg). The root cause was personnel error. MFP Exhibit 127 at 4. Corrective actions documented in the LER included both counseling of the. responsible technicians and, in order to prevent recurrence, management communications emphasizing the importance of procedural compliance and'self-verification. Id. at 5' I-245. MFP discussed these five personnel errors at length (MFP Proposed Findings 567-587), but we find that taken together these disparate incidents do not reflect a surveillance program problem. They were isolated errors, with no connwtion between them. They exist in a context of a very large number of surveillances prrformed at the plant. Tr.1572 (Vosburg); Tr. 834-35 (Crockett, Vosburg). l u >v- + v--- -rs. e-

O O I-246. PG&E reviewed each incident in detail. Despite the generally low safety significance of ESF actuations, each was reviewed by PG&E through the NCR process to fmd potential enhancements to minimize the likelihood of future similar actuations. Id. at 1572-74. The record reflects a healthy approach to evaluating and addressing these errors. Id. at 1574-75; 1582. HH. Limitoroue Valve Failum I-247. MFP offered two documents, MFP Exhibit 128, NCR DC2-92-EM-NO26 D8 (September 17,1992) and MFP Exhibit 129, LER 1-92-010-00 (October 15, 1992). MFP Exhibits 128 and 129 are a PG&E NCR and an LER addressing a single incident in June 1992 of failure of a Limitorque valve operator due to a personnel error during assembly of the spring pack. The failure occurred during a test of the motor operated valve. Tr. 1589-90 (Ortore). 1-248. PG&E determined that the failure occurred because the worm cartridge bearing locknut and set screw had not been adequately tightened. While this was a personnel error, PG&E concluded that maintenance instructions could be enhanced with respect to tightening the set screw on the worm shaft locknut. MFP Exhibit 129 at 4. PG&E has taken actions both to address the personnel error and to enhance the instructions. Id. at 7-8. PG&E also looked at other valves assembled by the same individual to assure that any discrepancies had been corrected. Tr.1590 (Ortore). 1-249. MFP would have us find that PG&E did not communicate adequately with the vendor; yet MFP offered no record basis for such a finding. MFP Proposed Finding 592. MFP also speculates about PG&E's requirements for verifying the material content

4 a g a _ 99 _ i of replacement parts, or lack thereof. Id. No evidence was introduced to substantiate the appropriate application of procedures for verifying the material content of the worm shaft. MFP further conjectures that PG&E's maintenance and surveillance program is deficient in these respects. However, this incident was an isolated occurrence, in that only the Limitorque valves serviced by an individual craftsman experienced any i problems. MFP Exhibit 128 at 15. Moreover, it was identified during surveillance testing, investigated and subsequently corrected. All suspect valves were determined to be capable of performing their specified safety function (s). MFP Exhibit 129 at 7. No - j plant safety feature was adversely affected by the locknut on the wormgear shaft of the Limitorque operator coming loose. MFP Exhibit 128 at 5. Therefore, the Board finds that this incident has no programmatic significance except insofar as it shows that the program worked, as the failure was identified in testing and corrected. II. Motor Pinion Keys in Limitoraue Motor Ooerators I-250. MFP offered one document, MFP Exhibit 132, LER l-91021-00 (August 28,1992) that addressed motor pinion keys in Limitorque motor operators. MFP Exhibit 132 is a voluntary LER documenting a September'1991 incident of a sheared motor pinion key in a Limitorque motor operator. Tr.1615 (Ortore). MFP i proposed findings that the inadequate key material was inadvertently discovered or found. only through luck. MFP Proposed Findings 604-05. However, the only evidence in the record was that PG&E did identify this during a test while the unit was shut down for refueling and the key sheared only under a much higher stress than would normally be the case during operations. Tr.1617 (Ortore). In addition, PG&E received information t

i I i 1 l l - 100 - i i l ? j from another nuclear power plant that had experienced similar Limitorque key shearing l i problems. MFP Exhibit 132 at 3. I-251. PG&E identified the root cause of the component failum to be that the key. material was inadequate. MFP Exhibit 132 at 4. The vendor believes the material to be l adequate, but is now using a harder or stronger material for the keys. Tr. at 1615-16 (Ortore). MFP proposed the finding that PG&E mlied on the defective components and speculated that the sheared pinion key might not have been found until the valve failed during operation. MFP Proposed Finding 603. MFP offers no basis for this finding. 4 t Metallurgical analysis of the key indicated that the key failed due to shear overload and additional cycling of the valve would not have sheared the key unless excessive force was. j applied to the key. MFP Exhibit 132 at 5. Regardless, PG&E analyzed this incident and. determined that valve operators with sheared keys are still able to perform their safety i function. Tr. 1616-18 (Ortore). 1 I-252. MFP proposed a finding that PG&E's inability to detect and correct hidden j defects might have a significant adverse effect on safety if the defects were present in l both trains of redundant safety systems. MFP Proposed Finding 605. MFP has provided - no record basis for this finding. The testimony was that the sheared keys were detected and corrected, and that the valves with sheared keys were still able to perform their i safety function. Following this incident, PG&E examined other Unit 2 valve operators l and found two with similar problems. Corrective maintenance was' performed. MFP Exhibit 132 at 6. The PG&E witness also testified that all keys made with the original j i material have been changed. Tr.1625 (Ottore). This incident does not represent a ' i a l .,, _. -. _... ~ _ -. _. ~. - ~ -

- - 101 - j pmgrammatic breakdown in communications (see MFP Pmposed Finding 604),~ but i demonstrates PG&E's ability to conduct a thorough _ root cause determination and implement effective corrective actions. There is nothing to support MFP's conclusions - that PG&E was not able to detect and correct hidden defects. MFP provides no basis for finding that a " blind spot" exists (MFP 'Pivpd Finding 605), or that there is a significant weakness in PG&E's maintenance program. JJ. Contml of LifHnr and Rieeine Devim 1 I-253. MFP offered three documents, MFP Exhibit 135, LER 1-91-00402, Special Report 91-02 R1,- Diesel Generator 'l-1 Failure to Imd within TS Limits (July 29,1992); MFP Exhibit 136, NCR DC1-91-MM-N028 (October 23, ~1991); and MFP Exhibit 137, NRC 1R 92-16 (July 7,1992). The NRC Staff objected.to the admission of MFP Exhibit 137 as that document formed one of the bases for MFP's Contention 2 concerning personnel error, which was rejected by the Licensing Board as lacking in basis. MFP's counsel then restricted the motion for admission of MFP Exhibit 37 to "the maintenance issues only" and to matters concerning " loss of offsite power." Tr.2254. I-254. MFP now proposes a number of diverse findings on this subject matter. MFP Proposed Findings 602-32. We cannot accept them, as the theory underlying their admission is inconsistent with the prehearing conference order admitting MFP as an' 1 intervenor and admitting two of MFP's contentions, while rejecting nine others, as well i as MFP's statements at the hearing. See Tr. 2254. i m

- 102 - I-255. The event concerning the rigging of a waste shipment cask occurred on May 28,1992. PG&E Exhibit 27, Enclosure 1 at 1. The LOOP event occurred on 1 March 7,1991. MFP Exhibit 136 at 1. The only connection between these two events was an observation by the NRC inspector: l l The inspector observed that this event involved riggers and contained ) elements similar to the loss of offsite power event in March 1991 which was caused by a boom crane under the Unit 1500kV lines. The Manager cf Maintenance Services stated that the boom crane event in 1991 had been caused by personnel who were not riggers, but who had been trained to use the equipment. The inspector acknowledged that two different groups - of giwnnel were involved, but noted that both events appeared to involve. waahs in the preplanning and control of lifting or rigging activities. MFP Exhibit 137 at 4. PG&E challenged this Staff observation in PG&E Exhibit 27, PG&E Reply to NOV in NRC IR 92-16 (August 5,1992). PG&E concluded that the " LOOP event does not have any commonality with... the rigging incident." PG&E Exhibit 27, Enclosure 2. The PG&E witness also testified that these two incidents are not related. Tr. 1630,1635 (Giffin). The NRC Staff considers the issue of the similarity between these two events as unresolved. Tr. 2248-49 (Miller). There is no convincing l record evidence linking the two events; in fact, the weight of the evidence is to the contrary. e I-256. The LOOP event was the result of a personnel error that occurred when l the equipment operator allowed the mobile crane boom to come too close to the power i lines during a refueling outage. It did not result in any NRC ' enforcement actian. Tr.1635-36 (Giffin). j \\ 4 .. _.. -.,.,.,.,,. ~., _.... - -.....

.= i 9 i l - 103 - I-257. The cask rigging event involved a health physics activity and was not a' ] maintenance activity. Tr. 2249 (Miller). In. adjusting the lid on a shipping cask, personnel used two chainfalls to level the lid and put it down in place. The two ] chainfalls used were rated for 2000 pounds, but in this operation each actually carried approximately 2400 pounds. Tr. 1628-29 (Giffin). i I-258. As stated above, we do not adopt MFP's proposed findings regarding this _ j i subject matter. MFP Proposed Findings 602-32. MFP would have us find that the l I i incident regardmg the use of chainfalls was a maintenance problem and that "PG&E's i insistence that it is not maintenance-related is yet another indication of a programmatic i problem at DCNPP, which is poor coordination arxi recognition of shared responsibility between maintenance and other departments." MFP Proposed Finding 626. However, it was MFP that identified the chainfalls incident as a personnel problem. See: Supplement, October 26,-1992 at 13,15. We rejected the contention. We noted that I four instances uncovered by the applicant,' including the mobile cranes coming too close to the 500kV powerlines, and the three based on StaffIRs, including a reported wanha=s _ f 1 in control of lifting and rigging devices for heavy loads, had no apparent common focus. l 37 NRC at 22-23. Further, MFP's finding ignores the fact that the chainfalls incident i was a health physics matter, not a maintenance matter. Tr. 2249 (Miller). MFP cannot .I repackage its concern with a periennel matter as a maintenance concern. De NRC Staff. I and PG&E focused their testimony on the admitted issue, the efficacy of PG&E's-maintenance program at DCPP. MFP. fails to show how these unrelated incidents bear I f m -m- = e r c,. e r#, 1-+- .,r v.,- -,.r,r = ,,-we,

l c + 1 - 104 - l i i ) upon PG&E's implementation of its maintenance and surveillance program at DCPP or l i } that they constitute evidence of a breakdown of that program. i l KK. Main Feedwater (MFW) Pump Speed Probes j I-259. MFP offered five documents, MFP Exhibit 138, NCR DCl-92-EM-N010 ' i (July 29,1992); MFP Exhibit 139, NRC IR 92-05 (April 17,1992); MFP Exhibit 140, i r i NRC Management Meeting, Report 92-13 (April 16,1992); MFP Exhibit 140A, LER f I 1-92-002-00 (April 3,1992); and MFP Exhibit 142, NCR DC1-91-TI-N045 (June 10,' j 1991), that address a sequence of events related to the inverter for the power supply to-i i the Imvejoy speed probes associated with the main feedwater (MFW) pump. Tr.1650 (Giffin). I I-260. The MFW pump is a nonsafety-related piece of equipment. Tr.1650 j (Giffin). 'Ihe inverter at issue is a power supply to the speed probes for the MFW pump. l

14. at 1651. In an effort to make the pump more reliable, PG&E~ installed newly j

designed inverters in 1989. Id.; MFP Exhibit 138 at 2. -{ I-261. Subsequent to installing the inverters of new design, the inverters failed on 'j i several occasions. The PG&E NCR addressing this subject recounts the history of nine l inverter failures over an approximately two-year period between 1990 and 1992. MFP Exhibit 138 at 2-4. MFP argues that the inverter failure was a long standing problem and - j t that corrective actions were ineffective and untimely. MFP Proposed Findmgs 643'and. l 650. However, MFP offers no connection between these findings and PG&E's l maintenance and surveillance program. MFP Proposed Finding 643,650. The issue of-- i l. concern to PG&E and the NRC related to the timeliness of PG&E's design engineering j i i I I -i I

i - 105 - efforts, not to the maintenance program. Tr. 2246-47 (Miller). PG&E has addressed j that timeliness concern in conjunction with both this issue and the CFCU issue. SpeciSc i 1 actions are also described in the NCR. See MFP Exhibit 138 at 12 (item V.B.2) and 13 j (item V.C.). Each time the inverter failed, the' issue was investigated by PG&E's engineering organization. Exhibit 138 at 2-4. 1-262. MFP, taking a statement from MFP Exhibit 138 at 18 out of context, would have us find that financial considerations influenced PG&E's corrective action. MFP Proposed Finding 648. However, contrary to the MFP proposed finding implying - that PG&E did not make repairs because of money considerations, the inverters were replaced, repaired, and/or modified. Id. In addition, there was a tendency to try to make the new inverter design work rather than reassess the design entirely. Tr.1652 (Giffin); MFP Exhibit 140 at 3. PG&E ultimately initiated a design change to address the issue. MFP Exhibit 138 at 13; Tr.1659 (Giffin). MFP attempts to raise utility rate issues which are not within thejurisdiction of the Commission. I-263. Without record support MFP argues that this is related to PG&E's maintenance and surveillance program. MFP Proposed Finding 654. However, MFP ignores NRC staff testimony that the inverter failures were an engineering matter and not t a maintenance matter. Tr. 2246-2247 (Miller). The Staffinspector also testified that this inverter issue was one of several clustered at a "very low level" of significance. Tr. 2216 (Miller). Moreover, MFP ignores PG&E's testimony' which characterizes this sequence of events, related to an attempted improvement to a nonsafety-related L

1 1 l j - 106 - i component, as not reflecting directly on either maintenance or surveillance. Tr.1653 l i j (Giffin). t I i I-264. We cannot accept MFP pmposed findings on this subject matter, as the ~ . j underlying problem was related to PG&E's engineering organization and not within the scope of the admitted contention on the maintenance and surveillance program. l LL. Containment Ventilatian Ienistian (CVII Sirnalm I-265. MFP offered twelve documents, MFP Exhibit 144, LER l1-92-005-01 j (July 20,1992); MFP Exhibit 145, NCR DCl-92-TI-N020 (June 24,1992); MFP Exhibit 146, LER 1-91-013-00 (September 6,1991); MFP Exhibit 146A, NCR DCl-91-TI-N068 (October 3,1991); MFP Exhibit 147, LER 2-91-001-00 (August 13,1991); MFP Exhibit - 148, NCR DC2-91-TI-N062 (August 9,1991); MFP Exhibit 149, LER-1-91-00600 j (April 25,1991); MFP Exhibit 149A, NCR DCl-91-EM-N041 (April 25,1991); MFP l Exhibit 150, LER l-90-019-00 (January 28,1991); MFP Exhibit 150A, NCR DC1-90. WP-N093 (January 18,1991); MFP Exhibit 151, LER 2-90-004-00 (May 17,.1990); and MFP Exhibit 151A, NCR DC2-90-TI-N025 (October 11,1990), relating to inadvertent . 1 CVIs due to personnel errors. Tr. 1667-68. The documents are matched pairs, an NCR and an LER, each pair addressing an i olated CVI event, six in all.- I-266. MFP introduces its findings on this matter by stating that CVIs occur more frequently at DCPP than at any other US power reactor. MFP Proposed Finding 658. MFP leaves out the statement in MFP Exhibit 146A at 11,' that: "However, due to - ) differences in reporting and interpretation of an ESF, the number of other CVIs at other i plants may not directly correlate to the number of CVIs at DCCP." ' .,.,__.4., _,..i. f,__

I C O - 107 - I-267. A CVI involves closure of a containment isolation valve and is, from a safety pe spective, a benign event. Tr.1670-71 (Vosburg). As reflected in these exhibits, CVIs were historically reportable to the NRC. However, recent regulatory changes have been made to reduce reports of CVIs in view of the lack of significance. Id. at 1668-69. I-268. The radiation monitors are extremely sensitive to the power supplied to them. Most of the CVIs documented in these exhibits resulted from personnel working on electrical equipment and inadvertently causing an arc or a spark to be generated. Any electrical noise, such as that generated by an arc or spark, can cause the radiation monitors to go into the safety-related mode, causing a CVI. Tr. 1669-70 (Vosburg). Examples of the causes of the CVIs in these exhibits include: personnel working inside a panel bumped a terminal with a pair of pliers, causing an arc (MFP Exhibit 150A at 2); a technician allowed a test probe to slip during troubleshooting, causing a short between hot and neutral legs (MFP Exhibit 151A at 3); a loose connector resulted in generation of electrical noise (MFP Exhibit 145 at 4); an I&C technician incorrectly installed ajumper, causing a voltage transient (MFP Exhibit 146 at 4); and the failure of a motor associated with a radiation monitor created electrical noise, which caused a CVI (Tr.1670 (Vosburg). I-269. These six events are unrelated, other than that they all involve CVIs. Tr.1669 (Vosburg). PG&E devotes substantial attention, through training and work practices, to preventing errors during work on energized instrumentation that could lead to CVIs. Id. at 1672-73. Following CVIs, PG&E initiates an NCR to identify and

C C - 108 - implement actions to prevent recurrence, and trains maintenance personnel on lessons learned from' these events. Tr.1673-74 (Giffin). - PG&E is'also in the process of installing a new, digital radiation monitoring system, which is less sensitive and'will' result in a reduction in the number of these events.' Id. MFP would have us find that: PG&E's response to a known deficiency in its radiation monitoring system is untimely. MFP Proposed Finding 684. However, MFP does not offer a recorded basis for this finding. MFP ignores the fact that the sensitivity of the radiation monitors causing the CVIs has little safety significance, since the valves go to their safety-related position and the wear and tear on the valves is not significant. Id. at 1670-1671. MFP provides no - basis for its claim that PG&E's upgrade program for the radiation monitoring system is warranted as a matter of safety, in a more timely manner. Rawl on the above, the Licensing Board does not adopt MFPs findings on this matter. I-270. MFP would have us find that corrective action taken by PG&E for previous similar events was ineffective (MFP Proposed Finding 663), yet MFP does not address the testimony characterizing the incidents as unrelated. Tr. at 1669 (Vosburg). MFP would also have us find that the mere occurrence of a CVI is enough to warrant stronger protective measures (MFP Proposed Finding 674), yet MFP ignores the fact that these incidents cause the equipment to go into its safety-related ' mode, they do not introduce transients and that wear and tear is negligible as the valves are designed to actuate hundreds of times. Id. at 1669-72 (Vosburg). I-271. The Board finds, in this context, that the CVIs identified by MFP represent six isolated cases, that they do not reflect a programmatic' weakness or a lack of-4 [ i

o - 109 - 1 communication and coordination, and that PG&E takes reasonable measures to minimize personnel errors that can result in inadvertent CVIs. For these reasons, the Board does l not adopt MFPs findings concerning this matter. 1 MM. Reactor Trio on Steam Generator Iow I2 vel - I-272. MFP offered two documents, MFP Exhibit 155, LER 1-91-002-01 (May 17,1991) and MFP Exhibit 156, NCR DC1-91-WP-N012 (May 13,1991), addressing a single reactor trip in February 1991. The cause of the trip was a low steam i generator level in two steam generators. The low steam generator level occurred due to { a personnel error by a carpenter while erecting a scaffold. Tr.1692 (Giffin). i I-273. The carpenter was carrying six foot planks for the scaffold. He inadvertently hit a valve with a plank, closing the valve and shutting off air supply to the main feedwater regulating and bypass valves, which in turn isolated feedwater flow to two steam generators causing the reactor trip. Id. at 1692-93; MFP Exhibit 155 at 1. I-274. MFP proposes that we find that this incident involving an individual component presents some concern regarding the adequacy of maintenance of PG&E's components. MFP Proposed Findings 690-704. MFP goes on to infer that a much more significant safety concern arises due to multiple failures and deficiencies associated with the incident. MFP Proposed Findings 691-92. However, MFP chooses to ignore PG&E's testimony characterizing the maintenance program as a living program that continually improves when operating experience such as this is incorporated as warranted. Tr.1700 (Vosburg). PG&E's testimony explains that the inadvertent closure of the air valve that caused the reactor trip did not in itself cause the other non-safety related ...v.- v

• ' ~

4 - 110 - components to fail. Tr. at 1697 (Giffin). MFP also ignores the fact that the four unrelated failures it lists in its proposed findings (MFP Proposed Finding 692), involved non-safety related equipment, and that the separate causes were investigated and-corrective actions taken nevertheless.14. at 1694-1696; MFP Exhibit 155 at 7-8. I-275. PG&E has addressed this event and identified actions to minimize future similar occurrences. MFP Exhibit 156 at 11-12; Tr.1693 (Giffin); Tr. 1701-02 (Vosburg, Giffin). As a result of this incident,'PG&E made enhancements to its maintenance program. Id. The Licensing Board does not adopt MFP's proposed findings regarding this matter, since this is an isolated personnel error that has been addressed and resolved. NN. Auxiliary Saltwater (ASW) Pump Crmetie Valve I-276. MFPExhibit168,NCRDCO-91-EM-N009(November 22,1991), concerns corrosion on a manual hand wheel for ASW pump crosstie valve SW-1-FCV-4%. Tr.1714-15 (Ortore). The ASW crosstie valve is located in the intake structure and subject to the salt environment. Tr.1732 (Vosburg). The corrosion on the hand wheel affected the manual operation of the valve. 'Ihe valve was capable of performing its function by remote operation from the control room. Tr.1714 (Ortore). This aspect of the valve's operability was included at the time in the surveillance program. Tr.1714 (Ortore). I-277. The corrosion on the hand wheel did not have any safety significance because another valve in series with this valve was capable of being closed, both

o o - 111 - manually and by remote operation. The valve operator that operates the valve is nonsafety-related. Tr. 1715-18,1725 (Ortore, Vosburg). I-278. Subsequent to the event reported in the NCR, PG&E enhanced its surveillance and preventive maintename programs to require manual testing of similar valves in the ASW system and to increase the frequency of preventive maintenance on the ASW crosstie valves. MFP Exhibit 168 at 9; Tr.1724 (Ortore); Tr. 1727, 1743 (Giffin). I-279. MFP would have us adopt findings iramporeting its reading of MFP Exhibit 168. MFP Proposed Findings 705-722. However, MFP's characterization of the - event recorded in MFP Exhibit 168 does not properly reflect either the testimony or the document itself. See MFP Proposed Findings 715-21. I-280. The testimony of the PG&E witnesses cited above was consistent with the - language of the safety analysis in the NCR, which in relevant part states: The function of the valves is to provide train separation to protect against the possibility of a passive failure, and the valves are not required to function to prevent or mitigate an accident. These valves are repositioned when the unit enters hot leg recirculation at 13 hours after the accident. The FSAR does not postulate a passive failure until 24 hours after the LOCA. This allows a maximum time window for re-aligning the valve of - 11 hours. Based on this time frame, and the fact that maintenance crews required approximately 15 seconds with a two man crew to remove paint from SW-1-FCV-495 to return the valve to manual operability, it was-determined that the trains of the ASW system could be separated within. the required time window, and the ASW system was determined operable ' with the valves in the as found condition. In addition, NECS stated that if a passive failure, defined not to exceed 50 gpm flow, were to occur with the crosstie valves open, the amount of flow lost with ASW pumps capable

'] i e o- - 112 - ) of discharging approximately 11,000 gpm would have no significant effect on the ability of the ASW system to cool the unit. MFP Exhibit 168 at 15-16. I-281. Thus, MFP's question cannot be. answered with a short answer. However, the operators were never inoperable; moreover, even had they been inoperable, the safety significance would have been minimal as they are not required to prevent or mitigate an l accident i I-282. MFP urges us to adopt a finding that PG&E was in error in reasoning that the "public health and safety was not affected by the manual inoperability of the valve i operator." MFP Proposed Finding 709. MFP provided no technical support for this i conclusion. Moreover, MFP's citation is to a passage that states that " maintenance 1 personnel demonstrated that the valves could be manually operated within the time { windows allowed for closure of the valves." MFP 168 at 8. I-283. With regard to FCV-495, MFP woule aim have us to adopt a finding that f PG&E is not specific about how long it took to close that valve. MFP Proposed Finding l 710. As noted above, the time to close the valve is stated in the NCR to be 15 seconds. Thus, we do not adopt the finding. I-284. We decline to adopt MFP's conclusion (see MFP's Proposed Finding 722) i because our finding is to the contrary: that as a result of the events documented in the i NCR PG&E did adapt its program to include appropriate surveillance and maintenance for the equipment addressed here. t i h l i

' j o - 113 - 1 OO. Testcock Valve on Diesel Generator 'I I-285. MFP Exhibit 172, NCR DCO-91-MM-N049 (October 2,1991), documents - l an isolated personnel error / equipment component problem identified by PG&E in 1991. l .I The NCR includes the specific corrective actions identified by PG&E to resolve this i matter. MFP Exhibit 172, at 7-13. The document does not suggest s' programmatic i i maintenance problem. The incident involves a maintenance worker tightening a tem =k : l l on a diesel generator cylinder head. The teewk is an instrumentation fitting for l compression readings. While performing 'a post-maintenance test,[the workx l 1 inadvertently went to the wrong testcock. When he tightened the teewk, it broke. l PG&E determined that the testcock broke due to fatigue. In consultation with the vendor, f i l PG&E determined that copper washers were needed to resolve the issue. Tr. 1746-47 i (Giffin). PG&E inspected all other diesel generators and found no similar fatigue crach. Id. at 1749. PG&E also revised a surveillance procedure to verify that testcocks are j i tight. Id. at 1751-52. I-286. It is apparent from MFP's Proposed Findings 723-27. that the issue it-l wishes to pursue is documented in repert-QE Q0008915 concerning the issue' of inadvertently working Testcock 2R instead of 8R rather than in MFP Exhibit 172, which i does not support the findings MFP urges us to adopt. We do not adopt MFP's findings = in this matter, which are based on conjecture rather than record evidence.. In any event,. .i no safety concern is raised here and the incident is entitled to no weight, as it does not - speak to the allegation of a programmatic deficiency in PG&E's surveillance and maintenance program.

1 o-o j - 114 - l i PP. Main Feedwater Check Valve l 1 1-287. MFP introduced four documents wihdig a leaking check valve: MFP Exhibit 190, NCR DCl-91-TN-N002 (February 18, 1991); MFP Exhibit 191, NCR i DCl-90-OP-N083 (February 8,1991): MFP Exhibit 192, LER 1-90-015-01 (January '25, ij 1991); and MFP Exhibit 193, NRC Review of LER 1-90-015-00 (January 18, 1991). 1 I-288. MFP Exhibit 191 is an NCR written to address an ESF actuation that ] I mund in December 1990. The particular ESF that actuated in this case'wasithe feedwater isolation system. Tr.1772 (Vosburg).- The root cause of this event was' l determined to be leakage through a feedwater regulating valve (FW-1-FCV-1430). A contributing cause was init ally identified as backleakage through a steam generator inlet ( feedwater check valve (FW-1-531) due to a slight misalignment of the check valve disc. i i MFP Exhibit 191, at 5. Corrective maintenance _ was performed on all three valves l following the event. lId. at 11. I-289. MFP Exhibit 190 concerns backleakage through the check valve FW-1-531. 1 Prior to the ESF actuation addressed in MFP Exhibit 191, the check valve leakage 'had l 4 been identified and evaluated by Operations in February 1990. Operations Ludned -{ t at that time that the condition would not cause any problems in plant operations. { l Maintenance on the valve was planned for the next refueling outage. Tr. 1773-74, 79 j l (Vosburg). The backleakage through check valve 1-FW-531 did not directly contribute i to the December 1992 ESF actuation. MFP Exhibit 190 at'3; Tr.1781-82 (Vosburg). I I-290. MFP Exhibit 193 is an NRC review of LER 1-90-015, Rev. O, which was i i introduced into evidence. In MFP Exhibit 193, the NRC questioned the root cause j 1

i- }- t o i f - - 115 - analysis in the original LER, citing the lack of operator knowledge of the leaking valves - l as a root cause of the ESF. The NRC also raised a question regarding' the safety analysis L l in the LER. ' Tr. 1776-77 (Vosburg). 'Ihe NRC's concerns were not related to i maintenance. Tr. 2256-58 (Narbut). PG&E subsequently revised the LER addressing this event. The revised LER is MFP Exhibit 192. The valve lealmge was analyzed as i being a condition that could be addressed by Operations'pending scheduled maintenance. The check valve leakage was corrected. - Tr. 1788-89 (Giffin). I-291. With respect to the leaking bypass and regulating valves implicated in the December 1990 ESF, these nonconforming conditions cannot be datacted with the unit on line. 'Ihey can be identified only in starting up the plant. Prior to this event, PG&E verified that the positioner was wustly set up following every outage. 'Ihe positioners are now checked prior to return to service 'any time the' unit has been shut down. Tr.1783-84 (Giffin); Tr.1786 (Vosburg). I-292. The central issue of this event was that Operations management should have told operators about the leaking valves and that their startup would be more difficult. Tr. 2259 (Narbut). There was no Notice of Violatior. Issued in this matter, which was not related to the maintensme of these valves but to an operations issue. Tr. 2259 (Miller); Tr. 2256-57 (Narbut). I-293. MFP would have us adopt findings that the lack of operator knowledge that was of concern to the NRC is part of an overall communication problem at'DCPP. MFP Proposed Findings 745-750. We do not adopt MFP's Proposed Findings on this matter because both MFP's exhibits and the testimony of record establish that the matter relates ?

- 116 - to operations and not to maintenance. We received these documents into evidence for the reasons given by Judge Kline at Tr. 696-98. However, having examined the documents, we cannot accept MFP's view that they relate to a maintenance problem. Further, they do not show any breakdown in the implementation of PG&E's surveillance and maintenance program. QQ. Auxiliary Saltwater (ASW) Pumo Vault Drain Check Valves I-294. MFP Exhibit 196, NCR DCO-91-MM-N067 D6 (January 15,1991), reports a minor issue identified by PG&E during a routine periodic inspection and refurbishment. MFP Exhibit 196 at 1. I-295. In performing maintenance on an ASW floor drain, PG&E found the check valves in the drain line to be partially stuck open due to debris. The check valve is designed to prevent backflow into the pump vault. However, if water were to reach the pump vault, there are level indicators that would alert the operators. Tr. 1798-99 (Giffin). PG&E's engineering evaluation determined that the check valve is unnecessary. Id. at 1800-01. I-296. The NCR was initiated to determine whether there was an operability issue l raised by Mechanical Maintenance's working on two ASW drains at the same time. l Engineering and Operations concurred that the ASW pumps were not inoperable and that this was not a nonconformance. MFP Exhibit 196 at 1; Tr.1795-97 (Giffm). I-297. We do not adopt MFP's proposed findings concerning this matter, MFP Proposed Findings 751-57, as they ignore the testimony of the PG&E witness who was

- 117 - responsible for the writing of the NCR. Mr. Giffin characterized the incident as a "nonproblem" that was not related to preventive maintenance. Tr. '1801 (Giffin). RR. SI-1-8805A Failed to Cvele on Ar*mdan Sirnal I-298. MFP Exhibit 210, NCR DCl-90-EM-N042 (June 27,1990), concerns a valve that failed to cycle during a test because a declutch fork in the Limitorque operator was installed upside down. The inconect installation caused stresses on some of the internal components leading, eventually, to'the valve's failure to cycle. Tr. at.1'809 (Ortore). I-299. The incorrect installation of the declutch fork was determined to have occurred in Damber 1982 during a maintenance overhaul. Between 1982 'and 1990, the valve had been tested and had performed as it should. Tr. 1809-10 (Ortore). I-300. At the time the incorrect declutch fork installation was made in 1982, maintenance instructions and training were not equivalent to the present standards. PG&E testified that it now has extensive procedures and training on the maintenance'and overhaul of Limitorque operators. Tr.1810 (Ortore). 1-301. MFP would have us repudiate PG&E's analysis of the event both in'the NCR and on the record of the 3nding and adopt MFP's unsupported interpretation (MFP Proposed Findings 758-65), which includes MFP's. view of the Commission's regulations regarding the single failure criterion. We need not entertain this extra-record evidence, however, as the issue can be resolved without addressing MFP's view' of Appendix A of 10 C.F.R. Part 50. The NCR concerns, as MFP conwily notes,

4 o s. 4 i i l - 118 - 1 1 l i 1 SI-1-8805A's failure to cyc!c, ot its failure to open.. The safety analysis offered in the l l NCR is as follows: i SI-1-8805A is'one of two parallel valves which open on a SI signal. I 4 'Ihese valves align the Refueling Water Storage Tank (RWST) with the i suction header for the centrifugal charging pumps to' allow injection of j borated water into the RCS. The operability of either 8805A or 8805B j will allow for sufficient flow from the RWST to the RCS. Additionally,- valve 8805A fully opened on an actuation signal, as required. In the event of an SI signal, SI-1-8805A would have performed its required safety ) function. Therefore, no adverse consequences resulted fmm this problem,. { and the health and safety of the public were not affected by this problem. 1 MFP Exhibit 210 at 5. I-302. There being no evidence to the contrary, we must adopt this view of the l safety sigrJficance of SI-I-8805A's failure to cycle. We cannot base a finding on MFP's - i i technical opinions and conclusions that first appear in findings, which no one had an opportunity to test at hearing. ' Moreover, the situation that gives rise to MFP's analysis. did not occur. Thus, we need not consider MFP's analysis. 1 SS. Fire in Flactriemi Panel-I-303. MFP Exhibit 216, NCR DCD-90-SE-N080 (January 28,1992), discusses a fire in an electrical panel in the security inverter room in November 1990. The most likely root cause of this incident was determined to be a faulty compression termination l during installation of ajumper in 1990. MFP Exhibit 216 at 1. I-304. PG&E testified that similar compression connections have been used in the plant for ten years, without any other. problems. Tr.1819-20 (Crockett). The-uncontroverted testimony is that the compression technique is adequate, even though in

1 + - 119 - one instance the application may have been deficient. M.'at 1820.1 Moreover, making -4 these connections is a task well within the skill of a journeyman electrician. " M. at 1821. I-305. As corrective actions, PG&E took several steps, including: -(1) scanning - ) similar panels with an infrared detector to identify any hot spots; (2) using lug crimp connectors to rebuild this one particular panel; and (3) issuing a bulletin to Electrical Maintenance on proper practices in making these connections. MFP Exhi2 216 at.7-8.- _l This adequately resolved the issue. I-306. There is no evidena to support MFP's views rather than those of the : i expert witnesses and the analyses in MFP Exhibit 216. See MFP Proposed Findings j 766-68. There was no evidence that would lead to the' conclusion that MFP urges, that the fire in the electrical panel would probably'have been avoided had PG&E upgraded -l the connectors to crimped-lug style terminations. MFP Proposed Finding 768. l i Mr. Crockett, who was Chairman of the Technical Review Group (TRG) that reviewed .t the incident, testified that no other compression connection in the plant had ever had problems, that the one at issue was the only one in the plant's history to have presented a problem. Tr. 1818-19 (Crockett). ) I-307. Although after the incident,' PG&E established a procedure for electrical - 4 connections, there was no real necessity for establishing such a procedure, asjourneymen. electricians know how to make connections from cables to terminals. M. ' No failure in PG&E's surveillance and maintenance program' was shown by the incident. 1

o 3 - 120 - 'IT. Chemical Volume Control System (CVCS) 12akage i I-308. LER 1-92-009, Rev.1 (January 11,1993) (PG&E Exhibit 28), reports an event that occurred in June 1992. During a routine radiation survey, PG&E noticed leakage from a CVCS valve bonnet in a heat-traced portion of the system. During routine plant operation, that system functions to maintain reactor coolant system inventory. Parts of the system are heat-traced to keep boric acid in solution. During a postulated accident, parts of the system are used to recirculate and supply water to mitigate the accident. PG&E Direct Testimony at 100-01 (Giffim). The leakage detected in June 1992 was determined to be outside the design basis for such leakage. The leak was stopped by tightening the valve body-to-bonnet nuts. Id. at 101. PG&E's investigation identified the cause of the leakage as thermal degradation of the valve diaphragm due to a malfunctioning heat trace thermostat that caused leakage through the body-to-bonnetjoint. Id. Corrective actions included lowering the heat trace temperature for the valve, reviewing every other valve in similar service to verify that there were no other potential problems, reviewing the entire heat trace system for proper setpoints and installation and improving the surveillance monitoring program. Id. 1-309. LER 2-91-009-01 (April 4,1992), PG&E Exhibit 29, reports a leak attributed to a different root cause. This leak was attributed to a failure of the preventive maintenance program to include the valve vendor's recommendation concerning bonnet nut torque and diaphragm replacement. The corrective action was to include this recommendation in the program. Id. at 101-02.

r + - 121 - I ~1 I-310. MFP urges the Licensing Board to adopt findings based on its reading of. these LERs. MFP Proposed Findings 776-84. However, the' proposed findings mischaracterize the testimony and present technical opinions and conclusions which are not in the record. We do not adopt them. I-311. MFP argues that these incidents had safety significance. MFP Fwyv4 Finding 781. MFP's citations to Mr. Gih's testimony do not support this statement. ] ) Mr. Giffin testified that leakage from an CVCS valve bonnet could potentially be safety j significant, not that the situation described in the LER had potential; for safety _ '2 significance. Tr.1829. Mr. Giffin's answer to the question was that if the leakage was ) large enough and the right circumstances existed,- it'could have the potential to be significant in that the dose in the control room would be larger than what it is required by the regulations to be.14. Mr. Giffin's answer was to counsel's hypothetical question. Mr. Giffin reiterated this response in answer to a question on redirect. Tr.1830. Thus, MFP's Proposed Finding is not permissible and there is nothing in the record to support MFP's theory that this event had any safety significance. Further, these events show how - PG&E's surveillance and maintenance program works in that PG&E reeagai=i and took action to stop the CVCS leakage; they do not show that the program is in any way deficient.

i - 122 - j i Contention V: Thermo ima Comnensatory Measures V-1. This contention was admitted pursuant to the Licensing Board's Prehearing Conference Order and alleges that PG&E has failed to implement and abide by the Commission's interim compensatory measures required for the use of Thermo-Lag fire barriers. Pactfe Gas & Electric Co. (Diablo Canyon Nuclear Power Plant, Units 1 and 2) LBP-93-1, 37 NRC 5, 27-28 (1993), as clanfed by " Memorandum and Order (Discovery and Hearing Schedules)," dated February 9,1993, at 2. V-2. The scope of this contention does not include whether fire watches, as a compensatory measure, are an adequate substitute for Thermo-Lag fire barriers declared inoperable. The issue is whether PG&E has implemented and will continue to implement the approved compensatory measures at DCPP. Tr. 1297,1299,1430 (Bechhoefer); Pacife Gas & Electric Co. (Diablo Canyon Nuclear Power Plant, Units 1 and 2), LBP-93-1, 37 NRC 5, 27-28 (1993), as clanfed by " Memorandum and Order (Discovery - and Hearing Schedules)," dated February 9,1993, at 2. V-3. Contention V does not concern inside containment applications (radiant energy shields) of Thermo-Lag fire barrier material. In rejecting a late-filed contention in this proceeding, we specifically ruled that the issue of radiant energy shields and the allegation that the Thermo-I2g material is itself a fire hazard in these applications was not admissible. Prehearing Conference Order, LBP-93-9,37 NRC at 444-45. V-4 MFP did not offer testimony to support Contention V. The NRC Staff ) presented testimony from Patrick M. Madden, Senior Fire Protection Engineer, Office of Nuclear Reactor Regulation, and Mary H. Miller, Senior Resident Inspector at DCPP.

q + - 123 - PG&E offered testimony from David K. Cosgrove, Supervisor of the Safety and Fire = -] i Protection Group at DCPP, and Robert P. Powers, Manager of the Nuclear Quality ] Services Department of PG&E's Nuclear Power Generation Business Unit. l V-5. The purpose of the NRC Staff's direct testimony, filed on July 30,1993,' was to address the adequacy of PG&E's adherence to interim fire protection measures to, l l' compensate for Thermo-Lag fire protection features declared ' operable.at-DCPP, m pending resolution of the generic 'Ihermo-Lag issue by the NRC Staff.~ NRC Staff Direct Testimony at 2 (Madden). l V-6. The problem with Thermo-lag fire barriers is generic in the nuclear power l plant industry. NRC Staff Direct Testimony at 2 (Madden).~ A majority of nuclear 1 1 power plants use Thermo lag to' satisfy NRC fire protection' requirements! ' Id. The: ]l potential problems of Thermo-Lag are that 1) Thermo-Lag fire barriers may not provide. the fire resistance necessary to satisfy NRC fire protection requirements; 2) 'Ihermo-Iag t may burn more readily than originally believed; and 3) the ampacity derating' factors used by licensees to derate power cables may not be great enough to account for the irrulating effects of the Thermo-lag material. Id. V-7. Only a moderate amount of Thermo-Iag is installed at DCPP. NRC Staff Direct Testimony at 2 (Madden). " Moderate" describes an installation inwiperding between 100 to 1,000 square feet or 100 to 1,000 linear feet of fire barrier material. Id. PG&E replaced the Thermo-Iag in the Unit 2 containment and will replace all the Thermo-Iag in the Unit I containment during the refueling-outage. scheduled for February 1994. NRC Staff Direct Testimony at 2-3 (Madden). i .l .._-..-_-,...__.J

J 4 3 - 124 - V-8. The NRC Staff has previously concluded that, on a generic basis, interim 4 compensatory measures provide the protection required by the NRC fire protection program pending resolution of the Thermo-Lag issue, due to the fact that nuclear power plant fire protection programs are based on the Defense-in-Depth philosophy. NRC Staff Direct Testimony at 3 (Madden). This approach contains three principles: 1) preventing fires; 2) quick detection and suppmssion of fires that occur despite the fire prevention program; and 3) physical separation requirements for safe shutdown functions to assure the ability to achieve and maintain safe shutdown conditions.- Id. The objective is to create a balance between these three factors, so that if a we*nass in the program occurs that affects one of these principles, the vulnerability can be compensated for by strengthening fire protection activities covered by one or both of the other principles. Id. V-9. NRC Bulletin 92-01, " Failure of Thermo-Lag 330-1 Barrier. System to Maintain Cabling and Wide Cable Trays in Small Conduits Free from Fire Damage," was issued on June 24,1992. It was issued subsequent to several NRC Information Notices that documented deficiencies in Thermo-Lag performance and fire endurance in tests performed by several utilities. V-10. NRC Bulletin 92-01 requested all nuclear plants utilizing Thermo-Lag as a fire barrier for conduits and cable trays to implement interim compensatory measures, 3 similar to plant Technical Specifications for impaired barriers associated with safe shutdown equipment or circuits, or for other inoperable fire protection systems. PG&E Direct Testimony at 4-5,13 (Cosgrove, Powers). Bulletin 92-01 relied on a determination that licensees can institute compensatory measures for any deficiency in the

.. ~ r - 125 - 1 fire puis; tion _due to Thermo-Lag, as with any impaired fire barrier, and be in) l t compliance with the NRC's AWit R regulations. Tr.1420-25 (Madden). l V-11. On June 29,1992, PG&E verified that interim compensatory measures were operating in DCPP fire areas where Hermo-Lag was present, in accordance with Bulletin 92-01. PG&E Direct Testimony at 13 (Cosgrove, Powers). 1 V-12. Supplement I to NRC Bulletin 92-01, dated August 28,1992, expanded Bulletin 92-01 to include further applications of the Thermo-Lag material. His included - 1 l-all sizes of conduits and cable trays, and all wall, ceiling, and equipment enclosures using i one-hour or three-hour pre-formed Hermo-Lag panels and conduit shapes. Licensees - were to identify these Thermo-Lag locations and institute interim compensatory measures. [ V-13. PG&E's response to NRC Bulletin 92-01, Supplement ILis-dated l 1 September 28,1992, and was admitted-as PG&E: Exhibit 3 with'.PG&E's Direct f Testimony on this contention. Tr.1277. Table 1 to that document identifies'the 11: ) specific Thermo-Lag fire areas at DCPP subject at that time to the interim compensatory measures. PG&E Direct Testimony at 13 (Cosgrove, Powers). De _ interim l compensatory measures adopted for these 11 areas were mplej by the NRC Staff. -Id. l The Staff's letter was dated October 27,1992, and was introduced as PG&E Exhibit F-1. l l Tr.1464. i V-14. PG&E's interim compensatory measures include fire watches like those in DCPP's Technical Specifications and the DCPP Fire Protection Program for instance: l i where fire barriers and other fire pivisction systems are inoperable. PG&E Direct il Testimony at 4-5 (Cosgrove, Powers). Dese compensatory measures involve either a i -l 1 __.......,;w...,,m,..,

4 o - 126 - continuous fire watch stationed at specific locations or a roving hourly fire watch. PG&E Direct Testimony at 5 (Cosgrove, Powers). A roving hourly fire watch is usually employed when the Thermo-Lag fire areas that must be covered have available fire detection devices. Id. Where fire detection devices are not available, a continuous rather than hourly fire watch is usually stationed at that Thermo-Lag location. Id. V-15. Implementing the interim compensatory fire watches at DCPP for the 11 Thermo-Lag fire areas involved adding to existing fire watch routes. PG&E Direct Testimony at 15 (Cosgrove, Powers). The routes were initially established when commercial operation at DCPP commenced. PG&E Direct Testimony at 6-7 (Cosgrove, Powers). V-16. Licensees use roving fire watch patrols to provide reasonable assurance that fire hazards in the area of the degraded fire barrier are minimized. NRC Staff Direct Testimony at 4 (Madden). Such alternative actions are utilized in programs associated with the operation of nuclear power plants to compensate for an inoperable condition. Id. These alternative actions have always been an integral part of NRC regulatory requirements. Id. V-17. The NRC previously evaluated the use of fire watches to compensate for degradation in the effectiveness of required fire barriers, and concluded that fire watches continue to assure adequate protection of the public health and safety. NRC Staff Direct Testimony at 4 (Madden). In addition to the Defense-in-Depth approach, NRC resident inspectors tour accessible areas of the plant every week to observe possible fire hazards and fire fighting equipment to verify operational safety. Id.

e j .'-l 127 - i .1 V-18. Fire watches are jwwce trained to inspect for the control of igrdtion .j 1 sources and combustible material, to look for conditions that may indicate an incipient.. fire, provide prompt notifications of fire hazards and' fires, and take actions when f -l appropriate to begin incipient fire suppression activities. -NRC Staff Direct Testimony -l at 4-5 (Madden). -l 4 l V-19. The two types of fire watches are the roving fire watch and the ' continuous fire watch. NRC Staff Direct Testimony at 5 (Madden). 'Ihe roving fire watch patrols ] i the area of the inoperable fire protection feature each hour. Id. The fire watch enhances j t plant fire prevention by identifying fire hazards in the. area of'the inoperable' fire f instion feature. Id.' A continuous fire watch is constantly posted in the area of the'- 1 impairment. Id. This fire watch enham fire prevention and the ability to' detect a fire there if one occurs. Id. Continuous fire watches are generally used to compensate for j fire protection features in plant areas where either automatic fire detection capability is i inoperable or not provided. Id. V-20. At DCPP, fire watch personnel receive over 40 hours of job training. i PG&E Direct Testimony at 8 (Cosgrove, Powers). This training involves general employee fire protection training, specific fire watch training including identification of f fire protection deficiencies like combustibles and impaired fire barriers, and training on i I using portable fire extinguishers Id. Fire watch wws4 are taught to identify and i i i control transient combustibles that may be present in the plant. Id. Before performing i i a fire watch, each new watch-stander accompanies the fire watch foremen on shifts for. "on the job training." Id. i l l l ~_...____,-,__.,_.,-,__.__._._,m_

r n 5 l - 128 - l

i V-21. Fire watch foremen receive additional training on fire fighting equipment a

and techniques, and complete a class involving training in the DCPP FSAR and all fire l protection systems. PG&E Direct Testimony at 8-9 (Cosgrove, Powers). 'Ihey are also - I i trained in fire barrier inspections and perform monthly; fire barrier ' inspection 'surveillances when they are not conducting actual mving and continuous fire watches. Id. f i .V-22. Responsibilities of fire watches include notifying the shift foreman of fires 1 i and sounding an alarm if==ary; extinguishing fires; conducting continuous and hourly roving fire watches when fire systems or barriers are impaired; performing compensatory ; requirements when combustible loading-exceeds commitment levels; and - having-f knowledge of " ignition source areas." PG&E Direct Testimony at 9-10 (Cosgrove,; j i Powers). V-23. Fire watch personnel also perform several collateral duties to document minor discrepant conditions like burned out light bulbs and ' door closure discrepancies. l PG&E Direct Testimony at 12 (Cosgrove, Powers). Discrepant conditions are noted on the logs. Tr.1399 (Powers); Tr. 1402-1404 (Cosgrove). These items are evaluated by ] the fire watch foreman and the plant fire protection specialist to ensure that adequate ~ corrective actions are implemented. PG&E Direct Testimony at 12 (Cosgrove, Powers). V-24. The NRC Staff concluded that the DCPP fire watch program-has been ) adequately implemented, because the evidence regarding the duties, training and responsibilities of fire watch personnel comports with NRC requirements. NRC Staff - Testimony at 10 (Miller). This testimony is uncontroverted and unimpeached.

o - 129 - V-25. The roving fire watch commences a fire watch tour at the top of the hour 3 i and walks an established route through each fire area within Units 1 and 2,-in addition to all common fire areas. PG&E Direct Testimony at 10 (Cosgrove, Powers). This tour covers fire areas where interim compensatory measures'for Thermo-Lag' have been implemented. Id. At the end of each tour (45 minutes) the watch-stander usually- !) exchanges with another watch-stander and the next round begins on the hour.

• Id. The tours are modified to include other areas when a degraded or impaired fire protection

) system or barrier has been identified.14. J V-26. Roving fire watch activities, exceptl for the ' intake stmeture,is,,, documented when the roving watch scans a bar code installed in locations along the tour - route with a portable electronic "bar code" reader wand. PG&E Direct Testimony at 11 (Cosgrove, Powers); Tr. 1282-85 (Cosgrove, Powers). Information from the hand held wand is downloaded to a personal computer at' the shift's end, and a pnntout of the' watch l activities, which includes the watch-stander's name, badge number, location of each area, and the time the watch was in the area, is made. Id. Any special conditions identified. l during the hourly watch inspection are also documented. Id. These watch activities are maintained on a computer disk and in a hard copy file for one year by the fire protection l specialist. Id. l is The intake structure fire location is not included in the roving fire watch because of its remote location. Thus, a dedicated fire watch is assigned there. PG&E Direct Testimony, at 10 (Cosgrove, Powers); Tr.1279 (Cosgrove).

+ - 2 I q - 130 - V-27. The fire watch contacts the fire watch foreman, the Industrial Fire Officer. or, ultimately, the Operation Shift Supervisor if a problem occurs during an hourly fire-f watch that causes a delay. PG&E Direct Testimony at 11-12 (Cosgrove, Powers); j i Tr.1305 (Cosgrove, Powers). Assistance can then be provided so the hourly fire patrol ( l can be completed. Id. j V-28. When exclusion areas are established for radiographic operations that~ f preclude the hourly fire watch from inspecting an area, the radiography examiners in the - area perform the fire watch activities there. PG&E Direct Testimony at 12 (Cosgrove, i { Powers); Tr.1321 (Powers). V-29. As part ofits interim compensatory measures, PG&E installed a portable. t fire detection system in conjunction with the hourly roving fire watch patrol in lieu of. posting a continuous fire watch at certain Thermo-lag' locations. PG&E Direct i l Testimony at 14 (Cosgrove, Powers); Tr. 1288-90 (Cosgrove, Powers); Tr.-1464 (Madden); PG&E Exhibit 3, Table 1. It utilizes smoke detectors and a dedicated plant 7 i phone system to immediately alert plant personnel in the event of a fire in the area. The ' NRC approved this system. Id. V-30. Since initiation of the interim compensatory measures for'Ihermo-Lag fire areas, the completion rate of hourly fire watches has been 100 percent. PG&E Direct j Testimony at 14-15 (Cosgrove, Powers); Tr.1320 (Powers). A successful tour of the - Thermo-Lag areas occurs when the fire watch enters the defined fire' area within the appointed hour. Tr.1307-08 (Powers). ' This entry, then,' demonstrates succe==ful f 'i implementation of the interim compensatory measures. Id. i l i

1 o ./- l - 131 - 1 I j V-31. On several occasions where a portion of the tour route was not accessible-' j (when radiography was being psfurusi), the fire watches did not log into every card 4 reader within a defined Thermo-lag fire area. Tr._ 1309. (Pceae). However, the )i Thermo-Lag fire area was actually entered and observed within the one hour period by j the fire watch at a separate location on the fire watch route, which'. satisfied the l compensatory requirement mandated by NRC Bulletin 92 01. Tr.1312-13 (Powers).- i j V-32. Whether all card readers were reached on a particular fire watch on these i j occasions is not significant. Tr. 1309-13 (Powers). Isolated instances of failure to reach ] 1 { every card reader on a fire watch route do not impact the adequacy of PG&E's: ' j 4 implementation of the fire watches because, first, the defined fire area was entered inside. l j the hour at a separate location and second, when this occurred (due to radiography in the L ] i j area), the radiographers observed the location. Tr.' 1321 (Powers). When this occurred i i ) j due to high radiation, radiation technicians were present. Id. heir presence satisfied a j the requirement for a human element in the area. Tr.1458 (Madden). Third, this issue ) only relates to two of the eleven Thermo-Lag areas-Units.1 and 2 GE-GW areas. 1 i j Tr.1312-13 (Powers). These areas have fire detection capability. Tr. at 1320 (Powers). j The fire watch simply augments the detection capability. Tr.1314 (Cosgrove).. Finally, ~ 1 si.ch failures are few in number. Tr.1315 (Cosgrove). Statistically, it does not : 4 i appreciably impact the success rate of the compensatory fire watches. Id. The success i rate in completing the entire route in Thermo-Lag areas was greater than 99 percent. Tr.1317 (Powers). l I j l i "y +vv e .t y ~g y -a?. + *wm-? --yr*- h a4ageti Mm

i - 4 i 132 -

i i

V-33. Since 1985, the roving fire watches at DCPP have successfidly completed ' l 1 i over 99.99 percent of the scheduled rounds. This evidence is uncontrove.ted and demonstrates that the interim compensatory measures have ' been.and are being [ t l implemented. V-34. The recording system for documenting fire weiss at DCPP provides i i reasonable assurance that the evidence of wmdul implementation of the watches'is l i a reliable. Tr.1392 (Cosgrove). The fire watch logs, except those for the intakc ' i ~ structure, are generated by computer. 14. The computer data is maintained on a . i computer disk that documents the fire watch's passage through and check at each of the - bar code reader locations. Tr.1283-86 (Cosgrove, Powers). - A harti copy of the computer operated fire watch information is downloaded from the computer at the end of each shift. Id. A diskette and a hard copy of the logs are stored in a locked and' l fireproof file cabinet to which access is limited. Tr.1286-87; 1414 (Cosgrove). f i V-35. Any alteration of. the fire watch logs is-not a credible scenario. Tr.1390-91,1394 (Powers). Such alteration would require access to the physical logs and the backup floppy disk. On the hard copy, there would likely be' detectable " white outs" or " cut outs." Tr.1397 (Cosgrove). j V-36. The fire watch logs can be cross-checked from information generated by i the plant security card key reader system. Tr. 14061409 (Cosgrove,- Powers). 'Ihis system documents the physical passage ofindividuals through' doors or areas of the plant. Id. The latter documentation is maintained by the plant security group in its own - computer and file system. Id. These two separate systems for recording passage of f E w - r e v-~ e-+z.-,-,---es-,- e-m .w v,.. e.,,, r-~~-nm e r-

o - 133 - personnel through the plant are ir%t of one another, and the two systems are staffed and managed by different groups at DCPP. Tr. 1393-94 (Powers, Cosgrove). i These systems provide separate methods and records which may be examined to confirm that roving fire patrol watches at DCPP are conducted pursuant to requirements. Id. V-37. There has been no evidence of tampering with the fire watch' logs or i records concerning documentation of fire watch rounds..Tr. 1390-91 (Powers and-Cosgrove); Tr. 1398-99 (Cosgrove, Powers); Tr. 1462-63 (Miller).- V-38. 'Ihere have been nine instances of " missed" fire ' watches and disabled fire i barriers at DCPP since 1990. NRC Staff Direct Testimony at 5 (Miller). - V-39. MFP introduced into evidence 'documentationt of several instances of. " missed" fire watches. MFP Exhibit F-1 A, LER 1-91-020 00 (March -31,.1993), ) i was addressed. in the NRC Staff's testimony as were the other LERs that MFP i I introduced. 'Ihese were MI :' F-2, LER 1-92-014-00 (October 7,1992); MFP. F-3, LER 1-092-008-00 (July 22,1992); MFP F-5, LER 2-92-006-00 (November 25,1992); and MFP F-6, LER 1-92-028-00 (December 28,1992). V-40. None of these instances of " missed" fire watches evidences a failure to perform an interim compensatory measure fire watch in a Thermo-Lag area. PG&E Direct Testimony at 18 (Cosgrove, Powers). Nor do they involve watches missed for a? known fire protection impairment. In all of these instances, the need for the watch had not been established, so that the fire watch was not " missed." Id. V-41. MFP Exhibit F-1A, LER 1-91-020-00, dated March 31, 1992, concerns a damaged ceiling tile in an identified fire area. NRC Staff Testimony ff. Tr.- 1417 at 7

= -,~. l l 4 + l l - 134 - (Miller). The ceiling-tile does not qualify as a fire barrier. Tr. : 1361 (Powers). However, the damaged tile had potential impact on performance of a Halon fire - suppression system. Tr. 1363-64 (Cosgrove). 'Ihis condition existed for approximately j l a { three weeks, because PG&E personnel did not initially realize the potential impact _on the i [ . Halon system. Tr.1362-63 (Powers). Consequently, an hourly fire watch was not .j posted during that three-week period. Id. However, the room fire ~ detection equipment was operable and the other manual fire suppression equipment was available. The control I room, the adjacent space, was continuously manned. NRC Staff Direct Testimony at 7 (Miller). This scenario differs from implementation of fire watches in Thermo-Lag fire' [ areas, because in Thermo-Lag fire areas the need for the fire watch was establiued by l NRC Bulletin 92-01. i V-42. MFP Exhibit F-2, LER 1-92-014-00, documents three incidents in which fire watches were not implemented as required where smoke detectors were inoperable l t or fire barrier impairments existed. MFP Exhibit F-2 at 2; NRC Staff Testimony at 7 (Miller), Tr.1292 (Powers). The root cause was determined to be personnel error._ j 1 NRC Staff Direct Testimony at 7-8 (Miller). V-43. MFP Exhibit F-3, LER 1-92-008-00, reports an incident in which a sprinkler system was unavailable for six hours because of a work tag out, a situation requiring that a continuous fire watch be posted. NRC Staff Testimony at 9-10 (Mdler). However, even though the continuous fire watch was not posted as required, detection was available in the area as well as an hourly fire watch. The Licensee's action, which -

e - 135 - . included adding fire watch considerations as a first step on tagz out-forms, 'was' apprcpriate. Id. V-44. MFP Exhibit F-5, LER 2-92-006-00, described two instances, one on' October 30,1992, and one on November 14,1992, in which operators did not recognize that required compensatory fire watches had not been established. 'Ihe operators had - apparently grown accustomed to spurious alarms from certain fire detectors. In the two-j i cases reported, the operators had assumed compensatory measures had been taken

j because there had been four previous instances of spurious alarms which had initiated compensatory measures. The action statement was exemlM for thirty-one (31) minutes in one instance, and about thirty-seven (37) hours in the other. In the second instance,.

j l although a continuous fire watch was not initiatM as ' required, an hourly roving watch. l l was in effect in the area. NRC Staff Testimony at 9 (Miller); Tr.1325 (Cosgrove)'. V-45. MFP Exhibit F-6, LER 2-92-028-00, described two fire detection con-puter failures: one on October 1,1992, and one on November 26,.1992. - The computer failures should have annunciated in the control room, but did not..This failure to annunciate was caused by a software error. The action statement, to establish-fire-watches, was exceeded for twenty-nine (29) minutes in one case and about fifteen (15) hours in the other. The missed compensatory action for the second instance was the establishment of a continuous fire watch in the control room, which is a continuously i manned space. The Licensee's actions were satisfactory and aimed at resolving the software deficiency and improving software quality assurance. NRC Staff Testimony at 8 ' (Miller); Tr. 1342-43 (Cosgrove). i .=,

s - 136 - V-46. MFP also asked questions about MFP Exhibit 137, an NRC inspection report admitted into evidence regarding Contention I. It was ncver formally offered as regards this contention. The inspection report at pages 8 and 9, documents a case in which an Appendix R emergency light was obscured by construction craft personnel, who hung a tarp to prevent dust and debris from entering a switchgear cabinet. Tr. 1355-56 (Cosgrove, Powers). This incident did not involve a Thermo-Lag compensatory measure. Tr.1357 (Cosgrove). No connection to Contention V was established, so that this document is not entitled to any evidentiary weight. V-47. PG&E has been successful in implementing compensatory measures. Tr.1467-69 (Miller). The aforementioned incidents do not demonstrate any inability by PG&E to implement or perform compensatory measures pursuant to NRC Bulletin 92-01. Nor do they suggest a legitimate doubt about PG&E's ability to implement a compensatory fire watch program. V-48. Each instance of a missed fire watch and disabled fire barriers was oflow safety significance. NRC Staff Direct Testimony at 10-11 (Miller). The instances of missed fire watches represent a very low percentage of the total number of fire watches conducted. Id. More importantly, however, Thermo-Lag fire watches involve the Defense-in-Depth approach, as fire watches in Thermo-I2g areas are an addition to existing detection and suppression systems. Id. No missed fire watch was associated with the compensatory measures the Licensee initiated due to the NRC's generic Thermo-Lag concerns. Id. In all of the other situations, PG&E's Defense-in-Depth approach was

=. I .O t i - 137 - operatmg. Id. Fire loading (combustibles) was low, and detection and suppression 4 l devices were available for most areas of concern. Id. I e j V-49. MFP would have us adopt a finding that certain of our rulings established f j that we allowed for discussion all aspects of the fire protection program at DCPP in all i fire areas, not exclusively the eleven areas where 'Ihermo-Lag material is used. MFP i }j i Proposed Finding 789. MFP reads our rulings much too expansively. When we ruled, I i Tr.1298-99, that all interim compensatory measures, including but not limited to fire o } watches, were open to question, we meantjust that. Similarly our ruling, Tr.1297,'that 1 j inquiry into deficiencies in the fire watch program would be permissible, does not invite i { the expanded reading that MFP offers in its Proposed Finding 789. Clearly, if we cannot inquire into the correctness of the Commission's decision concerning'the need for 4 compensatory measures in view of the declared inoperability of Thermo-Lag, then we ' 4 cannot examine the scattered failures in other aspects of PG&E's fire protection program that dictated a need for compensatory measures, a need that in the instances cited was not - l met. The NRC Staff's testimony discussed instances where the need for compensatory } measures was not recognized and, therefore, not met. It distinguished these instances i 4 from the situation with which we are confronted, that involving Thermo-lag, where the h j need was established by an NRC bulletin. The admitted contention concerns the adequacy of PG&E's implementation of the interim compensatory measures required by. the NRC in connection with the use of Thermo-Lag at DCPP. It does not concern the L substantive failures that established a need for compensatory measures.' L i ~..

- 138 - V-50. For the reasons discussed above, we do not adopt MFP's proposed findings concerning MFP Exhibit F-1 A, nor MFP's proposed findings concerning MFP Exhibit F-2. See MFP Proposed Findings 798-811. V-51. Regarding MFP Exhibit F-5, MFP would have us find that the events discussed in that exhibit illustrate "the fallibility of equipment and plant personnel at DCPP." MFP Proposed Finding 824. However, not only has MFP not established a basis for such a conclusion, such record evidence as exists on this matter is to the contrary. See PG&E Direct Testimony at 16-18 (Cosgrove, Powers). Even if we were generally to allow MFP to substitute its own post-hearing reading of an exhibit for the testimony of the witnesses, we could not do so in this instance, as there was virtually no cross-examination on the document. We cannot condone a practice where findings are based on extra-record evidence not subject to cross-examination, especially where record evidence is to the contrary. V-52. Similarly, we reject MFP's attempt to have us generahze from the " missed" fire watch involved in MFP Exhibit F-3 to all missed fire watches, including those required by the declared inoperability of Thermo-Lag barriers. MFP Proposed Findings 825-830. V-53. We do not adopt MFP's Proposed Findings regarding MFP Exhibit F-6, as those findings, too, extend far beyond the scope of the admitted contention. MFP Proposed Findings 831-838. MFP would have us conclude that "an inoperable fire detection system effects [ sic] the total fire protection program at DCNPP, including the areas containing Thermo-Lag fire barrier material." MFP Proposed Finding 838.

1 j a 1 'l - 139 - Although this finding might seem to be self-evident, the record evidence established that _ 1 the compensatory measures in place for 'Ihermo-Lag, notably the fire watches, were not i i affected by anomalies in the fire detection / system. *Ihe testimony of Ms Miller j regarding the significance of the events reported in the LERs addressed in the NRC Staff.- l Testimony - she testified that they were oflow safety significance - was uncontroverted. NRC Staff Testimony at 10 (Miller). - See also Tr. 1452,146849 (Miller). V-54. Contention V is without substance. PG&E has instituted the Thermo-lag - compensatory measures required by NRC Bulletin 92-01. There is nothing in the record t that would lead one to conclude that the interim compensatory measures are not being and a will not be implemented until the resolution of the generic Thermo-lag issue. Rather, PG&E has met its burden of proof in demonstrating that the interim compensatory measures have been more than adequately implemented and that there is every reason to j believe that this program will continue to operate as long as it is required. V. CONCLUSION I The Board has considered all the evidence submitted by the parties and the entire record in this proceeding. That record consists of the Commission's Notice of Hearing, - the pleadings filed herein, memoranda and orders formerly _ issued in this proceeding, the transcript of hearing and all exhibits received into evidence. On that basis we find that: 1. Our findings herein are supported by reliable, probative and substantial evidence as required by the Administrative Procedure Act,5 U.S.C. i 551 et seq., and ~ the Commission's Rules of Practice,10 C.F.R.~ Part 2,-and are based on the entire i record. e. m ..w .g....w.. c s - m..,.u.- n w.wm y,,.u -+-.e

=- i l ~ m l - 140 - 2. All issues, arguments or propcx4 findings presented by the parties,' but I not addressed herein have been found to be without merit or unmeary for this - l decision. .i 3. As to the Contentions admitted in this proceeding, PG&E has met its i burden of proof. 4. In regard to those C(ntentions there is reasonable assurance that if the l license amendments sought are granted, the activities authorized by those amendments which would permit a 40 year operating life for the Diablo Canyon Nuclear Power Plant, j Units 1 & 2, can be conducted without endangering the public health and safety, will not be inimical to the common defense and security or the environment,-and will be i conducted in compliance with the Commission's regulations. ORDER e WHEREFORE, IT IS SO ORDERED, on the basis of the forgoing, that the l ti Director, Nuclear Reactor Regulation is authorized, upon making requisite findings on i matters not at issue herein, to issue the license amendments sought by PG&E in its l I application of July 9,1992. l 'Ihis Initial Decision, in accordance with 10 C.F.R. I 2.760, will constitute the i final action of the Commission 40 days after its date 'of issuance, subject to review-s pursuant to the Commission's regulations. 1 Pursuant to 10 C.F.R. I 2.786, any petition for review of this Initial Decision i must be filed within 15 days after service of the decision. AnyLother party may file, within 10 days after service of a petition for review, an answer in support of, or in ' i _,,,,,.-,..-._.,c-._ .._,-._.,mn,,.m,,.. -.... ~. _ _., _. -,

-~ w - 141 - opposition to, the petition for review. The petition for review may be granted or denied - j in the discretion of the Commission under 10 C.F.R. I 2.786(b)(4). ) i Respectfully submitted, 1 hN .N bC,C Ann P. Hodgdon Q Counsel for NRC Staff Arlene(A. Jorgensen Counsel for NRC Staff Dated at Rockville, Maryland this 22nd day of December,1993

I l l ~ L q UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 93 DE 23 AS 16 l BEFORE THE ATOMIC SAFETY AND LICENSING BOARD.... tw< tn In the Matter of ) ~ ) Docket Nos. 50-275 OIA-2 PACIFIC GAS & ELECTRIC COMPANY ) 50-323 OLA-2 l ) l (Diablo Canyon Nuclear Power Plant, ) (Construction Period Recovery) Units 1 and 2) ) ) l l CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF'S FINDINGS OF FACT' AND CONCLUSIONS OF LAW IN THE FORM OF AN INITIAL DECISION" in the above captioned proceeding have been served on the following by deposit in the United States mail first class or, as indicated by an asterisk, by deposit in the Nuclear Regulatory l Commission's internal mail system, this 22nd day of December 1993: Charles Bechhoefer* Office of Commission Appellate Administrative Judge Adjudication

• Atomic Safety and Licensing Board Mail Stop: 16-G-15 OWFN Mail Stop: EW-439 U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 Washington, DC 20555 Adjudicatory File * (2)

Jerry R. Kline* Atomic Safety and Licensing Board Administrative Judge Panel Atomic Safety and Licensing Board Mail Stop: EW-439 Mail Stop: EW-439 U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 Washington, DC 20555 Atomic Safety and Licensing Board Frederick J. Shon* Panel

• Administrative Judge Mail Stop: EW-439 Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Mail Stop: EW-439 Washington, DC 20555 U.S. Nuclear Regulatory Commission Washington, DC 20555

1 \\ l \\ y i i ' l i Nancy Culver, President-Office of the Secretary * (2). San Luis Obispo Mothers for Peace Attn: Docketing and Service P.O. Box 164 Mail Stop: 16-G-15 OWFN Pismo Beach, CA 93448 U.S. Nuclear Regulatory Commission Washington, DC 20555 i Christopher J. Warner Richard F. Incke Joseph B. Knotts, Jr., Esq. Pacif i Gas & Electric Co. David A. Repka, Esq. 77 Beale Street Kathryn M. Kalowsky, Esq. San Francisco, CA 94106 Winston & Strawn 1400 L Street, N.W. Greg Minor Washington, DC 20005-3502 MHB Technical Assoc. j 1723 Hamilton Avenue, Suite K Truman Burns i San Jose, CA 9:MS Robert Kinosian Califomia Public Utilities Commission i l Diane Curran 505 Van Ness, Room 4103 i IEER San Francisco, CA 94102 6935 Laurel Avenue, Suite 204 Takoma Park, MD 20912 [0 e C Ann P.' Hodgdon '( ' Counsel for NRC Staff ( J I l . i = w wwe v-9 .w -'qv u 9 Wv}}