ML060660368

Issuance of Amendments 185 & 187 to Licenses Re Changes to Technical Specifications on Inoperable Snubbers

ML060660368
Person / Time
Site:	Diablo Canyon
Issue date:	03/07/2006
From:	Wang A Plant Licensing Branch III-2
To:	Kennan J Pacific Gas & Electric Co
Wang A, NRR/DLPM, 415-1445
References
TAC MC8878, TAC MC8879
Download: ML060660368 (24)

Text

March 7, 2006Mr. John S. Keenan Senior Vice President and CNO Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 770000 San Francisco, CA 94177-0001

SUBJECT:

DIABLO CANYON POWER PLANT, UNIT NOS. 1 AND 2 - ISSUANCE OFAMENDMENTS RE: INOPERABILITY OF SNUBBERS (TAC NOS. MC8878 AND MC8879)

Dear Mr. Keenan:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosedAmendment No.185 to Facility Operating License No. DPR-80 and Amendment No. 187 toFacility Operating License No. DPR-82 for the Diablo Canyon Power Plant, Unit Nos. 1 and 2,respectively. The amendments consist of changes to the Technical Specifications (TS) in response to your application dated October 19, 2005.The amendments revise the TSs by allowing a delay time for entering a support ed system TSwhen the inoperability is due solely to an inoperable snubber, if risk is assessed and managedconsistent with the program in place for complying with the requirements of Title 10 of the Codeof Federal Regulations, Paragraph 50.65(a)(4). Limiting Condition for Operation 3.0.8 is addedto the TSs to provide this allowance and define its requirements and limitations.A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included inthe Commission's next regular biweekly Federal Register notice.Sincerely,

/RA/Alan Wang, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket Nos. 50-275 and 50-323

Enclosures:

1. Amendment No. 185 to DPR-802. Amendment No. 187 to DPR-82

3. Safety Evaluationcc w/encls: See next page March 7, 2006Mr. John S. Keenan Senior Vice President and CNO Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 770000 San Francisco, CA 94177-0001

SUBJECT:

DIABLO CANYON POWER PLANT, UNIT NOS. 1 AND 2 - ISSUANCE OFAMENDMENTS RE: INOPERABILITY OF SNUBBERS (TAC NOS. MC8878 AND MC8879)

Dear Mr. Keenan:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosedAmendment No.185 to Facility Operating License No. DPR-80 and Amendment No. 187 toFacility Operating License No. DPR-82 for the Diablo Canyon Power Plant, Unit Nos. 1 and 2,respectively. The amendments consist of changes to the Technical Specifications (TS) in response to your application dated October 19, 2005.The amendments revise the TSs by allowing a delay time for entering a support ed system TSwhen the inoperability is due solely to an inoperable snubber, if risk is assessed and managedconsistent with the program in place for complying with the requirements of Title 10 of the Codeof Federal Regulations, Paragraph 50.65(a)(4). Limiting Condition for Operation 3.0.8 is addedto the TSs to provide this allowance and define its requirements and limitations.A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included inthe Commission's next regular biweekly Federal Register notice.Sincerely,/RA/

Alan Wang, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket Nos. 50-275 DISTRIBUTION

and 50-323PUBLICGHill (4)LPLIV ReadingTBoyce

Enclosures:

1. Amendment No. 185 to DPR-80RidsNrrDorl (CHaney/CHolden)2. Amendment No. 187 to DPR-82 RidsNrrDorlLplg (DTerao)RidsOgcRp

3. Safety EvaluationRidsNrrPMAWangRidsNrrLALFeizollahicc w/encls: See next pageRidsAcrsAcnwMailCenterRidsRegion4MailCenter (BJones)RidsNrrDprACCESSION NO.: ML060660368OFFICECLIIP LPMNRR/LPL4/PMNRR/LPL4/LATSBOGCNRR/LPL4/BCNAMEWReckleyAWangLFeizollahiBTjaderMSpencerDTeraoDATE1/23/062/15/062/14/062/17/062/24/062/28/06DOCUMENT NAME: E:\Filenet\ML060660368.wpdOFFICIAL RECORD COPY PACIFIC GAS AND ELECTRIC COMPANYDOCKET NO. 50-275DIABLO CANYON NUCLEAR POWER PLANT, UNIT NO. 1AMENDMENT TO FACILITY OPERATING LICENSEAmendment No. 185License No. DPR-801.The Nuclear Regulatory Commission (the Commission) has found that:A.The application for amendment by Pacific Gas and Electric Company (thelicensee) dated October 19, 2005, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I;B.The facility will operate in conformity with the application, the provisions of theAct, and the rules and regulations of the Commission;C.There is reasonable assurance (i) that the activities authorized by thisamendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with theCommission's regulations;D.The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public; and E.The issuance of this amendment is in accordance with 10 CFR Part 51 of theCommission's regulations and all applicable requirements have been satisfied.2.Accordingly, the license is amended by changes to the Technical Specifications asindicated in the attachment to this license amendment, and paragraph 2.C.(2) of FacilityOperating License No. DPR-80 is hereby amended to read as follows: (2)Technical SpecificationsThe Technical Specifications contained in Appendix A and the EnvironmentalProtection Plan contained in Appendix B, as revised through Amendment No. 185, are hereby incorporated in the license. Pacific Gas and Electric Company shall operate the facility in accordance with the TechnicalSpecifications and the Environmental Protection Plan, except where otherwisestated in specific license conditions.3.This license amendment is effective as of its date of issuance and shall be implementedwithin 120 days of issuance.FOR THE NUCLEAR REGULATORY COMMISSION

/RA/David Terao, ChiefPlant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical SpecificationsDate of Issuance: March 7, 2006 PACIFIC GAS AND ELECTRIC COMPANYDOCKET NO. 50-323DIABLO CANYON NUCLEAR POWER PLANT, UNIT NO. 2AMENDMENT TO FACILITY OPERATING LICENSEAmendment No. 187License No. DPR-821.The Nuclear Regulatory Commission (the Commission) has found that:A.The application for amendment by Pacific Gas and Electric Company (thelicensee) dated October 19, 2005, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I;B.The facility will operate in conformity with the application, the provisions of theAct, and the rules and regulations of the Commission;C.There is reasonable assurance (i) that the activities authorized by thisamendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with theCommission's regulations;D.The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public; and E.The issuance of this amendment is in accordance with 10 CFR Part 51 of theCommission's regulations and all applicable requirements have been satisfied.2.Accordingly, the license is amended by changes to the Technical Specifications asindicated in the attachment to this license amendment, and paragraph 2.C.(2) of FacilityOperating License No. DPR-82 is hereby amended to read as follows: (2)Technical SpecificationsThe Technical Specifications contained in Appendix A and the EnvironmentalProtection Plan contained in Appendix B, as revised through Amendment No. 187, are hereby incorporated in the license. Pacific Gas and Electric Company shall operate the facility in accordance with the TechnicalSpecifications and the Environmental Protection Plan, except where otherwisestated in specific license conditions.3.This license amendment is effective as of its date of issuance and shall be implementedwithin 120 days of issuance.FOR THE NUCLEAR REGULATORY COMMISSION

/RA/David Terao, ChiefPlant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical SpecificationsDate of Issuance: March 7, 2006 ATTACHMENT TO LICENSE AMENDMENT NO. 185 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 187 TO FACILITY OPERATING LICENSE NO. DPR-82DOCKET NOS. 50-275 AND 50-323Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. REMOVEINSERT3.0-13.0-13.0-23.0-2 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIONRELATED TO AMENDMENT NO. 185 TO FACILITY OPERATING LICENSE NO. DPR-80AND AMENDMENT NO. 187 TO FACILITY OPERATING LICENSE NO. DPR-82PACIFIC GAS AND ELECTRIC COMPANYDIABLO CANYON POWER PLANT, UNITS 1 AND 2DOCKET NOS. 50-275 AND 50-32

31.0INTRODUCTION

By application dated October 19, 2005 (Agencywide Documents Access Management SystemAccession No. ML053000316), Pacific Gas and Electric Company (the licensee) requested changes to the Technical Specifications (TSs, Appendix A to Facility Operating LicenseNos. DPR-80 and DPR-82) for the Diablo Canyon Power Plant, Units 1 and 2. The proposed change would add Limiting Condition for Operation (LCO) 3.0.8 to addressconditions where one or more snubbers are unable to perform their associated support function. The change is based on Technical Specification Task Force (TSTF) change traveler TSTF-372, Revision 4, that has been approved generically for the Standard TechnicalSpecifications (STS; NUREGs-1430 - 1434). A notice announcing the availability of thisproposed TS change using the consolidated line item improvement process (CLIIP) was published in the Federal Register on May 4, 2005 (70 FR 23252). A description of TSTF-372and its associated TS change, now follows.On April 23, 2004, the Nuclear Energy Institute Risk Informed Technical Specifications TaskForce (RITSTF) submitted a proposed change, TSTF-372, Revision 4, to the STS on behalf of the industry (TSTF-372, Revisions 1 through 3 were prior draft iterations). TSTF-372, Revision 4, is a proposal to add an LCO allowing a delay time for entering a supported systemTS, when the inoperability is due solely to an inoperable snubber, if risk is assessed andmanaged. The postulated seismic event requiring snubbers is a low-probability occurrence andthe overall TS system safety function would still be available for the vast majority of anticipated challenges.This proposal is one of the industry's initiatives being developed under the risk-informedtechnical specifications program. These initiatives are intended to maintain or improve safety through the incorporation of risk assessment and management techniques in TS, while reducing unnecessary burden and making technical specification requirements consistent with the Commission's other risk-informed regulatory requirements, in particular the Maintenance Rule. The proposed change adds a new limiting condition of operation, LCO 3.0.8, to the TS. LCO 3.0.8 allows licensees to delay declaring an LCO not met for equipment, supported by snubbers unable to perform their associated support functions, when risk is assessed and managed. This new LCO 3.0.8 states:When one or more required snubbers are unable to perform their associated supportfunction(s), any affected supported LCO(s) are not required to be declared not met solely for this reason if risk is assessed and managed, and:a.the snubbers not able to perform their associated support function(s) areassociated with only one train or subsystem of a multiple train or subsystemsupported system or are associated with a single train or subsystem supportedsystem and are able to perform their associated support function within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />;or b.the snubbers not able to perform their associated support function(s) areassociated with more than one train or subsystem of a multiple train orsubsystem supported system and are able to perform their associated supportfunction within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.At the end of the specified period the required snubbers must be able to perform theirassociated support function(s), or the affected support ed system LCO(s) shall bedeclared not met.

2.0 REGULATORY EVALUATION

In Section 50.36 of Title 10 of the Code of Federal Regulations (10 CFR), the Commissionestablished its regulatory requirements related to the content of the TSs. Pursuant to 10 CFR50.36, TSs are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings;(2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) designfeatures; and (5) administrative controls. The rule does not specify the particular requirements to be included in a plant's TSs. As stated in 10 CFR 50.36(c)(2)(i), the "Limiting conditions for operation are the lowest functional capability or performance levels of equipment required forsafe operation of the facility. When a limiting condition for operation of a nuclear reactor is notmet, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications ..." TS Section 3.0, on "LCO and SR Applicability," provides details orground rules for complying with the LCOs. Snubbers are chosen in lieu of rigid supports in areas where restricting thermal growth duringnormal operation would induce excessive stresses in the piping nozzles or other equipment.

Although they are classified as component standard supports, they are not designed to provide any transmission of force during normal plant operations. However, in the presence of dynamic transient loadings, which are induced by seismic events as well as by plant accidents and transients, a snubber functions as a rigid support. The location and size of the snubbers are determined by stress analysis based on different combinations of load conditions, depending on the design classification of the particular piping. Prior to the conversion to the improved STS, TS requirements applied directly to snubbers. These requirements included:

!A requirement that snubbers be functional and in service when the supported equipmentis required to be operable,!A requirement that snubber removal for testing be done only during plant shutdown,!A requirement that snubber removal for testing be done on a one-at-a-time basis whensupported equipment is required to be operable during shutdown, !A requirement to repair or replace within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> any snubbers, found to be inoperableduring operation in Modes 1 through 4, to avoid declaring any supported equipment inoperable,!A requirement that each snubber be demonstrated operable by periodic visualinspections, and

!A requirement to perform functional tests on a representative sample of at least 10% ofplant snubbers, at least once every 18 months during shutdown.In the late 1980s, a joint initiative of the NRC and industry was undertaken to improve the STS. This effort identified the snubbers as candidates for relocation to a licensee-controlled document based on the fact that the TS requirements for snubbers did not meet any of the fourcriteria in 10 CFR 50.36(c)(2)(ii) for inclusion in the improved STS. The NRC approved therelocation without placing any restriction on the use of the relocated requirements. However, this relocation resulted in different interpretations between the NRC and the industry regardingits implementation. The NRC has stated, that since snubbers are supporting safety equipmentthat is in the TS, the definition of OPERABILITY must be used to immediately evaluateequipment supported by a removed snubber and, if found inoperable, the appropriate TS required actions must be entered. This interpretation has in practice eliminated the 72-hour delay to enter the actions for the supported equipment that existed prior to the conversion to theimproved STS (the only exception is if the support ed system has been analyzed anddetermined to be OPERABLE without the snubber). The industry has argued that since the

NRC approved the relocation without placing any restriction on the use of the relocatedrequirements, the licensee controlled document requirements for snubbers should be invoked before the support ed system's TS requirements become applicable. The industry'sinterpretation would, in effect, restore the 72-hour delay to enter the actions for the supported equipment that existed prior to the conversion to the improved STS. The industry's proposalwould allow a time delay for all conditions, including snubber removal for testing at power. The option to relocate the snubbers to a licensee-controlled document, as part of the conversion to improved STS, has resulted in non-uniform and inconsistent treatment of snubbers. On the one hand, plants that have relocated snubbers from their TS are allowed to change the TSrequirements for snubbers under the auspices of 10 CFR 50.59, but they are not allowed a 72-hour delay before they enter the actions for the supported equipment. On the other hand,plants that have not converted to improved STS have retained the 72-hour delay if snubbersare found to be inoperable, but they are not allowed to use 10 CFR 50.59 to change TS requirements for snubbers. It should also be noted that a few plants that converted to theimproved STS chose not to relocate the snubbers to a licensee-controlled document and, thus, 1The following technical evaluation is presented in terms of the bounding assessment of thischange for all commercial nuclear power plants, including Diablo Canyon, performed as part ofthe approval of TSTF-372, Revision 4, and publication of the CLIIP notices.retained the 72-hour delay. In addition, it is important to note that unlike plants that have notrelocated, plants that have relocated can perform functional tests on the snubbers at power (aslong as they enter the actions for the supported equipment) and at the same time can reducethe testing frequency (as compared to plants that have not relocated) if it is justified by 10 CFR 50.59 assessments. Some potential undesirable consequences of this inconsistent treatmentof snubbers are:

!Performance of testing during crowded time period windows w hen the suppor ted systemis inoperable with the potential to reduce the snubber testing to a minimum since the snubber requirements that have been relocated from TSs are controlled by the licensee,!Performance of testing during crowded windows when the supported system isinoperable with the potential to increase the unavailability of safety systems, and

!Performance of testing and maintenance on snubbers affecting multiple trains of thesame supported system during the 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> allotted before entering MODE 3 underLCO 3.0.3.To remove the inconsistency in the treatment of snubbers among plants, the TSTF proposed arisk-informed TS change that introduces a delay time before entering the actions for the supported equipment, when one or more snubbers are found inoperable or removed for testing, if risk is assessed and managed. Such a delay time will provide needed flexibility in theperformance of maintenance and testing during power operation and at the same time willenhance overall plant safety by:

!Avoiding unnecessary unscheduled plant shutdowns and, thus, minimizing planttransition and realignment risks,!Avoiding reduced snubber testing and, thus, increasing the availability of snubbers to perform their supporting function,!Performing most of the required testing and maintenance during the delay time whenthe supported system is available to mitigate most challenges and, thus, avoidingincreases in safety system unavailability, and

!Providing explicit risk-informed guidance in areas in which that guidance currently doesnot exist, such as the treatment of snubbers impacting more than one redundant train of a support ed system.

3.0TECHNICAL EVALUATION

1The industry submitted TSTF-372, Revision 4, "Addition of LCO 3.0.8, Inoperability ofSnubbers," in support of the proposed TS change. This submittal (Reference 1) documents a risk-informed analysis of the proposed TS change. Probabilistic risk assessment (PRA) results and insights are used, in combination with deterministic and defense-in-depth arguments, toidentify and justify delay times for entering the actions for the supported equipment associatedwith inoperable snubbers at nuclear power plants. This is in accordance with guidance provided in Regulatory Guides (RGs) 1.174 and 1.177 (References 2 and 3, respectively).The risk impact associated with the proposed delay times for entering the TS actions for thesupported equipment can be assessed using the same approach as for allowed completion time (CT) extensions. Therefore, the risk assessment was performed following the three-tiered approach recommended in RG 1.177 for evaluating proposed extensions in currently allowed

CTs:!The first tier involves the assessment of the change in plant risk due to the proposed TSchange. Such risk change is expressed (1) by the change in the average yearly coredamage frequency (CDF) and the average yearly large early release frequency (LERF) and (2) by the incremental conditional core damage probability (ICCDP) andthe incremental conditional large early release probability (ICLERP). The assessedCDF and LERF values are compared to acceptance guidelines, consistent with theCommission's Safety Goal Policy Statement as documented in RG 1.174, so that theplant's average baseline risk is maintained within a minimal range. The assessed ICCDP and ICLERP values are compared to acceptance guidelines provided inRG 1.177, which aim at ensuring that the plant risk does not increase unacceptablyduring the period the equipment is taken out of service.

!The second tier involves the identification of potentially high-risk configurations thatcould exist if equipment in addition to that associated with the change were to be taken out of service simultaneously, or other risk-significant operational factors such as concurrent equipment testing were also involved. The objective is to ensure that appropriate restrictions are in place to avoid any potential high-risk configurations.

!The third tier involves the establishment of an overall configuration risk managementprogram (CRMP) to ensure that potentially risk-significant configurations resulting from maintenance and other operational activities are identified. The objective of the CRMP is to manage configuration-specific risk by appropriate scheduling of plant activities and/or appropriate compensatory measures.A simplified bounding risk assessment was performed to justify the proposed addition ofLCO 3.0.8 to the TS. This approach was necessitated by (1) the general nature of the proposed TS changes (i.e., they apply to all plants and are associated with an undetermined number of snubbers that are not able to perform their function), (2) the lack of detailed engineering analyses that establish the relationship between earthquake level and supportedsystem pipe failure probability when one or more snubbers are inoperable, and (3) the lack ofseismic risk assessment models for most plants. The simplified risk assessment is based on the following major assumptions, which the staff finds acceptable, as discussed below:

!The accident sequences contributing to the risk increase associated with the proposedTS changes are assumed to be initiated by a seismically-induced loss-of-offsite-power (LOOP) event with concurrent loss of all safety system trains supported by the out-of- service snubbers. In the case of snubbers associated with more than one train (orsubsystem) of the same system, it is assumed that all affected trains (or subsystems) ofthe supported system are failed. This assumption was introduced to allow theperformance of a simple bounding risk assessment approach with application to all plants. This approach was selected due to the lack of detailed plant-specific seismic risk assessments for most plants and the lack of fragility data for piping w hen one ormore supporting snubbers are inoperable.

!The LOOP event is assumed to occur due to the seismically-induced failure of theceramic insulators used in the power distribution systems. These ceramic insulators have a high confidence (95%) of low probability (5%) of failure (HCLPF) of about 0.1g,expressed in terms of peak ground acceleration. Thus, a magnitude 0.1g earthquake is conservatively assumed to have 5% probability of causing a LOOP initiating event. Thefact that no LOOP events caused by higher magnitude earthquakes were considered is justified because (1) the frequency of earthquakes decreases with increasing magnitude and (2) historical data (References 4 and 5) indicate that the mean seismic capacity of ceramic insulators (used in seismic PRAs), in terms of peak ground acceleration, is about 0.3g, which is significantly higher than the 0.1g HCLPF value. Therefore, thesimplified analysis, even though it does not consider LOOP events caused by earthquakes of magnitude higher than 0.1g, bounds a detailed analysis which would usemean seismic failure probabilities (fragilities) for the ceramic insulators.

!Analytical and experimental results obtained in the mid-eighties as part of the industry's"Snubber Reduction Program" (References 4 and 6) indicated that piping systems havelarge margins against seismic stress. The assumption that a magnitude 0.1g earthquake would cause the failure of all safety system trains supported by the out-of-service snubbers is very conservative because safety piping systems could withstandmuch higher seismic stresses even when one or more supporting snubbers are out of service. The actual piping failure probability is a function of the stress allowable and thenumber of snubbers removed for maintenance or testing. Since the licensee controlled testing is done on only a small (about 10%) representative sample of the total snubber population, typically only a few snubbers supporting a given safety system are out fortesting at a time. Furthermore, since the testing of snubbers is a planned activity, licensees have flexibility in selecting a sample set of snubbers for testing from a muchlarger population by conducting configuration-specific engineering and/or risk assessments. Such a selection of snubbers for testing provides confidence that the supported systems would perform their functions in the presence of a design-basisearthquake and other dynamic loads and, in any case, the risk impact of the activity willremain within the limits of acceptability defined in risk-informed RGs 1.174 and 1.177.

!The analysis assumes that one train (or subsystem) of all safety systems is unavailableduring snubber testing or maintenance (an entire system is assumed unavailable if aremoved snubber is associated with both trains of a two-train system). This is a veryconservative assumption for the case of corrective maintenance since it is unlikely that a visual inspection will reveal that one or more snubbers across all supported systems areinoperable. This assumption is also conservative for the case of the licensee-controlled testing of snubbers since such testing is performed only on a small representative sample. !In general, no credit is taken for recovery actions and alternative means of performing afunction, such as the function performed by a system assumed failed (e.g., whenLCO 3.0.8b applies). However, most plants have reliable alternative means of performing certain critical functions. For example, feed and bleed (F&B) can be used to remove heat in most pressurized-water reactors (PWRs) when auxiliary feedwater(AFW), the most important system in mitigating LOOP accidents, is unavailable. Similarly, if high pressure makeup (e.g., reactor core isolation cooling) and heat removal capability (e.g., suppression pool cooling) are unavailable in boiling-water reactors(BWRs), reactor depressurization in conjunction with low pressure makeup (e.g., low pressure coolant injection) and heat removal capability (e.g., shutdown cooling) can beused to cool the core. A 10% failure probability for recovery actions to provide corecooling using alternative means is assumed for Diablo Canyon, the only West Coast PWR plant with F&B capability, when a snubber impacting more than one train of theAFW system (i.e., when LCO 3.0.8b is applicable) is out of service. This failureprobability value is significantly higher than the value of 2.2E-2 used in Diablo Canyon'sPRA. Furthermore, Diablo Canyon has analyzed the impact of a single limiting snubber failure, and concluded that no single snubber failure would impact two trains of the AFW. No credit for recovery actions to provide core cooling using alternative means is necessary for West Coast PWR plants with no F&B capability because it has beendetermined that there is no single snubber whose non-functionality would disable two trains of an AFW in a seismic event of magnitude up to the plant's safe shutdown earthquake (SSE). It should be noted that a similar credit could have been applied to most Central and Eastern U.S. plants but this was not necessary to demonstrate the low-risk impact of the proposed TS change due to the lower earthquake frequencies at Central and Eastern U.S. plants as compared to West Coast plants.

!The earthquake frequency at the 0.1g level was assumed to be 1E-3/year for Centraland Eastern U.S. plants and 1E-1/year for West Coast plants. Each of these two values envelop the range of earthquake frequency values at the 0.1g level, for Eastern U.S.

and West Cost sites, respectively (References 5 and 7).

!The risk impact associated with non-LOOP accident sequences (e.g., seismicallyinitiated loss-of-coolant-accident (LOCA) or anticipated-transient-without-scram (ATWS)sequences) was not assessed. However, this risk impact is small compared to the risk impact associated with the LOOP accident sequences modeled in the simplified bounding risk assessment. Non-LOOP accident sequences, due to the ruggedness of nuclear power plant designs, require seismically-induced failures that occur at earthquake levels above 0.3g. Thus, the frequency of earthquakes initiating non-LOOP accident sequences is much smaller than the frequency of seismically-initiated LOOP events. Furthermore, because of the conservative assumption made for LOOP sequences that a 0.1g level earthquake would fail all piping associated with inoperable snubbers, non-LOOP sequences would not include any more failures associated with inoperable snubbers than LOOP sequences. Therefore, the risk impact of inoperable snubbers associated with non-LOOP accident sequences is small compared to the risk impact associated with the LOOP accident sequences modeled in the simplified bounding risk assessment.

!The risk impact of dynamic loadings other than seismic loads is not assessed. Theseshock-type loads include thrust loads, blowdown loads, waterhammer loads, steamhammer loads, LOCA loads and pipe rupture loads. However, there are someimportant distinctions between non-seismic (shock-type) loads and seismic loads which indicate that, in general, the risk impact of the out-of-service snubbers is smaller for non-seismic loads than for seismic loads. First, while a seismic load affects the entire plant, the impact of a non-seismic load is localized to a certain system or area of theplant. Second, although non-seismic shock loads may be higher in total force and the impact could be as much or more than seismic loads, generally they are of much shorter duration than seismic loads. Third, the impact of non-seismic loads is more plant specific, and thus harder to analyze generically, than for seismic loads. For these reasons, licensees will be required to confirm every time LCO 3.0.8a is used, that atleast one train of eac h system that is supported by the inoperable snubber(s) wouldremain capable of performing their required safety or support functions for postulated design loads other than seismic loads.3.1Risk Assessment Results and Insights The results and insights from the implementation of the three-tiered approach of RG 1.177 tosupport the proposed addition of LCO 3.0.8 to the TS are summarized and evaluated in the following Sections 3.1.1 to 3.1.3.3.1.1Risk ImpactThe bounding risk assessment approach, discussed in Section 3.0, was implementedgenerically for all U.S. operating nuclear power plants. Risk assessments were performed for two categories of plants, Central and East Coast plants and West Coast plants, based on historical seismic hazard curves (earthquake frequencies and associated magnitudes). The first category, Central and East Coast plants, includes the vast majority of the U.S. nuclear power plant population (Reference 7). For each category of plants, two risk assessments were performed:

!The first risk assessment applies to cases where all inoperable snubbers are associatedwith only one train (or subsystem) of the impacted safety systems. It was conservativelyassumed that a single train (or subsystem) of each safety system is unavailable. It wasalso assumed that the probability of non-mitigation using the unaffected redundanttrains (or subsystems) is 2%. This is a conservative value given that for core damage tooccur under those conditions, two or more failures are required.

!The second risk assessment applies to the case where one or more of the inoperablesnubbers are associated with multiple trains (or subsystems) of the same safetysystems. It was assumed in this bounding analysis that all safety systems areunavailable to mitigate the accident, except for West Coast PWR plants. Credit for using F&B to provide core cooling is taken for plants having F&B capability (e.g., DiabloCanyon) when a snubber impacting more than one train of the AFW system isinoperable. Credit for one AFW train to provide core cooling is taken for West Coast PWR plants with no F&B capability (e.g., San Onofre) because it has been determinedthat there is no single snubber whose non-functionality would disable two trains of the AFW in a seismic event of magnitude up to the plant's SSE. The results of the performed risk assessments, in terms of core damage and large earlyrelease risk impacts, are summarized in Table 1. The first row lists the conditional riskincrease, in terms of CDF (core damage frequency), RCDF, caused by the out-of-servicesnubbers (as assumed in the bounding analysis). The second and third rows list the I CCDP(incremental conditional core damage probability) and the ICLERP (incremental conditionallarge early release probability) values, respectively. The ICCDP for the case where allinoperable snubbers are associated with only one train (or subsystem) of the supported safetysystems, was obtained by multiplying the corresponding RCDF value by the time fraction of theproposed 72-hour delay to enter the actions for the supported equipment. The ICCDP for thecase where one or more of the inoperable snubbers are associated with multiple trains (orsubsystems) of the same safety system, was obtained by multiplying the corresponding R CDFvalue by the time fraction of the proposed 12-hour delay to enter the actions for the supportedequipment. The ICLERP values were obtained by multiplying the corresponding ICCDP valuesby 0.1 (i.e., by assuming that the ICLERP value is an order of magnitude less than the ICCDP). This assumption is conservative since containment bypass scenarios, such as steam generator tube rupture accidents and interfacing system loss-of-coolant accidents, would not be uniquelyaffected by the out-of-service snubbers. Finally, the fourth and fifth rows list the assessedCDF and LERF values, respectively. These values were obtained by dividing thecorresponding ICCDP and ICLERP values by 1.5 (i.e., by assuming that the snubbers aretested every 18 months, as was the case before the snubbers were relocated to a licensee-controlled document). This assumption is reasonable because (1) it is not expected that licensees would test the snubbers more often than what used to be required by the TS, and (2) testing of snubbers is associated with higher risk impact than the average corrective maintenance of snubbers found inoperable by visual inspection (testing is expected to involvesignificantly more snubbers out of service than corrective maintenance). The assessed CDF and LERF values are compared to acceptance guidelines, consistent with the Commission'sSafety Goal Policy Statement as documented in RG 1.174, so that the plant's average baselinerisk is maintained within a minimal range. This comparison indicates that the addition of LCO 3.0.8 to the existing TS would have an insignificant risk impact.Table 1Bounding Risk Assessment Results for Snubbers Impacting a Single Train andMultiple Trains of a Supported SystemCentral and East Coast PlantsWest Coast Plants Single Train Multiple Train Single Train Multiple Train RCDF/yr1E-65E-61E-45E-4 ICCDP8E-97E-98E-77E-7 ICLERP 8E-107E-108E-87E-8 CDF/yr 5E-95E-95E-75E-7LERF/yr 5E-105E-105E-85E-8The assessed CDF and LERF values meet the acceptance criteria of 1E-6/year and1E-7/year, respectively, based on guidance provided in RG 1.174. This conclusion is truewithout taking any credit for the removal of potential undesirable consequences associated with the current inconsistent treatment of snubbers (e.g., reduced snubber testing frequency, increased safety system unavailability and treatment of snubbers impacting multiple trains)discussed in Section 1 above, and given the bounding nature of the risk assessment. The assessed ICCDP and ICLERP values are compared to acceptance guidelines provided inRG 1.177, which aim at ensuring that the plant risk does not increase unacceptably during theperiod the equipment is taken out of service. This comparison indicates that the addition ofLCO 3.0.8 to the existing TS meets the RG 1.177 numerical guidelines of 5E-7 for ICCDP and5E-8 for ICLERP. The small deviations shown for West Coast plants are acceptable because of the bounding nature of the risk assessments, as discussed in Section 2.The risk assessment results of Table 1 are also compared to guidance provided in the revisedSection 11 of NUMARC 93-01, Revision 2 (Reference 8), endorsed by RG 1.182 (Reference 9),

for implementing the requirements of paragraph (a)(4) of the Maintenance Rule, 10 CFR 50.65. Such guidance is summarized in Table 2. Guidance regarding the acceptability of conditionalrisk increase in terms of CDF (i.e., RCDF) for a planned configuration is provided. Thisguidance states that a specific configuration that is associated with a CDF higher than1E-3/year should not be entered voluntarily. Since the assessed conditional risk increase,RCDF, is significantly less than 1E-3/year, plant configurations including out of service snubbersand other equipment may be entered voluntarily if supported by the results of the riskassessment required by 10 CFR 50.65(a)(4), by LCO 3.0.8, or by other TS.Table 2Guidance for Implementing 10 CFR 50.65(a)(4)RCDFGuidanceGreater than 1E-3/yearConfiguration should not normally be enteredvoluntarilyICCDPGuidance ICLERPGreater than 1E-5Configuration should not normally be entered voluntarilyGreater than 1E-61E-6 to 1E-5Assess non-quantifiable factors Establish risk management actions1E-7 to 1E-6Less than 1E-6Normal work controlsLess than1E-7 Guidance regarding the acceptability of ICCDP and ICLERP values for a specific plannedconfiguration and the establishment of risk management actions are also provided in NUMARC 93-01. This guidance, as shown in Table 2, states that a specific plant configurationthat is associated with ICCDP and ICLERP values below 1E-6 and 1E-7, respectively, isconsidered to require "normal work controls." Table 1 shows that for the majority of plants (i.e.,

for all plants in the Central and East Coast category) the conservatively assessed ICCDP andICLERP values are over an order of magnitude less than what is recommended as the threshold for the "normal work controls" region. For West Coast plants, the conservatively assessed ICCDP and ICLERP values are still within the "normal work controls" region. Thus,the risk contribution from out-of-service snubbers is within the normal range of maintenanceactivities carried out at a plant. Therefore, plant configurations involving out-of-service snubbers and other equipment may be entered voluntarily if supported by the results of the riskassessment required by 10 CFR 50.65(a)(4), by LCO 3.0.8, or by other TSs. However, this simplified bounding analysis indicates that for West Coast plants the provisions of LCO 3.0.8 must be used cautiously and in conjunction with appropriate management actions, especially when equipment other than snubbers is also inoperable, based on the results of configuration-specific risk assessments required by 10 CFR 50.65(a)(4), by LCO 3.0.8, or by other TS.The staff finds that the risk assessment results support the proposed addition of LCO 3.0.8 tothe TS. The risk increases associated with this TS change will be insignificant based onguidance provided in RGs 1.174 and 1.177 and within the range of risks associated with normalmaintenance activities. In addition, LCO 3.0.8 will remove potential undesirable consequencesstemming from the current inconsistent treatment of snubbers in the TS, such as reducedfrequency of snubber testing, increased safety system unavailability, and the treatment ofsnubbers impacting multiple trains.3.1.2Identification of High-Risk ConfigurationsThe second tier of the three-tiered approach recommended in RG 1.177 involves theidentification of potentially high-risk configurations that could exist if equipment, in addition tothat associated with the TS change, were to be taken out of service simultaneously. Insightsfrom the risk assessments, in conjunction with important assumptions made in the analysis and defense-in-depth considerations, were used to identify such configurations. To avoid these potentially high-risk configurations, specific restrictions to the implementation of the proposedTS changes were identified.For cases where all inoperable snubbers are associated with only one train (or s ubsystem) ofthe impacted systems (i.e., when LCO 3.0.8a applies), it was assumed in the analysis that therewill be unaffected redundant trains (or subsystems) available to mitigate the seismically-initiatedLOOP accident sequences. This assumption implies that there will be at least one successpath available when LCO 3.0.8a applies. Therefore, potentially high-risk configurations can beavoided by ensuring that such a success path exists when LCO 3.0.8a applies. Based on a review of the accident sequences that contribute to the risk increase associated with LCO3.0.8a, as modeled by the simplified bounding analysis (i.e., accident sequences initiated by a seismically-induced LOOP event with concurrent loss of all safety system trains supported bythe out-of-service snubbers), the following restrictions were identified to prevent potentiallyhigh-risk configurations: !For PWR plants, at least one AFW train (including a minimum set of supportingequipment required for its successful operation) not associated with the inoperable snubber(s), must be available when LCO 3.0.8a is used

!For BWR plants, one of the following two means of heat removal must be availablewhen LCO 3.0.8a is used:

SAt least one high pressure makeup path (e.g., using high pressure coolantinjection (HPCI) or reactor core isolation cooling (RCIC) or equivalent) and heat removal capability (e.g., suppression pool cooling), including a minimum set ofsupporting equipment required for success, not associated with the inoperable snubber(s), or SAt least one low pressure makeup path (e.g., low pressure coolant injection(LPCI) or core spray (CS)) and heat removal capability (e.g., suppression poolcooling or shutdown cooling), including a minimum set of supporting equipment required for success, not associated with the inoperable snubber(s).For cases where one or more of the inoperable snubbers are associated with multiple trains (orsubsystems) of the same safety system (i.e., when LCO 3.0.8b applies), it was assumed in thebounding analysis that all safety systems are unavailable to mitigate the accident, except forWest Coast plants. Credit for using F&B to provide core cooling is taken for plants having F&B capability (e.g., Diablo Canyon) when a snubber impacting more than one train of the AFWsystem is inoperable. Credit for one AFW train to provide core cooling is taken for West CoastPWR plants with no F&B capability (e.g., San Onofre) because it has been determined thatthere is no single snubber whose non-functionality would disable more than one train of the AFW in a seismic event of magnitude up to the plant's SSE. Based on a review of the accident sequences that contribute to the risk increase associated with LCO 3.0.8b (as modeled by thesimplified bounding analysis) and defense-in-depth considerations, the following restrictions were identified to prevent potentially high-risk configurations:

!LCO 3.0.8b cannot be used at West Coast PWR plants with no F&B capability w hen asnubber whose non-functionality would disable more than one train of AFW in a seismic event of magnitude up to the plant's SSE is inoperable (it should be noted, however, that based on information provided by the industry, there is no plant that falls in thiscategory),!When LCO 3.0.8b is used at PWR plants, at least one AFW train (including a minimumset of supporting equipment required for its successful operation) not associated with the inoperable snubber(s), or some alternative means of core cooling (e.g., F&B, firewater system or "aggressive secondary cooldown" using the steam generators) mustbe available, and

!When LCO 3.0.8b is used at BWR plants, it must be verified that at least one successpath exists, using equipment not associated with the inoperable snubber(s), to provide makeup and core cooling needed to mitigate LOOP accident sequences. 3.1.3Configuration Risk ManagementThe third tier of the three-tiered approach recommended in RG 1.177 involves theestablishment of an overall configuration risk management program (CRMP) to ensure that potentially risk-significant configurations resulting from maintenance and other operationalactivities are identified. The objective of the CRMP is to manage configuration-specific risk by appropriate scheduling of plant activities and/or appropriate compensatory measures. This objective is met by licensee programs to comply with the requirements of paragraph (a)(4) of the Maintenance Rule (10 CFR 50.65) to assess and manage risk resulting from maintenance activities, and by the TS requiring risk assessments and management using (a)(4) processes if no maintenance is in progress. These programs can support licensee decision making regarding the appropriate actions to manage risk whenever a risk-informed TS is entered.

Since the 10 CFR 50.65(a)(4) guidance, the revised (May 2000) Section 11 of NUMARC 93-01, does not currently address seismic risk, licensees adopting this change must ensure that the proposed LCO 3.0.8 is considered with respect to other plant maintenance activities and integrated into the existing 10 CFR 50.65(a)(4) process whether the process is invoked by a TSor (a)(4) itself.3.2Summary and Conclusions The option to relocate the snubbers to a licensee controlled document, as part of theconversion to Improved STS, has resulted in non-uniform and inconsistent treatment of snubbers. Some potential undesirable consequences of this inconsistent treatment of snubbers are:!Performance of testing during crowded windows when the supported system isinoperable with the potential to reduce the snubber testing to a minimum since the relocated snubber requirements are controlled by the licensee,!Performance of testing during crowded windows when the supported system isinoperable with the potential to increase the unavailability of safety systems, or

!Performance of testing and maintenance on snubbers affecting multiple trains of thesame supported system during the 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> allotted before entering MODE 3 underLCO 3.0.3.To remove the inconsistency among plants in the treatment of snubbers, licensees areproposing a risk-informed TS change which introduces a delay time before entering the actions for the supported equipment when one or more snubbers are found inoperable or removed for testing. Such a delay time will provide needed flexibility in the performance of maintenance andtesting during power operation and at the same time will enhance overall plant safety by(1) avoiding unnecessary unscheduled plant shutdowns, thus, minimizing plant transition and realignment risks; (2) avoiding reduced snubber testing, thus, increasing the availability ofsnubbers to perform their supporting function; (3) performing most of the required testing and maintenance during the delay time when the suppor ted system is available to mitigate mostchallenges, thus, avoiding increases in safety system unavailability; and (4) providing explicitrisk-informed guidance in areas in which that guidance currently does not exist, such as the treatment of snubbers impacting more than one redundant train of a support ed system. The risk impact of the proposed TS changes was assessed following the three-tiered approachrecommended in RG 1.177. A simplified bounding risk assessment was performed to justify the proposed TS changes. This bounding assessment assumes that the risk increase associated with the proposed addition of LCO 3.0.8 to the TS is associated with accident sequences initiated by a seismically-induced LOOP event with concurrent loss of all safety system trainssupported by the out-of-service snubbers. In the case of snubbers associated with more than one train, it is assumed that all affected trains of the supported system are failed. Thisassumption was introduced to allow the performance of a simple bounding risk assessment approach with application to all plants and was selected due to the lack of detailed plant-specific seismic risk assessments for most plants and the lack of fragility data for piping whenone or more supporting snubbers are inoperable. The impact from the addition of the proposed LCO 3.0.8 to the TS on defense-in-depth was also evaluated in conjunction with the risk assessment results.Based on this integrated evaluation, the staff concludes that the proposed addition of LCO 3.0.8to the TS would lead to insignificant risk increases, if any. Indeed, this conclusion is true without taking any credit for the removal of potential undesirable consequences associated with the current inconsistent treatment of snubbers, such as the effects of avoiding a potentialreduction in the snubber testing frequency and increased safety system unavailability.Consistent with the staff's approval and inherent in the implementation of TSTF-372, licenseesinterested in implementing LCO 3.0.8 must, as applicable, operate in accordance with the following stipulations:1.Appropriate plant procedures and administrative controls will be used to implement thefollowing Tier 2 Restrictions.(a) At least one AFW train (including a minimum set of supporting equipmentrequired for its successful operation) not associated with the inoperable snubber(s), must be available when LCO 3.0.8a is used at PWR plants.(b)At least one AFW train (including a minimum set of supporting equipmentrequired for its successful operation) not associated with the inoperable snubber(s), or some alternative means of core cooling (e.g., F&B, fire water system or "aggressive secondary cooldown" using the steam generators) mustbe available when LCO 3.0.8b is used at PWR plants.(c)LCO 3.0.8b cannot be used by West Coast PWR plants with no F&B capabilitywhen a snubber, whose non-functionality would disable more than one train of AFW in a seismic event of magnitude up to the plant's SSE, is inoperable.(d)BWR plants must verify, every time the provisions of LCO 3.0.8 are used, that atleast one success path, involving equipment not associated with the inoperablesnubber(s), exists to provide makeup and core cooling needed to mitigate LOOP accident sequences.(e)Every time the provisions of LCO 3.0.8 are used licensees will be required toconfirm that at least one train (or subsystem) of systems supported by theinoperable snubbers would remain capable of performing their required safety or support functions for postulated design loads other than seismic loads. LCO 3.0.8 does not apply to non-seismic snubbers. In addition, a record of thedesign function of the inoperable snubber (i.e., seismic vs. non-seismic),

implementation of any applicable Tier 2 restrictions, and the associated plant configuration shall be available on a recoverable basis for staff inspection.2. Should licensees implement the provisions of LCO 3.0.8 for snubbers, which includedelay times to enter the actions for the supported equipment when one or more snubbers are out of service for maintenance or testing, it must be done in accordance with an overall CRMP to ensure that potentially risk-significant configurations resulting from maintenance and other operational activities are identified and avoided, as discussed in the proposed TS Bases. This objective is met by licensee programs to comply with the requirements of paragraph (a)(4) of the Maintenance Rule, 10 CFR 50.65, to assess and manage risk resulting from maintenance activities or when this process is invoked by LCO 3.0.8 or other TS. These programs can support licensee decisionmaking regarding the appropriate actions to manage risk whenever a risk-informed TS is entered. Since the 10 CFR 50.65(a)(4) guidance, the revised (May 2000) Section 11 of NUMARC 93-01, does not currently address seismic risk, licensees adopting this change must ensure that the proposed LCO 3.0.8 is consideredin conjunction with other plant maintenance activities and integrated into the existing 10 CFR 50.65(a)(4) process. In the absence of a detailed seismic PRA, a bounding risk assessment, such as utilized in this Safety Evaluation, shall be followed.In its submittal, the licensee said that it reviewed the staff's evaluation, as well as theinformation provided to support TSTF-372, and has concluded that the justifications presented in the TSTF proposal and staff safety evaluation are applicable to the Diablo Canyon Power Plant, Unit Nos. 1 and 2, and justifies these amendments. Therefore, incorporating the changes into Diablo Canyon Power Plant, Unit Nos. 1 and 2's TSs is acceptable.

4.0STATE CONSULTATION

In accordance with the Commission's regulations, the California State official was notified of theproposed issuance of the amendments. The State official had no comments.

5.0ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of a facilitycomponent located within the restricted area as defined in 10 CFR Part 20. The NRC staff hasdetermined that the amendments involve no significant increase in the amounts, and nosignificant change in the types, of any effluents that may be released offsite, and that there isno significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration and there has been no public comment on such finding (70 FR 75495; December 20, 2005). Accordingly, the amendments meet the eligibility criteriafor categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) thereis reasonable assurance that the health and safety of the public will not be endangered byoperation in the proposed manner, (2) such activities will be conducted in compliance with theCommission's regulations, and (3) the issuance of the amendments will not be inimical to thecommon defense and security or to the health and safety of the public.

7.0REFERENCES

1.TSTF-372, Revision 4, "Addition of LCO 3.0.8, Inoperability of Snubbers," April 23, 2004.2.Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," USNRC, August 1998.3.Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-InformedDecisionmaking: Technical Specifications," USNRC, August 1998.4.Budnitz, R. J. et. al., "An Approach to the Quantification of Seismic Margins in NuclearPower Plants," NUREG/CR-4334, Lawrence Livermore National Laboratory, July 1985.5.Advanced Light Water Reactor Utility Requirements Document, Volume 2, ALWREvolutionary Plant, PRA Key Assumptions and Groundrules, Electric Power Research Institute, August 1990.6.Bier V. M. et. al., "Development and Application of a Comprehensive Framework forAssessing Alternative Approaches to Snubber Reduction," International Topical Conference on Probabilistic Safety Assessment and Risk Management PSA '87, SwissFederal Institute of Technology, Zurich, August 30-September 4, 1987.7.NUREG-1488, "Revised Livermore Seismic Hazard Estimates for Sixty-Nine NuclearPower Plant Sites East of the Rocky Mountains," April 1994.8.NEI, Revised Section 11 of Revision 2 of NUMARC 93-01, May 2000.

9.Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activitiesat Nuclear Power Plants," May 2000. Principal Contributors:T. TjaderW. ReckleyDate: March 7, 2006 February 2006Diablo Canyon Power Plant, Units 1 and 2 cc:NRC Resident InspectorDiablo Canyon Power Plant c/o U.S. Nuclear Regulatory Commission

P.O. Box 369 Avila Beach, CA 93424Sierra Club San Lucia ChapterATTN: Andrew Christie P.O. Box 15755 San Luis Obispo, CA 93406Ms. Nancy CulverSan Luis Obispo Mothers for Peace

P.O. Box 164 Pismo Beach, CA 93448ChairmanSan Luis Obispo County Board of Supervisors Room 370 County Government Center San Luis Obispo, CA 93408Mr. Truman BurnsMr. Robert Kinosian California Public Utilities Commission 505 Van Ness, Room 4102 San Francisco, CA 94102Diablo Canyon Independent Safety Committee ATTN: Robert R. Wellington, Esq. Legal Counsel 857 Cass Street, Suite D Monterey, CA 93940Regional Administrator, Region IVU.S. Nuclear Regulatory Commission Harris Tower & Pavillion 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064Richard F. Locke, Esq.Pacific Gas & Electric Company P.O. Box 7442 San Francisco, CA 94120City EditorThe Tribune 3825 South Higuera Street

P.O. Box 112 San Luis Obispo, CA 93406-0112Mr. Ed Bailey, ChiefRadiologic Health Branch State Department of Health Services P.O. Box 997414 (MS 7610)

Sacramento, CA 95899-7414Mr. James D. Boyd, CommissionerCalifornia Energy Commission 1516 Ninth Street (MS 31)

Sacramento, CA 95814Mr. James R. Becker, Vice PresidentDiablo Canyon Operations and Station Director Diablo Canyon Power Plant

P.O. Box 3 Avila Beach, CA 93424Jennifer TangField Representative United States Senator Barbara Boxer 1700 Montgomery Street, Suite 240 San Francisco, CA 94111