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Diablo Canyon, Units 1 and 2 - License Amendment Request 15-03 Application of Alternative Source Term. Part 1 of 8
	ML15176A527
Person / Time
Site:	Diablo Canyon
Issue date:	06/17/2015
From:	Welsch J M; Pacific Gas & Electric Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	DCL-15-069
	Download: ML15176A527 (119)
	v • d • e

Pacific Gas andElectric CompanyJames M. Welsch Diablo Canyon Power PlantSite Vice President Mail Code 104/6P. 0. Box 56Avila Beach, CA 93424June 17, 2015 805.545.3242Internal: 691.3242Fax: 895.545.4884PG&E Letter DCL-15-069U.S. Nuclear Regulatory Commission 10 CFR 50.90ATTN: Document Control DeskWashington, D.C. 20555-0001Diablo Canyon Units 1 and 2Docket No. 50-275, OL-DPR-80Docket No. 50-323, OL-DPR-82License Amendment Request 15-03Application of Alternative Source TermPursuant to 10 CFR 50.90, Pacific Gas and Electric Company (PG&E) herebyrequests approval of the enclosed proposed amendment to Facility OperatingLicense Nos. DPR-80 and DPR-82 for Units 1 and 2, respectively, of the DiabloCanyon Power Plant (DCPP). The enclosed license amendment request (LAR)proposes to revise the DCPP Units I and 2 licensing bases to adopt the alternativesource term (AST) as allowed by 10 CFR 50.67. The following TechnicalSpecification (TS) changes are required for AST implementation: TS 1.1 for thedefinition of Dose Equivalent 1-131; TS 3.4.16 to revise the noble gas activity limit;TS 3.6.3 to require the 48-inch containment purge supply and exhaust valves to besealed closed during MODES 1, 2, 3, and 4; TS 5.5.9 to revise the accident inducedleakage performance criterion; TS 5.5.11 to change the allowable methyl iodidepenetration testing criteria for the auxiliary building ventilation system charcoal filter:and TS 5.5.19 to replace "whole body or its equivalent to any part of the body," with"TEDE," which is the dose criteria specified in 10 CFR 50.67.The Enclosure provides a description of the proposed changes and supportingjustification including the determination of no significant hazards and environmentalconsiderations. Attachments to the Enclosure are described within.The changes in this LAR are not required to address an immediate safety concern.PG&E requests approval of this LAR no later than June 30, 2016. PG&E requeststhe license amendments be made effective upon NRC issuance, to be implementedwithin 365 days from the date of issuance.This communication contains five new regulatory commitments (as defined in NEI99-04) to be implemented following NRC approval of this LAR. The commitmentsare contained in Attachment 7 of the Enclosure.A member of the STARS (Strategic Teaming and Resource Sharing) AltianceCatlaway

Diablo Canyon

Palo Verde

Wolf Creek A w l Document Control Desk PG&E Letter DCL-15-069June 17, 2015Page 2In accordance with site administrative procedures and the Quality AssuranceProgram, the proposed amendment has been reviewed by the Plant Staff ReviewCommittee.Pursuant to 10 CFR 50.91, PG&E is sending a copy of this proposed amendment tothe California Department of Public Health.If you have any questions or require additional information, please contact HosseinHamzehee at 805-545-4720.I state under penalty of perjury that the foregoing is true and correct.Executed on June 17, 2015.Sincerely,J M. WelschSite Vice Presidentkjse/4328/50705089Enclosurecc: Diablo Distributioncc/enc: Marc L. Dapas, NRC Region IVThomas R. Hipschman, NRC Senior Resident InspectorSiva P. Lingham, NRR Project ManagerGonzalo L. Perez, Branch Chief, California Dept of Public HealthA member of the STARS (Strategic Teaming and Resource Sharing) AllianceCaLLaway

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Wolf Creek EnclosurePG&E Letter DCL-15-069Evaluation of the Proposed ChangeLicense Amendment Request 15-03Application of Alternative Source Term EnclosurePG&E Letter DCL-15-069Table of ContentsI. SUMMARY DESCRIPTION ............................................................................ I2. DETAILED DESCRIPTION ........................................................................ 22.1 Proposed Changes to Current Licensing Basis .................. I ..................... 22.2 Proposed Technical Specification Changes ........................................... 112.3 Technical Specification Bases Changes ................................................ 142.4 Plant Changes .... ................................................................... ...... 142.5 Procedure Changes ................................................................................ 152.6 Updated Final Safety Analysis Report Changes ..................................... 162.7 Presentation of Current Licensing Basis and Alternative Source TermAnalysis Inputs ....................................................................................... 163. TECHNICAL EVALUATION .................. .................................................... 174. REGULATORY EVALUATION .................................................................. 194.1 Applicable Regulatory Requirements/Criteria ........................................ 194.2 Precedent .............................................................................................. 254.3 No Significant Hazards Consideration .................................................... 264.4 Conclusions ........................................................................................... 325. ENVIRONMENTAL CONSIDERATION .................................................... 326. REFERENCES ......................................................................................... 34ii EnclosurePG&E Letter DCL-15-069List of Attachments1. Proposed Technical Specification Changes (MARKUP)2. Proposed Technical Specification Changes (RETYPED)3. Technical Specification Bases Markup (For Information Only)4. Diablo Canyon Power Plant Technical Assessment Prepared by Stone &Webster, Inc. (A CB&I Company) -Implementation of Alternative SourceTerms Summary of Dose Analyses and Results5. Regulatory Guide 1.183 Conformance Tables6. Diablo Canyon Power Plant Comparison to NRC Regulatory. InformationSummary (RIS) 2006-04 Experience with Implementation of Alternative.Source Terms7. Diablo Canyon Power Plant List of Regulatory Commitments for AlternativeSource Term Implementation8. Diablo Canyon Power Plant Updated Final Safety Analysis Report Markup(For Information Only)iii EnclosurePG&E Letter DCL-15-069EVALUATIONSUMMARY DESCRIPTIONThis license amendment request (LAR) would amend Operating Licenses DPR-80 and DPR-82 for Units 1 and 2 of the Diablo Canyon Power Plant (DCPP),respectively.Pacific Gas & Electric (PG&E) requests Nuclear Regulatory Commission (NRC)review and approval of a proposed revision to the licensing basis of DCPP Units1 and 2 that supports a full scope application of an alternative source term (AST)methodology as allowed by 10 CFR 50.67 (Reference 1).An application for the selective use of AST for the fuel handling accident (FHA) inthe fuel handling building (FHB) was reviewed and approved by the NRC in itsSafety Evaluation Report (SER) for License Amendment Nos. 163 and 165(Reference 2). However, the FHA in the FHB has also been reanalyzed with thisapplication and is included in this submittal to be consistent with revised inputs,as described in Attachment 4. Approval of this AST application will supersedethe FHA in the FHB dose analysis and results, as discussed in the SER forLicense Amendment Nos. 163 and 165.The AST methodology as established in Regulatory Guide (RG) 1.183,"Alternative Radiological Source Terms for Evaluating Design Basis Accidentsat Nuclear Power Reactors," July 2000 (Reference 3) is used to calculate theoffsite and Control Room radiological consequence for DCPP Units 1 and 2.Attachment 4 contains a summary of the analyses and results for the followingevents that are expected to produce the most limiting dose consequences.Conformance to RG 1.183 is provided in Attachment 5.* Loss of Coolant Accident (LOCA)* FHA in the Containment* FHAinthe FHB* Locked Rotor Accident (LRA)* Control Rod Ejection Accident (CREA)" Main Steam Line Break (MSLB)0 Steam Generator Tube Rupture (SGTR)" Loss-of-Load (LOL) EventIn addition to adopting AST for design basis accidents and the associated totaleffective dose equivalent (TEDE) dose criteria for offsite and Control Roomdoses, DCPP is adopting the TEDE dose criteria of 10 CFR 50.67 for theTechnical Support Center (TSC), as allowed by RG 1.183.1 EnclosurePG&E Letter DCL-15-069Full implementation of AST for DCPP Units 1 and 2 does not include revising thesource terms used for environmental qualification (EQ) of safety relatedequipment or NUREG-0737 responses associated with shielding and vital areaaccess. Section C.6 of RG 1.183 (Reference 3) discusses the position onperformance of required EQ analyses with respect to AST and TID-14844(Reference 4) source term assumptions. NUREG-0933, "Resolution of GenericSafety Issues," Section 3.0, Item 187 (Reference 5) resolved the issues relatedto the effect of increased cesium releases on EQ doses. The NRC staffconcluded that there is no clear basis for a requirement to modify the designbasis for EQ to adopt AST since there would be no discernible risk reductionassociated with adopting AST for EQ. In addition, post-accident vital area accessdose rates are not expected to be significantly impacted by the AST during thefirst 30 days following a LOCA based on an AST benchmarking study. The NRCSER for Fort Calhoun Station's implementation of AST (Reference 6) referencedthe SECY-98-154 (Reference 7) study as the source for the conclusion that theresults of analyses based on TID-14844 would be more limiting for a period up toone to four months after which time the AST results would be more limiting.Therefore, this LAR does not propose to modify the EQ design basis nor theshielding and vital area access dose rates to adopt AST.The proposed amendment revises Technical Specification (TS) definitions,requirements, and terminology related to the use of an AST associated withoffsite, Control Room, and TSC accident dose consequences. A markup ofaffected TS pages is included in Attachment 1 to this Enclosure.Regulatory Issue Summary (RIS) 2006-04, "Experience with Implementation ofAlternative Source Terms," dated March 7, 2006, (Reference 8) outlined twelveissues that the NRC staff has encountered during its review of AST submittals.Attachment 6 provides discussion on how DCPP has addressed the twelveissues identified in RIS 2006-04.2. DETAILED DESCRIPTION2.1 Proposed Changes to Current Licensing BasisThe dose consequence analyses addressed in this application have been revisedto incorporate the guidance provided in RG 1.183 (Reference 3) and to resolvethe findings of the Licensing Basis Verification Project (LBVP) which wasvoluntarily initiated by DCPP, as presented to NRC (ADAMS AccessionNo. ML15029A094). The LBVP findings have been addressed in promptoperability assessments, which include several temporary compensatorymeasures. The revised dose analyses address the LBVP findings, as well asimplements the following licensing basis changes.1. Implement RG 1.183, July 2000 (Reference 3), as the licensing basis forDCPP, as outlined in this LAR. RG 1.183 will replace DCPPs commitment2 EnclosurePG&E Letter DCL-:15-069to RG 1.195, "Methods and Assumptions for Evaluating RadiologicalConsequences of Design Basis Accidents at Light-Water Nuclear PowerReactors."2. Remove the "expected" accident dose consequence assessments that arein DCPP Updated Final Safety Analysis Report (UFSAR) Section 15.5.The original DCPP licensing application included two evaluations for eachaccident. The first evaluation, called the expected case, used estimates ofactual values expected to occur if the accident took place. The resultingdoses were close to the doses expected from an accident of this type.The second evaluation, the Design Basis Accident (DBA), used thecustomary conservative assumptions. The calculated doses for the DBA,while not a realistic estimate of expected doses, provided the basis fordetermining the design adequacy of the plant safety systems.Current NRC guidance related to expectations for a safety analysis reportfor a nuclear power plant, as provided in NUREG-0800 (Reference 9),does not require the inclusion of dose consequences from "expected"accident scenarios. Since these "expected" accident scenario evaluationsare not relevant for determining design adequacy of plant safety systems,PG&E is proposing to remove this information from its licensing basis.UFSAR markups showing the elimination of the "expected" cases areprovided for information only in Attachment 8.3. Eliminate the dose contribution of a containment purge via thecontainment hydrogen purge system following a LOCA for purposes ofhydrogen control.The NRC revised 10 CFR 50.44 (Reference 10) to acknowledge that theamount of combustible gas generated for the design basis LOCA was nota risk significant threat to containment integrity. Thus, with the exceptionof demonstrating the capability of ensuring a mixed atmosphere withincontainment, the requirements for hydrogen control pertaining to thedesign basis LOCA were eliminated. In the SER for License AmendmentNos. 168 and 169 to DCPP (Reference 11), the NRC confirmed theelimination of hydrogen release concerns associated with a design-basisLOCA, and the associated requirements that necessitated the need forhydrogen recombiners and backup hydrogen vent and purge systems.To ensure consistency with the current licensing basis, PG&E is proposingto eliminate the dose contribution due to the containment purge pathwaycurrently included in the LOCA dose consequence analysis in support ofhydrogen control.3 EnclosurePG&E Letter DCL-15-0694. Replace dose guidelines of 10 CFR 100.11 for whole body and thyroiddose with the TEDE acceptance criterion of 10 CFR 50.67(b)(2) andSection 4.4, Table 6 of RG 1.183 (Reference 3).As required by 10 CFR 50.67(b)(2) (Reference 1), Attachment 4 of thisLAR contains an evaluation of the consequences of applicable DBAspreviously analyzed in the DCPP UFSAR (Reference 12). The TEDEdose criteria will be applied to offsite locations, as well as the ControlRoom and the TSC.5. Replace General Design Criteria (GDC 19), 1971, with GDC 19, 1999 fordose only.10 CFR 50.67(b)(2)(iii) states that adequate radiation protection isprovided to permit access to and occupancy of the Control Room underaccident conditions without personnel receiving radiation exposures inexcess of 0.05 Sv.(5 rem) TEDE for the duration of the accident. As partof implementation of 10 CFR 50.67, the NRC also amended GDC 19,1971 (Reference 13) to reflect the 5 rem TEDE aspects. The DCPPlicense basis, from the original Final Safety Analysis Report throughAmendment 85, includes GDC 19, 1971, for Control Room dose only(Reference 12). DCPP will conform to GDC 19, 1999, for dose only, forControl Room dose limits of 5 rem TEDE upon implementation of AST.This change to GDC 19, 1999, for dose only, is consistent with theadoption of AST.6. Update the dose acceptance criterion for the TSC to 5 rem TEDE.The dose acceptance criterion for the TSC is based on Section 8.2.1,Item f of NUREG-0737, Supplement 1 (Reference 14), which states thatany person working in the TSC would not exceed 5 rem whole body, or itsequivalent to any part of the body, for the duration of the accident. Thedose acceptance criterion is modified to 5 rem TEDE to be consistent with10 CFR 50.67(b)(2) and GDC 19, 1999, for dose only. In accordance withDCPP current licensing basis, the TSC is only evaluated for the LOCA.Therefore, upon implementation of AST, the DCPP licensing basis forNUREG-0737, Item ll.B.2 and Ill.A.1.2 will be the AST acceptance criteriaspecified in 10 CFR 50.67(b) and GDC 19, 1999, for dose only. Thischange to GDC 19, 1999, is consistent with the adoption of AST7. Update computer codes that support the AST dose consequenceanalyses.Attachment 4, Section 3 provides the computer codes utilized in support ofthis application. These computer codes will now be included in DCPP's4 EnclosurePG&E Letter DCL-15-069licensing basis in the manner in which they are utilized in the doseconsequence analyses as outlined in Attachment 4. The codes used insupport of the AST application are recommended by RG 1.183 or havebeen used in priorAST applications that have been approved by NRC.8. Use inhalation dose conversion factors from Environmental ProtectionAgency (EPA) Federal Guidance Report (FGR) No. 11, 1988, "LimitingValues of Radionuclide Intake and Air Concentration and Dose ConversionFactors for Inhalation, Submersion, and Ingestion." (Reference 15).FGR No. 11 has been part of the licensing basis for DCPP in that TSDefinition 1.1 for Dose Equivalent 1-131 (DEI) allowed Table 2.1 of FGRNo. 11 to be used for determining DEI. FGR No. 11 is used in AST doseconsequence analyses for inhalation dose conversion factors, asrecommended by RG 1.183. See Section 2.2 for proposed TS changes.9. Update offsite atmospheric dispersion factors (x/Q) using recent 5-yearmeteorological data (2007 to 2011) and RG 1.145, Revision 1,"Atmospheric Dispersion Models for Potential Accident ConsequenceAssessments at Nuclear Power Plants," (Reference 16) methodology.The methodology outlined in RG 1.145, Revision 1, is used for calculatingground level releases to determine the short-term X/Q values for theExclusion Area Boundary (EAB) and the Low Population Zone (LPZ) fordesign basis radiological analyses. All releases are conservatively treatedas ground level releases, therefore Regulatory Positions C.1.3.2, C.2.1.2,and C.2.2.2 associated with elevated or stack releases are not applicableto DCPP.RG 1.111, "Methods for Estimating Atmospheric Transport and Dispersionof Gaseous Effluents in Routine Releases from Light Water CooledReactors," Regulatory Position C.1.c (Reference 17) is used to determinethe annual average XJQ values, which are used as input to develop theaccident X/Q values at the LPZ using RG 1.145 methodology.Attachment 4, Section 5 presents the development of the X/Q values.10. Update X/Q factors for on-site locations such as the Control Room and theTSC using recent 5-year meteorological data (2007 to 2011) and,"Atmospheric Relative Concentrations in Building Wakes," (ARCON96)methodology (Reference 18).Regulatory Guide 1.194, "Atmospheric Relative Concentrations for ControlRoom Radiological Habitability Assessments at Nuclear Power Plants,"dated June 2003 (Reference 19), Regulatory Position C.1 through C.3,and the adjustment factor for vertically orientated energetic releases from5 EnclosurePG&E Letter DCL-15-069steam relief valves and atmospheric dump valves (ADVs) allowed byRegulatory Position C.6 are used to determine short-term on-site X/Qvalues in support of design basis radiological habitability assessments.PG&E is requesting an exception to RG 1.194, Regulatory Position C.3.4,as part of this LAR. Two specific Control Room receptors are within10 meters of the release (9.4 meters and 7.8 meters). The x/Q values forthese two cases were developed to establish boundingX/Q values.However, the ,/Q values for these two locations were not the boundingvalues and therefore were not used in the dose consequence analyses.Attachment 4, Section 5.2 provides further discussion for the requestedexception.'Credit for DCPP's dual intake design for the Control Room pressurizationair intakes is taken per RG 1.194, Regulatory Position C.3.3.2.3. Inaddition, credit is taken for a reduction factor of 5 applied to the X/Q valuesfor energetic releases from the DCPP Main Steam Safety Valves (MSSVs)and the 10 percent ADVs, per RG 1.194, Regulatory Position C.6 due tothe velocity and orientation of the release. This credit is used for theMSLB, SGTR, LRA, CREA, and LOL events.Credit is also taken for the close proximity of the MSSVs/10 percent ADVsto the normal operating Control Room intake of the affected unit and thehigh vertical velocity of the steam discharge from the MSSVs/1 0 percentADVs resulting in the post-accident plume from the MSSVs/1 0 percentADVs not contaminating the normal operation Control Room intake of theaffected unit. This credit is used for the MSLB, SGTR, LRA, CREA, andLOL events.Attachment 4, Section 5 presents the development of the X/Q values.11. Update Control Room ventilation system (CRVS) parameters resultingfrom the installation of new back-draft damper in the Control Roomemergency filter recirculation lines.Back-draft dampers were installed to prevent reverse unfiltered flow intothe Control Room. The updated CRVS parameters have been included inthe new Control Room transport model discussed in Attachment 4, Section7.1.12. Update Control Room unfiltered inleakage.The updated Control Room unfiltered inleakage values, including back-draft damper leakage, have been included in the new Control Roomtransport model discussed in Attachment 4, Section 7.1. The updated6

'EnclosurePG&E Letter DCL-15-069Control Room unfiltered inleakage value bounds the unfiltered inleakagedetermined by the 2012 Control Room Tracer Gas Test (Reference. 20).13. Use containment spray in the recirculation mode following a LOCA forfission product cleanup.DCPP is designed and licensed to operate using containment spray in therecirculation mode. In accordance with the current licensing basis, and asdocumented in the NRC SER related to License Amendment No. 139 toFacility Operating License Nos. DPR-80 and DPR-82 (Reference 21),containment spray is not required per analyses to be actuated duringrecirculation, but may be actuated in accordance with the emergencyoperating procedures (EOPs) or at the discretion of the TSC.To address the delayed core damage sequence of a post-LOCAASTscenario and support fission product removal from the containmentatmosphere, credit is taken in the LOCA dose analysis for usingcontainment spray in the recirculation mode for dose mitigation. Thislicensing basis change to require containment spray during recirculation.for fission product cleanup does not affect the conclusions of the SERrelated to License Amendment No. 139 with respect to the other functionsof containment spray. The containment spray system will be operated in amanner consistent with the licensing basis established by the SER relatedto License Amendment No. 139 and no changes in operation are beingproposed, other than requiring its operation within, 12 minutes followingterminating injection spray, instead of being optional in accordance withEOPs or at the discretion of the TSC.The LOCA dose analysis also credits a time critical operator action(TCOA). A TCOA is a manual action or series of actions with a specifiedcompletion time limit to meet a plant licensing basis requirement. TheLOCA dose analysis assumes that containment spray is realigned fromthe injection mode to the recirculation mode within 12 minutes ofterminating injection spray to ensure that the duration of spray operation(injection + recirculation) exceeds 6.25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months following the event. TheTCOA will be implemented as part of AST implementation, using theguidelines provided in NRC Information Notice 97-78, "Crediting ofOperator Actions in Place of Automatic Actions and Modifications ofOperatorActions, Including Response Times." The required actions forthe new TCOA have been demonstrated on the simulator, showing thatthe 12 minute time requirement can be achieved with margin.14. Update allowable engineered safety features (ESF) system leakagevalues and associated release points following a LOCA.7 EnclosurePG&E Letter DCL-15-069In accordance with RG 1.183, the LOCA dose analysis assumes a releaseof two times the allowable ESF leakage values. As such, two times theallowable leakage values outlined in Section 2.5 are used to determinedoses following a LOCA, as discussed in Attachment 4, Section 7.2.3.3.The ESF leakage limits are controlled by TS 5.5.2, "Primary CoolantSources Outside Containment."15. Include environmental releases from the refueling water storage tank(RWST) vent due to sump water back-leakage following a LOCA.The LOCA dose analysis for AST includes the environmental releasesfrom the RWST due to sump water back-leakage following a LOCA, inaccordance with RG 1.183. Attachment 4, Section 7.2.3.5 presents theanalysis for this dose contribution.16. Include environmental releases from the Miscellaneous Equipment DrainTank (MEDT).The LOCA dose analysis for AST includes the environmental releasesfrom the MEDT following a LOCA, in accordance with RG 1.183.Attachment 4, Section 7.2.3.6 presents the analysis for this dosecontribution.17. Include environmental releases via the 12-inch containmentvacuum/pressure relief pathway prior to containment isolation following aLOCA.In accordance with RG 1.183, for containments such as DCPP that can beroutinely purged during normal operation, the dose consequence analysismust assume a release to the environment, through the purge pathway,occurs prior to containment isolation. As such, the LOCA dose analysisincludes a dose contribution from-the 12-inch containmentvacuum/pressure relief pathway prior to containment isolation.Attachment 4, Section 7.2.3.1 presents the analysis for this dosecontribution.18. Preclude environmental releases via the 48-inch containment purge andexhaust system pathway prior to containment isolation following a LOCA.In accordance with RG 1.183, for containments such as DCPP that can beroutinely purged during normal operation, the dose consequence analysismust assume a release to the environment, through the purge pathway,occurs prior to containment isolation. This release pathway is notapplicable to the containment purge and exhaust system because DCPPis requesting approval of a TS change for the 48-inch containment purgevalves to be sealed closed in accordance with Standard Review Plan8 EnclosurePG&E Letter DCL-15-069(SRP) Section 6.2.4, Revision 3, Items 11.6 and 11.14 during MODES 1,2,3, and 4. The 48-inch containment purge valves will be sealed closed byremoving motive power to the valve operators. With this proposed TSrevision, NUREG-0737, Item II.E.4.2, Position (6) (Reference 22) will besatisfied. Because the 48-inch containment purge valves will be requiredto be.sealed closed during MODES 1, 2, 3, and 4, DCPP will no longertake credit for a Phase A isolation signal for these valves, as outlined inresponse to NUREG-0737, November 1980 Item II.E.4.2, Position (7). Inaddition, piping classification for the associated containment penetrations(61 and 62) will change from Group A to Group E.See Section 2.2 for proposed TS changes associated with the 48-inchcontainment purge supply and exhaust valves.19. Define the portion of Room 506 of the Control Room which serves as aControl Room foyer between the Control Room Assistants' office and theShift Managers' office as a low occupancy, less frequented area.When determining the direct shine dose to the Control Room fromexternal and contained sources, the analysis presented in Attachment 4,Section 7.2.5.2 takes into consideration the function of Room 506.Room 506 is used as an area where occupancy is deemed to be minimal.Thus, an "occupancy adjustment" factor is utilized for Room 506 todetermine the maximum 30-day integrated dose in the Control Room (i.e.,the total direct shine dose in the Control Room includes the 30-day dosein Room 506 adjusted by occupancy factor).20. Define a minimum decay time prior to fuel movement as 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .As part of this application, DCPP proposes to revise the definition ofrecently irradiated fuel as fuel that has occupied part of a critical reactorcore within the previous 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This definition is used in the doseconsequence analysis of the FHA to determine the release following thepostulated event. Although the source term for the FHA will be slightlylarger with less decay of fuel prior to fuel movement, the doseconsequence analysis results show that the dose criteria are met, asshown in Table 1 and Attachment 4, Section 7.3.21. Credit the .redundant safety related gamma sensitive area radiationmonitors (1-RE 25/26, 2-RE 25/26) to initiate CRVS Mode 4 following aFHA.These monitors are located at the Control Room normal intakes and aredesigned to automatically isolate the normal CRVS intakes and shift toCRVS Mode 4 (pressurized filtered emergency ventilation). Thesemonitors are credited to perform their design function following a FHA in9 EnclosurePG&E Letter DCL-15-069the FHB or containment. See Section 2.4 for a description of a setpointchange for these radiation monitors.22. Credit the following existing administrative controls reflected in plantprocedures. These administrative controls ensure the FHB is maintainedat a negative pressure relative to atmosphere during movement ofirradiated fuel in the spent fuel pool, thus ensuring that the environmentalreleases occur via the Unit vent." The movable wall is in place and secured* No exit door from the FHB is propped open" At least one FHB Ventilation System exhaust fan is runningAttachment 4, Section 7.3 presents the FHA. Credit for the aboveadministrative controls is taken to facilitate that the post-accidentenvironmental release of radioactivity occurs via the plant vent.23. Update reported doses for other UFSAR Chapter 15 events with accidentsource terms to TEDE doses criteria.The DCPP licensing basis includes dose assessments at offsite locationsfor several Condition III and Condition IV events. RG 1.183 does notaddress Condition Ill and Condition IV events; therefore they have notbeen re-analyzed with this application. SRP 15.0.1 (Reference 23) statesthat a complete recalculation of all design basis radiological consequenceanalyses may not be required for an application to be acceptable.However, SRP 15.01 also states that a full AST implementation replacesthe previous accident source term used in all design basis radiologicalanalyses and incorporates the TEDE dose criteria. Therefore, DCPPperformed scoping evaluations, as allowed by RG 1.183, Section C.1.3.3(Reference 3), to demonstrate compliance with regulatory limits at theEAB and LPZ for the DCPP UFSAR Chapter 15 Condition III andCondition IV events. The evaluation compares the accident sequence,predicted fuel damage (if applicable), and resultant dose consequences ofthe DBAs analyzed forAST to those parameters of the Condition III andCondition IV events. These evaluations are presented in Attachment 4,Section 2.1.The tank rupture events presented in UFSAR Chapter 15.5 represent theaccidental release of radioactivity accumulated in tanks resulting fromnormal plant operations and are not affected by accident source termsassociated with AST. Therefore, the tank rupture events are notreanalyzed in support of this LAR and the dose acceptance criteria for thetank rupture events will remain unchanged.10 fEnclosurePG&E Letter DCL-15-069As part of its original licensing basis, DCPP provides an estimatedradiation exposure to the Control Room operator during egress andingress between the Control Room and the site boundary following aLOCA, as presented in UFSAR 15.5.17.10. Although RG 1.183 does notaddress or provide guidance for determining this dose contribution, DCPPis retaining this access dose in its licensing basis. The whole bodygamma dose and thyroid dose reported in the UFSAR are converted toreflect the estimated TEDE dose by using organ weighting factorsprovided in 10 CFR 20.1003. Attachment 4, Section 7.2.6 demonstratesthat the access dose, converted to TEDE dose, is minimal in that it is 1percent of the estimated operator dose due to Control Room occupancyfollowing a LOCA.It is noted that the dose received by the operator during transit outside theControl Room is not a measure of the "habitability" of the Control Room,which is defined by the radiation protection provided to the operator by theControl Room shielding and ventilation system design. Thus, theestimated dose to the operator during routine post-LOCA access to theControl Room is addressed separately from the Control Room occupancydose and is not included with the Control Room occupancy dose for thedemonstration of Control Room habitability.2.2 Proposed Technical Specification ChangesThe following TS changes are proposed to reflect the licensing basis changesoutlined in Section 2.1. Brief descriptions of the associated proposed TSchanges are provided below along with justification for each change. Thespecific wording changes to the TS-are provided in Attachments 1 and 2 to thisenclosure.* TS 1.1,,"Definitions," is revised to change the definition of Dose Equivalent1-131 (DEI).This TS provides a definition for DEI, which currently references Table IIIof TID-14844, AEC, 1962, "Calculation of Distance Factors for Power andTest Reactor Sites;" Table E-7 of Regulatory Guide 1.109, Revision 1,NRC, 1977; International Commission on Radiological ProtectionPublication 30, 1979, Supplement to Part 1, pages 192-212, Table titled"Committed Dose Equivalent in Target Organs or Tissues per Intake ofUnit Activity;" and Table 2.1 of EPA Federal Guidance Report No. 11,1988, "Limiting Values of Radionuclide Intake and Air Concentration andDose Conversion Factors for Inhalation, Submersion, and Ingestion."This TS change will be revised to only reference the committed thyroiddose equivalent conversion factors from Table 2.1 of FGR No. 1111 EnclosurePG&E Letter DCL-15-069(Reference 15). The change is consistent with the recommendations ofRG 1.183. Under AST, the doses are reported as TEDE dose.NRC RIS 2006-04 (Reference 8) states:"Although different references are available for dose conversionfactors, the TS definition should be based on the same doseconversion factors that are used in the determination of the reactorcoolant dose equivalent iodine curie content for the main steamlinebreak and steam generator tube rupture accident analyses."Dose conversion factors from Table 2.1 of FGR No. 11 (Reference 15) areused by DCPP to determine the reactor coolant dose equivalent iodinecurie content for the MSLB and SGTR accident analyses. Thus, the TSchange is consistent with item 10, Definition of Dose Equivalent 1-131, ofNRC RIS 2006-04.The TS change will remain consistent with the approved IndustryImproved Standard Technical Specification Traveler, TSTF-490(Reference 24).TS 3.4.16, "RCS Specific Activity," is revised to change the Noble gasactivity limit from less than or equal to 600 pCi/gm Dose Equivalent XE-133 (DEX) to less than or equal to 270 pCi/gm DEX.DEX limit is equivalent to approximately 0.5 percent fuel defects. Thecurrent limit of 600 pCi/gm DEX corresponds to approximately 1 percentfuel defects, which is the DCPP design basis value for system andshielding design. The limit is being reduced by DCPP to control the noblegas activity in the coolant to levels below the design basis values.* TS 3.6.3, "Containment Isolation Valves," is revised to modify Note 1 ofLimiting Condition of Operation (LCO) 3.6.3 concerning the 48-inchcontainment purge supply and exhaust valves. The TS currently allowsthe operation of these valves for less than 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months per year duringoperating MODES 1, 2, 3, and 4. The proposed revision eliminates theadministratively controlled operation of the 48-inch containment purgevalves during MODES 1, 2, 3, and 4. The proposed revision will nowrequire the 48-inch containment purge supply and exhaust valves toremain sealed closed during MODES 1, 2, 3, and 4. This change willeliminate a potential dose contribution due'to an open containment purgepathway at the initiation of a LOCA.The TS revision includes a new surveillance requirement for verifying the48-inch purge valves are sealed closed, removes the 48-inch purge valves12 EnclosurePG&E Letter DCL-1 5-069from Surveillance Requirements (SRs) 3.6.3.2 and modifies the frequencyfor SR 3.6.3.7.The proposed revision is consistent with NUREG-1431, Volume 1,Standard Technical Specifications, Westinghouse Plants (Reference 25).The 48-inch containment purge valves are to be sealed closed inaccordance with SRP Section 6.2.4, Revision 3, Item 11.6 and 11.14 duringMODES 1, 2, 3, and 4. The 48-inch containment purge valves will besealed closed by removing motive power to the valve operators. Withthis proposed TS revision, NUREG-0737, Item I1.E.4.2, Position (6)(Reference 22) will be satisfied. Because the 48-inch containment purgevalves will be required to be sealed closed during MODES 1, 2, 3, and 4,DCPP will no longer take credit for a Phase A isolation signal for thesevalves, as outlined in response to NUREG-0737, November 1980 ItemI1.E.4.2, Position (7).* TS 5.5.9, "Steam Generator (SG) Tube Inspection Program," is revised tolower the accident induced leakage performance criterion from 1 gallonper minute (gpm) per steam generator to 0.75 gpm total for all four steamgenerators. The accident induced leakage performance criterion shall notexceed the leakage rate assumed in the dose consequence analysis.A primary-to-secondary SG tube leakage of 0.75 gpm at standardtemperature and pressure is used in the dose consequence analysis forthe LRA, CREA, MSLB, SGTR, and LOL events. DCPP TS 3.4.13d limitsprimary-to-secondary SG tube leakage to 150 gallons per day (gpd) perSG for a total of 600 gpd for all 4 SGs. The revised testing criterion for theprimary-to-secondary leakage is more restrictive than the current testingcriteria and represents the leakage rate assumed in the doseconsequence analyses presented in Attachment 4. The 0.75 gpm from all4 SGs (or a total of 1080 gpd) conservatively bounds the TS 3.4.13.d limitin that the analyzed leakage rate accounts for higher leakage than theTS 3.4.13.d limit, and thus a higher analyzed release of radioactivity.* TS 5.5.11, "Ventilation Filter Testing Program (VFTP)," is revised tochange the allowable methyl iodide penetration testing criteria for theauxiliary building ventilation system (ABVS) charcoal filter from 15 percentto 5 percent.The allowable methyl iodide penetration is used to determine charcoalfilter efficiency for removing iodine from atmospheric releases. Credit forfiltration of the release of a residual heat removal (RHR) system pumpseal passive failure by the ABVS is taken in determining the dose13 EnclosurePG&E Letter DCL-15-069consequences to the public at the EAB and LPZ, and to personnel in theControl Room and TSC.* TS 5.5.19, "Control Room Envelope Habitability Program," is revised toreplace "whole body or its equivalent to any part of the body" with "TEDE,"which is the dose criteria specified in 10 CFR 50.67 (Reference 1).2.3 Technical Specification Bases ChangesThe TS Bases will be revised to reflect the licensing basis changes outlined inSection 2.1 and the TS changes identified in Section 2.2. A markup of the TSBases changes is provided for information only in Attachment 3 to this Enclosure.These TS Bases changes will be implemented in accordance with TS 5.5.14,"Technical Specification (TS) Bases Control Program," upon NRC approval of thisLAR.2.4 Plant ChangesThe following plant design modifications will be performed as part of ASTimplementation. These plant modifications support the AST analyses provided inAttachment 4.* Install shielding material, equivalent to that provided by the Control Roomouter walls, at the external concrete west wall of the Control Room briefingroom." Install a high efficiency particulate air filter (HEPA) in the TSC normalventilation system intake." Update setpoints for the redundant safety related gamma sensitive area,radiation monitors (1-RE 25/26, 2-RE 25/26). These monitors are locatedat the Control Room normal intakes and are designed to automaticallyisolate the normal CRVS intakes and shift to CRVS Mode 4 (pressurizedfiltered emergency ventilation). These monitors are relied upon to performtheir design function following a FHA in the FHB or Containment.Setpoints for these monitors are contained in the Offsite Dose CalculationManual, which is controlled by TS 5.5.1, requirements." Reclassify a portion of the 40-inch Containment Penetration Area(GE/GW) Ventilation line from PG&E Design Class II to PG&E DesignClass I and upgrade the damper actuators, pressure switches, and thedamper solenoid valves to PG&E Design Class I. See Attachment 4,Section 5.2 for further details." Reclassify a portion of the 2-inch gaseous radwaste system line whichconnects to the Plant Vent as PG&E Design Class I. This line is currently14 EnclosurePG&E Letter DCL-15-069classified as PG&E Design Class II. See Attachment 4, Section 5.2 forfurther details.2.5 Procedure ChangesAs part of AST implementation, the following procedural updates will include:* Update Equipment Control Guideline (ECG) 42.1, "Refueling Operations -Decay Time," to lower the, restriction on fuel movement from 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months to72 hours post-shutdown." Update ECG 42.5, "Refueling Operations -Water Level -Reactor Vessel(Control Rods)" to reflect the FHA AST analysis assumptions." Update-ECG 23.3, "Containment Ventilation System," to reflect changes toTS 3.6.3." Review and update, as necessary, the EOPs and operator trainingprocedures to ensure that the requisite steps to select the leastcontaminated CRVS pressurization intake are in use throughout the event.This review is to be performed as verification, since the EOPs currentlyinclude steps to select the least contaminated CRVS intake." Update Surveillance Test Procedure M-57, "Control Room VentilationSystem (CRVS) Tracer Gas Test," to include the new Control Roominleakage test acceptance criteria and the range of CRVS ventilation flowsdeemed acceptable by the AST dose consequence analyses.* Update the TSC administrative procedures to ensure that:a. The nominal normal operation TSC ventilation air intake flowrate is500 cubic feet per minute (cfm).b. Following a LOCA, the TSC will be manually placed in Mode 4operation such that filtered pressurization and recirculation can becredited within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of accident initiation.c. The nominal post-LOCA TSC ventilation filtered pressurization andrecirculation flowrates are 500 cfm, respectively." Review EOPs to verify valve alignment information to manually initiatecontainment spray in the recirculation mode. Update EOPs to includedirection to perform the realignment action within 12 minutes oftermination of injection spray to ensure that the duration of spray operation(injection plus recirculation) exceeds 6.25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months following the event. Anassociated TCOA will be implemented." Update ESF system leak testing procedures that are controlled byTS 5.5.2, "Primary Coolant Sources Outside Containment," to establishadministrative acceptance criteria to ensure:15 EnclosurePG&E Letter DCL-15-069a. The total as-tested leakage from ESF systems that recirculatesump fluid outside containment is less than 126 cubic centimetersper minute (cc/min), with the following breakdown:i. In areas covered by the ABVS, the as-tested leakage is lessthan 120 cc/min,ii. In the containment penetration area, the as-tested leakage isless than 6 cc/min.b. The total as-tested back leakage into the RWST from thecontainment recirculation sump is less than 1 gpm.c. The total as-tested flow hard piped to the MEDT is less than thefollowing values:i. Leakage from systems carrying non-radioactive fluids is lessthan 484 cc/min.ii. Leakage from ESF systems that recirculated sump fluids isless than 950 cc/min.Review and update the Emergency Plan to reflect AST, as necessary.There will be no change to the Emergency Planning Zone.2.6 Updated Final Safety Analysis Report ChangesThe UFSAR will be revised to reflect AST dose consequence analyses and theproposed licensing basis changes outlined in Section 2.1. UFSAR changes areprovided in Attachment 8, for information only.2.7 Presentation of Current Licensing Basis and Alternative Source Term AnalysisInputsThis section provides a summary of changes from current design and licensingbasis analysis input values to revised AST inputs for each analysis.The NRC's traditional methods for calculating the radiological consequences ofDBAs are described in a series of regulatory guides and SRP chapters. Thatguidance was developed to be consistent with TID-14844 source term and thewhole body and thyroid dose guidelines stated in 10 CFR 100.11. Many of thoseanalysis assumptions and methods are inconsistent with the AST and with theTEDE criteria provided in 10 CFR 50.67. In addition, many of DCPP's analysespre-date SRP guidance.RG 1.183 provides assumptions and methods that are acceptable to the NRCstaff for performing design basis radiological analyses using an AST. As stated inRG'1.183, the RG 1.183 guidance supersedes corresponding radiologicalanalysis assumptions provided in other regulatory guides and SRP chapters16 EnclosurePG&E Letter DCL-15-069when used in conjunction with an approved AST and TEDE criteria provided in 10CFR 50.67 (Reference 1).DCPP used the guidance provided by RG 1.183 for analysis assumptions andmethods for design basis radiological analyses. Conformance to RG 1.183guidance is presented in Attachment 5. A summary of design inputs,assumptions, and methodology used in the AST analyses is provided inChapter 7 of Attachment 4. Appendix B to Attachment 4 provides a comparisonbetween the design input values used in the current licensing basis doseconsequence analyses and the-values used to support the AST analysessupporting this LAR.As noted above and in Attachment 4, many of DCPP's analyses pre-date SRPguidance. Specifically, the DCPP's current licensing basis for the LRA, CREA,and LOL events are DCPP-specific with pre-SRP assumptions and only addressoffsite dose consequences. Updated analyses performed for AST now includeControl Room doses.3. TECHNICAL EVALUATIONThe DCPP Units 1 and 2 current licensing basis for the radiologicalconsequences analyses of accidents is based on source term methodology andassumptions derived from TID-14844 (Reference 4). An application for theselective use of AST for the FHA in the FHB was reviewed and approved by theNRC in its SER for License Amendment Nos. 163 and 165 (Reference 2). Thedesign basis accidents are discussed in Chapter 15 of the DCPP UFSAR(Reference 12). The current DCPP dose consequences of design basis events,other than the FHA in the FHB,, are based on acceptance criteria stated in10 CFR Part 100 and 10 CFR 50, AppendixA, GDC 19, 1971. The currentlicensing basis for the radiological consequences of the FHA in the FHB is10 CFR 50.67, "Accident Source Term."The AST and methodology described in NUREG-1465, "Accident Source Termsfor Light-Water Nuclear Power Plants," (Reference 25) and in RG 1.183(Reference 3), provide regulatory guidance for the implementation of the AST.Revision of a plant licensing basis from the TID-14844 (Reference 4) source termto an AST involves the preparation of dose consequence analyses.Demonstration that theresults satisfy the regulatory acceptance criteria and NRCapproval of the requested change establishes the acceptability of the use of theAST for DCPP.DCPP has performed radiological consequence analyses of the DBAsdocumented in Chapter 15 of the DCPP UFSAR that potentially result in the mostsignificant Control Room and offsite exposures. These analyses were performedto support full scope implementation of AST. The AST analyses have beenperformed in accordance with the guidance in RG 1.183 and SRP Section 15.0.117 EnclosurePG&E Letter DCL-15-069(Reference 23). Acceptance criteria consistent with those required by10 CFR 50.67 and RG 1.183, Table 6, were used to replace the current designbasis source term acceptance criteria. This represents a full implementation ofAST in which the RG 1.183 source term will become the licensing basis forDCPP DBAs.The technical justification for full implementation of the AST methodology, asdefined in RG 1.183, of the DCPP DBAanalyses is provided in Attachment 4.The following DBAs are addressed:" LOCA" FHA in the Containment* FHA in the FHB (reanalysis)* LRA* CREA* MSLB* SGTR* LOL EventConformance to RG 1.183 is documented in Attachment 5. The AST as definedin RG 1.183 has been incorporated into the DCPP site boundary and ControlRoom dose re-analyses discussed in Attachment 4. The estimated DCPP doseconsequences for all design basis events addressed in RG 1.183, meet theacceptance criteria specified in 10 CFR 50.67 and RG 1.183, as shown inTable 1 for offsite locations and Control Room personnel.In addition, the TSC dose is re-analyzed in Attachment 4. In accordance withcurrent licensing basis, the 30-day integrated dose to an operator in the TSC dueto immersion, inhalation, and direct shine is evaluated for the LOCA. Theresultant post-LOCA dose is estimated to be 4.1 rem TEDE, which is within theacceptance criteria of 5 rem TEDE.As stated in Section 2.4, some changes to the facility are required to implementAST for Control Room and TSC dose consequences. These changes includeadditional shielding for the Control Room, setpoint change for radiation monitors1-RE 25/26 and 2-RE 25/26, and component upgrades (damper actuators,pressure switches, and damper solenoid valves) for a portion of the 40-inchContainment Penetration Area (GE/GW) Ventilation system. A HEPA filter will beinstalled at the normal intake of the TSC ventilation system.The TSC is designed to meet NUREG-0696, Functional Criteria for EmergencyResponse Facilities, (Reference 27), which states that the TSC ventilationsystem need not be seismic Category 1 qualified, redundant, instrumented in theControl Room, or automatically activated to fulfill its role. Thus, the addition ofthe HEPA filter is not considered a safety-related component.18 EnclosurePG&E Letter DCL-15-069TheEngineering Change Package supporting AST Implementation will ensurethat any Design Class I equipment located adjacent to the ABVS and CRVSfilters and the new TSC filter are qualified to any potential increase in theestimated total integrated radiation dose resulting from the additional post-LOCAradiological release pathways and higher X/Q values addressed in thisapplication.Based on the preceding paragraphs, the methodology and dose consequenceanalyses presented in Attachment 4 do not rely on any newly installed safety-related systems, structure, or components. The containment spray system willnow be credited during sump water recirculation following a LOCA for dosemitigation, but DCPP is already licensed for recirculation containment sprayoperation. Therefore, there are no additions to the EQ list or RG 1.97instrumentation list.No changes have been made in the system responses to accidents. Therefore,there are no additional or new emergency diesel generator (EDG) loads and thetiming of the EDG loads did not change as a result of AST.4. REGULATORY EVALUATION4.1 Applicable Regulatory Requirements/CriteriaTitle 10 Code of Federal Regulations Section 50.36, "Technical specifications"10 CFR 50.36:(c) Technical specifications will include items in the following categories:2) Limiting conditions for operation.(i) Limiting conditions for operation are the lowest functionalcapability or performance levels of equipment required forsafe operation of the facility. When a limiting condition foroperation. of a nuclear reactor is not met, the licensee shallshut down the reactor or follow any remedial actionpermitted by the technical specifications until the conditioncan be met.LCO 3.4.16, "RCS Specific Activity," provides the limiting condition for operationof the Reactor Coolant System (RCS) DEI and DEX. The limit established forDEI is not being revised by this LAR; however, the definition of DEI in TS 1.1,"Definition," is being revised to reference Table 2.1 of FGR No. 11 (Reference 15)as the only acceptable dose conversion factors for determining DEI. Thus, thedefinition of DEI will reference the same dose conversion factors used to19 EnclosurePG&E Letter DCL-15-069determine the reactor coolant dose equivalent iodine curie content for the MSLBand SGTR analysis, as requested by NRC RIS 2006-04 (Reference 8).(ii) A technical specification limiting condition for operation of anuclear reactor must be established for each item meetingone or more of the following criteria:(A) Criterion 1. Installed instrumentation that is used todetect, and indicate in the Control Room, a significantabnormal degradation of the reactor coolant pressureboundary.(B) Criterion 2. A process variable, design feature, oroperating restriction that is an initial condition of a designbasis accident or transient analysis that either assumes thefailure of or presents a challenge to the integrity of a fissionproduct barrier.LCO 3.6.3, "Containment Isolation Valves," Note 1 is revised so that thecontainment purge supply and exhaust flow paths are sealed closed duringoperating MODES 1, 2, 3, and 4. This change is in support of the LOCA doseanalysis assumptions that the containment purge supply and exhaust paths areclosed and therefore, not a release path for radionuclides present in thecontainment following an accident prior to containment isolation. Therefore, thisproposed TS change will eliminate a potential radiological release pathway.(C) Criterion 3. A structure, system, or component that is partof the primary success path and which functions or actuatesto mitigate a design basis accident or transient that eitherassumes the failure of or integrity of a fission product barrier.LCO 3.6.3 is part of the success path which functions to mitigate a LOCA. Therevision to LCO 3.6.3, Note 1 support the LOCA dose analysis assumption thatthe 48-inch valves in the containment purge supply and exhaust paths are sealedclosed and therefore, are not a release path for radionuclides present in thecontainment following a LOCA prior to containment isolation.3) Surveillance requirements. Surveillance requirements arerequirements relating to test, calibration, or inspection to assurethat the necessary quality of systems and components ismaintained, that facility operation will be within safety limits, andthat the limiting conditions for operation will be met.SR 3.4.16.1 is revised to reduce the specific activity of DEX. The revised ASTanalyses that base the released radioactive source terms on RCS specificactivity uses a limit of less than or equal to 270pCi/gm DEX. The limit has been20 EnclosurePG&E Letter DCL-15-069reduced by DCPP to control the noble gas activity in the coolant to levels belowthe design basis values.SR 3.6.3.1, 3.6.3.2, and 3.6.3.7 are being revised to support the revision toLCO 3.6.3 to ensure that the LCO will be met.5) Administrative Controls. Administrative controls are the provisionsrelating to organization and management, procedures,recordkeeping, review and audit, and reporting necessary to assureoperation of the facility in a safe mannerTS 5.5.9, "Steam Generator (SG) Tube Inspection Program," is revised to changethe accident induced leakage performance criteria. The primary-to-secondaryaccident induced leakage rate for any design basis accident, other than a SGTR,shall not exceed the leakage rate assumed in the accident analysis in terms oftotal leakage rate for all SGs. Except during a SGTR, leakage is not to exceed0.75 gpm total for all four SGs. The revised testing criterion for the primary-to-secondary leakage is more restrictive than the current testing criteria andrepresents the leakage rate assumed in the AST dose consequence analyses.TS 5.5.11, "Ventilation Filter Testing Program (VFTP)," is revised to change theallowable methyl iodine penetration testing criteria for the ABVS charcoal filterfrom 15 percent to 5 percent. The allowable methyl iodide penetration is used todetermine charcoal filter efficiency for removing iodine from atmosphericreleases. This proposed revision will support the LOCA dose analysisassumptions with respect to the releases from an RHR pump seal passivefailure, and has been demonstrated to be acceptable.TS 5.5.19, "Control Room Envelope Habitability Program," is revised to replace"whole body or its equivalent to any part of the body" to "TEDE," to be consistentwith the dose criteria specified in 10 CFR 50.67 for AST.In summary, the changes proposed in this LAR in Section 2 support the ASTanalysis assumptions and have been demonstrated to be acceptable. The ASTanalysis results meet the dose criteria specified in 10 CFR 50.67 and Table 6 ofRG 1.183, therefore the requirements of 10 CFR 50.36 continue to be met.Title 10 Code of Federal Regulations Section 50.67, "Alternate Source Term"On December 23, 1999, the NRC published 10 CFR 50.67, "Accident SourceTerm," in the Federal Register. This regulation provides a mechanism forlicensed power reactors to replace the current accident source term used in theDBA analyses with an AST. The direction provided in 10 CFR 50.67 is thatlicensees who seek to revise their current accident source term in design basisradiological consequence analyses shall apply for a LAR under 10 CFR 50.90.Thus, this LAR meets 10 CFR 50.67.21 EnclosurePG&E Letter DCL-15-069General Design CriteriaThe construction of DCPP Units 1 and 2 was significantly complete prior toissuance of 10 CFR 50, Appendix A GDC. DCPP was designed and constructedto comply with Atomic Energy Commission GDC as proposed on July 10, 1967(AEC GDC), except as noted and described in the DCPP UFSAR Chapter 3.Criterion 19, 1971 -Control Room, describes provisions. for a Control Room thatprovides adequate radiation protection to permit access and occupancy underaccident conditions. The dose criterion of GDC 19 was modified to 5 rem TEDEin 1999 to be consistent with 10 CFR 50.67. The results from the dose analysesusing AST source terms and methodologies show that the predicted doseconsequence results are within the allowable regulatory limits of 10 CFR 50.67and GDC 19, 1999. DCPP will conform to GDC 19, 1999, for dose only, forControl Room dose limits of 5 rem TEDE upon implementation of AST. Thus,with the changes proposed in this LAR, DCPP will continue to meet therequirements of 10 CFR 50, Appendix A, GDC Criterion 19.Criterion 52, 1967- Containment Heat Removal Systems (Category A),describes two functions of the containment spray system. One function is tofacilitate heat removal from the containment following an accident. The secondfunction is to remove radioactive iodine isotopes from the containmentatmosphere should these fission products be released in the event of anaccident. The AST LOCA dose analysis assumes that the containment spraysystem now operates during the mode of operation that recirculates containmentsump water for dose mitigation. This change in containment spray systemoperation assumptions in the LOCA dose consequence analysis does not changethe function of the containment spray system or the actual operation ofcontainment spray system. DCPP is not crediting the containment spray systemin the recirculation mode for heat removal and thus not changing the licensingcommitment to Criterion 52, 1967. Therefore Criterion 52,1967, continues to bemet and the plant will continue to provide the basis for safe plant operation.Criterion 54, 1971 -Piping Systems Penetrating Containment, describesrequirements for isolation, including leak detection, and periodic testing. No newcontainment penetrations or lines penetrating the containment are beingproposed with this LAR. Changes to TS 3.6.3 concerning the containment purgesupply and exhaust paths isolation, including leak detection surveillances,includes an enhanced requirement to ensure that these lines remain sealedclosed during MODES 1, 2, 3, and 4 operation. Therefore Criterion 54, 1971,continues to be met and the piping systems penetrating containment will continueto provide the basis for safe plant operation.Criterion 56, 1971- Primary Containment Isolation, describes the provisions forproviding isolating lines that penetrate the containment. No new containment22 EnclosurePG&E Letter DCL-15-069penetrations or lines penetrating the containment are being proposed with thisLAR. Changes to TS 3.6.3 concerning the containment purge supply andexhaust paths isolation, includes an enhanced requirement to ensure that theselines remain sealed closed during MODES 1, 2, 3, and 4 operation. ThereforeCriterion 56, 1971, continues to be met.Criterion 58, 1967- Inspection of Containment Pressure-Reducing Systems(Category A), describes the design provision requirements to facilitate periodicphysical inspection of components of the containment pressure-reducingsystems. UFSAR 3.1.8.22 provides a discussion on how DCPP meetsCriterion 58, 1967, with a brief discussion of the containment pressure-reducingsystems. Containment spray is a containment pressure reducing system and thedescription states that during the recirculation phase, containment sprayoperation is not required. While this statement remains correct for thecontainment pressure-reducing function of containment spray, due to the timingof fission product releases associated with AST, containment spray will berequired during the recirculation phase for fission product cleanup. This LARdoes not change how DCPP meets Criterion 58, 1967; therefore, the criterioncontinues to be met.Criterion 59, 1967- Testing of Containment Pressure-Reducing SystemsComponents (Category A), discusses how the active components of thecontainment pressure-reducing systems are to be tested periodically foroperability and required functional performance. UFSAR 3.1.8.23 provides adiscussion on how DCPP meets Criterion 59, 1967. Containment spray is acontainment pressure reducing system; however, no changes are being made tothe system. The AST LOCA dose analysis now credits the operation of thesystem during recirculation for dose mitigation. Therefore, Criterion 59, 1967,continues to be met.Criterion 60, 1967-Testing of Containment Spray Systems (Category A),discusses the capability to test the delivery capability of the containment spraysystem at a position as close to the spray nozzles as practical. UFSAR 3.1.8.24provides a discussion on how DCPP meets Criterion 60, 1967. No changes arebeing made to the containment spray system. The only change is that the ASTLOCA dose analysis now credits the operation of the system during recirculationfor dose mitigation. Therefore, Criterion 60, 1967, continues to be met.Criterion 62, 1967- Inspection of Air Cleanup Systems (CategoryA), discussesthe physical inspection of containment air cleanup systems, such as ducts, filters,fans, and dampers. UFSAR 3.1.8.26 provides a discussion on how DCPP meetsCriterion 62, 1967. The containment spray system, using sodium hydroxide,serves as the air cleanup system. No changes are being made to thecontainment spray system. The only change is that the AST LOCA dose analysis23 Enclosure.PG&E Letter DCL-1 5-069now credits the operation of the system during recirculation for dose mitigation.Therefore, Criterion 62, 1967, continues to be met.Criterion 63, 1967-Testing of Air Cleanup Systems Components (Category A),discusses the provisions for testing containment air cleanup systems, such asducts, filters, fans, and dampers. UFSAR 3.1.8.27 provides a discussion on howDCPP meets Criterion 63, 1967. The containment spray system, using sodiumhydroxide, serves as the air cleanup system. No changes are being made to thecontainment spray system. The only change is that the AST LOCA dose analysisnow credits the operation of the system during recirculation for dose mitigation.Therefore, Criterion 63, 1967, continues to be met.Criterion 64, 1967-Testing of Air Cleanup Systems (Category A), discusses theprovisions for testing containment air cleanup systems. UFSAR 3.1.8.28provides a discussion on how DCPP meets Criterion 64, 1967. The containmentspray system serves as the air cleanup system. No changes are being made tothe containment spray system. The only change is that the AST LOCA doseanalysis now credits the operation of the system during recirculation for dosemitigation. Therefore, Criterion 64, 1967, continues to be met.RG 1.183, "Alternative Radiological Source Terms for Evaluating Design BasisAccidents at Nuclear Power Reactors," July 2000.The AST methodology used to perform the dose consequence analyses forDCPP is consistent with the guidance of RG 1.183 (Reference 3).Documentation of conformance to RG 1.183 is presented in Attachment 5, withcross-references to specific sections of Attachment 4, where more detail is,provided.RG 1.194, "Atmospheric Relative Concentrations for Control Room RadiologicalHabitability Assessments at Nuclear Power Plants," dated June 2003.RG 1.194, dated June 2003 (Reference 19), Regulatory Position C.1 throughC.3, and the adjustment factor for vertically orientated energetic releases fromsteam relief valves and ADVs allowed by Regulatory Position C.6 are used todetermine short-term onsite X/Q values in support of design basis radiologicalhabitability assessments. Credit is taken for DCPP's dual intake design for theControl Room pressurization air intakes per Regulatory Position C.3.3.2.3.PG&E takes an exception to Regulatory Position C.3.4, for two specific ControlRoom receptors (9.4 meters for Unit 1 containment building to Unit 1 ControlRoom normal intake and 7.8 meters for Unit 2 containment building to Unit 2Control Room normal intake). Use of ARCON96 methodology for these tworelease point-to-receptor distances is considered acceptable since thedominating factors in the calculation are building cross-sectional area and plumemeander and not the normal atmospheric dispersion coefficients. Note that the24 EnclosurePG&E Letter DCL-15-069X/Q values for these two release-receptor cases were developed to establish thebounding X/Q values. However, the y/Q values for these two cases were notbounding, and therefore not used in the dose consequence analyses. Seesection 2.1 and Attachment 4, Section 5.2 for further-detail. -RG 1.145, Revision 1, "Atmospheric Dispersion Models for Potential AccidentConsequence Assessments at Nuclear Power Plants," dated February 1983.The methodology outlined in RG 1.145, Revision 1, is used for calculating groundlevel releases to determine the short-term X/Q values for the EAB and LPZ fordesign basis radiological analyses. All releases are conservatively treated asground level releases, therefore Regulatory Positions C.1.3.2, C.2.1.2, andC.2.2.2 associated with elevated or stack releases are not applicable to DCPP.10 CFR 50.3410 CFR 50.34(b) specifies content requirements for the UFSAR includingevaluations required to show that accident dose criteria are met. Attachment 8contains UFSAR changes (for information only) to support AST implementation.Upon approval of this LAR, UFSAR changes will be made to fulfill theserequirements.4.2 PrecedentThe NRC has previously approved implementation of the AST methodology at anumber of nuclear power plants. In a LAR dated April 26, 2004, PSEG NuclearLLC, proposed to adopt the AST methodology for Salem Units 1 and 2(Reference 28, ADAMS Accession No. ML041280067). The DCPP LAR is similarto the PSEG submittal in that PSEG also proposed to credit recirculation spraysfollowing the LOCA for long term containment iodine removal. PSEG alsoadjusted the Control Room assumed in-leakage by replacing it with values basedupon their Tracer Gas Test. The NRC reviewed and approved theAST LAR forPSEG in a SER dated February 17, 2006 (Reference 29, ADAMS AccessionNo. ML060040322).In a LAR dated June 5, 2002, FirstEnergy Nuclear Operating Company proposedto adopt the AST methodology for Beaver Valley Power Station Units 1 and 2(Reference 30, ADAMS Accession No. ML021620298). The Beaver Valleyamendment used CBI S&W Proprietary computer codes, listed in Section 3 ofAttachment 4, in similar applications. The NRC reviewed and approved the ASTlicense amendment, including the use of the SBI S&W Proprietary computercodes, in a SER dated September 10, 2003 (Reference 31, ADAMS AccessionNo. ML032530204).In a LAR dated February 7, 2001, Omaha Public Power District proposed toadopt the AST methodology for Fort Calhoun Station Unit 1 (Reference 32,25 EnclosurePG&E Letter DCL-15-069ADAMS Accession No. ML010400079). The Fort Calhoun amendment used CBIS&W Proprietary computer codes, listed in Section 3 of Attachment 4, in similarapplications. The NRC reviewed and approved th'e AST license amendment,including the use of the SBI S&W Proprietary computer codes, in a SER datedDecember 5, 2001 (Reference 6, ADAMS Accession No. ML01 3030027).In a LAR dated September 26, 2002, Millstone Unit 2 proposed alternate non-LOCA gap fractions similar to the non-LOCA gap fractions DCPP proposes inSection 4.3 of Attachment 4, as enhancement to RG 1.183 for higher burnup fuel(Reference 33, ADAMS Accession No. ML023040334). The NRC reviewed andapproved the AST license amendment, including the use of the non-LOCA gapfractions, in a SER dated September 20, 2004 (Reference 34, ADAMS AccessionNo. ML042360671).Similarly, in a LAR dated June 5, 2002, Indian Point Unit 3 proposed alternatenon-LOCA gap fractions similar to the non-LOCA gap fractions DCPPproposes in Section 4.3 of Attachment 4 (Reference 35, ADAMS AccessionNo. ML021840136). The NRC reviewed and approved the AST licenseamendment, including the use of the non-LOCA gap fractions, in a SER datedMarch 17,.2003 (Reference 36, ADAMS Accession No. ML030760135).4.3 No Significant Hazards ConsiderationAs provided by 10 CFR 50.67, Pacific Gas & Electric (PG&E) is implementing theuse of an Alternative Source Term (AST) and the dose calculation methodologydescribed in Regulatory Guide (RG) 1.183 to calculate the accident doses to theControl Room, Technical Support Center (TSC), and offsite receptors followingpostulated design basis events that result in the release of radioactive materialfrom reactor fuel at Diablo Canyon Power Plant (DCPP) Units 1 and 2. The ASTand associated methodology for full implementation of AST define the amount,isotopic composition, physical and chemical characteristics, and timing ofradioactive material releases following postulated events. Transport of thematerial to the Control Room, TSC, and offsite areas is modeled, and theresulting Total Effective Dose Equivalent (TEDE) is determined. Regulatoryacceptance criteria account for the sum of the deep-dose equivalent (for externalexposures) and the committed effective dose equivalent (for internal exposures).In accordance with 10 CFR 50.67(b), licensees wishing to adopt an AST mustapply for a license amendment in accordance with 10 CFR 50.90.In support of the revised analyses applying AST, the following TechnicalSpecification (TS) changes are being made: the definition for Dose EquivalentIodine-1 31 (DEI) is revised to be consistent with AST dose conversion factorusage, the limit for reactor coolant system Dose Equivalent Xenon-133 (DEX)activity is decreased to control the noble gas activity in the coolant to .levelsbelow the design basis values, the requirement for containment penetrations is26 EnclosurePG&E Letter DCL-15-069revised to require the 48-inch containment purge supply and exhaust valves tobe sealed closed during operation MODES 1, 2, 3, and 4 eliminating a potentialdose contribution release path, the accident induced leakage performancecriterion for the steam generator tube inspection program is revised to be morerestrictive, and the testing requirement for.the auxiliary building ventilationsystem charcoal filter is also revised to be more restrictive. Other changes to theTSs involve the adoption of terminology on which AST is based.AST methods have been utilized in the analysis of the limiting design basisaccidents, as follows: loss of coolant accident (LOCA), fuel handing accident(FHA) in the containment and in the fuel handling building, locked rotor accident(LRA), control rod ejection accident (CREA), main steam line break (MSLB), andsteam generator tube rupture (SGTR). AST methods have also been utilized inthe analysis of the limiting Condition II event, the loss of load (LOL) accident.Other changes incorporated in the revised analyses include revising atmosphericdispersion factors (X/Q), reducing the minimum decay time before fuelmovement, adding shielding to the Control Room for additional. protection ofControl Room personnel and adding a high efficiency particulate air (HEPA) filterfor additional protection of TSC personnel. In addition, a portion of the 40-inch'Containment Penetration Area Ventilation line and a portion of the 2-inchgaseous radwaste system line which connect to the Plant Vent are beingreclassified from PG&E Design Class II to PG&E Design Class I. Because ASTmethodologies better represent the physical characteristics and timing of theradionuclide release following a postulated LOCA, containment spray is nowrelied upon during the recirculation of sump water for continued removal of iodineand particulate from the containment atmosphere for spray duration (injectionplus recirculation) greater than 6.25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months . In addition, setpoint changes arebeing made to the Control Room intake radiation monitors to incorporate theeffect of all possible release points from a FHA.PG&E has evaluated whether or not a significant hazards consideration isinvolved with the proposed amendment by focusing on the three standards setforth in 10 CFR 50.92, "Issuance of Amendment," as discussed below: .1. Does the proposed change involve a significant increase in the probabilityor consequences of an accident previously evaluated?Response: No.This license amendment does not physically impact any system, structure,or component (SSC) that is a potential initiator of an accident. Therefore,implementation of AST, the AST assumptions and inputs, the proposed TSchanges, and new X/Q values have no impact on the probability forinitiation of any design basis accident. Once the occurrence of anaccident has been postulated, the new accident source term and ,/Q27 EnclosurePG&E Letter DCL-15-069values are inputs to analyses that evaluate the radiological consequencesof the postulated events.Reactor coolant specific activity, testing criteria of charcoal filters, and theaccident induced primary-to-secondary system leakage performancecriterion are not initiators for any accident previously evaluated. Theproposed change to require the 48-inch containment purge valves to besealed closed during operating MODES 1, 2, 3, and 4 is not an accidentinitiator for any accident previously evaluated. The change in theclassifications of a portion of the 40-inch Containment Penetration AreaVentilation line and a portion of the 2-inch gaseous radwaste system lineis also not an accident initiator for any accident previously evaluated.Thus, the proposed TS changes and AST implementation will not increasethe probability of an accident.The change to the decay time prior to fuel movement is not an accidentinitiator. Decay time is used to determine the source term for the doseconsequence calculation following a potential FHA and has no effect onthe probability of the accident. Likewise, the change to the Control Roomradiation monitors setpoint cannot cause an accident and the operation ofcontainment spray during the recirculation phase is used for mitigation of aLOCA, and thus not an accident initiator.As a result, there are no proposed changes to the parameters orconditions that could contribute to the initiation of an accident previouslyevaluated in Chapter 15 of the Updated Final Safety Analysis Report(UFSAR). As such, the AST cannot affect the probability of an accidentpreviously evaluated.Regarding accident consequences, equipment and components affectedby the proposed changes are mitigative in nature and relied upon once theaccident has been postulated. The license amendment implements a newcalculation methodology for determining accident consequences and doesnot adversely affect any plant component or system that is credited tomitigate fuel damage. Subsequently, no conditions have been createdthat could significantly increase the consequences of any accidentspreviously evaluated.Requiring that the 48-inch cqntainment purge supply and exhaust valvesbe sealed closed during operating MODES 1, 2, 3, and 4 eliminates apotential path for radiological release following events that result inradioactive material releases to the containment, thus reducing potentialconsequences of the event. The steam generator tube inspection testingcriterion for accident induced leakage is being changed, resulting in lowerleakage rates, and thus less potential releases due to primary-to-28 EnclosurePG&E Letter DCL-15-069secondary leakage. The auxiliary building ventilation system allowablemethyl iodide penetration limit is being changed, which results in morestringent testing requirements, and thus higher filter efficiencies forreducing potential releases.Changes to the operation of the containment spray system to requireoperation during the recirculation mode are also mitigative in nature.While the plant design basis has always included the ability to implementcontainment spray during recirculation, this license amendment nowrequires operation of containment spray in the recirculation mode for dosemitigation. DCPP is designed and licensed to operate using containmentspray in the recirculation mode. As such, operation of containment sprayin the recirculation mode has already been analyzed,. evaluated, and iscurrently controlled by Emergency Operating Procedures. Usage ofrecirculation spray reduces the consequence of the postulated event.Likewise, the additional shielding to the Control Room and the addition ofa HEPA filter to the TSC ventilation system reduces the consequences ofthe postulated event to the Control Room and TSC personnel. Loweringthe limit for DEX lowers potential releases. By reclassifying a portion of,the 40-inch Containment Penetration Area Ventilation line and a portion ofthe 2-inch gaseous radwaste system line to PG&E Design Class I, theselines will be seismically qualified, thus assuring that post-LOCA releasepoints are the same as those used for determining X/Q values.The change to the decay time from 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months to 72 /hours prior to fuelmovement is an input to the FHA. Although less decay will result in higherreleased activity, the results of the FHA dose consequence analysisremain within the dose acceptance criteria of the event. Also, theradiation levels'to an operator from a raised fuel assembly may increasedue to a lower decay time, however, any exposure will continue to bemaintained under 10 CFR 20 limits by the'plant Radiation ProtectionProgram.Plant-specific radiological analyses have been performed using the ASTmethodology, assumption and inputs, as well as new x/Q values. Theresults of the dose consequences analyses demonstrate that theregulatory acceptance criteria are met for each analyzed event.Implementing the AST involves no facility equipment, procedure, orprocess changes that could significantly affect the radioactive materialactually released during an event. Subsequently, no conditions have beencreated that could significantly increase the consequences of any of theevents being evaluated.29 EnclosurePG&E Letter DCL-15-069Based on the above discussion, the proposed changes do not involve asignificant increase in the probability or consequences of an accidentpreviously evaluated.2. Does the proposed change create the possibility of a new or differentaccident from any accident previously evaluated?Response: No.This license amendment does not alter or place any SSC in aconfiguration outside its design or analysis limits and does not create anynew accident scenarios.The AST methodology is not an accident initiator, as it is a method used toestimate resulting postulated design basis accident doses. The proposedTS changes reflect the plant configuration that supports implementation ofthe new methodology and supports reduction in dose consequences.DCPP is designed and licensed to operate using containment spray in therecirculation mode. This change will not affect any operational aspect ofthe system or any other system, thus no new modes of operation areintroduced by the proposed change.The function of the radiation monitors has not changed; only the setpointhas changed as a result of an assessment of all potential releasepathways. The continued- operation of containment spray and theradiation monitor setpoint change do not create any new failure modes,alter the nature of events postulated in the UFSAR, nor introduce anyunique precursor mechanism.Requiring the 48-inch containment purge valves to be sealed closedduring operating MODES 1, 2, 3, and 4 does not introduce any newaccident precursor. This change only eliminates a potential release pathfor radionuclides following a LOCA.The proposed TS testing criteria for the auxiliary building ventilationsystem charcoal filters and the proposed performance criteria for steamgenerator tube integrity also cannot create an accident, but results inrequiring more efficient filtration of potentially released. iodine and lessallowable primary-to-secondary leakage. The proposed changes to theDEX activity limit, the TS terminology, and the decay time of the fuelbefore movement are also unrelated to accident initiators.The only physical changes to the plant being made in support of AST isthe addition of Control Room shielding in an area previously modified, theaddition of a HEPA filter at the intake of the TSC normal ventilationsystem, and the upgrade to the damper actuators, pressure switches, and30 EnclpsurePG&E Letter DCL-15-069damper solenoid valves to support reclassifying a portion of theContainment Penetration Area Ventilation line to PG&E Design Class I.Both Control Room shielding and HEPA filtration are mitigative in natureand do not have any impact on plant operation or system responsefollowing an accident. The Control Room modification for adding theshielding will meet applicable loading limits, so the addition of theshielding cannot initiate a failure. Upgrading damper actuators, pressureswitches, and damper solenoid valves involve replacing existingcomponents with components that are PG&E Design Class I. Therefore,the addition of shielding, a HEPA filter, and upgrading components cannotcreate a new or different kind of accident.Since the function of the SSCs has not changed for AST implementation,no new failure modes are created by this proposed change. The ASTchange itself does not have the capability to initiate accidents.Therefore, the proposed change does not create the possibility of a new ordifferent type of accident from any accident previously evaluated.3. Does the proposed change involve a significant reduction in a margin ofsafety?Response: No.Implementing the AST is relevant only to calculated dose consequences ofpotential design basis accidents evaluated in Chapter 15 of the UFSAR.The changes proposed in this license amendment involve the use of anew analysis methodology and related regulatory acceptance criteria.New atmospheric dispersion factors, which are based on site specificmeteorological data, were calculated in accordance with regulatoryguidelines. The proposed TS, TS Bases, and UFSAR changes reflect theplant configuration that will support implementation of the newmethodology and result in operation in accordance with regulatoryguidelines that support the revisions to the radiological analyses of thelimiting design basis accidents. Conservative methodologies, per theguidance of RG 1.183, have been used in performing the accidentanalyses. The radiological consequences of these accidents are all withinthe regulatory acceptance criteria associated with the use of ASTmethodology.The change to the minimum decay time prior to fuel movement results inhigher fission product releases after a FHA. However, the results of theFHA doseconsequence analysis remain within the dose acceptancecriteria of the event.31 EnclosurePG&E Letter DCL-15-069The proposed changes continue to ensure that the dose consequences ofdesign basis accidents at the exclusion area, low population zoneboundaries, in the TSC, and in the Control Room are within thecorresponding acceptance criteria presented in RG 1.183 and10 CFR 50.67. The margin of safety for the radiological consequences ofthese accidents is provided by meeting the applicable regulatory limits,which are set at or below the 10 CFR 50.67 limits. An acceptable marginof safety is inherent in these limits.Therefore, the proposed change does not involve a significant reduction ina margin of safety.Based on the above evaluation, PG&E concludes that the proposed change doesnot involve a significant hazards consideration under the standards set forth in10 CFR 50.92(c), and accordingly, a finding of "no significant hazardsconsideration" is justified.4.4 ConclusionsIn conclusion, based on the considerations discussed above, (1) there isreasonable assurance that the health and safety of the public will not beendangered by operation in the proposed manner, (2) such activities will beconducted in compliance with the Commission's regulations, and (3) theissuance of the amendment will not be inimical to the common defense andsecurity or to the health and safety of the public.5. ENVIRONMENTAL CONSIDERATIONPG&E has evaluated the proposed amendment and has determined that theproposed amendment does not involve (i) a significant hazards consideration, (ii)a significant change in the types or significant increase in the amounts of anyeffluents that may be released offsite, or (iii) a significant increase in individual orcumulative occupational radiation exposure.Based on the evaluation performed under the standards set forth in10 CFR 50.92(c), PG&E concludes that the proposed amendment does notinvolve a significant hazards consideration. AST only involves a change inaccident dose calculation inputs and methodology. Calculated doses meet TEDEcriteria.No aspect of implementing the AST involves facility equipment, procedure, orprocess changes that would increase actual onsite doses if an event were tooccur.The AST does not result in actual or calculated changes in the normal radiationlevels in the facility or in the type or quantity of radioactive materials processed32 EnclosurePG&E Letter DCL-15-069during normal operation. Implementation of.the AST also has no effect on theactual or calculated effluents arising from normal operation.Accordingly, the proposed amendment meets the eligibility criterion forcategorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to10 CFR 51.22(b), no environmental impact statement or environmentalassessment need be prepared in connection with the proposed amendment.33 EnclosurePG&E Letter DCL-15-0696. REFERENCES1. Code of Federal Regulations (CFR), 10 CFR 50.67, "Accident Source Terms."2. NRC Letter "Diablo Canyon Power Plant, Units 1 and 2 -Issuance ofAmendments RE: Control Room, Auxiliary Building, and Fuel Handling BuildingVentilation Systems (TAC Nos. MB8485 and MB8486)," dated February 27,2004.3. Regulatory Guide 1.183, "Alternative Radiological Source Terms for EvaluatingDesign Basis Accidents at Nuclear Power Reactors," July 2000.4. Technical Information Document 14844, "Calculation of Distance Factors forPower and Test Reactor Sites," 1962.5. NUREG-0933, "Resolution of Generic Safety Issues," dated December 2011.6. NRC Letter "Fort Calhoun Station, Unit No. 1 -Issuance of Amendment (TACNo. MB1221)," dated December 5, 2001 (ADAMS Accession No. ML013030027).7. SECY-98-154, "Results of the Revised (NUREG-1465) Source TermRebaselining For Operating Reactors," dated June 30, 1998.8. NRC Regulatory Issue Summary (RIS) 2006-04, "Experience withImplementation of Alternative Source Terms," dated March 7, 2006.9. NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reportsfor Nuclear Power Plants: LWR Edition," dated June 1987.10. Code of Federal Regulations 10 CFR 50.44, "Combustible Gas Control forNuclear Power Reactors."11. NRC Letter "Diablo Canyon Power Plant, Unit 1 (TAC No. MC1678) and UnitNo. 2 (TAC No. MC1679) -Issuance of Amendments RE: Elimination ofRequirements for Hydrogen Recombiners and Hydrogen Monitors," dated May 4,2004.12. Diablo Canyon Power Plant Updated Final Safety Analysis Report, Revision 21.13. Code of Federal Regulations, 10 CFR 50, Appendix A, General Design Criteria19, "Control Room," dated 1971.14. NUREG-0737, Supplement 1, "Clarification of TMI Action Plan Requirements,"dated January 1983.15. Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and AirConcentration and Dose Conversion Factors for lnhalation, Submersion, andIngestion," dated 1988.16. Regulatory Guide 1.145, Revision 1, "Atmospheric Dispersion Models forPotential Accident Consequence Assessments at Nuclear Power Plants," datedFebruary 1983.34 EnclosurePG&E Letter DCL-15-06917. Regulatory Guide 1.111, Revision 1, "Methods for Estimating AtmosphericTransport and Dispersion of Gaseous Effluents in Routine Releases from LightWater Cooled Reactors," dated July 1977.18. Ramsdell, J. V. Jr. and C. A. Simonen, "Atmospheric Relative Concentrations inBuilding Wakes." Prepared by Pacific Northwest Laboratory for the U.S. NuclearRegulatory Commission, PNL-10521, NUREG/CR-6331, Revision 1, May 1997.19. Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control RoomRadiological Habitability Assessments at Nuclear Power Plants," dated June2003.20. NUCON International Inc., "Control Room Habitability Tracer Gas Leak Testing atDiablo Canyon Power Plant," dated December 2012.21. NRC Letter "Diablo Canyon Power Plant, Unit Nos. 1 and 2 -Issuance ofAmendments RE: Containment Spray During the Recirculation Phase of a LOCA(TAC Nos. MA1408 and MA1409)," dated February 9, 2000.22. NUREG-0737, "Clarification of TMI Action Plan Requirements," dated November1980.23. NUREG-0800, Standard Review Plan 15.0.1, Revision 0, "RadiologicalConsequence Analyses using Alternative Source Terms," dated July 2000.24. Improved Standard Technical Specification Traveler, TSTF-490, Revision 0,"Deletion of E Bar Definition and Revision to RCS Specific Activity-Tech Spec,"dated September 13, 2005.25. NUREG-1431, Revision 4, Volume 1, "Standard Technical Specifications,Westinghouse Plants," dated April 2012.26. NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants,"dated February 1995.27. NUREG-0696, "Functional Criteria for Emergency Response Facilities," datedFebruary 1981.28. PSEG Nuclear LLC Letter "Implementation of Alternative Source Term (AST)Request for Changes to Technical Specifications and Updated Final SafetyAnalysis Report Salem Nuclear Generating Station, Units 1 and 2, FacilityOperating Licenses DBR-70 and DPR-75, Docket Nos. 50.272 and 50-311,"dated April 26, 2004 (ADAMS Accession No. ML041280067).29. NRC Letter "Salem Nuclear Generating Station, Unit Nos. I and 2, Issuance ofAmendments RE: Alternate Source Term (TAC Nos. MC3094 and MC3095),"dated February 17, 2006, (ADAMS Accession No. ML060040322).30. FirstEnergy Nuclear Operating Company Letter L-02-069 "Beaver Valley PowerStation, Unit 1 No. I and No. 2 BV-1 Docket No. 50-334, License No. DPR-66,BV-2 Docket No. 50-412, License No. NPF-73, License Amendment RequestNos. 300 and 172," dated June 5, 2002 (ADAMS Accession No. ML021620298).35 EnclosurePG&E Letter DCL-1 5-06931. NRC Letter "Beaver Valley Power Station, Unit Nos. I and 2 -Issuance ofAmendment RE: Selective Implementation of Alternate Source Term and ControlRoom Habitability Technical Specification Changes (TAC Nos. MB5303 andMB5304)," dated September 10, 2003 (ADAMS Accession No. ML032530204).32. Omaha Public Power District Letter LIC-01-0010, "Application for Amendment ofOperating License," dated February 7, 2001 (ADAMS Accession No.ML010400079).33. Dominion Nuclear Connecticut, Inc. Letter, "Millstone Power Station, Unit No. 2,License Basis Document Change Request (LBDCR) 2-18-02, SelectiveImplementation of the Alternative Source Term -Fuel Handling AccidentAnalyses," dated September 26, 2002 (ADAMS Accession No. ML023040334).34. NRC Letter "Millstone Power Station, Unit Nos. 2 -Issuance of Amendment RE:Selective Implementation of Alternate Source Term (TAC No. MB6479)," datedSeptember 20, 2004 (ADAMS Accession No. ML042360671).35. Entergy Nuclear Northeast Letter IPN-02-044, "Indian Point Nuclear GeneratingUnit No. 3, Docket No. 50-286, Proposed Changes to Technical Specifications:Selective Adoption of Alternate Source Term and Incorporation of GenericChanges; TSTF-51, TSTF-68, and TSTF-312," dated June 5, 2002 (ADAMSAccession No. ML021840136).36. NRC Letter "Indian Point Nuclear Generating Unit No. 3 -Issuance ofAmendment RE: Selective Adoption of Alternate Source Term (TACNo. MB5382)," dated March 17, 2003 (ADAMS Accession No. ML030760135).36 EnclosurePG&E Letter DCL-15-069Table 1 -AST Site Boundary and Control Room TEDE (rem)Regulatory Control RegulatoryAccident EAB(1) LPZ(2) Limit Room LimitLoss of Coolant Accident 5.6(') 1 25 3.7 (0.7) (3)5Fuel Handling Accident in Fuel 1.5 0.2 6.3 1.1 5Handling BuildingFuel Handling Accident in 47 5ContainmentLocked Rotor Accident 0.8 0.2 2.5 2.4 5Control Rod Ejection AccidentContainment Release 0.7 0.3 6.3 3.4 5Secondary Release 0.7 0.2 0.5Main Steam Line BreakPre-incident iodine Spike 0.1 <0.1 25 2.0 5Accident-Initiated Iodine Spike 0.7 0.2 2.5 4.1Steam Generator Tube RupturePre-incident iodine Spike 1.3 0.1 25 0.6 5Accident-Initiated Iodine Spike 0.7 <0.1 2.5 0.3Loss of LoadPre-incident iodine Spike <0.1 <0.1 2.5 <0.1 5Accident-Initiated Iodine Spike <0.1 <0.1 2.5 <0.1Notes(1) EAB doses are based on worst 2-hour period following onset of accident. Except as noted, the maximum2-hour dose period for the EAB dose for each of the accidents is the 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ' time period." LOCA: 24-26 hours (based on RHR Pump Seal Failure; see note 4 below for additional information)" LRA: 8.73 to 10.73 hours8.449074e-4 days 0.0203 hours 1.207011e-4 weeks 2.77765e-5 months " MSLB (Accident-Initiated Spike case): 7.6 to 9.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months " LOL (Accident-Initiated Spike case): 8.73 to 10.73 hours8.449074e-4 days 0.0203 hours 1.207011e-4 weeks 2.77765e-5 months .(2) LPZ Doses are based on the duration of the release.(3) The dose presented represents the operator dose due to occupancy. Value shown in parenthesis representsthat portion of the total dose reported that is the contribution of direct shine from contained sources/externalcloud. The dose to the Control Room operator during routine access for the 30 day duration of the accident isdiscussed in Attachment 4, Section 7.2.6 and summarized in the text of Attachment 4, Section 8.0.(4) The maximum 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months EAB dose is based on the assumed RHR pump seal passive failure resulting in a 50 gpmleak of sump water occurring at t=24 hours for 30 minutes. The RHR pump seal passive failure is considered apart of DCPP licensing basis with respect to passive system failure. If this assumed passive failure were notincluded, the maximum 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months dose at the EAB would occur between t=0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months to t=2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months (i.e., during thepost-LOCA ex-vessel release phase) and would be 3.4 rem.37 EnclosureAttachment 1PG&E Letter DCL-15-069ATTACHMENT IProposed Technical Specification Changes(MARKUP)Changes are proposed to the following Technical Specifications:1. Specification 1.1, Definitions, Dose Equivalent 1-1312. Specification 3.4.16, RCS Specific Activity3. Specification 3.6.3, Containment Isolation Valves4. Specification 5.5.9, Steam Generator (SG) Tube Inspection Program5. Specification 5.5.11, Ventilation Filter Testing Program (VFTP)6. Specification 5.5.19, Control Room Envelope Habitability Program Definitions1.11.1 Definitions (continued)DOSE EQUIVALENT 1-131DOSE EQUIVALENT 1-131 shall be that concentration of1-131 (microcuries per gram) that alone would produce thesame dose when inhaled as the combined activities ofiodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actuallypresent. The determination of DOSE EQUIVALENT 1-131shall be performed using thyroid dose conversion faGctorsfrom Table MI of TID 14844, AEC, 1962, "Calculation oGDistance for Power and Test Sites," orTable E 7 of Regulator; Guide 1.109, Revision 1, NRC,1977, Or International Commission on RadiologicalProtection (iCRP) PublicatinG 30), 1979, Supplement to Part1, pages 192 212, Table titled "CmmtedDse Equivalentin Target Organs or Tissues per Intake of Unit Acvity," o-,rthe committed thyroid dose conversion factors from Table2.1 of EPA Federal Guidance Report No. 11, 1988, "LimitingValues of Radionuclide Intake and Air Concentration andDose Conversion Factors for Inhalation, Submersion, andIngestion."DOSE EQUIVALENT XE-1 33 shall be that concentration ofXe-1 33 (microcuries per gram) that alone would produce thesame acute dose to the whole body as the combinedactivities of noble gas nuclides Kr-85m, Kr-87, Kr-88,Xe-133m, Xe-1 33, Xe-1 35m, Xe-1 35, and Xe-1 38 actuallypresent. If a specific noble gas nuclide is not detected, itshould be assumed to be present at the minimum detectableactivity. The determination of DOSE EQUIVALENT XE-1 33shall be performed using effective dose conversion factorsfor air submersion listed in Table 111.1 of EPA FederalGuidance Report No. 12, 1993, "External Exposure toRadionuclides in Air, Water, and Soil."DOSE EQUIVALENT XE-1 33(continued)1.1-3 Unit I -Amendment No. 135, 155,156, 192,Unit 2 -Amendment No. 135, 155,156, 193,DIABLO CANYON -UNITS 1 & 2Rev 9 Page 3 of 24 RCS Specific Activity3.4.16SURVEILLANCE REQUIREMENTSSURVEILLANCE FREQUENCYSR 3.4.16.1 ---------------------NOTE -------------------------- In accordance withOnly required to be performed in MODE 1. the Surveillance------------------------------------------------------------..- Frequency C ontrolVerify reactor coolant DOSE EQUIVALENT XE-133 Programspecific activity < 690-0270.0 pCi/gm.SR 3.4.16.2 --------------------NOTE -----------------Only required to be performed in MODE 1.Verify reactor coolant DOSE EQUIVALENT 1-131 In accordance withspecific activity < 1.0 pCi/gm. the SurveillanceFrequency ControlProgramANDBetween 2 and 6.hours after aTHERMALPOWER change of> 15% RTP withina 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period.DIABLO CANYON -UNITS 1 & 2Rev 10 Page 38 of 403.4-36 Unit I -Amendment No. 135,l-92,200,Unit 2 -Amendment No. 135,!93,201, Containment Isolation Valves3.6.33.6 CONTAINMENT SYSTEMS3.6.3 Containment Isolation ValvesLCO 3.6.3Each containment isolation valve shall be OPERABLE.APPLICABILITY:ACTIONSMODES 1, 2, 3, and 4.------------------------ N O TE S .1. Penetration flow path(s) except no mr.e than two of three flow paths for CoR.taincnt01 m/N 0, in!,- 14/ ,k-, 0+ Ain t-N,/ir/+ ;,il +11 IO InroO jrNrI F O+kO f+ fi/Nfi;;^r.. i il ......... Ln[JLi U L LJi i ........ ILJI48-inch purge valve flow paths, may be unisolated intermittently under administrativecontrols.2. Separate Condition entry is allowed for each penetration flow path.3. Enter applicable Conditions and Required Actions for systems made inoperable bycontainment isolation valves.4. Enter applicable Conditions and Required Actions of LCO 3.6.1, "Containment," whenisolation valve leakage results in exceeding the overall containment leakage rateacceptance criteria.CONDITION REQUIRED ACTION COMPLETION TIMEA. ---------NOTE--------- A.1 Isolate the affected 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months sOnly applicable to penetration flow path bypenetration flow paths with use of at least onetwo containment isolation closed and de-activatedvalves. automatic valve, closed------------------------------------ m anual valve, blindOne or more penetration flange, or check valveflow paths with one with flow through thecontainment isolation valve valve secured.inoperable except for acontainment purge supplyand exhaust valve orpressure/vacuum reliefvalve leakage not withinlimit. (continued)DIABLO CANYON -UNITS I & 2Rev 6 Page 5 of 183.6-5Unit 1 -Amendment No. -435,Unit 2 -Amendment No. 4-35, Containment Isolation Valves3.6.3SURVEILLANCE REQUIREMENTSSURVEILLANCE FREQUENCYSR 3.6.3.1 Net-usedVerify each 48 inch purge valve is In accordance with thesealed closed, except for one purge valve in a Surveillance Frequencypenetration flow path while in Condition D of this Control ProgramLCO.SR 3.6.3.2 Verify each 48 inch containment purge supply In accordance with theand exhaust and 12 inch vacuum/pressure relief Surveillance Frequencyvalve is closed, except when these valves are Control Programopen for pressure control, ALARA or air qualityconsiderations for personnel entry, or forSurveillances that require the valves to be open.SR 3.6.3.3 ----------------------------- NOTE ---------------------------Valves and blind flanges in high radiation areasmay be verified by use of administrative controls.Verify each containment isolation manual valve In accordance with theand blind flange that is located outside Surveillance Frequencycontainment and not locked, sealed or otherwise Control Programsecured and required to be closed duringaccident conditions is closed, except forcontainment isolation valves that are open underadministrative controls.SR 3.6.3.4 ----------------------------- NOTE --------------Valves and blind flanges in high radiation areasmay be verified by use of administrative means.Verify each containment isolation manual valve Prior to entering MODEand blind flange that is located inside 4 from MODE 5 if notcontainment and not locked, sealed or otherwise performed within thesecured and required to be closed during previous 92 daysaccident conditions is closed, except forcontainment isolation valves that are open underadministrative controls.SR 3.6.3.5 Verify the isolation time of each automatic power In accordance with theoperated containment isolation valve is within Inservice Testinglimits. ProgramSR 3.6.3.6 Not used(continued)DIABLO CANYON -UNITS 1 & 2Rev 6 Page 9 of 183.6-9 Unit 1 -Amendment. No. 135,200,Unit 2 -Amendment No. 135,20!,

Containment Isolation Valves3.6.3SURVEILLANCE REQUIREMENTS (continued)SURVEILLANCEFREQUENCYSR 3.6.3.7----------------------------- NOTE --------------This surveillance is not required when thepenetration flow path is isolated by a leak testedblank flange.Perform leakage rate testing for containmentpurge supply and exhaust and vacuum/pressurerelief valves with resilient seals.In accordance with theSurveillance FrequencyControl ProgramANDPFo containment purFgesupply and exhaustvalves only, Within 92days opening theSR 3.6.3.8 Verify each automatic containment isolation valve In accordance with thethat is not locked, sealed or otherwise secured in Surveillance Frequencyposition, actuates to the isolation position on an Control Programactual or simulated actuation signal.SR 3.6.3.9 Not usedSR 3.6.3.10 Verify each 12 inch containment In accordance with thevacuum/pressure relief valve is blocked to restrict Surveillance Frequencythe valve from opening > 50'. Control ProgramSR 3.6.3.11 Not usedDIABLO CANYON -UNITS 1 & 2Rev 6 Page10 of 183.6-10Unit 1 -Amendment No. !35,!75,200 ,Unit 2 -Amendment No. 4,35477,20!,

Programs and Manuals5.55.5 Programs and Manuals (continued)5.5.9 Steam Generator (SG) Tube Inspection ProgramA Steam Generator Program shall be established and implemented to ensure that SGtube integrity is maintained. In addition, the Steam Generator Program shall include thefollowing provisions:a. Provisions for condition monitoring assessments.Condition monitoring assessment means an evaluation of the "as found"condition of the tubing with respect to the performance criteria for structuralintegrity and accident induced leakage. The "as found" condition refers to thecondition of the tubing during an SG inspection outage, as determined from theinservice inspection results or by other means, prior to the plugging of tubes.Condition monitoring assessments shall be conducted during each outage duringwhich the SG tubes are inspected or plugged to confirm that the performancecriteria are being met.b. Performance criteria for SG tube integrity.SG tube integrity shall be maintained by meeting the performance criteria fortube structural integrity, accident induced leakage, and operational LEAKAGE.1. Structural integrity performance criterion: All in-service steam generatortubes shall retain structural integrity over the full range of normal operatingconditions (including startup, operation in the power range, hot standby,and cool down and all anticipated transients included in the designspecification) and design basis accidents. This includes retaining a safetyfactor of 3.0 against burst under normal steady state full power operationprimary-to-secondary pressure differential and a safety factor of 1.4against burst applied to the design basis accident primary-to-secondarypressure differentials. Apart from the above requirements, additionalloading conditions associated with the design basis accidents, orcombination of accidents in accordance with the design and licensingbasis, shall also be evaluated to determine if the associated loadscontribute significantly to burst or collapse. In the assessment of tubeintegrity, those loads that do significantly affect burst or collapse shall bedetermined and assessed in combination with the loads due to pressurewith a safety factor of 1.2 on the combined primary loads and 1.0 on axialsecondary loads.2. Accident induced leakage performance criterion: The primary to secondaryaccident induced leakage rate for any design basis accident, other than aSG tube rupture, shall not exceed the leakage rate assumed in theaccident analysis in terms of total leakage rate for all SGs and leakagerate for an indi-idual SG. Except during a SG tube rupture, leakage is alsonot to exceed 40.75 gallon per minute per-total for all four SGs.(continued)DIABLO CANYON -UNITS 1 & 2 5.0-10 Unit 1 -Amendment No. 4-98,Rev29 Page10of27 Unit 2 -Amendment No. 4-99, Programs and Manuals5.55.5 Programs and Manuals5.5.11 Ventilation Filter Testing Program (VFTP) (continued)c. Demonstrate for each of the ESF systems that a laboratory test of a sample ofthe charcoal absorber, when obtained as described in Regulatory Guide 1.52,Revision 2, shows the methyl iodide penetration less than the value specifiedbelow when tested in accordance with ASTM D3803-1989 at a temperature of300C and at the relative humidity specified below. Laboratory testing shall becompleted at least once per 24 months and after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoaloperation.ESF Ventilation System Penetration RHControl Room 2.5% 95%Auxiliary Building 45.0% 95%Fuel Handling Building 15.0% 95%d. Demonstrate for each of the ESF systems that the pressure drop across thecombined HEPA filters and the charcoal adsorbers is less than the valuespecified below when tested in accordance with ANSI N510-1980 at the systemflowrate specified below +/- 10% at least once per 24 months.ESF Ventilation System Delta P FlowrateControl Room 3.5 in. WG 2100 cfmAuxiliary Building 3.7 in. WG 73,500 cfmFuel Handling Building 4.1 in. WG 35,750 cfmThe provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring ProgramThis program provides controls for potentially explosive gas mixtures contained in theWaste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks,and the quantity of radioactivity contained in temporary unprotected outdoor liquidstorage tanks.The gaseous radioactivity quantities shall be determined following the methodology inRegulatory Guide 1.24 "Assumptions Used For Evaluating the Potential RadiologicalConsequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure."The liquid radwaste quantities shall be maintained such that 10 CFR Part 20 limits aremet.(continued)DIABLO CANYON -UNITS I & 2 5.0-13 Unit 1 -Amendment No. 135,142,163,-0,Rev 29 Page 13 of 27 Unit 2- Amendment No. 135,142,165,199, Programs and Manuals5:55.5 Programs and Manuals (continued)5.5.19 Control Room Envelope Habitability ProqramA Control Room Envelope (CRE) Habitability Program shall be established andimplemented to ensure that CRE habitability is maintained such that, with anOPERABLE Control Room Ventilation System (CRVS), CRE occupants can control thereactor safely under normal conditions and maintain it in a safe condition following aradiological event, hazardous chemical release, or a smoke challenge. The programshall ensure that adequate radiation protection is provided to permit access andoccupancy of the CRE under design basis accident (DBA) conditions without personnelreceiving radiation exposures in excess of 5 rem whole-body TEDE or its cquivalent toany part of the body for the duration of the accident. The program shall include thefollowing elements:a. The definition of the CRE and the CRE boundary.b. Requirements for maintaining the CRE boundary in its design condition, includingconfiguration control and preventive maintenance.c. Requirements for (i) determining the unfiltered air inleakage past the CREboundary into the CRE in accordance with the testing methods and at theFrequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197,"Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors,"Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequenciesspecified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.d. Measurement, at designated locations, of the CRE pressure relative to allexternal areas adjacent to the CRE boundary during the pressurization mode ofoperation by one train of the CRVS, operating at the flow rate required by theVFTP, at a Frequency of 24 months on a STAGGERED TEST BASIS. Theresults shall be trended and used as part of the 24 month assessment of theCRE boundary.e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shallbe stated in a manner to allow direct comparison to the unfiltered air inleakagemeasured by the testing described in paragraph c. The unfiltered air inleakagelimit for radiological challenges is the inleakage flow rate assumed in thelicensing basis analyses of DBA consequences. Unfiltered air inleakage limits forhazardous chemicals must ensure that exposure of CRE occupants to thesehazards will be within the assumptions in the licensing basis.f. The provisions of SR 3.0.2 are applicable to the Frequencies required byparagraphs c and d for determining CRE unfiltered inleakage and assessing CREhabitability, and measuring CRE pressure and assessing the CRE boundary.DIABLO CANYON -UNITS 1 & 2 5.0-17a Unit 1 -Amendment No. 2-14-,Rev 29 Page 18 of 27 Unit 2 -Amendment No. 2-2, EnclosureAttachment 2PG&E Letter DCL-15-069ATTACHMENT 2Proposed Technical Specification Changes(RETYPED)Remove Pages1.1-33.4-363.6-53.6-93.6-105.0-105.0-135.0-17aInsert Pages1.1-33.4-363.6-53.6-93.6-105.0-105.0-135.0-17a Definitions1.11.1 Definitions (continued)DOSE EQUIVALENT 1-131DOSE EQUIVALENT XE-1 33DOSE EQUIVALENT 1-131 shall be that concentration of1-131 (microcuries per gram) that alone would produce thesame dose when inhaled as the combined activities ofiodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actuallypresent. The determination of DOSE EQUIVALENT 1-131shall be performed using the committed thyroid doseconversion factors from Table 2.1 of EPA Federal GuidanceReport No. 11, 1988, "Limiting Values of RadionuclideIntake and Air Concentration and Dose Conversion Factorsfor Inhalation, Submersion, and Ingestion."DOSE EQUIVALENT XE-1 33 shall be that concentration ofXe-1 33 (microcuries per gram) that alone would produce thesame acute dose to the whole body as the combinedactivities of noble gas nuclides Kr-85m, Kr-87, Kr-88,Xe-1 33m, Xe-133, Xe-135m, Xe-1 35, and Xe-1 38 actuallypresent. If a specific noble gas nuclide is not detected, itshould be assumed to be present at the minimum detectableactivity. The determination of DOSE EQUIVALENT XE-133shall be performed using effective'dose conversion factorsfor air submersion listed in Table 111.1 of EPA FederalGuidance Report No. 12, 1993, "External Exposure toRadionuclides in Air, Water, and Soil."I(continued)1.1-3 Unit 1 -Amendment No. 135, 155,456, 192,Unit 2 -Amendment No. 135, 155,1546, 193,DIABLO CANYON -UNITS 1 & 2 RCS Specific Activity3.4.16SURVEILLANCE REQUIREMENTSSURVEILLANCE FREQUENCYSR 3.4.16.1 -------------------------------- NOTE -------------------------- In accordance withOnly required to be performed in MODE 1. the Surveillance------------------------------------------------------------..- F re quency C ontrolVerify reactor coolant DOSE EQUIVALENT XE-133 Programspecific activity -270.0 pCi/gm.SR 3.4.16.2 ----------------------------- NOTE -----------------Only required to be performed in MODE 1.Verify reactor coolant DOSE EQUIVALENT 1-131 In accordance withspecific activity -1.0 pCi/gm. the SurveillanceFrequency ControlProgramANDBetween 2 and 6.hours after aTHERMALPOWER change of> 15% RTP withina 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period.IDIABLO CANYON -UNITS 1 & 23.4-36 Unit 1 -Amendment No. 135,4 92,20o,Unit 2 -Amendment No. 4-35,93,20,1 Containment Isolation Valves3.6.33.6 CONTAINMENT SYSTEMS3.6.3 Containment Isolation ValvesLCO 3.6.3Each containment isolation valve shall be OPERABLE.APPLICABILITY:ACTIONSMODES 1, 2, 3, and 4.--NOTES1. Penetration flow path(s) except for 48-inch purge valve flow paths, may be unisolatedintermittently under administrative controls.2. Separate Condition entry is allowed for each penetration flow path.3. Enter applicable Conditions and Required Actions for systems made inoperable bycontainment isolation valves.4. Enter applicable Conditions and Required Actions of LCO 3.6.1, "Containment," whenisolation valve leakage results in exceeding the overall containment leakage rateacceptance criteria.CONDITION REQUIRED ACTION COMPLETION TIMEA. -------------- NOTE--------- A.1 Isolate the affected 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months sOnly applicable to penetration flow path bypenetration flow paths with use of at least onetwo containment isolation closed and de-activatedvalves. automatic valve, closed------------------------------------ m anual valve, blindOne or more penetration flange, or check valveflow paths with one with flow through thecontainment isolation valve valve secured.inoperable except for acontainment purge supplyand exhaust valve orpressure/vacuum reliefvalve leakage not within (continued)limit. (continued)DIABLO CANYON -UNITS I & 23.6-5Unit 1 -Amendment No. 4-35,Unit 2 -Amendment No. 4-35, Containment Isolation Valves3.6.3SURVEILLANCE REQUIREMENTSSURVEILLANCE FREQUENCYSR 3.6.3.1 Verify each 48 inch purge valve is sealed closed, In accordance with theexcept for one purge valve in a penetration flow Surveillance Frequencypath while in Condition D of this LCO. Control ProgramSR 3.6.3.2 Verify each 12 inch vacuum/pressure relief valve In accordance with theis closed, except when these valves are open for Surveillance Frequencypressure control, ALARA or air quality Control Programconsiderations for personnel entry, or forSurveillances that require the valves to be open.SR 3.6.3.3 --------------------NOTE --------------Valves and blind flanges in high radiation areasmay be verified by use of administrative controls.Verify each containment isolation manual valve In accordance with theand blind flange that is located outside Surveillance Frequencycontainment and not locked, sealed or otherwise Control Programsecured and required to be closed duringaccident conditions is closed, except forcontainment isolation valves that are open underadministrative controls.SR 3.6.3.4 --------------------NOTE --------------Valves and blind flanges in high radiation areasmay be verified by use of administrative means.Verify each containment isolation manual valve Prior to entering MODEand blind flange that is located inside 4 from MODE 5 if notcontainment and not locked, sealed or otherwise performed within thesecured and required to be closed during previous 92 daysaccident conditions is closed, except forcontainment isolation valves that are open underadministrative controls.SR 3.6.3.5 Verify the isolation time of each automatic power In accordance with theoperated containment isolation valve is within Inservice Testinglimits. ProgramSR 3.6.3.6 Not used(continued)IDIABLO CANYON -UNITS 1 & 23.6-9 Unit 1 -Amendment No. !35,200,Unit 2 -Amendment No. 135,20!,

Containment Isolation Valves3.6.3SURVEILLANCE REQUIREMENTS (continued)SURVEILLANCE FREQUENCYSR 3.6.3.7 --------------------NOTE --------------This surveillance is not required when thepenetration flow path is isolated by a leak testedblank flange.Perform leakage rate testing for containment In accordance with thepurge supply and exhaust and vacuum/pressure Surveillance Frequencyrelief valves with resilient seals. Control ProgramSR 3.6.3.8 Verify each automatic containment isolation valve In accordance with thethat is not locked, sealed or otherwise secured in Surveillance Frequencyposition, actuates to the isolation position on an Control Programactual or simulated actuation signal.SR 3.6.3.9 Not usedSR 3.6.3.10 Verify each 12 inch containment In accordance with thevacuum/pressure relief valve is blocked to restrict Surveillance Frequencythe valve from opening > 50'. Control ProgramSR 3.6.3.11 Not used-tDIABLO CANYON -UNITS 1 & 23.6-10Unit 1 -Amendment No. 135,!75,200,Unit 2 -Amendment No. 135,!77,201, Programs and Manuals5.55.5 Programs and Manuals (continued)5.5.9 Steam Generator (SG) Tube Inspection ProgramA Steam Generator Program shall be established and implemented to ensure that SGtube integrity is maintained. In addition, the Steam Generator Program shall include thefollowing provisions:a. Provisions for condition monitoring assessments.Condition monitoring assessment means an evaluation of the "as found"condition of the tubing with respect to the performance criteria for structuralintegrity and accident induced leakage. The "as found" condition refers to thecondition of the tubing during an SG inspection outage, as determined from theinservice inspection results or by other means, prior to the plugging of tubes.Condition monitoring assessments shall be conducted during each outage duringwhich the SG tubes are inspected or plugged to confirm that the performancecriteria are being met.b. Performance criteria for SG tube integrity.SG tube integrity shall be maintained by meeting the performance criteria fortube structural integrity, accident induced leakage, and operational LEAKAGE.1. Structural integrity performance criterion: All in-service steam generator.tubes shall retain structural integrity over the full range of normal operatingconditions (including startup, operation in the power range, hot standby,and cool down and all anticipated transients included in the designspecification) and design basis accidents. This includes retaining a safetyfactor of 3.0 against burst under normal steady state full power operationprimary-to-secondary pressure differential and a safety factor of 1.4against burst applied to the design basis accident primary-to-secondarypressure differentials. Apart from the above requirements, additionalloading conditions associated with the design basis accidents, orcombination of accidents in accordance with the design and licensingbasis, shall also be evaluated to determine if the associated loadscontribute significantly to burst or collapse. In the assessment of tubeintegrity, those loads that do significantly affect burst or collapse shall bedetermined and assessed in combination with the loads due to pressurewith a safety factor of 1.2 on the combined primary loads and 1.0 on axialsecondary loads.2. Accident induced leakage performance criterion: The primary to secondaryaccident induced leakage rate for any design basis accident, other than aSG tube rupture, shall not exceed the leakage rate assumed in theaccident analysis in terms of total leakage rate for all SGs. Except during aSG tube rupture, leakage is not to exceed 0.75 gallon per minute total forall four SGs.(continued)DIABLO CANYON -UNITS 1 & 2 5.0-10 Unit 1 -Amendment No. --98,Unit 2 -Amendment No. 99, Programs and Manuals5.55.5 Programs and Manuals5.5.11 Ventilation Filter Testing Program (VFTP) (continued)c. Demonstrate for each of the ESF systems that a laboratory test of a sample ofthe charcoal absorber, when obtained as described in Regulatory Guide 1.52,Revision 2, shows the methyl iodide penetration less than the value specifiedbelow when tested in accordance with ASTM D3803-1989 at a temperature of300C and at the relative humidity specified below. Laboratory testing shall becompleted at least once per 24 months and after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoaloperation.ESF Ventilation System Penetration RHControl Room 2.5% 95%Auxiliary Building 5.0% 95%Fuel Handling Building 15.0% 95%d. Demonstrate for each of the ESF systems that the pressure drop across thecombined HEPA filters and the charcoal adsorbers is less than the valuespecified below when tested in accordance with ANSI N510-1980 at the systemflowrate specified below +/- 10% at least once per 24 months.ESF Ventilation System Delta P FlowrateControl Room 3.5 in. WG 2100 cfmAuxiliary Building 3.7 in. WG 73,500 cfmFuel Handling Building 4.1 in. WG 35,750 cfmThe provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring ProgramThis program provides controls for potentially explosive gas mixtures contained in theWaste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks,and the quantity of radioactivity contained in temporary unprotected outdoor liquidstorage tanks.The gaseous radioactivity quantities shall be determined following the methodology inRegulatory Guide 1.24 "Assumptions Used For Evaluating the Potential RadiologicalConsequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure."The liquid radwaste quantities shall be maintained such that 10 CFR Part 20 limits aremet.(continued)DIABLO CANYON -UNITS 1 & 2 5.0-13 Unit 1 -Amendment No. 135,112,163,198,Unit 2 -Amendment No. 135,142,1654199, Programs and Manuals5.55.5 Programs and Manuals (continued)5.5.19 Control Room Envelope Habitability ProgramA Control Room Envelope (CRE) Habitability Program shall be established andimplemented to ensure that CRE habitability is maintained such that, with anOPERABLE Control Room Ventilation System (CRVS), CRE occupants can control thereactor safely under normal conditions and maintain it in a safe condition following aradiological event, hazardous chemical release, or a smoke challenge. The programshall ensure that adequate radiation protection is provided to permit access andoccupancy of the CRE under design basis accident (DBA) conditions without personnelreceiving radiation exposures in excess of 5 rem TEDE for the duration of the accident.The program shall include the following elements:a. The definition of the CRE and the CRE boundary.b. Requirements for maintaining the CRE boundary in its design condition, includingconfiguration control and preventive maintenance.c. Requirements for (i) determining the unfiltered air inleakage past the CREboundary into the CRE in accordance with the testing methods and at theFrequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197,"Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors,"'Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequenciesspecified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.d. Measurement, at designated locations, of the CRE pressure relative to allexternal areas adjacent to the CRE boundary during the pressurization mode ofoperation by one train of the CRVS, operating at the flow rate required by theVFTP, at a Frequency of 24 months on a STAGGERED TEST BASIS. Theresults shall be trended and used as part of the 24 month assessment of theCRE boundary.e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shallbe stated in a manner to allow direct comparison to the unfiltered air inleakagemeasured by the testing described in paragraph c. The unfiltered air inleakagelimit for radiological challenges is the inleakage flow rate assumed in thelicensing basis analyses of DBA consequences. Unfiltered air inleakage limits forhazardous chemicals must ensure that exposure of CRE occupants to thesehazards will be within the assumptions in the licensing basis.f. The provisions of SR 3.0.2 are applicable to the Frequencies required byparagraphs c and d for determining CRE unfiltered inleakage and assessing CREhabitability, and measuring CRE pressure and assessing the CRE boundary.DIABLO CANYON -UNITS I & 2 5.0-17a Unit I -Amendment No. 20-4-,Unit 2 -Amendment No. 2-02, EnclosureAttachment 3PG&E Letter DCL-15-069ATTACHMENT 3Technical Specification Bases Markup(For Information Only)

RCS Pressure SLB 2.1.2B 2.0 SAFETY LIMITS (SLs)B 2.1.2 Reactor Coolant System (RCS) Pressure SLBASESBACKGROUNDThe SL on RCS pressure protects the integrity of the RCS againstoverpressurization. In the event of fuel cladding failure, fissionproducts are released into the reactor coolant. The RCS then servesas the primary barrier in preventing the release of fission products intothe atmosphere. By establishing an upper limit on RCS pressure, thecontinued integrity of the RCS is ensured. According to 10 CFR 50,Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," andGDC 15, "Reactor Coolant System Design" (Ref. 1), the reactorpressure coolant boundary (RCPB) design conditions are not to beexceeded during normal operation and anticipated operationaloccurrences (AOOs). Also, in accordance with GDC 28, "ReactivityLimits" (Ref. 1), reactivity accidents, including rod ejection, do notresult in damage to the RCPB greater than limited local yielding.The design pressure of the RCS is 2500 psia. During normal operationand AOOs, RCS pressure is limited from exceeding the designpressure by more than 10%, in accordance with Section III of theASME Code (Ref. 2). To ensure system integrity, all RCS componentswere hydrostatically tested at 125% of design pressure, according tothe ASME Code requirements prior to initial operation when there is nofuel in the core. Following inception of unit operation, RCScomponents shall be pressure tested, in accordance with therequirements of ASME Code,Section XI (Ref. 3).Overpressurization of the RCS could result in a breach of the RCPB. Ifsuch a breach occurs in conjunction with a fuel cladding failure, fissionproducts could enter the containment atmosphere, raising concernsrelative to limits on radioactive releases specified in 10 CFR 4-050.67,"Reactor Site CriteriaAccident Source Term" (Ref. 4).(continued)DIABLO CANYON -UNITS 1 & 2Rev 8 Page 4 of 6 RCS Pressure SLB 2.1.2BASES (continued)APPLICABILITYSL 2.1.2 applies in MODES 1, 2, 3, 4, and 5 because this SL could beapproached or exceeded in these MODES due to overpressurizationevents. The SL is not applicable in MODE 6 because the reactorvessel head closure bolts are not fully tightened, or the reactor vesselis sufficiently vented, making it unlikely that the RCS can bepressurized.SAFETY LIMITVIOLATIONSIf the RCS pressure SL is violated when the reactor is in MODE 1 or 2,the requirement is to restore compliance and be in MODE 3 withinI hour.Exceeding the RCS pressure SL may cause immediate RCS failureand create a potential for radioactive releases in excess of10 CFR 4-050.67, "Reactor Site CriteriaAccident Source Term" limits(Ref. 4).The allowable Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance ofreducing power level to a MODE of operation where the potential forchallenges to safety systems is minimized.If the RCS pressure SL is exceeded in MODE 3, 4, or 5, RCS pressuremustbe restored to within the SL value within 5 minutes. Exceedingthe RCS pressure SL in MODE 3, 4, or 5 is more severe thanexceeding this SL in MODE I or 2, since the reactor vesseltemperature may be lower and the vessel material, consequently, lessductile. As such, pressure must be reduced to less than the SL within5 minutes. The action does not require reducing MODES, since thiswould require reducing temperature, which would compound theproblem by adding thermal gradient stresses to the existing pressurestress.REFERENCES1. 10 CFR 50, Appendix A, GDC 14 (associated with 1967 GDC 9per FSAR Appendix 3.1A), GDC 15 (no direct correlation to 1967GDC; however, intent of 1971 GDC is per met per FSARAppendix 3.1A), and GDC 28 (associated with 1967 GDC 30 perFSAR Appendix 3.1A).2. ASME, Boiler and Pressure Vessel Code,Section III, Summer1969.3. ASME, Boiler and Pressure Vessel Code, Section Xl.4. 10 CFR 4-050.67.5. FSAR, Section 7.2.6. DCM S-7, 3.4.1.DIABLO CANYON -UNITS 1 & 2Rev 8 Page 6 of 6 SDMB 3.1.1BASESAPPLICABLESAFETYANALYSIS(continued)In the boron dilution analysis, the required SDM defines the reactivitydifference between an initial subcritical boron concentration and thecorresponding critical boron concentration. These values, inconjunction with the configuration of the RCS and the assumed dilutionflow rate, directly affect the results of the analysis. This event is mostlimiting at the beginning of core life, when critical boron concentrationsare highest.Depending on the system initial conditions and reactivity insertion ratethe uncontrolled rod withdrawal transient is terminated by either a highpower level trip or a high pressurizer pressure trip. In all cases, powerlevel, RCS pressure, linear heat rate, and the DNBR do not exceedallowable limits.The ejection of a control rod rapidly adds reactivity to the reactor coreicausing both the core power level and heat flux to increase withcorresponding increases in reactor coolant temperatures and pressure.The ejection of a rod also produces a time dependent redistribution ofcore power.The startup of an inactive RCP in MODES 1 or 2 is precluded. InMODE 3, the startup of an inactive RCP cannot result in a "cold water"criticality, even if the maximum difference in temperature existsbetween the SG and the core. The maximum positive reactivityaddition that can occur due to an inadvertent start is less than half theminimum required SDM. Startup of an idle RCP cannot, therefore,produce a return to power from the hot standby condition.SDM satisfies Criterion 2 of IOCFR50.36(c)(2)(ii). Even though it is notdirectly observed from the control room, SDM is considered an initialcondition process variable because it is periodically monitored toensure that the unit is operating within the bounds of accident analysisassumptions.LCOSDM is a core design condition that can be ensured during operationthrough control rod positioning (control and shutdown banks) andthrough the soluble boron concentration.The MSLB (Ref. 2) and the boron dilution (Ref. 3) accidents are themost limiting analyses that establish the SDM value of the LCO. ForMSLB accidents, if the LCO is violated, there is a potential to exceedthe DNBR limit and to exceed 10 CFR 4-0050.67, "ReaGtOF-SiteGrt-e4aAccident Source Term," limits (Ref. 4). For the boron dilutionaccident, if the LCO is violated, the minimum required time assumedfor operator action to terminate dilution may no longer be sufficient.The required SDM is specified in the COLR.(contnued(continued)DIABLO CANYON -UNITS 1 & 2Rev 8A Page 3 of 45 SDMB 3.1.1BASESSURVEILLANCE SR 3.1.1.1 (continued)REQUIREMENTS Using the ITC accounts for Doppler reactivity in this calculationbecause the reactor is subcritical, and the fuel temperature will bechanging at the same rate as the RCS.The Surveillance Frequency is based on operating experience,equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.REFERENCES 1 10 CFR 50, Appendix A, GDC 26.2 FSAR, Chapter 15, Section 15.4.2.1.3 FSAR, Chapter 15, Section 15.2.4.4 10 CFR 40050.67.5 FSAR, Chapter 15, Section 15.4.6.1.6.DIABLO CANYON -UNITS 1 & 2Rev8A Page5of45 RTS InstrumentationB 3.3.1B 3.3 INSTRUMENTATIONB 3.3.1 Reactor Trip System (RTS) InstrumentationBASESBACKGROUND The RTS initiates a unit shutdown, based on the values of selected unitparameters, to protect against violating the core fuel design limits andReactor Coolant System (RCS) pressure boundary during anticipatedoperational occurrences (AOOs) and to assist the Engineered SafetyFeatures (ESF) Systems in mitigating accidents.The protection and monitoring systems have been designed to assuresafe operation of the reactor. This is achieved by specifying limitingsafety system settings (LSSS) in terms of parameters directlymonitored by the RTS, as well as specifying LCOs on other reactorsystem parameters and equipment performance.The LSSS, as defined in 10 CFR 50.36, are defined in this specificationas the Allowable Values, and in conjunction with the LCOs, establishthe threshold for protective system action to prevent exceedingacceptable limits during Design Basis Accidents (DBAs).During AOOs, which are those events expected to occur more thanonce during the unit life, the acceptable limits are:1. The Departure from Nucleate Boiling Ratio (DNBR) shall bemaintained above the Safety Limit (SL) value to prevent departurefrom nucleate boiling (DNB);2. Fuel centerline melt shall not occur; and3. The RCS pressure SL of 2735 psig shall not be exceeded.Operation within the SLs of Specification 2.0, "Safety Limits (SLs),"also maintains the above values and assures that offsite dose will bewithin the 10 CFR 50 and 10 CFR 4-0050.67 criteria during AOOs.Accidents are events that are analyzed even though they are notexpected to occur during the unit life. The acceptable limit duringaccidents is that offsite dose shall be maintained within an acceptablefraction of 10 CFR 050.67 limits. Different accident categories areallowed a different fraction of these limits, based on probability ofoccurrence. Meeting the acceptable dose limit for an accident category.is considered having acceptable consequences, for that event.(continued)DIABLO CANYON -UNITS I & 2Rev 8 Page I of 167 ESFAS InstrumentationB 3.3.2BASESBACKGROUND During AOOs, which are those events expected to occur one or more(continued) times during the unit life, the acceptable limits are:1. The Departure from Nucleate Boiling Ratio (DNBR) shall bemaintained above the Safety Limit (SL) value to prevent departurefrom nucleate boiling (DNB).2. Fuel centerline melt shall not occur, and3. The RCS pressure SL of 2750 psia shall not be exceeded.Operation within the SLs of Specification 2.0, "Safety Limits (SLs),"also maintains the above values and assures that offsite dose will bewithin the 10 CFR 50 and 10 CFR 4-1050.67 criteria during AOOs.Accidents are events that are analyzed even though they are notexpected to occur during the unit life. The acceptable limit duringaccidents is that offsite dose shall be maintained within an acceptablefraction of 10 CFR 40050.67 limits. Different accident categories areallowed a different fraction of these limits, based on probability ofoccurrence. Meeting the acceptable consequences for that event isconsidered having acceptable consequences for that event. However,these values and their associated NTSPs are not considered to beLSSS as defined in 10 CFR 50.36.The ESFAS instrumentation is segmented into three distinct butinterconnected modules as identified below:Field transmitters or process sensors and instrumentation: providea measurable electronic signal based on the physicalcharacteristics of the parameter being measured;Signal processing equipment including digital protection system,field contacts, and protection channel sets: provide signalconditioning, bistable setpoint comparison, process algorithmactuation, compatible electrical signal output to protection systemdevices, and control board/control room/miscellaneous indications;andSolid State Protection System (SSPS) including input, logic, andoutput bays: initiates the proper unit shutdown or engineeredsafety feature (ESF) actuation in accordance with the defined logicand based on the bistable outputs from the signal process controland protection system. The residual heat removal pump trip orrefueling water storage tank level-low signal is not processed by theSSPS. The associated relays are located in the residual heatremoval pumps control system.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8 Page 69 of 167 Containment Ventilation Isolation InstrumentationB 3.3.6BASESAPPLICABLE They are also the primary means for automatically isolatingSAFETY containment in the event of a fuel handling accident or any otherANALYSES source within containment during shutdown. Containment isolation in(continued) turn ensures meeting the containment leakage rate assumptions of thesafety analyses, and ensures that the calculated accidental offsiteradiological doses are below 10 CFR 41-050.67 (Ref. 1) limits. Due toradioactive decay, containment is only required to isolate during fuelhandling accidents involving handling recently irradiated fuel (i.e., fuelthat has occupied part of a critical reactor core within the previous* 4-072 hours.)The containment ventilation isolation instrumentation satisfiesCriterion 3 of 10 CFR 50.36(c)(2)(ii).LCO The LCO requirements ensure that the instrumentation necessary toinitiate Containment Ventilation Isolation, listed in Table 3.3.6-1, isOPERABLE.1. Manual Initiation -Not used2. Automatic Actuation Lociic and Actuation RelaysThe LCO requires two trains of Automatic Actuation Logic andActuation Relays OPERABLE to ensure that no single random failurecan prevent automatic actuation.Automatic Actuation Logic and Actuation Relays consist of the samefeatures and operate in the same manner as described for ESFASFunction 1.b, SI, and ESFAS Function 3.a, Containment Phase AIsolation. The applicable MODES and specified conditions for theContainment Ventilation Isolation portion of these Functions aredifferent and less restrictive than those for their Phase A isolation andSI roles. If one or more of the SI or Phase A isolation Functionsbecomes inoperable in such a manner that only the ContainmentVentilation Isolation Function is affected, the Conditions applicable totheir SI and Phase A isolation Functions need not be entered. The lessrestrictive Actions specified for inoperability of the ContainmentVentilation Isolation Functions specify Sufficient compensatorymeasures for this case.3. Containment RadiationThe LCO specifies two required channels of radiation monitors toensure that the radiation monitoring instrumentation necessary toinitiate Containment ventilation Isolation remains OPERABLE inMODES 1-4.The LCO only requires one monitor to be OPERABLE duringmovement of recently irradiated fuel assemblies in containment. Inorder to provide the CVI function under these conditions withoutplacing the entire SSPS in service, an alternate circuit is provided topower the output relays and provide logic actuation signalsindependent of the SSPS.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8 Page 151 of 167 Containment Ventilation Isolation Instrumentation.B 3.3.6BASESSURVEILLANCEREQUIREMENTSSR 3.3.6.7 (continued)The test verifies that the channel responds to a measured parameterwithin the necessary range and accuracy. The Surveillance Frequencyis based on operating experience, equipment reliability, and plant riskand is controlled under the Surveillance Frequency Control Program.SR 3.3.6.8This SR assures that the individual channel RESPONSE TIMES for theCVI from Containment Purge Radiation Gaseous and Particulatefunction are less than'or equal to the maximum values assumed in theaccident analysis. Response Time testing acceptance criteria areincluded in ECG 38.2. Individual component response times are notmodeled in the analyses. The analyses model the overall or elapsedtime, from the point at which the parameter exceeds the Trip Setpointvalue at the sensor, to the point at which the equipment in both trainsreaches the required functional state (e.g., valves in full closedposition). The response time may be measured by a series ofoverlapping tests such that the entire response time is measured.The Surveillance Frequency is based on operating experience,equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.REFERENCES 1. 10 CFR44-.0A450.67.2. NUREG-1366, December 1992.3. DCM No. T-16, Containment Function.4. WCAP-1 5376-P-A, Revision 1, "Risk-Informed Assessment of theRTS and ESFAS Surveillance Test Intervals and Reactor TripBreaker Test and Completion Times," March 2003.5. License Amendment 184/186, January 3, 2006.IDIABLO CANYON -UNITS I & 2Rev 8 Page 156 of 167 CRVS Actuation instrumentationB 3.3.7BASESBACKGROUND The CRVS has two additional manually selected emergency operating(continued) modes; smoke removal and recirculation. Neither of these modes arerequired for the CRVS to be OPERABLE, but they are useful for certainnon-DBA circumstances.APPLICABLE The control room must be kept habitable for the operators stationedSAFETY there during accident recovery and post accident operations.ANALYSES The CRVS acts to terminate the supply of unfiltered outside air to thecontrol room, initiate filtration, and pressurize the control room. Theseactions are necessary to ensure the control room is kept habitable forthe operators stationed there during accident recovery and postaccident operations by minimizing the radiation exposure of controlroom personnel.In MODES 1, 2, 3, and 4, the radiation monitor (located at the controlroom intakes) actuation of the CRVS is a backup for the Phase Asignal actuation. This ensures initiation of the CRVS during a loss ofcoolant accident, er--Steam generator tube rupture, control rod ejectionaccident and Main Steam Line Break involving a releasc of radioActi-.ve.materials.The radiation monitor actuation of the CRVS in MODES 5 and 6, duringmovement of recently irradiated fuel assemblies (i.e., fuel that hasoccupied part of a critical reactor core within the previous 4-G072hours), is the primary means to ensure control room habitability in theevent of a fuel handling or waste gas decay tank rupture accident.This actuation is credited in the FHA. The CRVS pressurization systemactuation instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).In MODES 1. 2. 3, 4. 5 and 6, credit is taken for the dual ventilationintake design of the CR Dressurization air intakes. Based on theavailability of redundant PG&E Design Class I radiation monitors ateach pressurization intake location, the DCPP design has the capabilityof initial selection of the cleaner intake, but does not have the capabilityof automatic selection of the clean intake throughout the event. Basedon the CRVS pressurization intake design, and the expectation that theoperator will manually make the proper intake selection throughout theevent, and oer RG 1.194, June 2003, Regulatory Position C.3.3.2.3,when the CRVS is in Mode 4, the X/Q values for the more favorableCR intake is reduced by a factor of 4 and utilized to estimate the doseconsequences.LCO The LCO requirements ensure that instrumentation necessary toinitiate the CRVS pressurization system is OPERABLE.1. Manual InitiationThe LCO requires two trains OPERABLE. The operator caninitiate the CRVS pressurization mode at any time by using eitherof two switches in the control room. This action will causeactuation of all components in the same manner as any of theDIABLO CANYON -UNITS 1 & 2Rev 8 Page 158 of 167 CRVS Actuation InstrumentationB 3.3.7automatic actuation signals.The LCO for Manual Initiation ensures the proper amount ofredundancy is maintained in the manual actuation circuitry toensure the operator has manual initiation capability.2. Automatic Actuation RelaysThe LCO requires two trains of Actuation Relays OPERABLE toensure that no single random failure can prevent automaticactuation of the pressurization system. Since each unit has onetrain of Actuation Relays consisting of two sets of actuation logic,each unit must have at least one logic set for both trains to beconsidered OPERABLE.(continued)(Spillover from previous page.)DIABLO CANYON -UNITS I & 2Rev 8 Page 158 of 167 CRVS Actuation InstrumentationB 3.3.7BASESSURVEILLANCEREQUIREMENTS(continued)SR 3.3.7.5SR 3.3.7.5 is the performance of a SLAVE RELAY TEST. This testenergizes the Slave Relays and verifies actuation of the equipment tothe pressurization mode. Although there are no "Slave Relays" as inthe SSPS, this surveillance was maintained to preserve the format ofthe standard specification. The surveillance is intended to ensure thatthe actuation relays, downstream of the logic, function to actuate thepressurization mode equipment. Since the radiation monitors directlyactuate the actuation relays, this test is performed as part of theperformance of SR 3.3.7.2.The Surveillance Frequency is based on operating experience,equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.SR 3.3.7.6SR 3.3.7.6 is the performance of a TADOT. This test is a check of theManual Actuation Functions. Each Manual Actuation Function is testedup to, and including, the master relay coils. In some instances, the testincludes actuation of the end device (i.e., pump starts, valve cycles,etc.).The Surveillance Frequency is based on operating experience,equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program. The SR is modified by aNote that excludes verification of setpoints during the TADOT. TheFunctions tested have no setpoints associated with them.SR 3.3.7.7CHANNEL CALIBRATION is a complete check of the instrument loop,including the sensor. The test verifies that the channel responds to ameasured parameter within the necessary range and accuracy.The Surveillance Frequency is based on operating experience,equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.REFERENCES 1. WCAP-13878, "Reliability of Potter & Brumfield MDR Relays",June 1994.2. WCAP-1 3900, "Extension of Slave Relay Surveillance TestIntervals", April 1994.3. License Amendment 184/186, January 3, 2006.4. RG 1.194, "Atmospheric Relative Concentrations for Control RoomRadiological Habitability Assessments at Nuclear Power Plants,"June 2003.DIABLO CANYON -UNITS I & 2Rev 8 Page 163 of 167 FBVS Actuation InstrumentationB 3.3.8B 3.3 INSTRUMENTATIONB 3.3.8 Fuel Building Ventilation System (FBVS) Actuation InstrumentationBASESBACKGROUNDThe FBVS ensures that radioactive materials in the fuel buildingatmosphere following a fuel handling accident involving handlingrecently irradiated fuel (i.e., fuel that has occupied part of a criticalreactor core within the previous 40072 hours) are filtered and adsorbedprior to exhausting to the environment. The system is described in theBases for LCO 3.7.13, "Fuel Handling Building Ventilation System."The system initiates filtered ventilation of the fuel building automaticallyfollowing receipt of a high radiation signal from the Spent Fuel PoolMonitor or from the New Fuel Storage Vault Monitor. Initiation mayalso be performed manually as needed from the main control room orfuel handling building.High radiation, from either of the two monitors, provides FBVSinitiation. These actions function to prevent exfiltration of contaminatedair by initiating filtered ventilation, which imposes a negative pressureon the fuel building.APPLICABLE The FBVS ensures that radioactive materials in the fuel buildingSAFETY atmosphere following a fuel handling accident involving handlingANALYSES recently irradiated fuel are filtered and adsorbed prior to beingexhausted to the environment. This action reduces the radioactivecontent in the fuel building exhaust following a fuel handling accidentso that offsite doses Fremain within the limits spccified in 10 CFR 100(Ref~.1)The FBVS actuation instrumentation satisfies Criterion 3 of 10 CFR50.36(c)(2)(ii).LCO The LCO requirements ensure that instrumentation necessary toinitiate the FBVS is OPERABLE.1. Manual InitiationThe LCO requires two channels OPERABLE. The operator caninitiate the FBVS at any time by using either of two switches, onein the control room and another in the fuel handling building. Thisaction will cause actuation of all components in the same manneras any of the automatic actuation signals.The LCO for Manual Initiation ensures the proper amount ofredundancy is maintained in the manual actuation circuitry toensure the operator has manual initiation capability.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8 Page 164 of 167 FBVS Actuation InstrumentationB 3.3.8BASESSURVEILLANCE SR 3.3.8.1 (continued)REQUIREMENTSThe CHANNEL CHECK supplements less formal, but more frequent,checks of channels during normal operational use of the displaysassociated with the LCO required channels. The SurveillanceFrequency is based on operating experience, equipment reliability, andplant risk and is controlled under the Surveillance Frequency ControlProgram.SR 3.3.8.2A CFT is performed on each required channel to ensure the entirechannel will perform the intended function. This test verifies thecapability of the instrumentation to provide the FBACS actuation. TheSurveillance Frequency is based on operating experience, equipmentreliability, and plant risk and is controlled under the SurveillanceFrequency Control Program.SR 3.3.8.3 -Not usedSR 3.3.8.4SR 3.3.8.4 is the performance of a TADOT. This test is a check of themanual actuation functions. Each manual actuation function is testedup to, and including, the master relay coils. In some instances, the testincludes actuation of the end device (e.g., pump starts, valve cycles,etc.). The Surveillance Frequency is based on operating experience,equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program. The SR is modified by aNote that excludes verification of setpoints during the TADOT. TheFunctions tested have no setpoints associated with them.SR 3.3.8.5CHANNEL CALIBRATION is a complete check of the instrument loop,including the sensor. The test verifies that the channel responds to ameasured parameter within the necessary range and accuracy. TheSurveillance Frequency is based on operating experience, equipmentreliability, and plant risk and is controlled under the SurveillanceFrequency Control Program.REFERENCES 1. 10CFR 100.!!Not used.2. License Amendment 184/186, January 3, 2006.3. PG&E Letter DCL-05-124DIABLO CANYON -UNITS 1 & 2Rev 8 Page 167 of 167 RCS Operational LEAKAGEB 3.4.13BASES (continued)APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses do notSAFETY address operational LEAKAGE. However, other operational LEAKAGEANALYSES is related to the safety analyses for LOCA; the amount of leakage canaffect the probability of such an event. Safety analyses for designbasis events that model primary to secondary LEAKAGE result insteam discharge to the atmosphere. The safety analysis for the S1LBevent assumes that prirnar; to secoendary LEAKAGE is 10.5 gpm (rootemperature conditions) from the faulted SG r incr.eases to 10.5 gp,,as a resiult of accident indced conditions, and 0.1 gpm (roomtemperature co.ditions) from eac. h intact SG. The safety analyses forevents resulting in steam discharge to the atmosphere, ethe-4 haRSGTR aRd SLB, assume that primary to secondary LEAKAGE from allSGs is 0.75 gpm (hot ......ie.sStandard Temperature and Pressure)under accident conditions. For conservatism, the SLB assumes that thetotal 0.75 gpm tube leakage is assigned to the faulted steam generatorand the SGTR assumes that the total 0.75 gpm tube leakage isassigned to the 3 intact steam generators. The LCO requirement tolimit primary to secondary LEAKAGE through any one SG to less thanor equal to 150 gallons per day is significantly less than the conditionsassumed in the SLB safety analysis for the faulted SG.Primary to secondary LEAKAGE is a factor in the dose releasesoutside containment resulting from a steam line break (SLB) accident.To a lesser extent, other accidents or transients involve secondarysteam release to the atmosphere, such as a steam generator tuberupture (SGTR). The leakage contaminates the secondary fluid.The SGTR (Ref. 3) is more limiting for radiological releases at the siteboundary. The radiological dose analysis assumes loss of off-sitepower at the time of reactor trip with no subsequent condenser coolingavailable. The SGTR assumes that the total 0.75 gpm tube leakage isassigned to the 3 intact steam generators. The steam generator (SG)PORV for the SG that has sustained the tube rupture is assumed to failopen for 30 minutes, at which time the operator closes the block valveto the PORV. The dose consequences resulting from the SGTRaccident are within the limits defined in 10 CFR 40050.67 (Ref. 6).'.The SLB is more limiting for site radiation releases for events otherthan SGTR. The safety a.alysis for the SLB accident assumes10.5 gpM primary to secondary LEAKA~GE is thro)ugh the faulted SG.The dose consequences resulting from the SLB accident are wellwithin the limits defined in10 CFR 400-5067- approvedlicensing basis (i.e., small fraction of these limits).The safety analysis for RCS main loop piping for GDC-4 (Ref. 1)assumes 1 gpm unidentified leakage and monitoring per RG 1.45(Ref. 2) are maintained (Ref. 4 and 5).The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).(continued)DIABLO CANYON -UNITS 1 & 2Rev 8A Page 71 of 105 RCS Operational LEAKAGEB 3.4.13BASES (continued)REFERENCES1.2.3.4.5.6.7.8.10 CFR 50, Appendix A, GDC 4 and 30.Regulatory Guide 1.45, May 1973.FSAR, Section 15.FSAR, Section 3.NUREG-1061, Volume 3, November, 1984.10 CFR -1050.67.NEI 97-06, "Steam Generator Program Guidelines."EPRI, "Pressurized Water Reactor Primary-to-Secondary LeakGuidelines."DIABLO CANYON -UNITS 1 & 2Rev 8A_ Page 77 of 105 RCS Specific ActivityB 3.4.16B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.16 RCS Specific ActivityBASESBACKGROUNDThe maximum dose to the whole body and the thyroid that an individualat the exclusion area boundary can receive for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following anaccident or at the low population zone outer boundary for theradiological release duration, is, specified in 10 CFR 410._-_50.67 (Ref.1). Doses to the control room operators must be limited per GDC 19.The limits on specific activity ensure that the doses are appropriatelylimited during analyzed transients and accidents.The RCS specific activity LCO limits the allowable concentration levelof radionuclides in the reactor coolant. The LCO limits are establishedto minimize the dose consequences in the event of a steam line break(SLB) or steam generator tube rupture (SGTR) accident.The LCO contains specific activity limits for both DOSE EQUIVALENT1-131 and DOSE EQUIVALENT XE-133. The allowable levels areintended to ensure that offsite and control room doses meet theappropriate acceptance criteria in the Standard Review Plan RG 1.183(Ref. 2).IAPPLICABLESAFETYANALYSESThe LCO limits on the specific activity of the reactor coolant ensuresthat the resulting offsite and control room doses meet the appropriateSRP acceptance criteria following a SLB or a SGTR accident. Thesafety analyses (Refs. 3 and 4) assume the specific activity of thereactor coolant is at or, more conseprative than the LCO limits, and anexisting reactor coolant steam generator (SG) tube leakage rate of40.75 gpm-exists. The safety analyses assume the specific activity ofthe secondary coolant is at its limit of 0.1 pCi/gm DOSE EQUIVALENT1-131 from LCO 3.7.18, "Secondary Specific Activity."IThe analysis for the SLB and SGTR accidents establish theacceptance limits for RCS specific activity. Reference to theseanalyses is used to assess changes to the unit that could affect RCSspecific activity, as they relate to the acceptance limits.The analyses consider two cases of reactor coolant specific activity.One case assumes specific activity at 1.0 pCi/gm DOSE EQUIVALENT1-131 with a concurrent large iodine spike that increases the rate ofrelease of iodine from the fuel rods containing cladding defects to theprimary coolant immediately after a SLB (by a factor of 500) or SGTR(by a factor of 335), respectively.(continued)DIABLO CANYON -UNITS I & 2Rev 8A Page 93 of 105 RCS Specific ActivityB 3.4.16BASESAPPLICABLESAFETYANALYSES(continued)The second case assumes the initial reactor coolant iodine activity at60.0 pCi/gm DOSE EQUIVALENT 1-131 due to a pre-accident iodinespike caused by an RCS transient. In both cases, the noble gasspecific activity is assumed to be 651-270 pCi/gm DOSE EQUIVALENTXE-133.The SGTR analysis also assumes a loss of offsite power at the sametime as the reactor trip. The SGTR causes a reduction in reactorcoolant inventory. The reduction initiates a reactor trip from a lowpressurizer pressure signal or an RCS overtemperature AT signal.The loss of offsite power causes the steam dump valves to close toprotect the condenser. The rise in pressure in the ruptured SGdischarges radioactively contaminated steam to the atmospherethrough the SG power operated relief valves and the main steam safetyvalves. The unaffected SGs remove core decay heat by venting steamto the atmosphere until the cooldown ends and the RHR system isplaced in service.Operation with iodine specific activity levels greater than the LCO limitis permissible, if the activity levels do not exceed 60.0 pCi/gm DOSEEQUIVALENT 1-131, for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .The limits on RCS specific activity are also used for establishingstandardization in radiation shielding and plant personnel radiationprotection practices.RCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).LCO 'The iodine specific activity in the reactor coolant is limited to1.0 pCi/gm DOSE EQUIVALENT 1-131, and the noble gas specificactivity in the reactor coolant is limited to 609.0270.0 pCi/gm DOSEEQUIVALENT XE-133, as contained in SR 3.4.16.2 and SR 3.4.16.1respectively. The limits on specific activity ensure that offsite andcontrol room doses will meet the appropriate SRR acceptance criteria(Refs. 1 and 2).(continued)IIDIABLO CANYON -UNITS I & 2Rev 8A Page 94 of 105 RCS Specific ActivityB 3.4.1,6BASESSURVEILLANCE The definition of DOSE EQUIVALENT XE-133 in Specification 1.1,REQUIREMENTS "Definitions," requires that the determination of DOSE EQUIVALENT(continued) XE-133 shall be performed using the effective dose conversion factorsfor air submersion listed in Table 111.1 of EPA Federal Guidance ReportNo. 12, 1993, "External Exposure to Radionuclides in Air, Water, andSoil." These dose conversion factors are consistent with the doseconversion factors used in the applicable dose consequence analyses.The Note modifies this SR to allow entry into and operation in MODE 4,MODE 3, and MODE 2 prior to performing the SR. This allows theSurveillance to be performed in those MODES, prior to enteringMODE 1.SR 3.4.16.2This Surveillance is performed to ensure iodine specific activity remainswithin the LCO limit during normal operation and following fast powerchanges when iodine spiking is more apt to occur. The SurveillanceFrequency is based on operating experience, equipment reliability, andplant risk and is controlled under the Surveillance Frequency Control "Program. The Frequency, between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change-> 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, is established because the iodinelevels peak during this time following iodine spike initiation; samples atother times would provide inaccurate results.The definition of DOSE EQUIVALENT 1-131 in Specification 1.1,"Definitions," specifies the thyroid dose conversion factors which may beused to determine DOSE EQUIVALENT 1-131. The thyraid doseconversion factors used to determine DOSE EQUIVALENT 1-131 arethe committed thyroid dose conversion factors from Table 2.1 of EPAFederal Guidance Report No. 11, 1988, "Limiting Values ofRadionuclide Intake and Air Concentration and Dose ConversionFactors for Inhalation, Submersion and Ingestion." and arete-beconsistent with the dose conversion factors used in the applicable doseconsequence analyses, Or be .. .cnSative with reSPect to the doseconcrsonfactors used in the applicablc dose coensequence analyses..uch that a higher DOSE EQU/I'ALENT 1 131 is deteFrmined.The Note modifies this SR to allow entry into and operation in MODE 4,MODE 3, and MODE 2 prior to performing the SR. This allows theSurveillance to be performed in those MODES, prior to enteringMODE 1.(continued)DIABLO CANYON -UNITS 1 & 2Rev8A Page 97 of 105 RCS Specific ActivityB 3.4.16BASES (continued)REFERENCES 1. 10 CFR 100.41, 197350.67.2. Standard Review Plan (SRP), Section 6.4 (SLe and SGT-R controroomn dose limits), Section 4 5. 1.5Appendix A (SL=B offsite doselimits) and Section 5.6.3 (SGT-R offsite dose imits).RegulatoryGuide 1.183, July 2000.3. FSAR, Sections 15.4.3 and 15.5.20.4. FSAR Section 15.15.5.18.DIABLO CANYON -UNITS I & 2Rev 8A Page 98 of 105 Steam Generator (SG) Tube IntegrityB 3.4.17BASES (continued)APPLICABLESAFETYANALYSESThe steam generator tube rupture (SGTR) accident is the limitingdesign basis event for SG tubes and avoiding an SGTR is the basis forthis Specification. The analysis of a SGTR event assumes a teta-primary to secondary LEAKAGE rate of 40.75 gpm from the intact SGsplus the leakage rate associated with a double-ended rupture of asingle tube. The SGTR radiological dose analysis assumes loss of off-site power at the time of reactor trip with no subsequent condensercooling available. The SG PORV for the SG that has sustained thetube rupture is assumed to fail open for 30 minutes, at which time theoperator closes the block valve to the PORV. The SGTR radiologicaldose analysis assumes the contaminated secondary fluid is releasedbriefly to the atmosphere from all the PORVs following reactor trip, isreleased from the ruptured SG PORV for 30 minutes, is released fromthe intact SG PORVs during the cooldown, and is released from allPORVs following cooldown until termination of the event.The analysis for design basis accidents and transients other than aSGTR assume the SG tubes retain their structural integrity (i.e., theyare assumed not to rupture.) For the SLB event, the prim.ar; toSecondary L-E.AK1A.GE is 10.5 gpm from the faulted SGn o-r is assumedto increase to 10. 5 gp. .as a resuit of accident indu.ed conditions, and0.1 gpm from each intact SG. For other events, theThe steamdischarge to the atmosphere is based on the total primary to secondaryLEAKAGE from all SGs of 0.75 gpm under accident conditions. Foraccidents that do not involve fuel damage, the primary coolant activitylevel of DOSE EQUIVALENT 1-131 is assumed to be equal to the LCO3.4.16, "RCS Specific Activity," limits. For accidents that assume fueldamage, the primary coolant activity is a function of the amount ofactivity released from the damaged fuel. The dose consequences ofthese events are within the limits of GDC 19 (Ref. 2), 10 CFR 41-050.67(Ref. 3) or the NRC approved licensing basis (e.g., a small fraction atthese-limit and RG 1.183 (Ref. 7).Steam generator tube integrity satisfies Criterion 2 of10 CFR 50.36(c)(2)(ii).LCO The LCO requires that SG tube integrity be maintained. The LCO alsorequires that all SG tubes that satisfy the repair criteria be plugged inaccordance with the Steam Generator Program.During an SG inspection, any inspected tube that satisfies the SteamGenerator Program repair criteria is removed from service by plugging.If a tube was determined to satisfy the repair criteria but was notplugged, the tube may still have tube integrity.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8A Page 100 of 105 Steam Generator (SG) Tube IntegrityB 3.4.17BASESLCO Structural integrity requires that the primary membrane stress intensity(continued) in a tube not exceed the yield strength for all ASME Code, Section Ill,Service Level A (normal operating conditions) and Service Level B(upset or abnormal conditions) transients included in the designspecification. This includes safety factors and applicable design basisloads based on ASME Code,Section III, Subsection NB (Ref. 4) andDraft Regulatory Guide 1.121 (Ref. 5).The accident induced leakage performance criterion ensures (a) thatthe primary to secondary LEAKAGE caused by a design basisaccident, other than a SGTR, is within the accident analysisassumptions, and (b) that the primary to secondary LEAKAGE will notexceed 40.75 gpm total for all four peW SGs (except for specific types ofdegradation at specific locations where the NRC has approved greateraccident induced leakage) to ensure that the potential for inducedleakage during severe accidents will be maintained at a level that willnot increase risk. The accident analysis for the SLB event, the SGTRevent and other events resulting in steam release to the atmosphereassumes that accident induced leakage does not exceed 10 gpm in thefaulted SG and 0.1 gpmn in each intact SG. For the faulted SG in theSLB event, 10.5 gpm is the accid.nt induced leakage limit, of Which nomr..e thaRn 1 gpM can come from sour.es not specifically exempted bythe NRC fromn this I gpmn limit. The accident analyses for events otherthan SGTR aRd SLBI assume that leakage does not exceed 0.75 gpmtotal under accident conditions. The accident induced leakage rateincludes any primary to secondary LEAKAGE existing prior to theaccident in addition to primary to secondary LEAKAGE induced duringthe accident.The operational LEAKAGE performance criterion provides anobservable indication of SG tube conditions during plant operation.The limit on operational LEAKAGE is contained in LCO 3.4.13, "RCSOperational LEAKAGE," and limits primary to secondary LEAKAGEthrough any one SG to 150 gallons per day. This limit is based on theassumption that a single crack leaking this amount would notpropagate to a SGTR under the stress conditions of a LOCA or a mainsteam line break. If this amount of LEAKAGE is due to more than onecrack, the cracks are very small, and the above assumption isconservative.APPLICABILITY Steam generator tube integrity is challenged when the pressuredifferential across the tubes is large. Large differential pressuresacross SG tubes can only be experienced in MODE 1, 2, 3, or 4.RCS conditions are far less challenging in MODES 5 and 6 than duringMODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondarydifferential pressure is low, resulting in lower stresses and reducedpotential for LEAKAGE.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8A Page 102 of 105 Steam-Generator (SG) Tube IntegrityB 3.4.17BASESSURVEILLANCE In addition, Specification 5'.5.9.contains prescriptive requirementsREQUIREMENTS concerning inspection intervals to provide added assurance that the SG(continued) performance criteria will be met between scheduled inspections.SR 3.4.17.2During an SG inspection, any inspected tube that satisfies the SteamGenerator Program repair criteria is removed from service by plugging.The tube repair criteria delineated in Specification 5.5.9 are intended toensure that tubes accepted for continued service satisfy the SGperformance criteria with allowance for error in the flaw sizemeasurement and for future flaw growth. In addition, the tube repaircriteria, in conjunction with other elements of the Steam GeneratorProgram, ensure that the SG performance criteria will continue to bemet until the next inspection of the subject tube(s). Reference Iprovides guidance for performing operational assessments to verify thatthe tubes remaining in service will continue to meet the SG performancecriteria.The Frequency of prior to entering MODE 4 following a SG inspection.ensures that the Surveillance has been completed and all tubes meetingthe repair criteria are plugged prior to subjecting the SG tubes tosignificant primary to secondary pressure differential.REFERENCES 1. NEI 97-06, "Steam Generator Program Guidelines."2. 10 CFR 50 Appendix A, GDC 19 1_999.3. 10 CFR -I050.67.4. ASME Boiler and Pressure Vessel Code,Section III, SubsectionNB.5. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded SteamGenerator Tubes," August 1976.6. EPRI, "Pressurized Water Reactor Steam Generator ExaminationGuidelines."7. Regulatory Guide 1.183, July 2000.DIABLO CANYON -UNITS 1 & 2Rev8A Page105 of 105 Containment Isolation ValvesB 3.6.3BASESBACKGROUND(continued) ,Containment Puroe System (48 inch ourae valves)The Containment Purge System operates to supply outside air into thecontainment for ventilation and cooling or heating needed forprolonged containment access following.a shutdown and duringrefueling. The system may also be u.Sed to r-,ed-ucl e the concentration o,noble gases wfithinq containmnent priorto and duFrig personn~el accr-ess.The supply and exhaust lines each contain two isolation valves. The418 inch Containment Purge valves are qualified for: automnatic closurFefroM their open position under DBA c.nditions. The safety analysesassume that the 48-inch supply and exhaust line valves are closed atthe start of the DBA. Therefore, the 48 inch Containment Purge supplyand exhaust isolation valves are ...m.lly maintained sealed closed inMODES 1, 2, 3, and 4 to ensure the containment boundary ismaintained. The Purge Supply and Exhaust .selationvalves are supplied with an internalrblock which prevents opening thevalve beyond 80 degrees. This block was provided by the mnanufadtureto allow lim.iting the valve's opening. GCalcuations peormed duringqualification to B-ranc~h Technical Position CSB3 6 41 showed the block tobe unnecessary to assure clsu-re time within 2 seconds under DBAconditions (SSER 9, Juno 1980 and M 661). Adjustmntsof this block to values greater than or less than 80 degrees will nota.... Mnc valve s ability W. Gies., Tis des:gn ass .r.s that conEtanmentboundary is m.aintained. These valves may be opened as necessarya. Reduce noble gases within containment prior to and duringpersonnel access, andb. Mitigate the effects of controller leakage and other sources whichmay effect the habitabiity of the containment for personnelOperation in Modes 1, 2, 3, o .with the 48 inch purge valves or the12 incnh vacuu/pressure relief valves open providing a flow path islimnited t o Fnomoe than 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months per calendar year-.Containment Pressure/Vacuum Relief (12 inch isolation valves)rhThe Containment Pressure/Vacuum Relief valves are operated asnecessary to:a. Reduce the concentration of noble gases within containment priorto and during personnel access, andb. Equalize containment internal and external pressures.Since the 12 inch Containment Pressure/Vacuum Relief valves aredesigned to meet the requirements for automatic containment isolationwithin 5 seconds if mechanical blocks are installed to prevent openingmore then 500, these valves may be opened as needed in MODES 1,2, 3, and 4.(continued)DIABLO CANYON -UNITS I & 2Rev 8C Page 13 of 50 Containment Isolation ValvesB 3.6.3BASES (continued)APPLICABLE The containment isolation valve LCO was derived from theSAFETY assumptions related to minimizing the loss of reactor coolant inventoryANALYSES and establishing the containment boundary during major accidents. Aspart of the containment boundary, containment isolation valveOPERABILITY supports leak tightness of the containment. Therefore,the safety analyses of any event requiring isolation of containment isapplicable to this LCO.The DBA that results in a release of radioactive material withincontainment in MODES 1, 2, 3, or 4 is a loss of coolant accident(LOCA) (Ref. 1). In the analyses for this accident, it is assumed thatcontainment isolation valves are either closed or function to closewithin the required isolation time following event initiation. Thisensures that potential paths to the environment through containmentisolation valves (including the Containment Purge, and ContainmentVacuum/Pressure Relief valves) are minimized. The safety analysesassume that the 48 inch purge valves are closed at event initiation. Ifthe 48 inch ContaiRme t Purge supply and exhaust valVes close within2 seccnds and the 12 inch pressure/vacuum relief valves close within5 seconds after the DBA initiation, the safety analysis shows thatoffsite dose release will be less than 10 CFR-41O-50.67 guidelines.The DBA analysis assumes that containment isolation occurs andleakage is prevented except for the design leakage rate, La.The LOCA offsite dose analysis assumes leakage from thecontainment at a maximum leak rate of 0.10 percent of thecontainment volume per day for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and at 0.05 percentof the containment volume per day for the duration of the accident.The single failure criterion required to be imposed in the conduct ofplant safety analyses was considered in the original design of the41 R inc Centainment Purge supply and exhaust and the 12 inchContainment Pressure/Vacuum Relief valves. Two valves in seriesprovide assurance that the flow paths can be isolated even if a singlefailure occurred. The inboard and outboard isolation valves areprovided with diverse power sources and are pneumatically operatedspring closed valves that will fail closed on the loss of power or air.The 48 inGh Purge supply and exhaust and 12 inchContainment PressureNacuum Relief valves are able to close in theenvironment following a LOCA. Therefore, each of the GGstaimentPurge supply and exhaust and Containment Vacuum/pressure Reliefvalves may be opened to provide a flow path. The 48il4ehContainment Purge supply and exhaust valves 12-inchvacuum/pressure relief valves may be open no more then 200 hour0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months sper calendar year-while in MODES 1, 2, 3, and 4. Additionally, enly tweof the three flow paths (containment purge supply and exhaust, andcontainmnent vacuum/Wpressure relieD may be open at one time.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8C Page 14 of 50 Containment Isolation ValvesB 3.6.3BASESAPPLICABLESAFETYANALYSES(continued)The system is designed to preclude a single failure from compromisingthe containment boundary as long as the system is operated inaccordance with the subject LCO.The 48 inch Containment Purge supply and exhaust valves may beunable to close in the environment following a LOCA in sufficient timeto support DBA acceptance criteria. Therefore, each of the purgevalves is required to remain sealed closed during MODES 1. 2. 3. and4. In this case, the single failure criterion remains applicable to thecontainment purge valves due to failure in the control circuit associatedwith each valve. Again, the purge system valve design precludes asingle failure from compromising the containment boundary as long asthe system is operated in accordance with the subject LCO.The containment isolation valves satisfy Criterion 3 of1 OCFR50.36(c)(2)(ii).LCOContainment isolation valves form a part of the containment boundary.The containment isolation valves' safety function is related tominimizing the loss of reactor coolant inventory and establishing thecontainment boundary during a DBA. The automatic power operatedisolation valves are required to have isolation times within limits and toactuate on an automatic isolation signal. The 48 inch ContainmentPurge supply and exhaust valves aid 4he must be sealed closed duringMODES 1. 2, 3, and 4. The Pressure/Vacuum Relief valves must haveblocks installed to prevent full opening. These blocked valves alsoactuate on an automatic isolation signal. The valves covered by thisLCO are listed along with their associated stroke times in PlantProcedure AD13.DC1 (Ref. 5).Normally closed passive containment isolation valves/devices areconsidered OPERABLE when manual valves are closed, automaticvalves are de-activated and secured in their closed position, blindflanges are in place, and closed systems are intact. These passiveisolation valves/devices are those listed in Reference 5.Containment Purge supply and exhaust valves, and ContainmentPressure/Vacuum Relief valves with resilient seals must meetadditional leakage rate surveillance frequency requirements. The othercontainment isolation valve leakage rates are addressed by LCO 3.6.1,"Containment."This LCO provides assurance that the containment isolation valves andthe Containment Purge supply and exhaust, and ContainmentPressure/Vacuum Relief valves will perform their designed safetyfunction to minimize the loss of reactor coolant inventory and establishthe containment boundary during accidents.IAPPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactivematerial to containment. In MODES 5 and 6, the probability andconsequences of these events are reduced due to the pressure andtemperature limitations of these MODES. Therefore, the containmentDIABLO CANYON -UNITS 1 & 2Rev 8C Page 15 of 50 Containment Isolation ValvesB 3.6.3isolation valves are not required to be OPERABLE in MODE 5. Therequirements for containment isolation valves during MODE 6 areaddressed in LCO 3.9.4, "Containment Penetrations."(continued)(Spillover from previous page)DIABLO CANYON -UNITS 1 & 2Rev8C Page 15of 50 Containment Isolation ValvesB 3.6.3BASES (continued)ACTIONS The ACTIONS are modified by a Note allowing penetration flow paths,except 48-inch purge valve flow paths. to be unisolated intermittentlyunder administrative controls. These administrative controls consist ofstationing a person at the valve controls, who is in continuouscommunication with the control room. In this way, the penetration canbe rapidly isolated when a need for containment isolation is indicated.Due to the size of the containment purge line penetration and the factthat those penetrations exhaust directly from the containmentatmosphere to the environment, the penetration flow path containinqthese valves may not be opened under administrative controls. Asingle purge valve in a penetration flow path may be opened to effectrepairs to an inoperable valve, as allowed by SR 3.6.3.1.This Note alsolimits operation of the nrE)mally isolated Containment Supply andExhaust valves (2 penetration flow paths) and the VacUUM/PccsSUreRelief valves (I penetration flow path) to no more thanP 2 of 13penetrationR flow paths open at one time.A second Note has been added to provide clarification that, for thisLCO, separate Condition entry is allowed for each penetration flowpath. This is acceptable, since the Required Actions for eachCondition provide appropriate compensatory actions for eachinoperable containment isolation valve. Complying with the RequiredActions may allow for continued operation, and subsequent inoperablecontainment isolation valves are governed by subsequent Conditionentry and application of associated Required Actions.The ACTIONS are further modified by a third Note, which ensuresappropriate remedial actions are taken, if necessary, if the affectedsystems are rendered inoperable by an inoperable containmentisolation valve.In the event the containment isolation valve leakage results inexceeding the overall containment leakage rate acceptance criteria,Note 4 directs entry into the applicable Conditions and RequiredActions of LCO 3.6.1.Plant Procedure AD13.DC1 Attachment 7.7 (Ref. 5) provides theapplicable CONDITION to enter for each containment isolation valve ifthe valve is inoperable.A.1 and A.2In the event one containment isolation valve in one or more penetrationflow paths requiring isolation following a DBA is inoperable except forContainment Purge supply and exhaust, and ContainmentPressureNacuum Relief isolation valve leakage not within limit, theaffected penetration flow path must be isolated. The method ofisolation must include the use of at least one isolation barrier that(continued)DIABLO CANYON -UNITS I & 2Rev8C Page16of50 Containment Isolation ValvesB 3.6.3BASESACTIONSD.1, D.2, and D.3 (continued)condition, it is prudent to perform the SR more often. Therefore, aFrequency of once per 92 days was chosen and has been shown to beacceptable based on operating experience.Required Action D.2 is modified by two Notes. Note 1 applies to valvesand blind flanges located in high radiation areas and allows thesedevices to be verified closed by use of administrative means. Allowingverification by administrative means is considered acceptable, sinceaccess to these areas is typically restricted. Note 2 applies to isolationdevices that are locked, sealed, or otherwise secured in position andallows these devices to be verified closed by use of administrativemeans. Allowing verification by administrative means is consideredacceptable, since the function of locking, sealing, or securingcomponents is to ensure that these devices are not inadvertentlyrepositioned. Therefore, the probability of misalignment of thesevalves, once they have been verified to be in the proper position, issmall.E.1 and E.2If the Required Actions and associated Completion Times are not met,the plant must be brought to a MODE in which the LCO does not apply.To achieve this status, the plant must be brought to at least MODE 3within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowedCompletion Times are reasonable, based on operating experience, toreach the required plant conditions from full power conditions in anorderly manner and without challenging plant systems.SURVEILLANCEREQUIREMENTSSR 3.6.3.1Ne4-Used Each 48 inch Containment Purge supply and exhaust valve isrequired to be verified sealed closed. This Surveillance is designed toensure that a gross breach of containment is not caused by aninadvertent or spurious opening of a Containment Purge valve. Thesevalves are assumed to be closed at the start of a DBA. Therefore.these valves are required to be in the sealed closed position duringMODES 1, 2, 3, and 4. A Containment Purge valve that is sealedclosed must have motive power to the valve operator removed. Thiscan be accomplished by de-energizing the source of electric power orby removing the air supply to the valve operator. In the event the purgevalve leakage requires entry into Condition D, the surveillance permitsopening one purge valve in a penetration flow path to perform repairs.The Surveillance Frequency is controlled under the SurveillanceFrequency Control Program.SR 3.6.3.2This SR ensures that the 48 inch Containment Purge supply dexhaust-and the 12 inch Containment PressureNacuum Relief valvesare closed as required or, if open, open for an allowable reason. If aDIABLO CANYON -UNITS 1 & 2Rev 8C Page 21 of 50 Containment Isolation ValvesB 3.6.3purge or pressure relief valve is open in violation of this SR, the valveis considered inoperable. If the inoperable valve is not otherwiseknown to have excessive leakage when closed, it is not considered tohave leakage outside of limits. The SR is not required to be met whenthe CoRtainment Purge supply and exhaust or Containment PressureRelief valves are open for the reasons stated. The valves may beopened for pressure control, ALARA or air quality considerations forpersonnel entry, or for Surveillances that require the valves to be open.The Containment Purge supply and exhaust or ContainmentPressureNacuum Relief valves are capable of closing in the(continued)(Spillover from previous page.)DIABLO CANYON -UNITS I & 2Rev 8C Page 21 of 50 Containment Spray and Cooling SystemsB 3.6.6BASESBACKGROUND Containment Spray System (continued)In the recirculation mode of operation, containment spray is supplied bymanual realignment of the residual heat removal (RHR) pumps afterthe RWST is empty.The Containment Spray System provides a spray of cold borated watermixed with sodium hydroxide (NaOH) from the spray additive tank intothe upper regions of containment to reduce the containment pressureand temperature, and to reduce fission products from the containmentatmosphere during a DBA. The RWST solution temperature is animportant factor in determining the heat removal capability of theContainment Spray System during the injection phase. In therecirculation mode of operation, heat is removed from the containmentsump water by the RHR heat exchangers. Each train of theContainment Spray System provides adequate spray coverage to meetthe system design requirements for containment atmospheric heatremoval.The Spray Additive System injects an NaOH solution into the spray.The resulting alkaline pH of the spray enhances the ability of the sprayto scavenge fission products from the containment atmosphere. TheNaOH added in the spray also ensures an alkaline pH for the solutionrecirculated in the containment sump. The alkaline pH of thecontainment sump water maximizes the retention of iodine andminimizes the occurrence of chloride and caustic stress corrosion onmechanical systems and components exposed to the fluid.The Containment Spray System is actuated either automatically by acontainment High-High pressure signal or manually. If an "S" signal ispresent, the High-High pressure signal automatically starts the twocontainment spray pumps, opens the containment spray pumpdischarge valves, opens the spray additive tank outlet valves, initiates aphase "B" isolation signal, and begins the injection phase. A manualactuation of the Containment Spray System will begin the samesequence and can be initiated by operator action from the main controlboard. The injection phase of containment spray continues until anRWST Low-Low level alarm is received. The Low-Low level alarm forthe RWST signals the operator to manually secure the system. Afterre-alignment of the RHR system to the containment recirculation sump,the associated RHR spray header isolation valve may-beis opened toallow continued spray operation of one train of spray utilizing the RHRpump to supply flow. The LOCA dose analysis takes credit for thismanual initiation of Containment Spray during recirculation to takeplace within 12 minutes following the termination of Containment Sprayduring the iniection phase.Containment Spray is fiat required to be actuated during therecirculation phase of a LOCA, but may be at the discretion oGthe Technical Support Center. Containment Spray operation (injectionplus recirculation) is credited until 6.25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months following initiation of aDIABLO CANYON -UNITS 1 & 2Rev8C Page 34 of 50 Containment Spray and Cooling SystemsB 3.6.6LOCA. During the recirculation phase of a LOCA, the ContainmentSpray System must be capable of(continued)(Spillover from previous page.)DIABLO CANYON -UNITS I & 2Rev 8C Page 34 of 50 Containment Spray and Cooling SystemsB 3.6.6BASESAPPLICABLESAFETYANALYSES(continued)Analyses and evaluation show that containment spray is not requiredduring the recirculation phase of a LOCA for containment pressure andtemperature control (Ref. 7). However, for dose consequences.containment spray is required during the recirculation phase of a LOCAfor removing radioactive iodine and particulates from the containmentatmosphere.If only one RHR pump is available during the recirculation phase of aLOCA, it may not be possible to obtain significant containment spraywithout closing valves 8809A or B. If recirculation spray is used withonly one train of RHR in operation, ECCS flow to the reactor will bereduced, but analysis has shown that the flow to the reactor in thissituation is still in excess of that needed to supply the required corecooling.The effect of an inadvertent containment spray actuation has beenanalyzed. An inadvertent spray actuation results in a -1.8.0 psidcontainment pressure decrease and is based on a sudden coolingeffect of 70'F in the interior of the leak tight containment. Additionaldiscussion is provided in the Bases for LCO 3.6.4.The modeled Containment Spray System actuation from thecontainment analysis is based on a response time associated withexceeding the containment High-High pressure setpoint to achievingfull flow through the containment spray nozzles. The ContainmentSpray System total response time includes diesel generator (DG)startup (for loss of offsite power),(continued)DIABLO CANYON -UNITS I & 2Rev 8C Page 37 of 50 Containment Spray and Cooling SystemsB 3.6.6BASESAPPLICABLE sequenced loading of equipment, containment spray pump startup, andSAFETY spray line filling (Ref. 4). The CFCUs performance for post accidentANALYSES conditions is given in Reference 4. The result of the analysis is that(continued) each train (two CFCUs) combined with one train of containment spraycan provide 100% of the required peak cooling capacity during the postaccident condition.The modeled Containment Cooling System actuation from thecontainment analysis is based upon a response time associated withexceeding the containment High-High pressure setpoint to achievingfull Containment Cooling System air and safety grade cooling waterflow. The Containment Cooling System total response time includessignal delay, DG startup (for loss of offsite power), and componentcooling water pump startup times.The Containment Spray System and the Containment Cooling Systemsatisfies Criterion 3 of 10CFR50.36(c)(2)(ii).LCO During a DBA LOCA, a minimum of two CFCUs and one containmentspray train are required to maintain the containment peak pressure andtemperature below the design limits (Refs. 4). Additionally, onecontainment spray train is also required to remove radioactive iodineand particulates from the containment atmosphere and maintainconcentrations below those assumed in the safety analysis. To ensurethat these requirements are met, two containment spray trains and theCFCU system consisting of four CFCUs or thre6 CFCUs each suppliedby a different vital bus must be OPERABLE. Therefore, in the event ofan accident, at least one train of containment spray and two CFCUsoperate, assuming the worst case single active failure occurs. EachContainment Spray train typically includes a spray pump, sprayheaders, nozzles, valves, piping, instruments, and controls to ensurean OPERABLE flow path capable of taking suction from the RWSTupon an ESF actuation signal. Upon actuation of the RWST Low-Lowalarm, the containment spray pumps are secured. Containment sprayGeuldis then be-supplied as Fequired by an RHR pump taking suctionfrom the containment sump for a total spray operation (injection andrecirculation) of 6.25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months .Each CFCU includes cooling coils, dampers, fans, instruments, andcontrols to ensure an OPERABLE flow path.APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactivematerial to containment and an increase in containment pressure andtemperature requiring the operation of the containment spray trainsand CFCUs.In MODES 5 and 6, the probability and consequences of these eventsare reduced due to the pressure and temperature limitations of theseMODES. Thus, the Containment Spray System and the ContainmentCooling System are not required to be OPERABLE in MODES 5 and 6.DIABLO CANYON -UNITS I & 2Rev 8C Page 38 of 50 Containment Spray and Cooling SystemsB 3.6.6(continued)DIABLO CANYON -UNITS 1 & 2Rev 8C Page 38 of 50 MSIVsB 3.7.2BASESAPPLICABLE c. A break downstream of the MSIVs will be isolated by the closure ofSAFETY the MSIVs.ANALYSES(continued) d. Following a steam generator tube rupture, closure of the MSIVsisolates the ruptured steam generator from the intact steamgenerators to minimize radiological releases.e. The MSIVs are also utilized during other events such as afeedwater line break. This event is less limiting so far as MSIVOPERABILITY is concerned.The MSIVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(i).LCO This LCO requires that four MSIVs in the steam lines be OPERABLE.The MSIVs are considered OPERABLE when the isolation times arewithin limits, and they close on an isolation actuation signal.This LCO provides assurance that the MSIVs will perform their designsafety function to mitigate the consequences of accidents that couldresult in offsite exposures comparable to the 10 CFR -0050.67 (Ref. 4)limits or the NRC staff approved licensing basis.APPLICABILITY The MSIVs must be OPERABLE in MODE 1, and in MODES 2 and 3except when closed and de-activated (vented or prevented fromopening), when there is significant mass and energy in the RCS andsteam generators. When the MSIVs are closed, they are alreadyperforming the safety function.In MODE 4, the steam generator energy is low, thus OPERABILITY inMODE 4 is not required.In MODE 5 or 6, the steam generators do not contain much energybecause their temperature is below the boiling point of water; therefore,the MSIVs are not required for isolation of potential high energysecondary system pipe breaks in these MODES.ACTIONS A.1With one MSIV inoperable in MODE 1, action must be taken to restoreOPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Some repairs to the MSIV can bemade with the unit hot. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is reasonable,considering the low probability of an accident occurring during this timeperiod that would require a closure of the MSIVs.The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is greater than that normally allowed forcontainment isolation valves because the MSIVs are valves that isolatea closed system penetrating containment. These valves differ fromother containment isolation valves in that the closed system providesan additional means for containment isolation.(continued)IDIABLO CANYON -UNITS I & 2Rev8D Page9of87 MSIVsB 3.7.2BASESSURVEILLANCEREQUIREMENTSSR 3.7.2.1 (continued)analyses. This Surveillance is normally performed upon returning theunit to operation following a refueling outage. The MSIVs should notbe tested at power, since even a part stroke exercise increases the riskof a valve closure when the unit is generating power.As the MSIVs are not tested at power, they are exempt from the ASMECode,Section XI (Ref. 5), requirements during operation in MODE Ior 2.The Frequency is in accordance with the Inservice Testing Program.This test may be conducted in MODE 3 with the unit at operatingtemperature and pressure. This SR is modified by a Note that allowsentry into and operation in MODE 3 prior to performing the SR.However, the test is normally conducted in MODE 5 as permitted bythe cold shutdown frequency justification provided in the InserviceTesting Program (IST) and as permitted by Reference 6,Subsection ISTC-3521 (c).SR 3.7.2.2This SR verifies that each MSIV can close on an actual or simulatedactuation signal. This Surveillance is normally performed uponreturning the plant to operation following a refueling outage. TheSurveillance Frequency is based on operating experience, equipmentreliability, and plant risk and is controlled under the SurveillanceFrequency Control Program.REFERENCES 1. FSAR, Section 10.3.2. FSAR, Section 6, Appendix 6.2 D.3. FSAR, Section 15.4.2.4. 10 CFR I4--.14,50.67.5. ASME, Boiler and Pressure Vessel Code,Section XI.6. ASME Code for Operation and Maintenance of Nuclear PowerPlants, 2001 Edition including 2002 and 2003 Addenda.DIABLO CANYON -UNITS I & 2Rev 8D Page 11 of 87 CRVSB 3.7.10BASESBACKGROUND Redundant supply and recirculation trains provide the required filtration(continued) should an excessive pressure drop develop across the other filter train.Normally open isolation dampers are arranged in series pairs so thatthe failure of one damper to shut will not result in a breach of isolation.The CRVS is designed in accordance with Seismic Category lrequirements.The CRVS is designed to maintain a habitable environment in the CREfor the duration of the most severe Design Basis Accident (DBA)without exceeding a 5 rem wheGe-bedyTEDE dose or its equivalent toany part of the body.APPLICABLE The CRVS components are arranged in redundant, safety relatedSAFETY ventilation trains. The location of components and ducting within theANALYSES CRE ensures an adequate supply of filtered air to all areas requiringaccess. The CRVS provides airborne radiological protection for theCRE occupants, as demonstrated by the CRE occupant dose analysesfor the most limiting design basis accident, fission product releasepresented in the FSAR, Chapter 15 (Ref. 2).There are no offsite or onsite hazardous chemicals that would pose acredible threat to control room habitability. Consequently, engineeredcontrols for the control room are not required to ensure habitabilityagainst a hazardous chemical threat. The amount Of CRE unfilteredinleakage is not incorporated into PG&E's hazardous chemicalassessment.The evaluation of -a smoke challenge demonstrated that smoke will notresult in the inability of the CRE occupants to control the reactor eitherfrom the control room or from the remote shutdown panels (Ref. 1).The assessment verified that a fire or smoke event anywhere within theplant would not simultaneously render the Hot Shutdown Panel (HSP)and the CRE uninhabitable, nor would it prevent access from the CREto the HSP in the event remote shutdown is required. No CRVSautomatic actuation is required for hazardous chemical releases orsmoke and no Surveillance Requirements are required to verifyoperability in cases of hazardous chemicals or smoke.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8D Page 54 of 87 CRVSB 3.7.10BASESAPPLICABLE The worst case single active failure of a component of the CRVS,SAFETY assuming a loss of offsite power, does not impair the ability of theANALYSES system to perform its design function.(continued) The CRVS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).LCO Two independent and redundant CRVS trains are required to beOPERABLE to ensure that at least one is available if a single activefailure disables the other train. The redundant train means a secondtrain from the other unit (Ref. 5). Total system failure, such as from aloss of both ventilation trains or from an inoperable CRE boundary,could result in exceeding a dose of 5 rem whTlebodyTEDE er-its,equivalent to any part of the body to the CRE occupants in the event ofa large radioactive release.Each CRVS train is considered OPERABLE when the individualcomponents necessary to limit CRE occupant exposure areOPERABLE. A CRVS train is OPERABLE when the associated:a. main supply fan (one), filter booster fan (one) and pressurizationfan (one) are OPERABLE;b. HEPA filters and charcoal adsorbers are not excessively restrictingflow, and are capable of performing their filtration functions; andc. Ductwork, valves, and dampers are OPERABLE, and aircirculation can be maintained.In order for the CRVS trains to be considered OPERABLE, the CREboundary must be maintained such that the CRE occupant dose from alarge radioactive release does not exceed the calculated dose in thelicensing basis consequence analyses for DBAs. In the event of aninoperable CRE boundary in MODES 1, 2, 3, or 4, mitigating actionsare required to ensure CRE occupants are protected from hazardouschemicals and smoke.DCPP does not have CRVS automatic actuation for hazardouschemicals or smoke. Current practices at DCPP do not utilizechemicals in sufficient quantity to present a chemical hazard to thecontrol room. Smoke is not considered in the DCPP safety analyses.Therefore, there are no specific limits at DCPP for hazardouschemicals or smoke.(continued)DIABLO CANYON -UNITS I & 2Rev 8D Page 55 of 87 CRVSB 3.7.10BASES (continued)APPLICABILITY In MODES 1, 2, 3, 4, 5, and 6, and during movement of recentlyirradiated fuel assemblies (i.e., fuel that has occupied part of a criticalreactor core within the previous 1-0072 hours) the CRVS must beOPERABLE to ensure that the CRE will remain habitable during andfollowing a DBA or the release from the rupture of an outside wastegas tank.During movement of recently irradiated fuel assemblies, the CRVSmust be OPERABLE to cope with the release from a fuel handlingaccident involving handling recently irradiated fuel.CRVS OPERABILITY requires that for MODE 5 and 6 and duringmovement of recently irradiated fuel assemblies in either unit, whenthere is only one OPERABLE train of CRVS, the OPERABLE CRVStrain must be capable of being powered from an OPERABLE dieselgenerator that is directly associated with the bus which is energizingthe OPERABLE CRVS train. This is an exception to LCO 3.0.6.ACTIONS The ACTIONS are modified by a NOTE that states that ACTIONSapply simultaneously to both units. The CRVS is common to bothunits.A.1When one CRVS train is inoperable for reasons other than aninoperable CRE boundary, action must be taken to restore OPERABLEstatus within 7 days. In this Condition, the remaining OPERABLECRVS train is adequate to perform the CRE occupant protectionfunction. However, the overall reliability is reduced because a singlefailure in the OPERABLE CRVS train could result in loss of CRVSfunction. The 7 day Completion Time is based on the low probability ofa DBA occurring during this time period, and ability of the remainingtrain to provide the required capability.B.1, 8.2, and B.3The CRE boundary is inoperable if unfiltered inleakage past the CREboundary can result in CRE occupant radiological dose greater thanthe calculated dose of the licensing basis analyses of DBAconsequences (allowed to be up to 5 rem whole body Or its equival ,nto any o-f the body"TEDE).In the event of an inoperable CRE boundary in MODES 1, 2, 3, or 4,action must be initiated to implement mitigating actions to lessen theeffect on CRE occupants from the potential hazards of a radiological orchemical event or a challenge from smoke. Actions must be takenwithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to verify that in the event of a DBA, the mitigatingactions will ensure that CRE occupant radiological exposures will notexceed the calculated dose of the licensing basis analyses of DBAconsequences, and that CRE occupants are protected from potentialsmoke and chemical hazards.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8D Page 57 of 87 CRVSB 3.7.10BASESSURVEILLANCE SR 3.7.10.3REQUIREMENTS This SR verifies that the required CRVS testing is performed in(continued) accordance with the Ventilation Filter Testing Program (VFTP). TheCRVS filter tests are in accordance with ANSI N510-1980 (Ref. 3).The VFTP includes testing the performance of the HEPA filter,charcoal adsorber efficiency, minimum flow rate, and the physicalproperties of the activated charcoal. Specific test Frequencies andadditional information are discussed in detail in the VFTP.SR 3.7.10.4This 'SR verifies that each CRVS train automatically starts andoperates in the pressurization mode on an actual or simulatedactuation signal generated from a Phase "A" Isolation. TheSurveillance Frequency is based on operating experience, equipmentreliability, and plant risk and is controlled under the SurveillanceFrequency Control Program.SR 3.7.10.5This SR verifies the OPERABILITY of the CRE boundary by testing forunfiltered air inleakage past the CRE boundary and into the CRE. Thedetails of the testing are specified in the Control Room EnvelopeHabitability Program. Any changes to the most limiting configuration ofthe CRVS testing alignment for determining unfiltered air inleakagepast the CRE boundary into the CRE must be made using aconservative decision making process (References 11-13).The CRE is considered habitable when the radiological dose to CREoccupants calculated in the licensing basis analyses of DBAconsequences is no more than 5 rem whole body Or its equivai.. t toany of th b and the CRE occupants are protected fromhazardous chemicals and smoke. For DCPP, there is no CRVSautomatic actuation for hazardous chemical releases or smoke andthere are no CRVS Surveillance Requirements that verify operability incases of hazardous chemicals or smoke. This SR verifies that theunfiltered air inleakage into the CRE is no greater than the flow rateassumed in the licensing basis analyses of DBA consequences. Whenunfiltered air inleakage is greater than the assumed flow rate,Condition B must be entered. Required Action B.3 allows time torestore the CRE boundary to OPERABLE status provided mitigatingactions can ensure that the CRE remains within the licensing basishabitability limits for the occupants following an accident.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8D Page 60 of 87 ABVSB 3.7.12BASESBACKGROUND(continued)The ABVS is discussed in the FSAR, Sections 9.4 2, and 15.5 (Refs. 1,and 2, respectively) since it may be used for normal, as well as postaccident, ventilation and atmospheric cleanup functions. The primarypurpose of the single manually initiated heater is to maintain therelative humidity at an acceptable level, consistent with iodine removalefficiencies per ASTM D 3803-1989 (Ref. 3). There is no redundantheater since the failure of the charcoal adsorber and heater train wouldconstitute a second failure in addition to the RHR pump seal failureassumed in conjunction with a LOCA (Ref.7). The heaters are notrequired for ABVS operability.APPLICABLESAFETYANALYSESThe design basis of the ABVS is established by the large break LOCA.The system evaluation assumes a passive failure of the ECCS outsidecontainment, such as an RHR pump seal failure, during therecirculation mode. In such a case, the system limits radioactiverelease to within the 10 CFR 41-050.67 (Ref. 5) limits. The analysis ofthe effects and consequences of a large break LOCA is presented inReference 2. The ABVS also functions, following a LOCA, in thosecases where the ECCS goes into the recirculation mode of long termcooling, to clean up releases of smaller leaks, such as from valve stempacking.The ventilation flow is also required to maintain the temperatures of theoperating ECCS motors within allowable limits. The ventilation functionhas been designed for single failure and the system will continue tofunction to provide its ECCS motor cooling function.Two types of system failures are considered in the accident analysis forradiation release: complete loss of function of one train, and excessiveRHR pump seal LEAKAGE. Either type of failure may result in a lowerefficiency of removal for any gaseous and particulate activity releasedto the ECCS pump rooms following a LOCA.The ABVS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).LCOTwo trains of the ABVS are required to be OPERABLE to ensure thatat least one is available, assuming that a single failure disables theother train coincident with loss of offsite power. Total system failurecould result in the atmospheric release from the ECCS pump roomexceeding 10 CFR-40050.67 limits in the event of a Design BasisAccident (DBA).ABVS is considered OPERABLE when the individual componentsnecessary to maintain the ECCS pump room filtration and temperatureare OPERABLE in both trains.(continued)DIABLO CANYON -UNITS I & 2Rev8D Page 65 of 87 ABVSB 3.7.12BASESSURVEILLANCE SR 3.7.12.6REQUIREMENTS This SR verifies the leak tightness of dampers that isolate flow to the(continued) normally operating filter train. This SR assures that the flow from theauxiliary building passes through the HEPA filter and charcoal adsorberunit when the ABVS Buildings and Safeguards or Safeguards Onlymodes have been actuated coincident with an SI. The SurveillanceFrequency is based on operating experience, equipment reliability, andplant risk and is controlled under the Surveillance Frequency ControlProgram.REFERENCES 1. FSAR, Section 9.4.2.2. FSAR, Section 15.5.3. ASTM D 3803-19894. ANSI N510-19805. 10 CFR 40-.450.67.6. NUREG-0800, Section 6.5.1, Rev. 2, July 1981.7. DCM S-23B, "Main Auxiliary Building Heating and VentilationSystem".DIABLO CANYON -UNITS 1 & 2Rev 8D Page 69 of 87 FHBVSB 3.7.13BASESAPPLICABLESAFETYANALYSES(continued)FHBVS is only required to isolate during fuel handling accidentsinvolving the handling of recently irradiated fuel (i.e., fuel that hasoccupied part of a critical reactor core within the previous4--72 hours). In accordance with assumptions made in the fuelhandling accident analysis, loss of offsite power is not consideredconcurrent with a fuel handling accident. However. less of pe.w.eicnVcoped by the fuel handling accident aRalysis,. To maximize FHBVScapability to mitigate-the consequences of a fuel handling accident, atleast one of the FHBVS trains must be capable of being supplied froman operable emergency diesel generator at all times whenevermovement of recently irradiated fuel is taking place in the spent fuelpool. These assumptions and the analysis follow the guidanceprovided in Regulatory Guide 4-.2-51.183 (Ref. 3).The FHBVS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).ILCOTwo independent and redundant trains of the FHBVS are required tobe OPERABLE to ensure that at least one train is available, assuminga single failure that disables the other train. In accordance withassumptions made in the fuel handling accident analysis, loss of offsitepower is not considered concurrent with a fuel handling accident.However, loss of power is enveloped by the fuel handling accidentaalysi&s-This requires that when two trains of the FHBVS areOPERABLE, at least one train of the FHBVS must be capable of beingpowered from an OPERABLE diesel generator that is directlyassociated with the bus which energizes the FHBVS train. When onlyone train is OPERABLE, an OPERABLE diesel generator must bedirectly associated with the bus which energizes that one OPERABLEFHBVS train. Total system failure could result in the atmosphericrelease from the fuel handling building exceeding the 10 CFR 40050.67(Ref. 4) limits in the event of a fuel handling accident.The FHBVS is considered OPERABLE when the individualcomponents necessary to control releases from fuel handling buildingare OPERABLE in both trains. An FHBVS train is consideredOPERABLE when its associated:a. Exhaust fan is OPERABLE;b. HEPA filter and charcoal adsorber are not excessively restrictingflow, and are capable of performing their filtration function; andc. Ductwork, valves, and dampers are OPERABLE, and aircirculation can be maintained.(continued)DIABLO CANYON -UNITS I & 2Rev 8D Page 71 of 87 FHBVSB 3.7.13BASESSURVEILLANCEREQUIREMENTS(continued)SR 3.7.13.4This SR verifies the integrity of the fuel handling building enclosure.The ability of the fuel handling building to maintain negative pressurewith respect to potentially uncontaminated adjacent areas isperiodically tested to verify proper function of the FHBVS. During thepost accident mode of operation, the FHBVS is designed to maintain aslight negative pressure in the fuel handling building, to preventunfiltered LEAKAGE. The FHBVS is designed to maintain the buildingpressure < -0.125 inches water gauge with respect to atmosphericpressure. The Surveillance Frequency is based on operatingexperience, equipment reliability, and plant risk and is controlled underthe Surveillance Frequency Control Program.SR 3.7.13.5Operation of damper M-29 is necessary to ensure that the systemfunctions properly. The operability of damper M-29 is verified if it canbe closed. The Surveillance Frequency is based on operatingexperience, equipment reliability, and plant risk and is controlled underthe Surveillance Frequency Control Program.REFERENCES1. FSAR, Section 9.4.4.2. FSAR, Section 15.5.3.4.5.6.7.8.9.10.11.Regulatory Guide 4-251.183. July 2000.10 CFRI- 050.67.ASTM D 3802-1989ANSI N510-1980.NUREG-0800, Section 6.5.1, Rev. 2, July 1981.DCM S-23D, "Fuel handling Building HVAC System."Not usedLicense Amendment 184/186, January 3, 2006.PG&E Letter DCL-05-124DIABLO CANYON -UNITS 1 & 2Rev 8D Page 74 of 87 Spent Fuel Storage Pool Water LevelB 3.7.15B 3.7 PLANT SYSTEMSB 3.7.15 Spent Fuel Storage Pool Water LevelBASESBACKGROUNDThe minimum water level in the spent fuel pool meets the assumptionsof iodine decontamination factors following a fuel handling accident.The specified w~ater level shields and minimizes the general area dosewhen the storage racks are filled to their maximum capacity. Thewater also provides shielding during the movement of spent fuel.A general description of the spent fuel pool design is given in theFSAR, Section 9.1.2 (Ref. 1). A description of~the Spent Fuel PoolCooling and Cleanup System is given in the FSAR, Section 9.1.3(Ref. 2). The assumptions of the fuel handling accident are given inthe FSAR, Section 9.1.4.3.4, 15.4.5 and 15.5.22 (Ref. 3).APPLICABLESAFETYANALYSESThe minimum water level in the spent fuel pool meets the assumptionsof the fuel handling accident described in Regulatory Guide 4--251.183(Ref. 4). The resultant 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose per person at the exclusion.area boundary is a small fraction of the 10 CFR 4GQ50.67 (Ref. 5)limits.According to Reference 4, there is 23 ft of water between the top of thedamaged fuel rods and the fuel pool surface during a fuel handlingaccident. With 23 ft of water, the assumptions of Reference 4 can beused directly. Although there are other spent fuel pool elevationswhere fuel handling accidents can occur, the design basis fuel handlingaccident, which uses the conservative assumptions of RG 4-Z51.183, isexpected to be bounding. To add conservatism, the analysis assumesthat all fuel rods of the damaged fuel assembly fail.In practice, the water level maintained for fuel handling provides morethan 23 feet of water over the top of irradiated fuel assemblies seatedin the storage racks. FSAR Section 9.1.4.3.4 requires the water levelprovide a minimum of 8 feet of water shielding during fuel handling.This assures more than 24 feet 6 inches of water shielding over the topof the fuel assemblies in the racks and more than. 23 feet of watershielding over a fuel assembly lying horizontally on top of the racks.The spent fuel pool water level satisfies Criterion 2 of10 CFR 50.36(c)(2)(ii).ILCOThe spent fuel pool water level is required to be > 23 ft over the top ofirradiated fuel assemblies seated in the storage racks. The specifiedwater level preserves the assumptions of the fuel handling accidentanalysis (Ref. 3). As such, it is the minimum required for fuel storageand movement within the fuel storage pool.(continued)DIABLO CANYON -UNITS I & 2Rev 8D Page 76 of 87 Spent Fuel Storage Pool Water LevelB 3.7.15BASES (continued)APPLICABILITYThis LCO applies during movement of irradiated fuel assemblies in thespent fuel pool, since the potential for a release of fission productsexists.ACTIONSA.1IRequired Action A.1 is modified by a Note indicating that LCO 3.0.3does not apply.When the initial conditions for prevention of an accident cannot be met,steps should be taken to preclude the accident from occurring. Whenthe spent fuel pool water level is lower than the required level, themovement of irradiated fuel assembly in the spent fuel pool isimmediately suspended. This does not preclude movement of a fuelassembly to a safe position.If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3would not specify any action. If moving irradiated fuel assemblies whilein MODES 1, 2, 3, and 4, the fuel movement is independent of reactoroperations. Therefore, inability to suspend movement of irradiated fuelassemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCEREQUIREMENTSSR 3.7.15.1This SR is done during the movement of irradiated fuel assemblies asstated in the Applicability. This SR verifies sufficient fuel storage poolwater is available in the event of a fuel handling accident. The waterlevel in the spent fuel pool must be checked periodically. TheSurveillance Frequency is based on operating experience, equipmentreliability, and plant risk and is controlled under the SurveillanceFrequency Control Program.During refueling operations, the level in the spent fuel pool is inequilibrium with the refueling canal, and the level in the refueling canalis checked daily in accordance with SR 3.9.7.1.REFERENCES 1. FSAR, Section 9.1.2.2. FSAR, Section 9.1.3.3. FSAR, Section 9.1.4.3.4, 15.4.5 and 15.5.22.4. Regulatory Guide 41--5.183, July 2000.5. 10 CFR 4-1-.450.67.DIABLO CANYON -UNITS 1 & 2Rev8D Page 77 of 87 Secondary Specific ActivityB 3.7.18B 3.7 PLANT SYSTEMSB 3.7.18 Secondary Specific ActivityBASESBACKGROUNDActivity in the secondary coolant results from steam generator tubeoutleakage from the Reactor Coolant System (RCS). Under steadystate conditions, the activity is primarily iodines with relatively short halflives and, thus, indicates current conditions. During transients, 1-131spikes have been observed as well as increased releases of somenoble gases. Other fission product isotopes, as well as activatedcorrosion products in lesser amounts, may also be found in thesecondary coolant.A limit on secondary coolant specific activity during power operationminimizes releases to the environment because of normal operation,anticipated operational occurrences, and accidents.This limit is lower than the activity value that might be expected from a4--0.75 gpm tube leak (LCO 3.4.13, "RCS Operational LEAKAGE") ofprimary coolant at the limit of 1.0 pCi/gm (LCO 3.4.16, "RCS Specific'Activity"). The steam line failure is assumed to result in the release ofthe noble gas and iodine activity contained in the steam generatorinventory, the feedwater, and the reactor coolant LEAKAGE. Most ofthe iodine isotopes have short half lives, (i.e., < 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months ). Operatingat or below 0.1 pCi/gm ensures that in the event of a DBA, offsitedoses will be less than 10 CFR 4-0050.67_requirements.IIAPPLICABLESAFETYANALYSESThe accident analysis of the main steam line break (MSLB), asdiscussed in the FSAR, Chapter 15 (Ref. 2) assumes the initialsecondary coolant specific activity to have a radioactive isotopeconcentration of 0.10 pCi/gm DOSE EQUIVALENT 1-131. Thisassumption is used in the analysis for determining the radiologicalconsequences of the postulated accident. The accident analysis,based on this and other assumptions, shows that the radiologicalconsequences of an MSLB do not exceed 10 CFR 40050.67 limits(Ref. 1) foF whole body and thyroid dose ra. Ft .With the loss of offsite power, the remaining steam generators areavailable for core decay heat dissipation by venting steam to theatmosphere through the MSSVs and steam generator atmosphericdump valves (ADVs). The Auxiliary Feedwater System supplies thenecessary makeup to the steam generators. Venting continues untilthe reactor coolant temperature and pressure have decreasedsufficiently for the Residual Heat Removal System to complete thecooldown.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8D Page 85 of 87 Secondary Specific ActivityB 3.7.18BASESACTIONS A.1 and A.2 (continued)least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Theallowed Completion Times are reasonable, based on operatingexperience, to reach the required unit conditions from full powerconditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.18.1REQUIREMENTS This SR verifies that the secondary specific activity is within the limitsof the accident analysis. A gamma isotopic analysis of the secondarycoolant, which determines DOSE EQUIVALENT 1-131, confirms thevalidity of the safety analysis assumptions as to the source terms inpost accident releases. It also serves to identify and trend any unusualisotopic concentrations that might indicate changes in reactor coolantactivity or LEAKAGE. The Surveillance Frequency is based onoperating experience, equipment reliability, and plant risk and iscontrolled under the Surveillance Frequency Control Program.REFERENCES 1. 10 CFR 409.450.67.2. FSAR, Chapter 15..IDIABLO CANYON -UNITS I & 2Rev 8D Page 87 of 87 AC Sources -ShutdownB 3.8.2B 3.8 ELECTRICAL POWER SYSTEMSB 3.8.2 AC Sources-ShutdownBASESBACKGROUND A description of the AC sources is provided in the Bases for LCO 3.8.1,"AC Sources -Operating."APPLICABLE The OPERABILITY of the minimum AC sources during MODES 5 andSAFETY 6 and during movement of recently irradiated fuel assemblies ensuresANALYSES that:a. The unit can be maintained in the shutdown or refuelingcondition for extended periods;b. Sufficient instrumentation and control capability is available formonitoring and maintaining the unit status; andc. Adequate AC electrical power is provided to mitigate eventspostulated during shutdown, such as a fuel handling accidentinvolving handling recently irradiated fuel. Due to radioactivedecay, AC electrical power is only required to mitigate fuelhandling accidents involving recently irradiated fuel (i.e., fuelthat has occupied part of a critical reactor core within theprevious 40072 hours).In general, when the unit is shut down, the Technical Specificationsrequirements ensure that the unit has the capability to mitigate theconsequences of postulated accidents. However, assuming a singlefailure and concurrent loss of all offsite or all onsite power is notrequired. The rationale for this is based on the fact that many DesignBasis Accidents (DBAs) that are analyzed in MODES 1, 2, 3, and 4have no specific analyses in MODES 5 and 6. Worst case boundingevents are deemed not credible in MODES 5 and 6 because theenergy contained within the reactor pressure boundary, reactor coolanttemperature and pressure, and the corresponding stresses result in theprobabilities of occurrence being significantly reduced or eliminated,and in minimal consequences. These deviations from DBA analysisassumptions and designrequirements during shutdown conditions areallowed by the LCO for required systems.During MODES 1, 2, 3, and 4, various deviations from the analysisassumptions and design requirements are allowed within the RequiredActions. This allowance is in recognition that certain testing andmaintenance activities must be conducted, provided an acceptablelevel of risk is not exceeded.' During MODES 5 and 6, performance ofa significant number of required testing and maintenance activities isalso required. In MODES 5 and 6, the activities are generally plannedand administratively controlled. Relaxations from MODE 1, 2, 3, and 4LCO requirements are acceptable during shutdown modes based on:a. The fact that time in an outage is limited. This is a risk prudentgoal as well as a utility economic consideration.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8B Page 34 of 90 DC Sources -ShutdownB 3.8.5B 3.8 ELECTRICAL POWER SYSTEMSB 3.8.5 DC Sources-ShutdownBASESBACKGROUNDA description of the DC sources is provided in the Bases for LCO 3.8.4,"DC Sources-Operating."APPLICABLESAFETYANALYSESThe initial conditions of Design Basis Accident and transient analysesin the FSAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assume thatEngineered Safety Feature systems are OPERABLE. The DCelectrical power system provides normal and emergency DC electricalpower for the diesel generators, emergency auxiliaries, and control andswitching during all MODES of operation.The OPERABILITY of the DC subsystems is consistent with the initialassumptions of the accident analyses and the requirements for thesupported systems' OPERABILITY.The OPERABILITY of the minimum DC electrical power sources duringMODES 5 and 6 and during movement of recently irradiated fuelassemblies ensures that:a. The unit can be maintained in the shutdown or refuelingcondition for extended periods;b. Sufficient instrumentation and control capability is available formonitoring and maintaining the unit status; andc. Adequate DC electrical power is provided to mitigate eventspostulated during shutdown, such as a fuel handling accidentinvolving handling recently irradiated fuel. Due to radioactivedecay, DC electrical power is only required to mitigate fuelhandling accidents involving recently irradiated fuel (i.e., fuelthat has occupied part of a critical reactor core within theprevious 4-072 hours).The DC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).LCO The DC electrical power subsystems, each subsystem consisting ofone battery, one battery charger per battery, and the correspondingcontrol equipment and interconnecting class 1 E cabling within thesubsystem, are required to be OPERABLE to support required trains ofthe distribution systems required OPERABLE by LCO 3.8.10,"Distribution Systems-Shutdown." An OPERABLE subsystem consistsof a DC bus connected to a battery with an OPERABLE battery chargerwhich is fed from an OPERABLE AC vital bus (Ref B.3.8.10).(continued)DIABLO CANYON -UNITS I & 2Rev 8B Page 60 of 90 Inverters -ShutdownB 3.8.8B 3.8 ELECTRICAL POWER SYSTEMSB 3.8.8 Inverters-ShutdownBASESBACKGROUNDA description of the inverters is provided in the Bases for LCO 3.8.7,"Inverters -Operating."APPLICABLESAFETYANALYSESThe initial conditions of Design Basis Accident (DBA) and transientanalyses in the FSAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2),assume Engineered Safety Feature systems are OPERABLE. TheClass 1 E UPS inverters are designed to provide the required capacity,capability, redundancy, and reliability to ensure the availability ofnecessary power to the Reactor Protective System and EngineeredSafety Features Actuation System instrumentation and controls so thatthe fuel, Reactor Coolant System, and containment design limits arenot exceeded.The OPERABILITY of the inverters is consistent with the initialassumptions of the accident analyses and the requirements for thesupported systems' OPERABILITY.The OPERABILITY of the minimum inverters to each 120 VAC vital busduring MODES 5 and 6 and during movement of recently irradiated fuelassemblies ensures that:a. The unit can be maintained in the shutdown or refuelingcondition for extended periods;b. Sufficient instrumentation and control capability is available formonitoring and maintaining the unit status; andc. Adequate power is available to mitigate events postulatedduring shutdown, such as a fuel handling accident involvinghandling recently irradiated fuel. Due to radioactive decay, ACand DC inverters are only required to mitigate fuel handlingaccidents involving recently irradiated fuel (i.e., fuel that hasoccupied part of a critical reactor core within the previous40072 hours).The inverters were previously identified as part of the distributionsystem and, as such, satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).LCOThis ensures the availability of sufficient inverter power sources tooperate the unit in a safe manner and to mitigate the consequences ofpostulated events during shutdown (e.g., fuel handling accidentsinvolving handling recently irradiated fuel).(continued)DIABLO CANYON -UNITS 1 & 2Rev 8B Page 75 of 90 Distribution Systems -ShutdownB 3.8.10B 3.8 ELECTRICAL POWER SYSTEMSB 3.8.10 Distribution Systems -ShutdownBASESBACKGROUNDA description of the Class 1E AC, DC, and 120 VAC vital bus electricalpower distribution systems is provided in the Bases for LCO 3.8.9,"Distribution Systems -Operating."APPLICABLESAFETYANALYSESThe initial conditions of Design Basis Accident and transient analysesin the FSAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assumeEngineered Safety Feature (ESF) systems are OPERABLE. The ClassI E AC, DC, and 120 VAC vital bus electrical power distribution systemsare designed to provide sufficient capacity, capability, redundancy, andreliability to ensure the availability of necessary power to ESF systemsso that the fuel, Reactor Coolant System, and containment designlimits are not exceeded.The OPERABILITY of the Class 1E AC, DC, and 120 VAC vital buselectrical power distribution system is consistent with the initialassumptions of the accident analyses and the requirements for thesupported systems' OPERABILITY.The OPERABILITY of the minimum Class IE AC, DC, and 120 VACvital bus electrical power distribution subsystems during MODES 5 and6, and during movement of recently irradiated fuel assemblies ensuresthat:a. The unit can be maintained in the shutdown or refuelingcondition for extended periods;b. Sufficient instrumentation and control capability is available formonitoring and maintaining the unit status; andc. Adequate power is provided to mitigate events postulatedduring shutdown, such as a fuel handling accident involvinghandling recently irradiated fuel. Due to radioactive decay, ACand DC electrical power is only required to mitigate fuelhandling accidents involving handling recently irradiated fuel(i.e., fuel that has occupied part of a critical reactor core withinthe previous 4-072 hours).The Class IE AC, DC, and 120 VAC electrical power distributionsystems satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).ILCOVarious combinations of subsystems, equipment, and components arerequired OPERABLE by other LCOs, depending on the specific plantcondition. An OPERABLE AC subsystem shall consist of a 4kV vitalbus powered from at least one energized offsite power source with thecapability of being powered from an OPERABLE DG. The DG may bethe DG associated with that bus or, with administrative controls inplace, a DG that can be cross-tied (via the startup cross-tie feederbreakers) to another bus. However, credit for this cross-tie capability(continued)DIABLO CANYON -UNITS I & 2Rev 8B Page 87 of 90 Containment PenetrationsB 3.9.4B 3.9 REFUELING OPERATIONSB 3.9.4 Containment PenetrationsBASESBACKGROUNDIn MODES 1, 2, 3, and 4, the containment serves to contain fissionproduct radioactivity that may be released from the reactor corefollowing an accident, such that offsite radiation exposures aremaintained well within the requirements of 10 CFR 4-050.67.Additionally, in all operating modes the containment provides radiationshielding from the fission products that may be present in thecontainment atmosphere following accident conditions. Howeverduring CORE ALTERATIONS or movement of irradiated fuelassemblies within containment, the potential for containmentpressurization as a result of an accident is not likely; therefore,requirements to maintain the pressure boundary can be less stringent.An analysis has been performed that shows by meeting the LCO,during CORE ALTERATION and movement of irradiated fuelassemblies in containment, the potential release as a result of a fuelhandling accident (FHA) will remain well-within the requirements of10 CFR 4-0050.67 limits.The containment equipment hatch, which is part of the containmentpressure boundary, provides a means for moving large equipment andcomponents into and out of containment. The LCO requires thatduring CORE ALTERATIONS or the movement of irradiated fuelassemblies the equipment hatch must be capable of being closed andheld in place by at least four bolts. Good engineering practice dictatesthat the bolts required by this LCO be approximately equally spaced.The containment Personnel Air Lock (PAL) and Emergency Air Lock(EAL), which are also part of the containment pressure boundary,provide a means for personnel and emergency access duringMODES 1, 2, 3, and 4 unit operation in accordance with LCO 3.6.2,"Containment Air Locks." Each of these air locks has a door at bothends. The doors are normally interlocked to prevent simultaneousopening when containment OPERABILITY is required. During periodsof unit shutdown when the PAL and EAL are not required to be closed,the door interlock mechanisms may be disabled, allowing both doors ofeach of the air locks to remain open for extended periods whenfrequent containment entry is necessary.(continued)IDIABLO CANYON -UNITS 1 & 2Rev 8A Page 10 of 26 Containment PenetrationsB 3.9.4BASESBACKGROUND Per the FHA inside containment analysis, there are no closure(continued) restrictions required to limit any release to well within the requirementsof 10 CFR 4GG50.67 limits for offsite dose as the result of a fuelhandling accident during refueling. The LCO requirements forcontainment penetration closure are not provided to meet regulatoryrequirements, but rather to reduce the potential volume of the releaseof fission product radioactivity within containment to the environment.The Containment Purge and Exhaust System includes twosubsystems. The normal subsystem includes a 48 inch purgepenetration and a 48 inch exhaust penetration in Which the flow path islimited to beiRg open 200 hour0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months or less per calendar year. The secondsubsystem, a pressure equalization system provides a single 12 inchsupply and exhaust penetration. The three valves in the 12 inchpressure equalization penetration can be opened intermittently-. Eachof these systems are qualified to closed automatically by theEngineered Safety Features Actuation System .(ESFAS). Neither of-tsubsystems is subject to a Specification in MODE 5.In MODE 6, large air exchanges are necessary to conduct refuelingoperations. The normal 48 inch purge system is used for this purpose,and all four valves are closed by the ESFAS in accordance withLCO 3.3.6, "Containment Purge and Exhaust IsolationInstrumentation."The pressure equalization system is disassembled and used inMODE 6 for other outage functions.The other, containment penetrations that provide direct access fromcontainment atmosphere to outside atmosphere must be isolated on atleast one side if they are not opened under administrative controls.Isolation may be achieved by an OPERABLE automatic isolation valve,or by a manual isolation valve, blind flange, or equivalent. The fueltransfer tube is open but closure is provided by an equivalent isolationof a water loop seal. Equivalent isolation methods must be approvedand may include use of a material that can provide a temporary,ventilation barrier for the other containment penetrations during fuelmovements (Ref. 1).Although the historic severe weather patterns for DCPP do not requireconsideration of tornados as part of the design basis, severe weatherconditions might occur at the site that could necessitate closure ofopen penetrations with direct access to the outside atmosphere duringrefueling operations with core alterations or irradiated fuel movementinside containment. As a result, administrative procedures shallrequire that closure of these penetrations be initiated immediately ifsevere weather warnings are in effect. All fuel handling activities insidecontainment shall be suspended until closure of the equipment hatch iscompleted.(continued)DIABLO CANYON -UNITS 1 & 2Rev8A Pagellof26 Containment PenetrationsB 3.9.4BASES (continued)APPLICABLESAFETYANALYSISDuring CORE ALTERATIONS, or movement of irradiated fuelassemblies within containment, the most severe radiologicalconsequences result from a fuel handling accident. The fuel handlingaccident is a postulated event that involves damage to irradiated fuel(Ref. 2). Fuel handling accident inside the containment is based ondropping a single irradiated fuel assembly of which all 264 fuel rodsrupture. In addition the analysis assumes free and rapidcommunication of air from the containment to the outside environment;the accident occurs 40072 hours after reactor shutdown; almostinstantaneous release of the entire containment volume to the outsideatmosphere; thyroid dose con..rsion factors based on ICRP 30(Ref. 4); a radial peaking factor of 1.65 based on 105% full poweroperation; and the other guidance from RG 4-_251.183. (Ref 5)..The requirements of LCO 3.9.7, "Refueling Cavity Water Level," andthe minimum decay time of 1-4072 hours prior to CORE ALTERATIONSensure that the release of fission product radioactivity, subsequent to afuel handling accident, results in doses that are well within theguideline values specified in 10 CFR 100. Standard Review Pl4an-,Secti on 15.7.4,Rev. 1 (Ref. 3), defines 2wellwithin" 10 CFR 100 to be 259o or less bfthe 10 CFR 100 values. The acceptance limits for offsitc radiationexposure... .ill be 25% of 10 C FR 100 values. less than the accidentdose criteria specified in Table 6 of RG 1.183 (Ref. 5).Containment penetrations satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).(continued)DIABLO CANYON -UNITS 1 & 2Rev 8A Page 12 of 26 Containment PenetrationsB 3.9.4BASES (continued)REFERENCES 1. Design Criteria Memorandum T-16, Containment Functions.2. FSAR, Section 15.4.5 and 15.5.22.3. NUREG 0800, Section 15.7.4, Rev. 1, July 198!Not Used.4. international Comm~iSSion on Radiological Pro~tectioPublication 30, "Limits for intakes of Radionuclides byWorkers," 1 Not Used.5. RG --.2-51.183, July 2000.DIABLO CANYON -UNITS 1 & 2Rev8A Page 16 of 26 Refueling Cavity Water LevelB 3.9.7B 3.9 REFUELING OPERATIONSB 3.9.7 Refueling Cavity Water LevelBASESBACKGROUNDThe movement of irradiated fuel assemblies within containment requiresa minimum water level of 23 ft above the top of the reactor vesselflange. During refueling, this maintains sufficient water level in thecontainment, refueling canal, fuel transfer canal, refueling cavity, andspent fuel pool. Sufficient water is necessary to retain iodine fissionproduct activity in the water in the event of a fuel handling accident(Refs. 2-aad-61 and 2). Sufficient iodine activity would be retained tolimit offsite doses from the accident to < 25% of 10 CFR 100 limits, asprovided by the guidance of Reference 3the acceptance criteria of 10CFR 50.67 (Ref. 4) and RG 1.183 (Ref. 1).APPLICABLESAFETYANALYSISDuring CORE ALTERATIONS and movement of irradiated fuelassemblies, the water level in the refueling canal and the refuelingcavity is an initial condition design parameter in the analysis of a fuelhandling accident in containment, as postulated by Regulatory Guide4-.2-5 1.183 (Ref. 1). A minimum water level of 23 ft allows adecontamination factor of 200 (Appendix B (2) of Ref. 61 approved inRef. 7) to be used in the accident analysis for iodine. This relates to theassumption that 99.5% of the total iodine released from the pellet tocladding gap of all the dropped fuel assembly rods is retained by therefueling cavity water. The fuel pellet to cladding gap is assumed tocontain 41-012% of 1-131 and 10% of corethe tota- fuel rod iodineinventory of all other iodine isotopes (Ref. 42).I1The fuel handling accident analysis inside containment is described inReference 2. With a minimum water level of 23 ft and a minimum decaytime of 1-0072 hours prior to fuel handling, the analysis and testprograms demonstrate that the iodine release due to a postulated fuelhandling accident is adequately captured by the water and offsite dosesare maintained well within allowable limits (Refs. 1 and 4, and 5).Refueling cavity water level satisfies Criterion 2 of10 CFR 50.36(c)(2)(ii).IILCOA minimum refueling cavity water level of 23 ft above the reactor vesselflange is required to ensure that the radiological consequences of apostulated fuel handling accident inside containment are withinacceptable limits, as provided by the guidance of Reference 31.I(continued)DIABLO CANYON -UNITS 1 & 2Rev 8A Page 25 of 26 Refueling Cavity Water LevelB 3.9.7BASES (continued)APPLICABILITYLCO 3.9.7 is applicable during CORE ALTERATIONS, except duringlatching and unlatching of control rod drive shafts, and when movingirradiated fuel assemblies within containment. The LCO minimizes thepossibility of a fuel handling accident in containment that is beyond theassumptions of the safety analysis. If irradiated fuel assemblies arenot present in containment, there can be no significant radioactivityrelease as a result of a postulated fuel handling accident.Requirements for fuel handling accidents in the spent fuel pool arecovered by LCO 3.7.15, "Fuel Storage Pool Water Level."ACTIONS A..1With a water level of < 23 ft above the top of the reactor vessel flange,all operations involving movement of irradiated fuel assemblies withinthe, containment shall be suspended immediately to ensure that a fuelhandling accident cannot occur.The suspension of fuel movement shall not preclude completion ofmovement of a component to a safe position.SURVEILLANCE SR 3.9.7.1REQUIREMENTS Verification of a minimum water level of 23 ft above the top of thereactor vessel flange ensures that the design basis for the analysis ofthe postulated fuel handling accident during refueling operations ismet. Water at the required level above the top of the reactor vesselflange limits the consequences of damaged fuel rods that arepostulated to result from a fuel handling accident inside containment(Ref. 2).The Surveillance Frequency is based on operating experience,equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.REFERENCES 1. Regulatory Guide 1.25, March 23, 19721.183, July 2000.2. FSAR, Section 15.4.5 and 15.5.22.3. NUREG 0800, Section 15.7.4. Not Used4. 10 CFR 4004050.67.5. Malmn'e. .ki, D. D., Bell, M. j., Duhn, E., and Locante, J.,V\GAP_ 828, .O.sequencc. of a Fuel HandlingAccident, December 197-1.- Not Used6. Appendix B (2) of Regulator. , Guide i.183, july 2000Not Used.7. License Amendment 155/155, October 2\1, 2002DIABLO CANYON -UNITS I & 2Rev 8A Page 26 of 26

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+{{#Wiki_filter:Pacific Gas andElectric CompanyJames M. Welsch Diablo Canyon Power PlantSite Vice President Mail Code 104/6P. 0. Box 56Avila Beach, CA 93424June 17, 2015 805.545.3242Internal: 691.3242Fax: 895.545.4884PG&E Letter DCL-15-069U.S. Nuclear Regulatory Commission 10 CFR 50.90ATTN: Document Control DeskWashington, D.C. 20555-0001Diablo Canyon Units 1 and 2Docket No. 50-275, OL-DPR-80Docket No. 50-323, OL-DPR-82License Amendment Request 15-03Application of Alternative Source TermPursuant to 10 CFR 50.90, Pacific Gas and Electric Company (PG&E) herebyrequests approval of the enclosed proposed amendment to Facility OperatingLicense Nos. DPR-80 and DPR-82 for Units 1 and 2, respectively, of the DiabloCanyon Power Plant (DCPP). The enclosed license amendment request (LAR)proposes to revise the DCPP Units I and 2 licensing bases to adopt the alternativesource term (AST) as allowed by 10 CFR 50.67. The following TechnicalSpecification (TS) changes are required for AST implementation: TS 1.1 for thedefinition of Dose Equivalent 1-131; TS 3.4.16 to revise the noble gas activity limit;TS 3.6.3 to require the 48-inch containment purge supply and exhaust valves to besealed closed during MODES 1, 2, 3, and 4; TS 5.5.9 to revise the accident inducedleakage performance criterion; TS 5.5.11 to change the allowable methyl iodidepenetration testing criteria for the auxiliary building ventilation system charcoal filter:and TS 5.5.19 to replace "whole body or its equivalent to any part of the body," with"TEDE," which is the dose criteria specified in 10 CFR 50.67.The Enclosure provides a description of the proposed changes and supportingjustification including the determination of no significant hazards and environmentalconsiderations. Attachments to the Enclosure are described within.The changes in this LAR are not required to address an immediate safety concern.PG&E requests approval of this LAR no later than June 30, 2016. PG&E requeststhe license amendments be made effective upon NRC issuance, to be implementedwithin 365 days from the date of issuance.This communication contains five new regulatory commitments (as defined in NEI99-04) to be implemented following NRC approval of this LAR. The commitmentsare contained in Attachment 7 of the Enclosure.A member of the STARS (Strategic Teaming and Resource Sharing) AltianceCatlaway
+* Diablo Canyon
+* Palo Verde
+* Wolf Creek A w l Document Control Desk PG&E Letter DCL-15-069June 17, 2015Page 2In accordance with site administrative procedures and the Quality AssuranceProgram, the proposed amendment has been reviewed by the Plant Staff ReviewCommittee.Pursuant to 10 CFR 50.91, PG&E is sending a copy of this proposed amendment tothe California Department of Public Health.If you have any questions or require additional information, please contact HosseinHamzehee at 805-545-4720.I state under penalty of perjury that the foregoing is true and correct.Executed on June 17, 2015.Sincerely,J M. WelschSite Vice Presidentkjse/4328/50705089Enclosurecc: Diablo Distributioncc/enc: Marc L. Dapas, NRC Region IVThomas R. Hipschman, NRC Senior Resident InspectorSiva P. Lingham, NRR Project ManagerGonzalo L. Perez, Branch Chief, California Dept of Public HealthA member of the STARS (Strategic Teaming and Resource Sharing) AllianceCaLLaway
+* Diablo Canyon
+* Palo Verde
+* Wolf Creek EnclosurePG&E Letter DCL-15-069Evaluation of the Proposed ChangeLicense Amendment Request 15-03Application of Alternative Source Term EnclosurePG&E Letter DCL-15-069Table of ContentsI. SUMMARY DESCRIPTION ............................................................................ I2. DETAILED DESCRIPTION ........................................................................ 22.1 Proposed Changes to Current Licensing Basis .................. I ..................... 22.2 Proposed Technical Specification Changes ........................................... 112.3 Technical Specification Bases Changes ................................................ 142.4 Plant Changes .... ................................................................... ...... 142.5 Procedure Changes ................................................................................ 152.6 Updated Final Safety Analysis Report Changes ..................................... 162.7 Presentation of Current Licensing Basis and Alternative Source TermAnalysis Inputs ....................................................................................... 163. TECHNICAL EVALUATION .................. .................................................... 174. REGULATORY EVALUATION .................................................................. 194.1 Applicable Regulatory Requirements/Criteria ........................................ 194.2 Precedent .............................................................................................. 254.3 No Significant Hazards Consideration .................................................... 264.4 Conclusions ........................................................................................... 325. ENVIRONMENTAL CONSIDERATION .................................................... 326. REFERENCES ......................................................................................... 34ii EnclosurePG&E Letter DCL-15-069List of Attachments1. Proposed Technical Specification Changes (MARKUP)2. Proposed Technical Specification Changes (RETYPED)3. Technical Specification Bases Markup (For Information Only)4. Diablo Canyon Power Plant Technical Assessment Prepared by Stone &Webster, Inc. (A CB&I Company) -Implementation of Alternative SourceTerms Summary of Dose Analyses and Results5. Regulatory Guide 1.183 Conformance Tables6. Diablo Canyon Power Plant Comparison to NRC Regulatory. InformationSummary (RIS) 2006-04 Experience with Implementation of Alternative.Source Terms7. Diablo Canyon Power Plant List of Regulatory Commitments for AlternativeSource Term Implementation8. Diablo Canyon Power Plant Updated Final Safety Analysis Report Markup(For Information Only)iii EnclosurePG&E Letter DCL-15-069EVALUATIONSUMMARY DESCRIPTIONThis license amendment request (LAR) would amend Operating Licenses DPR-80 and DPR-82 for Units 1 and 2 of the Diablo Canyon Power Plant (DCPP),respectively.Pacific Gas & Electric (PG&E) requests Nuclear Regulatory Commission (NRC)review and approval of a proposed revision to the licensing basis of DCPP Units1 and 2 that supports a full scope application of an alternative source term (AST)methodology as allowed by 10 CFR 50.67 (Reference 1).An application for the selective use of AST for the fuel handling accident (FHA) inthe fuel handling building (FHB) was reviewed and approved by the NRC in itsSafety Evaluation Report (SER) for License Amendment Nos. 163 and 165(Reference 2). However, the FHA in the FHB has also been reanalyzed with thisapplication and is included in this submittal to be consistent with revised inputs,as described in Attachment 4. Approval of this AST application will supersedethe FHA in the FHB dose analysis and results, as discussed in the SER forLicense Amendment Nos. 163 and 165.The AST methodology as established in Regulatory Guide (RG) 1.183,"Alternative Radiological Source Terms for Evaluating Design Basis Accidentsat Nuclear Power Reactors," July 2000 (Reference 3) is used to calculate theoffsite and Control Room radiological consequence for DCPP Units 1 and 2.Attachment 4 contains a summary of the analyses and results for the followingevents that are expected to produce the most limiting dose consequences.Conformance to RG 1.183 is provided in Attachment 5.* Loss of Coolant Accident (LOCA)* FHA in the Containment* FHAinthe FHB* Locked Rotor Accident (LRA)* Control Rod Ejection Accident (CREA)" Main Steam Line Break (MSLB)0 Steam Generator Tube Rupture (SGTR)" Loss-of-Load (LOL) EventIn addition to adopting AST for design basis accidents and the associated totaleffective dose equivalent (TEDE) dose criteria for offsite and Control Roomdoses, DCPP is adopting the TEDE dose criteria of 10 CFR 50.67 for theTechnical Support Center (TSC), as allowed by RG 1.183.1 EnclosurePG&E Letter DCL-15-069Full implementation of AST for DCPP Units 1 and 2 does not include revising thesource terms used for environmental qualification (EQ) of safety relatedequipment or NUREG-0737 responses associated with shielding and vital areaaccess. Section C.6 of RG 1.183 (Reference 3) discusses the position onperformance of required EQ analyses with respect to AST and TID-14844(Reference 4) source term assumptions. NUREG-0933, "Resolution of GenericSafety Issues," Section 3.0, Item 187 (Reference 5) resolved the issues relatedto the effect of increased cesium releases on EQ doses. The NRC staffconcluded that there is no clear basis for a requirement to modify the designbasis for EQ to adopt AST since there would be no discernible risk reductionassociated with adopting AST for EQ. In addition, post-accident vital area accessdose rates are not expected to be significantly impacted by the AST during thefirst 30 days following a LOCA based on an AST benchmarking study. The NRCSER for Fort Calhoun Station's implementation of AST (Reference 6) referencedthe SECY-98-154 (Reference 7) study as the source for the conclusion that theresults of analyses based on TID-14844 would be more limiting for a period up toone to four months after which time the AST results would be more limiting.Therefore, this LAR does not propose to modify the EQ design basis nor theshielding and vital area access dose rates to adopt AST.The proposed amendment revises Technical Specification (TS) definitions,requirements, and terminology related to the use of an AST associated withoffsite, Control Room, and TSC accident dose consequences. A markup ofaffected TS pages is included in Attachment 1 to this Enclosure.Regulatory Issue Summary (RIS) 2006-04, "Experience with Implementation ofAlternative Source Terms," dated March 7, 2006, (Reference 8) outlined twelveissues that the NRC staff has encountered during its review of AST submittals.Attachment 6 provides discussion on how DCPP has addressed the twelveissues identified in RIS 2006-04.2. DETAILED DESCRIPTION2.1 Proposed Changes to Current Licensing BasisThe dose consequence analyses addressed in this application have been revisedto incorporate the guidance provided in RG 1.183 (Reference 3) and to resolvethe findings of the Licensing Basis Verification Project (LBVP) which wasvoluntarily initiated by DCPP, as presented to NRC (ADAMS AccessionNo. ML15029A094). The LBVP findings have been addressed in promptoperability assessments, which include several temporary compensatorymeasures. The revised dose analyses address the LBVP findings, as well asimplements the following licensing basis changes.1. Implement RG 1.183, July 2000 (Reference 3), as the licensing basis forDCPP, as outlined in this LAR. RG 1.183 will replace DCPPs commitment2 EnclosurePG&E Letter DCL-:15-069to RG 1.195, "Methods and Assumptions for Evaluating RadiologicalConsequences of Design Basis Accidents at Light-Water Nuclear PowerReactors."2. Remove the "expected" accident dose consequence assessments that arein DCPP Updated Final Safety Analysis Report (UFSAR) Section 15.5.The original DCPP licensing application included two evaluations for eachaccident. The first evaluation, called the expected case, used estimates ofactual values expected to occur if the accident took place. The resultingdoses were close to the doses expected from an accident of this type.The second evaluation, the Design Basis Accident (DBA), used thecustomary conservative assumptions. The calculated doses for the DBA,while not a realistic estimate of expected doses, provided the basis fordetermining the design adequacy of the plant safety systems.Current NRC guidance related to expectations for a safety analysis reportfor a nuclear power plant, as provided in NUREG-0800 (Reference 9),does not require the inclusion of dose consequences from "expected"accident scenarios. Since these "expected" accident scenario evaluationsare not relevant for determining design adequacy of plant safety systems,PG&E is proposing to remove this information from its licensing basis.UFSAR markups showing the elimination of the "expected" cases areprovided for information only in Attachment 8.3. Eliminate the dose contribution of a containment purge via thecontainment hydrogen purge system following a LOCA for purposes ofhydrogen control.The NRC revised 10 CFR 50.44 (Reference 10) to acknowledge that theamount of combustible gas generated for the design basis LOCA was nota risk significant threat to containment integrity. Thus, with the exceptionof demonstrating the capability of ensuring a mixed atmosphere withincontainment, the requirements for hydrogen control pertaining to thedesign basis LOCA were eliminated. In the SER for License AmendmentNos. 168 and 169 to DCPP (Reference 11), the NRC confirmed theelimination of hydrogen release concerns associated with a design-basisLOCA, and the associated requirements that necessitated the need forhydrogen recombiners and backup hydrogen vent and purge systems.To ensure consistency with the current licensing basis, PG&E is proposingto eliminate the dose contribution due to the containment purge pathwaycurrently included in the LOCA dose consequence analysis in support ofhydrogen control.3 EnclosurePG&E Letter DCL-15-0694. Replace dose guidelines of 10 CFR 100.11 for whole body and thyroiddose with the TEDE acceptance criterion of 10 CFR 50.67(b)(2) andSection 4.4, Table 6 of RG 1.183 (Reference 3).As required by 10 CFR 50.67(b)(2) (Reference 1), Attachment 4 of thisLAR contains an evaluation of the consequences of applicable DBAspreviously analyzed in the DCPP UFSAR (Reference 12). The TEDEdose criteria will be applied to offsite locations, as well as the ControlRoom and the TSC.5. Replace General Design Criteria (GDC 19), 1971, with GDC 19, 1999 fordose only.10 CFR 50.67(b)(2)(iii) states that adequate radiation protection isprovided to permit access to and occupancy of the Control Room underaccident conditions without personnel receiving radiation exposures inexcess of 0.05 Sv.(5 rem) TEDE for the duration of the accident. As partof implementation of 10 CFR 50.67, the NRC also amended GDC 19,1971 (Reference 13) to reflect the 5 rem TEDE aspects. The DCPPlicense basis, from the original Final Safety Analysis Report throughAmendment 85, includes GDC 19, 1971, for Control Room dose only(Reference 12). DCPP will conform to GDC 19, 1999, for dose only, forControl Room dose limits of 5 rem TEDE upon implementation of AST.This change to GDC 19, 1999, for dose only, is consistent with theadoption of AST.6. Update the dose acceptance criterion for the TSC to 5 rem TEDE.The dose acceptance criterion for the TSC is based on Section 8.2.1,Item f of NUREG-0737, Supplement 1 (Reference 14), which states thatany person working in the TSC would not exceed 5 rem whole body, or itsequivalent to any part of the body, for the duration of the accident. Thedose acceptance criterion is modified to 5 rem TEDE to be consistent with10 CFR 50.67(b)(2) and GDC 19, 1999, for dose only. In accordance withDCPP current licensing basis, the TSC is only evaluated for the LOCA.Therefore, upon implementation of AST, the DCPP licensing basis forNUREG-0737, Item ll.B.2 and Ill.A.1.2 will be the AST acceptance criteriaspecified in 10 CFR 50.67(b) and GDC 19, 1999, for dose only. Thischange to GDC 19, 1999, is consistent with the adoption of AST7. Update computer codes that support the AST dose consequenceanalyses.Attachment 4, Section 3 provides the computer codes utilized in support ofthis application. These computer codes will now be included in DCPP's4 EnclosurePG&E Letter DCL-15-069licensing basis in the manner in which they are utilized in the doseconsequence analyses as outlined in Attachment 4. The codes used insupport of the AST application are recommended by RG 1.183 or havebeen used in priorAST applications that have been approved by NRC.8. Use inhalation dose conversion factors from Environmental ProtectionAgency (EPA) Federal Guidance Report (FGR) No. 11, 1988, "LimitingValues of Radionuclide Intake and Air Concentration and Dose ConversionFactors for Inhalation, Submersion, and Ingestion." (Reference 15).FGR No. 11 has been part of the licensing basis for DCPP in that TSDefinition 1.1 for Dose Equivalent 1-131 (DEI) allowed Table 2.1 of FGRNo. 11 to be used for determining DEI. FGR No. 11 is used in AST doseconsequence analyses for inhalation dose conversion factors, asrecommended by RG 1.183. See Section 2.2 for proposed TS changes.9. Update offsite atmospheric dispersion factors (x/Q) using recent 5-yearmeteorological data (2007 to 2011) and RG 1.145, Revision 1,"Atmospheric Dispersion Models for Potential Accident ConsequenceAssessments at Nuclear Power Plants," (Reference 16) methodology.The methodology outlined in RG 1.145, Revision 1, is used for calculatingground level releases to determine the short-term X/Q values for theExclusion Area Boundary (EAB) and the Low Population Zone (LPZ) fordesign basis radiological analyses. All releases are conservatively treatedas ground level releases, therefore Regulatory Positions C.1.3.2, C.2.1.2,and C.2.2.2 associated with elevated or stack releases are not applicableto DCPP.RG 1.111, "Methods for Estimating Atmospheric Transport and Dispersionof Gaseous Effluents in Routine Releases from Light Water CooledReactors," Regulatory Position C.1.c (Reference 17) is used to determinethe annual average XJQ values, which are used as input to develop theaccident X/Q values at the LPZ using RG 1.145 methodology.Attachment 4, Section 5 presents the development of the X/Q values.10. Update X/Q factors for on-site locations such as the Control Room and theTSC using recent 5-year meteorological data (2007 to 2011) and,"Atmospheric Relative Concentrations in Building Wakes," (ARCON96)methodology (Reference 18).Regulatory Guide 1.194, "Atmospheric Relative Concentrations for ControlRoom Radiological Habitability Assessments at Nuclear Power Plants,"dated June 2003 (Reference 19), Regulatory Position C.1 through C.3,and the adjustment factor for vertically orientated energetic releases from5 EnclosurePG&E Letter DCL-15-069steam relief valves and atmospheric dump valves (ADVs) allowed byRegulatory Position C.6 are used to determine short-term on-site X/Qvalues in support of design basis radiological habitability assessments.PG&E is requesting an exception to RG 1.194, Regulatory Position C.3.4,as part of this LAR. Two specific Control Room receptors are within10 meters of the release (9.4 meters and 7.8 meters). The x/Q values forthese two cases were developed to establish boundingX/Q values.However, the ,/Q values for these two locations were not the boundingvalues and therefore were not used in the dose consequence analyses.Attachment 4, Section 5.2 provides further discussion for the requestedexception.'Credit for DCPP's dual intake design for the Control Room pressurizationair intakes is taken per RG 1.194, Regulatory Position C.3.3.2.3. Inaddition, credit is taken for a reduction factor of 5 applied to the X/Q valuesfor energetic releases from the DCPP Main Steam Safety Valves (MSSVs)and the 10 percent ADVs, per RG 1.194, Regulatory Position C.6 due tothe velocity and orientation of the release. This credit is used for theMSLB, SGTR, LRA, CREA, and LOL events.Credit is also taken for the close proximity of the MSSVs/10 percent ADVsto the normal operating Control Room intake of the affected unit and thehigh vertical velocity of the steam discharge from the MSSVs/1 0 percentADVs resulting in the post-accident plume from the MSSVs/1 0 percentADVs not contaminating the normal operation Control Room intake of theaffected unit. This credit is used for the MSLB, SGTR, LRA, CREA, andLOL events.Attachment 4, Section 5 presents the development of the X/Q values.11. Update Control Room ventilation system (CRVS) parameters resultingfrom the installation of new back-draft damper in the Control Roomemergency filter recirculation lines.Back-draft dampers were installed to prevent reverse unfiltered flow intothe Control Room. The updated CRVS parameters have been included inthe new Control Room transport model discussed in Attachment 4, Section7.1.12. Update Control Room unfiltered inleakage.The updated Control Room unfiltered inleakage values, including back-draft damper leakage, have been included in the new Control Roomtransport model discussed in Attachment 4, Section 7.1. The updated6
+'EnclosurePG&E Letter DCL-15-069Control Room unfiltered inleakage value bounds the unfiltered inleakagedetermined by the 2012 Control Room Tracer Gas Test (Reference. 20).13. Use containment spray in the recirculation mode following a LOCA forfission product cleanup.DCPP is designed and licensed to operate using containment spray in therecirculation mode. In accordance with the current licensing basis, and asdocumented in the NRC SER related to License Amendment No. 139 toFacility Operating License Nos. DPR-80 and DPR-82 (Reference 21),containment spray is not required per analyses to be actuated duringrecirculation, but may be actuated in accordance with the emergencyoperating procedures (EOPs) or at the discretion of the TSC.To address the delayed core damage sequence of a post-LOCAASTscenario and support fission product removal from the containmentatmosphere, credit is taken in the LOCA dose analysis for usingcontainment spray in the recirculation mode for dose mitigation. Thislicensing basis change to require containment spray during recirculation.for fission product cleanup does not affect the conclusions of the SERrelated to License Amendment No. 139 with respect to the other functionsof containment spray. The containment spray system will be operated in amanner consistent with the licensing basis established by the SER relatedto License Amendment No. 139 and no changes in operation are beingproposed, other than requiring its operation within, 12 minutes followingterminating injection spray, instead of being optional in accordance withEOPs or at the discretion of the TSC.The LOCA dose analysis also credits a time critical operator action(TCOA). A TCOA is a manual action or series of actions with a specifiedcompletion time limit to meet a plant licensing basis requirement. TheLOCA dose analysis assumes that containment spray is realigned fromthe injection mode to the recirculation mode within 12 minutes ofterminating injection spray to ensure that the duration of spray operation(injection + recirculation) exceeds 6.25 hours following the event. TheTCOA will be implemented as part of AST implementation, using theguidelines provided in NRC Information Notice 97-78, "Crediting ofOperator Actions in Place of Automatic Actions and Modifications ofOperatorActions, Including Response Times." The required actions forthe new TCOA have been demonstrated on the simulator, showing thatthe 12 minute time requirement can be achieved with margin.14. Update allowable engineered safety features (ESF) system leakagevalues and associated release points following a LOCA.7 EnclosurePG&E Letter DCL-15-069In accordance with RG 1.183, the LOCA dose analysis assumes a releaseof two times the allowable ESF leakage values. As such, two times theallowable leakage values outlined in Section 2.5 are used to determinedoses following a LOCA, as discussed in Attachment 4, Section 7.2.3.3.The ESF leakage limits are controlled by TS 5.5.2, "Primary CoolantSources Outside Containment."15. Include environmental releases from the refueling water storage tank(RWST) vent due to sump water back-leakage following a LOCA.The LOCA dose analysis for AST includes the environmental releasesfrom the RWST due to sump water back-leakage following a LOCA, inaccordance with RG 1.183. Attachment 4, Section 7.2.3.5 presents theanalysis for this dose contribution.16. Include environmental releases from the Miscellaneous Equipment DrainTank (MEDT).The LOCA dose analysis for AST includes the environmental releasesfrom the MEDT following a LOCA, in accordance with RG 1.183.Attachment 4, Section 7.2.3.6 presents the analysis for this dosecontribution.17. Include environmental releases via the 12-inch containmentvacuum/pressure relief pathway prior to containment isolation following aLOCA.In accordance with RG 1.183, for containments such as DCPP that can beroutinely purged during normal operation, the dose consequence analysismust assume a release to the environment, through the purge pathway,occurs prior to containment isolation. As such, the LOCA dose analysisincludes a dose contribution from-the 12-inch containmentvacuum/pressure relief pathway prior to containment isolation.Attachment 4, Section 7.2.3.1 presents the analysis for this dosecontribution.18. Preclude environmental releases via the 48-inch containment purge andexhaust system pathway prior to containment isolation following a LOCA.In accordance with RG 1.183, for containments such as DCPP that can beroutinely purged during normal operation, the dose consequence analysismust assume a release to the environment, through the purge pathway,occurs prior to containment isolation. This release pathway is notapplicable to the containment purge and exhaust system because DCPPis requesting approval of a TS change for the 48-inch containment purgevalves to be sealed closed in accordance with Standard Review Plan8 EnclosurePG&E Letter DCL-15-069(SRP) Section 6.2.4, Revision 3, Items 11.6 and 11.14 during MODES 1,2,3, and 4. The 48-inch containment purge valves will be sealed closed byremoving motive power to the valve operators. With this proposed TSrevision, NUREG-0737, Item II.E.4.2, Position (6) (Reference 22) will besatisfied. Because the 48-inch containment purge valves will be requiredto be.sealed closed during MODES 1, 2, 3, and 4, DCPP will no longertake credit for a Phase A isolation signal for these valves, as outlined inresponse to NUREG-0737, November 1980 Item II.E.4.2, Position (7). Inaddition, piping classification for the associated containment penetrations(61 and 62) will change from Group A to Group E.See Section 2.2 for proposed TS changes associated with the 48-inchcontainment purge supply and exhaust valves.19. Define the portion of Room 506 of the Control Room which serves as aControl Room foyer between the Control Room Assistants' office and theShift Managers' office as a low occupancy, less frequented area.When determining the direct shine dose to the Control Room fromexternal and contained sources, the analysis presented in Attachment 4,Section 7.2.5.2 takes into consideration the function of Room 506.Room 506 is used as an area where occupancy is deemed to be minimal.Thus, an "occupancy adjustment" factor is utilized for Room 506 todetermine the maximum 30-day integrated dose in the Control Room (i.e.,the total direct shine dose in the Control Room includes the 30-day dosein Room 506 adjusted by occupancy factor).20. Define a minimum decay time prior to fuel movement as 72 hours.As part of this application, DCPP proposes to revise the definition ofrecently irradiated fuel as fuel that has occupied part of a critical reactorcore within the previous 72 hours. This definition is used in the doseconsequence analysis of the FHA to determine the release following thepostulated event. Although the source term for the FHA will be slightlylarger with less decay of fuel prior to fuel movement, the doseconsequence analysis results show that the dose criteria are met, asshown in Table 1 and Attachment 4, Section 7.3.21. Credit the .redundant safety related gamma sensitive area radiationmonitors (1-RE 25/26, 2-RE 25/26) to initiate CRVS Mode 4 following aFHA.These monitors are located at the Control Room normal intakes and aredesigned to automatically isolate the normal CRVS intakes and shift toCRVS Mode 4 (pressurized filtered emergency ventilation). Thesemonitors are credited to perform their design function following a FHA in9 EnclosurePG&E Letter DCL-15-069the FHB or containment. See Section 2.4 for a description of a setpointchange for these radiation monitors.22. Credit the following existing administrative controls reflected in plantprocedures. These administrative controls ensure the FHB is maintainedat a negative pressure relative to atmosphere during movement ofirradiated fuel in the spent fuel pool, thus ensuring that the environmentalreleases occur via the Unit vent." The movable wall is in place and secured* No exit door from the FHB is propped open" At least one FHB Ventilation System exhaust fan is runningAttachment 4, Section 7.3 presents the FHA. Credit for the aboveadministrative controls is taken to facilitate that the post-accidentenvironmental release of radioactivity occurs via the plant vent.23. Update reported doses for other UFSAR Chapter 15 events with accidentsource terms to TEDE doses criteria.The DCPP licensing basis includes dose assessments at offsite locationsfor several Condition III and Condition IV events. RG 1.183 does notaddress Condition Ill and Condition IV events; therefore they have notbeen re-analyzed with this application. SRP 15.0.1 (Reference 23) statesthat a complete recalculation of all design basis radiological consequenceanalyses may not be required for an application to be acceptable.However, SRP 15.01 also states that a full AST implementation replacesthe previous accident source term used in all design basis radiologicalanalyses and incorporates the TEDE dose criteria. Therefore, DCPPperformed scoping evaluations, as allowed by RG 1.183, Section C.1.3.3(Reference 3), to demonstrate compliance with regulatory limits at theEAB and LPZ for the DCPP UFSAR Chapter 15 Condition III andCondition IV events. The evaluation compares the accident sequence,predicted fuel damage (if applicable), and resultant dose consequences ofthe DBAs analyzed forAST to those parameters of the Condition III andCondition IV events. These evaluations are presented in Attachment 4,Section 2.1.The tank rupture events presented in UFSAR Chapter 15.5 represent theaccidental release of radioactivity accumulated in tanks resulting fromnormal plant operations and are not affected by accident source termsassociated with AST. Therefore, the tank rupture events are notreanalyzed in support of this LAR and the dose acceptance criteria for thetank rupture events will remain unchanged.10 fEnclosurePG&E Letter DCL-15-069As part of its original licensing basis, DCPP provides an estimatedradiation exposure to the Control Room operator during egress andingress between the Control Room and the site boundary following aLOCA, as presented in UFSAR 15.5.17.10. Although RG 1.183 does notaddress or provide guidance for determining this dose contribution, DCPPis retaining this access dose in its licensing basis. The whole bodygamma dose and thyroid dose reported in the UFSAR are converted toreflect the estimated TEDE dose by using organ weighting factorsprovided in 10 CFR 20.1003. Attachment 4, Section 7.2.6 demonstratesthat the access dose, converted to TEDE dose, is minimal in that it is 1percent of the estimated operator dose due to Control Room occupancyfollowing a LOCA.It is noted that the dose received by the operator during transit outside theControl Room is not a measure of the "habitability" of the Control Room,which is defined by the radiation protection provided to the operator by theControl Room shielding and ventilation system design. Thus, theestimated dose to the operator during routine post-LOCA access to theControl Room is addressed separately from the Control Room occupancydose and is not included with the Control Room occupancy dose for thedemonstration of Control Room habitability.2.2 Proposed Technical Specification ChangesThe following TS changes are proposed to reflect the licensing basis changesoutlined in Section 2.1. Brief descriptions of the associated proposed TSchanges are provided below along with justification for each change. Thespecific wording changes to the TS-are provided in Attachments 1 and 2 to thisenclosure.* TS 1.1,,"Definitions," is revised to change the definition of Dose Equivalent1-131 (DEI).This TS provides a definition for DEI, which currently references Table IIIof TID-14844, AEC, 1962, "Calculation of Distance Factors for Power andTest Reactor Sites;" Table E-7 of Regulatory Guide 1.109, Revision 1,NRC, 1977; International Commission on Radiological ProtectionPublication 30, 1979, Supplement to Part 1, pages 192-212, Table titled"Committed Dose Equivalent in Target Organs or Tissues per Intake ofUnit Activity;" and Table 2.1 of EPA Federal Guidance Report No. 11,1988, "Limiting Values of Radionuclide Intake and Air Concentration andDose Conversion Factors for Inhalation, Submersion, and Ingestion."This TS change will be revised to only reference the committed thyroiddose equivalent conversion factors from Table 2.1 of FGR No. 1111 EnclosurePG&E Letter DCL-15-069(Reference 15). The change is consistent with the recommendations ofRG 1.183. Under AST, the doses are reported as TEDE dose.NRC RIS 2006-04 (Reference 8) states:"Although different references are available for dose conversionfactors, the TS definition should be based on the same doseconversion factors that are used in the determination of the reactorcoolant dose equivalent iodine curie content for the main steamlinebreak and steam generator tube rupture accident analyses."Dose conversion factors from Table 2.1 of FGR No. 11 (Reference 15) areused by DCPP to determine the reactor coolant dose equivalent iodinecurie content for the MSLB and SGTR accident analyses. Thus, the TSchange is consistent with item 10, Definition of Dose Equivalent 1-131, ofNRC RIS 2006-04.The TS change will remain consistent with the approved IndustryImproved Standard Technical Specification Traveler, TSTF-490(Reference 24).TS 3.4.16, "RCS Specific Activity," is revised to change the Noble gasactivity limit from less than or equal to 600 pCi/gm Dose Equivalent XE-133 (DEX) to less than or equal to 270 pCi/gm DEX.DEX limit is equivalent to approximately 0.5 percent fuel defects. Thecurrent limit of 600 pCi/gm DEX corresponds to approximately 1 percentfuel defects, which is the DCPP design basis value for system andshielding design. The limit is being reduced by DCPP to control the noblegas activity in the coolant to levels below the design basis values.* TS 3.6.3, "Containment Isolation Valves," is revised to modify Note 1 ofLimiting Condition of Operation (LCO) 3.6.3 concerning the 48-inchcontainment purge supply and exhaust valves. The TS currently allowsthe operation of these valves for less than 200 hours per year duringoperating MODES 1, 2, 3, and 4. The proposed revision eliminates theadministratively controlled operation of the 48-inch containment purgevalves during MODES 1, 2, 3, and 4. The proposed revision will nowrequire the 48-inch containment purge supply and exhaust valves toremain sealed closed during MODES 1, 2, 3, and 4. This change willeliminate a potential dose contribution due'to an open containment purgepathway at the initiation of a LOCA.The TS revision includes a new surveillance requirement for verifying the48-inch purge valves are sealed closed, removes the 48-inch purge valves12 EnclosurePG&E Letter DCL-1 5-069from Surveillance Requirements (SRs) 3.6.3.2 and modifies the frequencyfor SR 3.6.3.7.The proposed revision is consistent with NUREG-1431, Volume 1,Standard Technical Specifications, Westinghouse Plants (Reference 25).The 48-inch containment purge valves are to be sealed closed inaccordance with SRP Section 6.2.4, Revision 3, Item 11.6 and 11.14 duringMODES 1, 2, 3, and 4. The 48-inch containment purge valves will besealed closed by removing motive power to the valve operators. Withthis proposed TS revision, NUREG-0737, Item I1.E.4.2, Position (6)(Reference 22) will be satisfied. Because the 48-inch containment purgevalves will be required to be sealed closed during MODES 1, 2, 3, and 4,DCPP will no longer take credit for a Phase A isolation signal for thesevalves, as outlined in response to NUREG-0737, November 1980 ItemI1.E.4.2, Position (7).* TS 5.5.9, "Steam Generator (SG) Tube Inspection Program," is revised tolower the accident induced leakage performance criterion from 1 gallonper minute (gpm) per steam generator to 0.75 gpm total for all four steamgenerators. The accident induced leakage performance criterion shall notexceed the leakage rate assumed in the dose consequence analysis.A primary-to-secondary SG tube leakage of 0.75 gpm at standardtemperature and pressure is used in the dose consequence analysis forthe LRA, CREA, MSLB, SGTR, and LOL events. DCPP TS 3.4.13d limitsprimary-to-secondary SG tube leakage to 150 gallons per day (gpd) perSG for a total of 600 gpd for all 4 SGs. The revised testing criterion for theprimary-to-secondary leakage is more restrictive than the current testingcriteria and represents the leakage rate assumed in the doseconsequence analyses presented in Attachment 4. The 0.75 gpm from all4 SGs (or a total of 1080 gpd) conservatively bounds the TS 3.4.13.d limitin that the analyzed leakage rate accounts for higher leakage than theTS 3.4.13.d limit, and thus a higher analyzed release of radioactivity.* TS 5.5.11, "Ventilation Filter Testing Program (VFTP)," is revised tochange the allowable methyl iodide penetration testing criteria for theauxiliary building ventilation system (ABVS) charcoal filter from 15 percentto 5 percent.The allowable methyl iodide penetration is used to determine charcoalfilter efficiency for removing iodine from atmospheric releases. Credit forfiltration of the release of a residual heat removal (RHR) system pumpseal passive failure by the ABVS is taken in determining the dose13 EnclosurePG&E Letter DCL-15-069consequences to the public at the EAB and LPZ, and to personnel in theControl Room and TSC.* TS 5.5.19, "Control Room Envelope Habitability Program," is revised toreplace "whole body or its equivalent to any part of the body" with "TEDE,"which is the dose criteria specified in 10 CFR 50.67 (Reference 1).2.3 Technical Specification Bases ChangesThe TS Bases will be revised to reflect the licensing basis changes outlined inSection 2.1 and the TS changes identified in Section 2.2. A markup of the TSBases changes is provided for information only in Attachment 3 to this Enclosure.These TS Bases changes will be implemented in accordance with TS 5.5.14,"Technical Specification (TS) Bases Control Program," upon NRC approval of thisLAR.2.4 Plant ChangesThe following plant design modifications will be performed as part of ASTimplementation. These plant modifications support the AST analyses provided inAttachment 4.* Install shielding material, equivalent to that provided by the Control Roomouter walls, at the external concrete west wall of the Control Room briefingroom." Install a high efficiency particulate air filter (HEPA) in the TSC normalventilation system intake." Update setpoints for the redundant safety related gamma sensitive area,radiation monitors (1-RE 25/26, 2-RE 25/26). These monitors are locatedat the Control Room normal intakes and are designed to automaticallyisolate the normal CRVS intakes and shift to CRVS Mode 4 (pressurizedfiltered emergency ventilation). These monitors are relied upon to performtheir design function following a FHA in the FHB or Containment.Setpoints for these monitors are contained in the Offsite Dose CalculationManual, which is controlled by TS 5.5.1, requirements." Reclassify a portion of the 40-inch Containment Penetration Area(GE/GW) Ventilation line from PG&E Design Class II to PG&E DesignClass I and upgrade the damper actuators, pressure switches, and thedamper solenoid valves to PG&E Design Class I. See Attachment 4,Section 5.2 for further details." Reclassify a portion of the 2-inch gaseous radwaste system line whichconnects to the Plant Vent as PG&E Design Class I. This line is currently14 EnclosurePG&E Letter DCL-15-069classified as PG&E Design Class II. See Attachment 4, Section 5.2 forfurther details.2.5 Procedure ChangesAs part of AST implementation, the following procedural updates will include:* Update Equipment Control Guideline (ECG) 42.1, "Refueling Operations -Decay Time," to lower the, restriction on fuel movement from 100 hours to72 hours post-shutdown." Update ECG 42.5, "Refueling Operations -Water Level -Reactor Vessel(Control Rods)" to reflect the FHA AST analysis assumptions." Update-ECG 23.3, "Containment Ventilation System," to reflect changes toTS 3.6.3." Review and update, as necessary, the EOPs and operator trainingprocedures to ensure that the requisite steps to select the leastcontaminated CRVS pressurization intake are in use throughout the event.This review is to be performed as verification, since the EOPs currentlyinclude steps to select the least contaminated CRVS intake." Update Surveillance Test Procedure M-57, "Control Room VentilationSystem (CRVS) Tracer Gas Test," to include the new Control Roominleakage test acceptance criteria and the range of CRVS ventilation flowsdeemed acceptable by the AST dose consequence analyses.* Update the TSC administrative procedures to ensure that:a. The nominal normal operation TSC ventilation air intake flowrate is500 cubic feet per minute (cfm).b. Following a LOCA, the TSC will be manually placed in Mode 4operation such that filtered pressurization and recirculation can becredited within 2 hours of accident initiation.c. The nominal post-LOCA TSC ventilation filtered pressurization andrecirculation flowrates are 500 cfm, respectively." Review EOPs to verify valve alignment information to manually initiatecontainment spray in the recirculation mode. Update EOPs to includedirection to perform the realignment action within 12 minutes oftermination of injection spray to ensure that the duration of spray operation(injection plus recirculation) exceeds 6.25 hours following the event. Anassociated TCOA will be implemented." Update ESF system leak testing procedures that are controlled byTS 5.5.2, "Primary Coolant Sources Outside Containment," to establishadministrative acceptance criteria to ensure:15 EnclosurePG&E Letter DCL-15-069a. The total as-tested leakage from ESF systems that recirculatesump fluid outside containment is less than 126 cubic centimetersper minute (cc/min), with the following breakdown:i. In areas covered by the ABVS, the as-tested leakage is lessthan 120 cc/min,ii. In the containment penetration area, the as-tested leakage isless than 6 cc/min.b. The total as-tested back leakage into the RWST from thecontainment recirculation sump is less than 1 gpm.c. The total as-tested flow hard piped to the MEDT is less than thefollowing values:i. Leakage from systems carrying non-radioactive fluids is lessthan 484 cc/min.ii. Leakage from ESF systems that recirculated sump fluids isless than 950 cc/min.Review and update the Emergency Plan to reflect AST, as necessary.There will be no change to the Emergency Planning Zone.2.6 Updated Final Safety Analysis Report ChangesThe UFSAR will be revised to reflect AST dose consequence analyses and theproposed licensing basis changes outlined in Section 2.1. UFSAR changes areprovided in Attachment 8, for information only.2.7 Presentation of Current Licensing Basis and Alternative Source Term AnalysisInputsThis section provides a summary of changes from current design and licensingbasis analysis input values to revised AST inputs for each analysis.The NRC's traditional methods for calculating the radiological consequences ofDBAs are described in a series of regulatory guides and SRP chapters. Thatguidance was developed to be consistent with TID-14844 source term and thewhole body and thyroid dose guidelines stated in 10 CFR 100.11. Many of thoseanalysis assumptions and methods are inconsistent with the AST and with theTEDE criteria provided in 10 CFR 50.67. In addition, many of DCPP's analysespre-date SRP guidance.RG 1.183 provides assumptions and methods that are acceptable to the NRCstaff for performing design basis radiological analyses using an AST. As stated inRG'1.183, the RG 1.183 guidance supersedes corresponding radiologicalanalysis assumptions provided in other regulatory guides and SRP chapters16 EnclosurePG&E Letter DCL-15-069when used in conjunction with an approved AST and TEDE criteria provided in 10CFR 50.67 (Reference 1).DCPP used the guidance provided by RG 1.183 for analysis assumptions andmethods for design basis radiological analyses. Conformance to RG 1.183guidance is presented in Attachment 5. A summary of design inputs,assumptions, and methodology used in the AST analyses is provided inChapter 7 of Attachment 4. Appendix B to Attachment 4 provides a comparisonbetween the design input values used in the current licensing basis doseconsequence analyses and the-values used to support the AST analysessupporting this LAR.As noted above and in Attachment 4, many of DCPP's analyses pre-date SRPguidance. Specifically, the DCPP's current licensing basis for the LRA, CREA,and LOL events are DCPP-specific with pre-SRP assumptions and only addressoffsite dose consequences. Updated analyses performed for AST now includeControl Room doses.3. TECHNICAL EVALUATIONThe DCPP Units 1 and 2 current licensing basis for the radiologicalconsequences analyses of accidents is based on source term methodology andassumptions derived from TID-14844 (Reference 4). An application for theselective use of AST for the FHA in the FHB was reviewed and approved by theNRC in its SER for License Amendment Nos. 163 and 165 (Reference 2). Thedesign basis accidents are discussed in Chapter 15 of the DCPP UFSAR(Reference 12). The current DCPP dose consequences of design basis events,other than the FHA in the FHB,, are based on acceptance criteria stated in10 CFR Part 100 and 10 CFR 50, AppendixA, GDC 19, 1971. The currentlicensing basis for the radiological consequences of the FHA in the FHB is10 CFR 50.67, "Accident Source Term."The AST and methodology described in NUREG-1465, "Accident Source Termsfor Light-Water Nuclear Power Plants," (Reference 25) and in RG 1.183(Reference 3), provide regulatory guidance for the implementation of the AST.Revision of a plant licensing basis from the TID-14844 (Reference 4) source termto an AST involves the preparation of dose consequence analyses.Demonstration that theresults satisfy the regulatory acceptance criteria and NRCapproval of the requested change establishes the acceptability of the use of theAST for DCPP.DCPP has performed radiological consequence analyses of the DBAsdocumented in Chapter 15 of the DCPP UFSAR that potentially result in the mostsignificant Control Room and offsite exposures. These analyses were performedto support full scope implementation of AST. The AST analyses have beenperformed in accordance with the guidance in RG 1.183 and SRP Section 15.0.117 EnclosurePG&E Letter DCL-15-069(Reference 23). Acceptance criteria consistent with those required by10 CFR 50.67 and RG 1.183, Table 6, were used to replace the current designbasis source term acceptance criteria. This represents a full implementation ofAST in which the RG 1.183 source term will become the licensing basis forDCPP DBAs.The technical justification for full implementation of the AST methodology, asdefined in RG 1.183, of the DCPP DBAanalyses is provided in Attachment 4.The following DBAs are addressed:" LOCA" FHA in the Containment* FHA in the FHB (reanalysis)* LRA* CREA* MSLB* SGTR* LOL EventConformance to RG 1.183 is documented in Attachment 5. The AST as definedin RG 1.183 has been incorporated into the DCPP site boundary and ControlRoom dose re-analyses discussed in Attachment 4. The estimated DCPP doseconsequences for all design basis events addressed in RG 1.183, meet theacceptance criteria specified in 10 CFR 50.67 and RG 1.183, as shown inTable 1 for offsite locations and Control Room personnel.In addition, the TSC dose is re-analyzed in Attachment 4. In accordance withcurrent licensing basis, the 30-day integrated dose to an operator in the TSC dueto immersion, inhalation, and direct shine is evaluated for the LOCA. Theresultant post-LOCA dose is estimated to be 4.1 rem TEDE, which is within theacceptance criteria of 5 rem TEDE.As stated in Section 2.4, some changes to the facility are required to implementAST for Control Room and TSC dose consequences. These changes includeadditional shielding for the Control Room, setpoint change for radiation monitors1-RE 25/26 and 2-RE 25/26, and component upgrades (damper actuators,pressure switches, and damper solenoid valves) for a portion of the 40-inchContainment Penetration Area (GE/GW) Ventilation system. A HEPA filter will beinstalled at the normal intake of the TSC ventilation system.The TSC is designed to meet NUREG-0696, Functional Criteria for EmergencyResponse Facilities, (Reference 27), which states that the TSC ventilationsystem need not be seismic Category 1 qualified, redundant, instrumented in theControl Room, or automatically activated to fulfill its role. Thus, the addition ofthe HEPA filter is not considered a safety-related component.18 EnclosurePG&E Letter DCL-15-069TheEngineering Change Package supporting AST Implementation will ensurethat any Design Class I equipment located adjacent to the ABVS and CRVSfilters and the new TSC filter are qualified to any potential increase in theestimated total integrated radiation dose resulting from the additional post-LOCAradiological release pathways and higher X/Q values addressed in thisapplication.Based on the preceding paragraphs, the methodology and dose consequenceanalyses presented in Attachment 4 do not rely on any newly installed safety-related systems, structure, or components. The containment spray system willnow be credited during sump water recirculation following a LOCA for dosemitigation, but DCPP is already licensed for recirculation containment sprayoperation. Therefore, there are no additions to the EQ list or RG 1.97instrumentation list.No changes have been made in the system responses to accidents. Therefore,there are no additional or new emergency diesel generator (EDG) loads and thetiming of the EDG loads did not change as a result of AST.4. REGULATORY EVALUATION4.1 Applicable Regulatory Requirements/CriteriaTitle 10 Code of Federal Regulations Section 50.36, "Technical specifications"10 CFR 50.36:(c) Technical specifications will include items in the following categories:2) Limiting conditions for operation.(i) Limiting conditions for operation are the lowest functionalcapability or performance levels of equipment required forsafe operation of the facility. When a limiting condition foroperation. of a nuclear reactor is not met, the licensee shallshut down the reactor or follow any remedial actionpermitted by the technical specifications until the conditioncan be met.LCO 3.4.16, "RCS Specific Activity," provides the limiting condition for operationof the Reactor Coolant System (RCS) DEI and DEX. The limit established forDEI is not being revised by this LAR; however, the definition of DEI in TS 1.1,"Definition," is being revised to reference Table 2.1 of FGR No. 11 (Reference 15)as the only acceptable dose conversion factors for determining DEI. Thus, thedefinition of DEI will reference the same dose conversion factors used to19 EnclosurePG&E Letter DCL-15-069determine the reactor coolant dose equivalent iodine curie content for the MSLBand SGTR analysis, as requested by NRC RIS 2006-04 (Reference 8).(ii) A technical specification limiting condition for operation of anuclear reactor must be established for each item meetingone or more of the following criteria:(A) Criterion 1. Installed instrumentation that is used todetect, and indicate in the Control Room, a significantabnormal degradation of the reactor coolant pressureboundary.(B) Criterion 2. A process variable, design feature, oroperating restriction that is an initial condition of a designbasis accident or transient analysis that either assumes thefailure of or presents a challenge to the integrity of a fissionproduct barrier.LCO 3.6.3, "Containment Isolation Valves," Note 1 is revised so that thecontainment purge supply and exhaust flow paths are sealed closed duringoperating MODES 1, 2, 3, and 4. This change is in support of the LOCA doseanalysis assumptions that the containment purge supply and exhaust paths areclosed and therefore, not a release path for radionuclides present in thecontainment following an accident prior to containment isolation. Therefore, thisproposed TS change will eliminate a potential radiological release pathway.(C) Criterion 3. A structure, system, or component that is partof the primary success path and which functions or actuatesto mitigate a design basis accident or transient that eitherassumes the failure of or integrity of a fission product barrier.LCO 3.6.3 is part of the success path which functions to mitigate a LOCA. Therevision to LCO 3.6.3, Note 1 support the LOCA dose analysis assumption thatthe 48-inch valves in the containment purge supply and exhaust paths are sealedclosed and therefore, are not a release path for radionuclides present in thecontainment following a LOCA prior to containment isolation.3) Surveillance requirements. Surveillance requirements arerequirements relating to test, calibration, or inspection to assurethat the necessary quality of systems and components ismaintained, that facility operation will be within safety limits, andthat the limiting conditions for operation will be met.SR 3.4.16.1 is revised to reduce the specific activity of DEX. The revised ASTanalyses that base the released radioactive source terms on RCS specificactivity uses a limit of less than or equal to 270pCi/gm DEX. The limit has been20 EnclosurePG&E Letter DCL-15-069reduced by DCPP to control the noble gas activity in the coolant to levels belowthe design basis values.SR 3.6.3.1, 3.6.3.2, and 3.6.3.7 are being revised to support the revision toLCO 3.6.3 to ensure that the LCO will be met.5) Administrative Controls. Administrative controls are the provisionsrelating to organization and management, procedures,recordkeeping, review and audit, and reporting necessary to assureoperation of the facility in a safe mannerTS 5.5.9, "Steam Generator (SG) Tube Inspection Program," is revised to changethe accident induced leakage performance criteria. The primary-to-secondaryaccident induced leakage rate for any design basis accident, other than a SGTR,shall not exceed the leakage rate assumed in the accident analysis in terms oftotal leakage rate for all SGs. Except during a SGTR, leakage is not to exceed0.75 gpm total for all four SGs. The revised testing criterion for the primary-to-secondary leakage is more restrictive than the current testing criteria andrepresents the leakage rate assumed in the AST dose consequence analyses.TS 5.5.11, "Ventilation Filter Testing Program (VFTP)," is revised to change theallowable methyl iodine penetration testing criteria for the ABVS charcoal filterfrom 15 percent to 5 percent. The allowable methyl iodide penetration is used todetermine charcoal filter efficiency for removing iodine from atmosphericreleases. This proposed revision will support the LOCA dose analysisassumptions with respect to the releases from an RHR pump seal passivefailure, and has been demonstrated to be acceptable.TS 5.5.19, "Control Room Envelope Habitability Program," is revised to replace"whole body or its equivalent to any part of the body" to "TEDE," to be consistentwith the dose criteria specified in 10 CFR 50.67 for AST.In summary, the changes proposed in this LAR in Section 2 support the ASTanalysis assumptions and have been demonstrated to be acceptable. The ASTanalysis results meet the dose criteria specified in 10 CFR 50.67 and Table 6 ofRG 1.183, therefore the requirements of 10 CFR 50.36 continue to be met.Title 10 Code of Federal Regulations Section 50.67, "Alternate Source Term"On December 23, 1999, the NRC published 10 CFR 50.67, "Accident SourceTerm," in the Federal Register. This regulation provides a mechanism forlicensed power reactors to replace the current accident source term used in theDBA analyses with an AST. The direction provided in 10 CFR 50.67 is thatlicensees who seek to revise their current accident source term in design basisradiological consequence analyses shall apply for a LAR under 10 CFR 50.90.Thus, this LAR meets 10 CFR 50.67.21 EnclosurePG&E Letter DCL-15-069General Design CriteriaThe construction of DCPP Units 1 and 2 was significantly complete prior toissuance of 10 CFR 50, Appendix A GDC. DCPP was designed and constructedto comply with Atomic Energy Commission GDC as proposed on July 10, 1967(AEC GDC), except as noted and described in the DCPP UFSAR Chapter 3.Criterion 19, 1971 -Control Room, describes provisions. for a Control Room thatprovides adequate radiation protection to permit access and occupancy underaccident conditions. The dose criterion of GDC 19 was modified to 5 rem TEDEin 1999 to be consistent with 10 CFR 50.67. The results from the dose analysesusing AST source terms and methodologies show that the predicted doseconsequence results are within the allowable regulatory limits of 10 CFR 50.67and GDC 19, 1999. DCPP will conform to GDC 19, 1999, for dose only, forControl Room dose limits of 5 rem TEDE upon implementation of AST. Thus,with the changes proposed in this LAR, DCPP will continue to meet therequirements of 10 CFR 50, Appendix A, GDC Criterion 19.Criterion 52, 1967- Containment Heat Removal Systems (Category A),describes two functions of the containment spray system. One function is tofacilitate heat removal from the containment following an accident. The secondfunction is to remove radioactive iodine isotopes from the containmentatmosphere should these fission products be released in the event of anaccident. The AST LOCA dose analysis assumes that the containment spraysystem now operates during the mode of operation that recirculates containmentsump water for dose mitigation. This change in containment spray systemoperation assumptions in the LOCA dose consequence analysis does not changethe function of the containment spray system or the actual operation ofcontainment spray system. DCPP is not crediting the containment spray systemin the recirculation mode for heat removal and thus not changing the licensingcommitment to Criterion 52, 1967. Therefore Criterion 52,1967, continues to bemet and the plant will continue to provide the basis for safe plant operation.Criterion 54, 1971 -Piping Systems Penetrating Containment, describesrequirements for isolation, including leak detection, and periodic testing. No newcontainment penetrations or lines penetrating the containment are beingproposed with this LAR. Changes to TS 3.6.3 concerning the containment purgesupply and exhaust paths isolation, including leak detection surveillances,includes an enhanced requirement to ensure that these lines remain sealedclosed during MODES 1, 2, 3, and 4 operation. Therefore Criterion 54, 1971,continues to be met and the piping systems penetrating containment will continueto provide the basis for safe plant operation.Criterion 56, 1971- Primary Containment Isolation, describes the provisions forproviding isolating lines that penetrate the containment. No new containment22 EnclosurePG&E Letter DCL-15-069penetrations or lines penetrating the containment are being proposed with thisLAR. Changes to TS 3.6.3 concerning the containment purge supply andexhaust paths isolation, includes an enhanced requirement to ensure that theselines remain sealed closed during MODES 1, 2, 3, and 4 operation. ThereforeCriterion 56, 1971, continues to be met.Criterion 58, 1967- Inspection of Containment Pressure-Reducing Systems(Category A), describes the design provision requirements to facilitate periodicphysical inspection of components of the containment pressure-reducingsystems. UFSAR 3.1.8.22 provides a discussion on how DCPP meetsCriterion 58, 1967, with a brief discussion of the containment pressure-reducingsystems. Containment spray is a containment pressure reducing system and thedescription states that during the recirculation phase, containment sprayoperation is not required. While this statement remains correct for thecontainment pressure-reducing function of containment spray, due to the timingof fission product releases associated with AST, containment spray will berequired during the recirculation phase for fission product cleanup. This LARdoes not change how DCPP meets Criterion 58, 1967; therefore, the criterioncontinues to be met.Criterion 59, 1967- Testing of Containment Pressure-Reducing SystemsComponents (Category A), discusses how the active components of thecontainment pressure-reducing systems are to be tested periodically foroperability and required functional performance. UFSAR 3.1.8.23 provides adiscussion on how DCPP meets Criterion 59, 1967. Containment spray is acontainment pressure reducing system; however, no changes are being made tothe system. The AST LOCA dose analysis now credits the operation of thesystem during recirculation for dose mitigation. Therefore, Criterion 59, 1967,continues to be met.Criterion 60, 1967-Testing of Containment Spray Systems (Category A),discusses the capability to test the delivery capability of the containment spraysystem at a position as close to the spray nozzles as practical. UFSAR 3.1.8.24provides a discussion on how DCPP meets Criterion 60, 1967. No changes arebeing made to the containment spray system. The only change is that the ASTLOCA dose analysis now credits the operation of the system during recirculationfor dose mitigation. Therefore, Criterion 60, 1967, continues to be met.Criterion 62, 1967- Inspection of Air Cleanup Systems (CategoryA), discussesthe physical inspection of containment air cleanup systems, such as ducts, filters,fans, and dampers. UFSAR 3.1.8.26 provides a discussion on how DCPP meetsCriterion 62, 1967. The containment spray system, using sodium hydroxide,serves as the air cleanup system. No changes are being made to thecontainment spray system. The only change is that the AST LOCA dose analysis23 Enclosure.PG&E Letter DCL-1 5-069now credits the operation of the system during recirculation for dose mitigation.Therefore, Criterion 62, 1967, continues to be met.Criterion 63, 1967-Testing of Air Cleanup Systems Components (Category A),discusses the provisions for testing containment air cleanup systems, such asducts, filters, fans, and dampers. UFSAR 3.1.8.27 provides a discussion on howDCPP meets Criterion 63, 1967. The containment spray system, using sodiumhydroxide, serves as the air cleanup system. No changes are being made to thecontainment spray system. The only change is that the AST LOCA dose analysisnow credits the operation of the system during recirculation for dose mitigation.Therefore, Criterion 63, 1967, continues to be met.Criterion 64, 1967-Testing of Air Cleanup Systems (Category A), discusses theprovisions for testing containment air cleanup systems. UFSAR 3.1.8.28provides a discussion on how DCPP meets Criterion 64, 1967. The containmentspray system serves as the air cleanup system. No changes are being made tothe containment spray system. The only change is that the AST LOCA doseanalysis now credits the operation of the system during recirculation for dosemitigation. Therefore, Criterion 64, 1967, continues to be met.RG 1.183, "Alternative Radiological Source Terms for Evaluating Design BasisAccidents at Nuclear Power Reactors," July 2000.The AST methodology used to perform the dose consequence analyses forDCPP is consistent with the guidance of RG 1.183 (Reference 3).Documentation of conformance to RG 1.183 is presented in Attachment 5, withcross-references to specific sections of Attachment 4, where more detail is,provided.RG 1.194, "Atmospheric Relative Concentrations for Control Room RadiologicalHabitability Assessments at Nuclear Power Plants," dated June 2003.RG 1.194, dated June 2003 (Reference 19), Regulatory Position C.1 throughC.3, and the adjustment factor for vertically orientated energetic releases fromsteam relief valves and ADVs allowed by Regulatory Position C.6 are used todetermine short-term onsite X/Q values in support of design basis radiologicalhabitability assessments. Credit is taken for DCPP's dual intake design for theControl Room pressurization air intakes per Regulatory Position C.3.3.2.3.PG&E takes an exception to Regulatory Position C.3.4, for two specific ControlRoom receptors (9.4 meters for Unit 1 containment building to Unit 1 ControlRoom normal intake and 7.8 meters for Unit 2 containment building to Unit 2Control Room normal intake). Use of ARCON96 methodology for these tworelease point-to-receptor distances is considered acceptable since thedominating factors in the calculation are building cross-sectional area and plumemeander and not the normal atmospheric dispersion coefficients. Note that the24 EnclosurePG&E Letter DCL-15-069X/Q values for these two release-receptor cases were developed to establish thebounding X/Q values. However, the y/Q values for these two cases were notbounding, and therefore not used in the dose consequence analyses. Seesection 2.1 and Attachment 4, Section 5.2 for further-detail. -RG 1.145, Revision 1, "Atmospheric Dispersion Models for Potential AccidentConsequence Assessments at Nuclear Power Plants," dated February 1983.The methodology outlined in RG 1.145, Revision 1, is used for calculating groundlevel releases to determine the short-term X/Q values for the EAB and LPZ fordesign basis radiological analyses. All releases are conservatively treated asground level releases, therefore Regulatory Positions C.1.3.2, C.2.1.2, andC.2.2.2 associated with elevated or stack releases are not applicable to DCPP.10 CFR 50.3410 CFR 50.34(b) specifies content requirements for the UFSAR includingevaluations required to show that accident dose criteria are met. Attachment 8contains UFSAR changes (for information only) to support AST implementation.Upon approval of this LAR, UFSAR changes will be made to fulfill theserequirements.4.2 PrecedentThe NRC has previously approved implementation of the AST methodology at anumber of nuclear power plants. In a LAR dated April 26, 2004, PSEG NuclearLLC, proposed to adopt the AST methodology for Salem Units 1 and 2(Reference 28, ADAMS Accession No. ML041280067). The DCPP LAR is similarto the PSEG submittal in that PSEG also proposed to credit recirculation spraysfollowing the LOCA for long term containment iodine removal. PSEG alsoadjusted the Control Room assumed in-leakage by replacing it with values basedupon their Tracer Gas Test. The NRC reviewed and approved theAST LAR forPSEG in a SER dated February 17, 2006 (Reference 29, ADAMS AccessionNo. ML060040322).In a LAR dated June 5, 2002, FirstEnergy Nuclear Operating Company proposedto adopt the AST methodology for Beaver Valley Power Station Units 1 and 2(Reference 30, ADAMS Accession No. ML021620298). The Beaver Valleyamendment used CBI S&W Proprietary computer codes, listed in Section 3 ofAttachment 4, in similar applications. The NRC reviewed and approved the ASTlicense amendment, including the use of the SBI S&W Proprietary computercodes, in a SER dated September 10, 2003 (Reference 31, ADAMS AccessionNo. ML032530204).In a LAR dated February 7, 2001, Omaha Public Power District proposed toadopt the AST methodology for Fort Calhoun Station Unit 1 (Reference 32,25 EnclosurePG&E Letter DCL-15-069ADAMS Accession No. ML010400079). The Fort Calhoun amendment used CBIS&W Proprietary computer codes, listed in Section 3 of Attachment 4, in similarapplications. The NRC reviewed and approved th'e AST license amendment,including the use of the SBI S&W Proprietary computer codes, in a SER datedDecember 5, 2001 (Reference 6, ADAMS Accession No. ML01 3030027).In a LAR dated September 26, 2002, Millstone Unit 2 proposed alternate non-LOCA gap fractions similar to the non-LOCA gap fractions DCPP proposes inSection 4.3 of Attachment 4, as enhancement to RG 1.183 for higher burnup fuel(Reference 33, ADAMS Accession No. ML023040334). The NRC reviewed andapproved the AST license amendment, including the use of the non-LOCA gapfractions, in a SER dated September 20, 2004 (Reference 34, ADAMS AccessionNo. ML042360671).Similarly, in a LAR dated June 5, 2002, Indian Point Unit 3 proposed alternatenon-LOCA gap fractions similar to the non-LOCA gap fractions DCPPproposes in Section 4.3 of Attachment 4 (Reference 35, ADAMS AccessionNo. ML021840136). The NRC reviewed and approved the AST licenseamendment, including the use of the non-LOCA gap fractions, in a SER datedMarch 17,.2003 (Reference 36, ADAMS Accession No. ML030760135).4.3 No Significant Hazards ConsiderationAs provided by 10 CFR 50.67, Pacific Gas & Electric (PG&E) is implementing theuse of an Alternative Source Term (AST) and the dose calculation methodologydescribed in Regulatory Guide (RG) 1.183 to calculate the accident doses to theControl Room, Technical Support Center (TSC), and offsite receptors followingpostulated design basis events that result in the release of radioactive materialfrom reactor fuel at Diablo Canyon Power Plant (DCPP) Units 1 and 2. The ASTand associated methodology for full implementation of AST define the amount,isotopic composition, physical and chemical characteristics, and timing ofradioactive material releases following postulated events. Transport of thematerial to the Control Room, TSC, and offsite areas is modeled, and theresulting Total Effective Dose Equivalent (TEDE) is determined. Regulatoryacceptance criteria account for the sum of the deep-dose equivalent (for externalexposures) and the committed effective dose equivalent (for internal exposures).In accordance with 10 CFR 50.67(b), licensees wishing to adopt an AST mustapply for a license amendment in accordance with 10 CFR 50.90.In support of the revised analyses applying AST, the following TechnicalSpecification (TS) changes are being made: the definition for Dose EquivalentIodine-1 31 (DEI) is revised to be consistent with AST dose conversion factorusage, the limit for reactor coolant system Dose Equivalent Xenon-133 (DEX)activity is decreased to control the noble gas activity in the coolant to .levelsbelow the design basis values, the requirement for containment penetrations is26 EnclosurePG&E Letter DCL-15-069revised to require the 48-inch containment purge supply and exhaust valves tobe sealed closed during operation MODES 1, 2, 3, and 4 eliminating a potentialdose contribution release path, the accident induced leakage performancecriterion for the steam generator tube inspection program is revised to be morerestrictive, and the testing requirement for.the auxiliary building ventilationsystem charcoal filter is also revised to be more restrictive. Other changes to theTSs involve the adoption of terminology on which AST is based.AST methods have been utilized in the analysis of the limiting design basisaccidents, as follows: loss of coolant accident (LOCA), fuel handing accident(FHA) in the containment and in the fuel handling building, locked rotor accident(LRA), control rod ejection accident (CREA), main steam line break (MSLB), andsteam generator tube rupture (SGTR). AST methods have also been utilized inthe analysis of the limiting Condition II event, the loss of load (LOL) accident.Other changes incorporated in the revised analyses include revising atmosphericdispersion factors (X/Q), reducing the minimum decay time before fuelmovement, adding shielding to the Control Room for additional. protection ofControl Room personnel and adding a high efficiency particulate air (HEPA) filterfor additional protection of TSC personnel. In addition, a portion of the 40-inch'Containment Penetration Area Ventilation line and a portion of the 2-inchgaseous radwaste system line which connect to the Plant Vent are beingreclassified from PG&E Design Class II to PG&E Design Class I. Because ASTmethodologies better represent the physical characteristics and timing of theradionuclide release following a postulated LOCA, containment spray is nowrelied upon during the recirculation of sump water for continued removal of iodineand particulate from the containment atmosphere for spray duration (injectionplus recirculation) greater than 6.25 hours. In addition, setpoint changes arebeing made to the Control Room intake radiation monitors to incorporate theeffect of all possible release points from a FHA.PG&E has evaluated whether or not a significant hazards consideration isinvolved with the proposed amendment by focusing on the three standards setforth in 10 CFR 50.92, "Issuance of Amendment," as discussed below: .1. Does the proposed change involve a significant increase in the probabilityor consequences of an accident previously evaluated?Response: No.This license amendment does not physically impact any system, structure,or component (SSC) that is a potential initiator of an accident. Therefore,implementation of AST, the AST assumptions and inputs, the proposed TSchanges, and new X/Q values have no impact on the probability forinitiation of any design basis accident. Once the occurrence of anaccident has been postulated, the new accident source term and ,/Q27 EnclosurePG&E Letter DCL-15-069values are inputs to analyses that evaluate the radiological consequencesof the postulated events.Reactor coolant specific activity, testing criteria of charcoal filters, and theaccident induced primary-to-secondary system leakage performancecriterion are not initiators for any accident previously evaluated. Theproposed change to require the 48-inch containment purge valves to besealed closed during operating MODES 1, 2, 3, and 4 is not an accidentinitiator for any accident previously evaluated. The change in theclassifications of a portion of the 40-inch Containment Penetration AreaVentilation line and a portion of the 2-inch gaseous radwaste system lineis also not an accident initiator for any accident previously evaluated.Thus, the proposed TS changes and AST implementation will not increasethe probability of an accident.The change to the decay time prior to fuel movement is not an accidentinitiator. Decay time is used to determine the source term for the doseconsequence calculation following a potential FHA and has no effect onthe probability of the accident. Likewise, the change to the Control Roomradiation monitors setpoint cannot cause an accident and the operation ofcontainment spray during the recirculation phase is used for mitigation of aLOCA, and thus not an accident initiator.As a result, there are no proposed changes to the parameters orconditions that could contribute to the initiation of an accident previouslyevaluated in Chapter 15 of the Updated Final Safety Analysis Report(UFSAR). As such, the AST cannot affect the probability of an accidentpreviously evaluated.Regarding accident consequences, equipment and components affectedby the proposed changes are mitigative in nature and relied upon once theaccident has been postulated. The license amendment implements a newcalculation methodology for determining accident consequences and doesnot adversely affect any plant component or system that is credited tomitigate fuel damage. Subsequently, no conditions have been createdthat could significantly increase the consequences of any accidentspreviously evaluated.Requiring that the 48-inch cqntainment purge supply and exhaust valvesbe sealed closed during operating MODES 1, 2, 3, and 4 eliminates apotential path for radiological release following events that result inradioactive material releases to the containment, thus reducing potentialconsequences of the event. The steam generator tube inspection testingcriterion for accident induced leakage is being changed, resulting in lowerleakage rates, and thus less potential releases due to primary-to-28 EnclosurePG&E Letter DCL-15-069secondary leakage. The auxiliary building ventilation system allowablemethyl iodide penetration limit is being changed, which results in morestringent testing requirements, and thus higher filter efficiencies forreducing potential releases.Changes to the operation of the containment spray system to requireoperation during the recirculation mode are also mitigative in nature.While the plant design basis has always included the ability to implementcontainment spray during recirculation, this license amendment nowrequires operation of containment spray in the recirculation mode for dosemitigation. DCPP is designed and licensed to operate using containmentspray in the recirculation mode. As such, operation of containment sprayin the recirculation mode has already been analyzed,. evaluated, and iscurrently controlled by Emergency Operating Procedures. Usage ofrecirculation spray reduces the consequence of the postulated event.Likewise, the additional shielding to the Control Room and the addition ofa HEPA filter to the TSC ventilation system reduces the consequences ofthe postulated event to the Control Room and TSC personnel. Loweringthe limit for DEX lowers potential releases. By reclassifying a portion of,the 40-inch Containment Penetration Area Ventilation line and a portion ofthe 2-inch gaseous radwaste system line to PG&E Design Class I, theselines will be seismically qualified, thus assuring that post-LOCA releasepoints are the same as those used for determining X/Q values.The change to the decay time from 100 hours to 72 /hours prior to fuelmovement is an input to the FHA. Although less decay will result in higherreleased activity, the results of the FHA dose consequence analysisremain within the dose acceptance criteria of the event. Also, theradiation levels'to an operator from a raised fuel assembly may increasedue to a lower decay time, however, any exposure will continue to bemaintained under 10 CFR 20 limits by the'plant Radiation ProtectionProgram.Plant-specific radiological analyses have been performed using the ASTmethodology, assumption and inputs, as well as new x/Q values. Theresults of the dose consequences analyses demonstrate that theregulatory acceptance criteria are met for each analyzed event.Implementing the AST involves no facility equipment, procedure, orprocess changes that could significantly affect the radioactive materialactually released during an event. Subsequently, no conditions have beencreated that could significantly increase the consequences of any of theevents being evaluated.29 EnclosurePG&E Letter DCL-15-069Based on the above discussion, the proposed changes do not involve asignificant increase in the probability or consequences of an accidentpreviously evaluated.2. Does the proposed change create the possibility of a new or differentaccident from any accident previously evaluated?Response: No.This license amendment does not alter or place any SSC in aconfiguration outside its design or analysis limits and does not create anynew accident scenarios.The AST methodology is not an accident initiator, as it is a method used toestimate resulting postulated design basis accident doses. The proposedTS changes reflect the plant configuration that supports implementation ofthe new methodology and supports reduction in dose consequences.DCPP is designed and licensed to operate using containment spray in therecirculation mode. This change will not affect any operational aspect ofthe system or any other system, thus no new modes of operation areintroduced by the proposed change.The function of the radiation monitors has not changed; only the setpointhas changed as a result of an assessment of all potential releasepathways. The continued- operation of containment spray and theradiation monitor setpoint change do not create any new failure modes,alter the nature of events postulated in the UFSAR, nor introduce anyunique precursor mechanism.Requiring the 48-inch containment purge valves to be sealed closedduring operating MODES 1, 2, 3, and 4 does not introduce any newaccident precursor. This change only eliminates a potential release pathfor radionuclides following a LOCA.The proposed TS testing criteria for the auxiliary building ventilationsystem charcoal filters and the proposed performance criteria for steamgenerator tube integrity also cannot create an accident, but results inrequiring more efficient filtration of potentially released. iodine and lessallowable primary-to-secondary leakage. The proposed changes to theDEX activity limit, the TS terminology, and the decay time of the fuelbefore movement are also unrelated to accident initiators.The only physical changes to the plant being made in support of AST isthe addition of Control Room shielding in an area previously modified, theaddition of a HEPA filter at the intake of the TSC normal ventilationsystem, and the upgrade to the damper actuators, pressure switches, and30 EnclpsurePG&E Letter DCL-15-069damper solenoid valves to support reclassifying a portion of theContainment Penetration Area Ventilation line to PG&E Design Class I.Both Control Room shielding and HEPA filtration are mitigative in natureand do not have any impact on plant operation or system responsefollowing an accident. The Control Room modification for adding theshielding will meet applicable loading limits, so the addition of theshielding cannot initiate a failure. Upgrading damper actuators, pressureswitches, and damper solenoid valves involve replacing existingcomponents with components that are PG&E Design Class I. Therefore,the addition of shielding, a HEPA filter, and upgrading components cannotcreate a new or different kind of accident.Since the function of the SSCs has not changed for AST implementation,no new failure modes are created by this proposed change. The ASTchange itself does not have the capability to initiate accidents.Therefore, the proposed change does not create the possibility of a new ordifferent type of accident from any accident previously evaluated.3. Does the proposed change involve a significant reduction in a margin ofsafety?Response: No.Implementing the AST is relevant only to calculated dose consequences ofpotential design basis accidents evaluated in Chapter 15 of the UFSAR.The changes proposed in this license amendment involve the use of anew analysis methodology and related regulatory acceptance criteria.New atmospheric dispersion factors, which are based on site specificmeteorological data, were calculated in accordance with regulatoryguidelines. The proposed TS, TS Bases, and UFSAR changes reflect theplant configuration that will support implementation of the newmethodology and result in operation in accordance with regulatoryguidelines that support the revisions to the radiological analyses of thelimiting design basis accidents. Conservative methodologies, per theguidance of RG 1.183, have been used in performing the accidentanalyses. The radiological consequences of these accidents are all withinthe regulatory acceptance criteria associated with the use of ASTmethodology.The change to the minimum decay time prior to fuel movement results inhigher fission product releases after a FHA. However, the results of theFHA doseconsequence analysis remain within the dose acceptancecriteria of the event.31 EnclosurePG&E Letter DCL-15-069The proposed changes continue to ensure that the dose consequences ofdesign basis accidents at the exclusion area, low population zoneboundaries, in the TSC, and in the Control Room are within thecorresponding acceptance criteria presented in RG 1.183 and10 CFR 50.67. The margin of safety for the radiological consequences ofthese accidents is provided by meeting the applicable regulatory limits,which are set at or below the 10 CFR 50.67 limits. An acceptable marginof safety is inherent in these limits.Therefore, the proposed change does not involve a significant reduction ina margin of safety.Based on the above evaluation, PG&E concludes that the proposed change doesnot involve a significant hazards consideration under the standards set forth in10 CFR 50.92(c), and accordingly, a finding of "no significant hazardsconsideration" is justified.4.4 ConclusionsIn conclusion, based on the considerations discussed above, (1) there isreasonable assurance that the health and safety of the public will not beendangered by operation in the proposed manner, (2) such activities will beconducted in compliance with the Commission's regulations, and (3) theissuance of the amendment will not be inimical to the common defense andsecurity or to the health and safety of the public.5. ENVIRONMENTAL CONSIDERATIONPG&E has evaluated the proposed amendment and has determined that theproposed amendment does not involve (i) a significant hazards consideration, (ii)a significant change in the types or significant increase in the amounts of anyeffluents that may be released offsite, or (iii) a significant increase in individual orcumulative occupational radiation exposure.Based on the evaluation performed under the standards set forth in10 CFR 50.92(c), PG&E concludes that the proposed amendment does notinvolve a significant hazards consideration. AST only involves a change inaccident dose calculation inputs and methodology. Calculated doses meet TEDEcriteria.No aspect of implementing the AST involves facility equipment, procedure, orprocess changes that would increase actual onsite doses if an event were tooccur.The AST does not result in actual or calculated changes in the normal radiationlevels in the facility or in the type or quantity of radioactive materials processed32 EnclosurePG&E Letter DCL-15-069during normal operation. Implementation of.the AST also has no effect on theactual or calculated effluents arising from normal operation.Accordingly, the proposed amendment meets the eligibility criterion forcategorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to10 CFR 51.22(b), no environmental impact statement or environmentalassessment need be prepared in connection with the proposed amendment.33 EnclosurePG&E Letter DCL-15-0696. REFERENCES1. Code of Federal Regulations (CFR), 10 CFR 50.67, "Accident Source Terms."2. NRC Letter "Diablo Canyon Power Plant, Units 1 and 2 -Issuance ofAmendments RE: Control Room, Auxiliary Building, and Fuel Handling BuildingVentilation Systems (TAC Nos. MB8485 and MB8486)," dated February 27,2004.3. Regulatory Guide 1.183, "Alternative Radiological Source Terms for EvaluatingDesign Basis Accidents at Nuclear Power Reactors," July 2000.4. Technical Information Document 14844, "Calculation of Distance Factors forPower and Test Reactor Sites," 1962.5. NUREG-0933, "Resolution of Generic Safety Issues," dated December 2011.6. NRC Letter "Fort Calhoun Station, Unit No. 1 -Issuance of Amendment (TACNo. MB1221)," dated December 5, 2001 (ADAMS Accession No. ML013030027).7. SECY-98-154, "Results of the Revised (NUREG-1465) Source TermRebaselining For Operating Reactors," dated June 30, 1998.8. NRC Regulatory Issue Summary (RIS) 2006-04, "Experience withImplementation of Alternative Source Terms," dated March 7, 2006.9. NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reportsfor Nuclear Power Plants: LWR Edition," dated June 1987.10. Code of Federal Regulations 10 CFR 50.44, "Combustible Gas Control forNuclear Power Reactors."11. NRC Letter "Diablo Canyon Power Plant, Unit 1 (TAC No. MC1678) and UnitNo. 2 (TAC No. MC1679) -Issuance of Amendments RE: Elimination ofRequirements for Hydrogen Recombiners and Hydrogen Monitors," dated May 4,2004.12. Diablo Canyon Power Plant Updated Final Safety Analysis Report, Revision 21.13. Code of Federal Regulations, 10 CFR 50, Appendix A, General Design Criteria19, "Control Room," dated 1971.14. NUREG-0737, Supplement 1, "Clarification of TMI Action Plan Requirements,"dated January 1983.15. Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and AirConcentration and Dose Conversion Factors for lnhalation, Submersion, andIngestion," dated 1988.16. Regulatory Guide 1.145, Revision 1, "Atmospheric Dispersion Models forPotential Accident Consequence Assessments at Nuclear Power Plants," datedFebruary 1983.34 EnclosurePG&E Letter DCL-15-06917. Regulatory Guide 1.111, Revision 1, "Methods for Estimating AtmosphericTransport and Dispersion of Gaseous Effluents in Routine Releases from LightWater Cooled Reactors," dated July 1977.18. Ramsdell, J. V. Jr. and C. A. Simonen, "Atmospheric Relative Concentrations inBuilding Wakes." Prepared by Pacific Northwest Laboratory for the U.S. NuclearRegulatory Commission, PNL-10521, NUREG/CR-6331, Revision 1, May 1997.19. Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control RoomRadiological Habitability Assessments at Nuclear Power Plants," dated June2003.20. NUCON International Inc., "Control Room Habitability Tracer Gas Leak Testing atDiablo Canyon Power Plant," dated December 2012.21. NRC Letter "Diablo Canyon Power Plant, Unit Nos. 1 and 2 -Issuance ofAmendments RE: Containment Spray During the Recirculation Phase of a LOCA(TAC Nos. MA1408 and MA1409)," dated February 9, 2000.22. NUREG-0737, "Clarification of TMI Action Plan Requirements," dated November1980.23. NUREG-0800, Standard Review Plan 15.0.1, Revision 0, "RadiologicalConsequence Analyses using Alternative Source Terms," dated July 2000.24. Improved Standard Technical Specification Traveler, TSTF-490, Revision 0,"Deletion of E Bar Definition and Revision to RCS Specific Activity-Tech Spec,"dated September 13, 2005.25. NUREG-1431, Revision 4, Volume 1, "Standard Technical Specifications,Westinghouse Plants," dated April 2012.26. NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants,"dated February 1995.27. NUREG-0696, "Functional Criteria for Emergency Response Facilities," datedFebruary 1981.28. PSEG Nuclear LLC Letter "Implementation of Alternative Source Term (AST)Request for Changes to Technical Specifications and Updated Final SafetyAnalysis Report Salem Nuclear Generating Station, Units 1 and 2, FacilityOperating Licenses DBR-70 and DPR-75, Docket Nos. 50.272 and 50-311,"dated April 26, 2004 (ADAMS Accession No. ML041280067).29. NRC Letter "Salem Nuclear Generating Station, Unit Nos. I and 2, Issuance ofAmendments RE: Alternate Source Term (TAC Nos. MC3094 and MC3095),"dated February 17, 2006, (ADAMS Accession No. ML060040322).30. FirstEnergy Nuclear Operating Company Letter L-02-069 "Beaver Valley PowerStation, Unit 1 No. I and No. 2 BV-1 Docket No. 50-334, License No. DPR-66,BV-2 Docket No. 50-412, License No. NPF-73, License Amendment RequestNos. 300 and 172," dated June 5, 2002 (ADAMS Accession No. ML021620298).35 EnclosurePG&E Letter DCL-1 5-06931. NRC Letter "Beaver Valley Power Station, Unit Nos. I and 2 -Issuance ofAmendment RE: Selective Implementation of Alternate Source Term and ControlRoom Habitability Technical Specification Changes (TAC Nos. MB5303 andMB5304)," dated September 10, 2003 (ADAMS Accession No. ML032530204).32. Omaha Public Power District Letter LIC-01-0010, "Application for Amendment ofOperating License," dated February 7, 2001 (ADAMS Accession No.ML010400079).33. Dominion Nuclear Connecticut, Inc. Letter, "Millstone Power Station, Unit No. 2,License Basis Document Change Request (LBDCR) 2-18-02, SelectiveImplementation of the Alternative Source Term -Fuel Handling AccidentAnalyses," dated September 26, 2002 (ADAMS Accession No. ML023040334).34. NRC Letter "Millstone Power Station, Unit Nos. 2 -Issuance of Amendment RE:Selective Implementation of Alternate Source Term (TAC No. MB6479)," datedSeptember 20, 2004 (ADAMS Accession No. ML042360671).35. Entergy Nuclear Northeast Letter IPN-02-044, "Indian Point Nuclear GeneratingUnit No. 3, Docket No. 50-286, Proposed Changes to Technical Specifications:Selective Adoption of Alternate Source Term and Incorporation of GenericChanges; TSTF-51, TSTF-68, and TSTF-312," dated June 5, 2002 (ADAMSAccession No. ML021840136).36. NRC Letter "Indian Point Nuclear Generating Unit No. 3 -Issuance ofAmendment RE: Selective Adoption of Alternate Source Term (TACNo. MB5382)," dated March 17, 2003 (ADAMS Accession No. ML030760135).36 EnclosurePG&E Letter DCL-15-069Table 1 -AST Site Boundary and Control Room TEDE (rem)Regulatory Control RegulatoryAccident EAB(1) LPZ(2) Limit Room LimitLoss of Coolant Accident 5.6(') 1 25 3.7 (0.7) (3)5Fuel Handling Accident in Fuel 1.5 0.2 6.3 1.1 5Handling BuildingFuel Handling Accident in 47 5ContainmentLocked Rotor Accident 0.8 0.2 2.5 2.4 5Control Rod Ejection AccidentContainment Release 0.7 0.3 6.3 3.4 5Secondary Release 0.7 0.2 0.5Main Steam Line BreakPre-incident iodine Spike 0.1 <0.1 25 2.0 5Accident-Initiated Iodine Spike 0.7 0.2 2.5 4.1Steam Generator Tube RupturePre-incident iodine Spike 1.3 0.1 25 0.6 5Accident-Initiated Iodine Spike 0.7 <0.1 2.5 0.3Loss of LoadPre-incident iodine Spike <0.1 <0.1 2.5 <0.1 5Accident-Initiated Iodine Spike <0.1 <0.1 2.5 <0.1Notes(1) EAB doses are based on worst 2-hour period following onset of accident. Except as noted, the maximum2-hour dose period for the EAB dose for each of the accidents is the 0 to 2 hours' time period." LOCA: 24-26 hours (based on RHR Pump Seal Failure; see note 4 below for additional information)" LRA: 8.73 to 10.73 hours" MSLB (Accident-Initiated Spike case): 7.6 to 9.6 hours" LOL (Accident-Initiated Spike case): 8.73 to 10.73 hours.(2) LPZ Doses are based on the duration of the release.(3) The dose presented represents the operator dose due to occupancy. Value shown in parenthesis representsthat portion of the total dose reported that is the contribution of direct shine from contained sources/externalcloud. The dose to the Control Room operator during routine access for the 30 day duration of the accident isdiscussed in Attachment 4, Section 7.2.6 and summarized in the text of Attachment 4, Section 8.0.(4) The maximum 2 hour EAB dose is based on the assumed RHR pump seal passive failure resulting in a 50 gpmleak of sump water occurring at t=24 hours for 30 minutes. The RHR pump seal passive failure is considered apart of DCPP licensing basis with respect to passive system failure. If this assumed passive failure were notincluded, the maximum 2 hour dose at the EAB would occur between t=0.5 hours to t=2.5 hours (i.e., during thepost-LOCA ex-vessel release phase) and would be 3.4 rem.37 EnclosureAttachment 1PG&E Letter DCL-15-069ATTACHMENT IProposed Technical Specification Changes(MARKUP)Changes are proposed to the following Technical Specifications:1. Specification 1.1, Definitions, Dose Equivalent 1-1312. Specification 3.4.16, RCS Specific Activity3. Specification 3.6.3, Containment Isolation Valves4. Specification 5.5.9, Steam Generator (SG) Tube Inspection Program5. Specification 5.5.11, Ventilation Filter Testing Program (VFTP)6. Specification 5.5.19, Control Room Envelope Habitability Program Definitions1.11.1 Definitions (continued)DOSE EQUIVALENT 1-131DOSE EQUIVALENT 1-131 shall be that concentration of1-131 (microcuries per gram) that alone would produce thesame dose when inhaled as the combined activities ofiodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actuallypresent. The determination of DOSE EQUIVALENT 1-131shall be performed using thyroid dose conversion faGctorsfrom Table MI of TID 14844, AEC, 1962, "Calculation oGDistance for Power and Test Sites," orTable E 7 of Regulator; Guide 1.109, Revision 1, NRC,1977, Or International Commission on RadiologicalProtection (iCRP) PublicatinG 30), 1979, Supplement to Part1, pages 192 212, Table titled "CmmtedDse Equivalentin Target Organs or Tissues per Intake of Unit Acvity," o-,rthe committed thyroid dose conversion factors from Table2.1 of EPA Federal Guidance Report No. 11, 1988, "LimitingValues of Radionuclide Intake and Air Concentration andDose Conversion Factors for Inhalation, Submersion, andIngestion."DOSE EQUIVALENT XE-1 33 shall be that concentration ofXe-1 33 (microcuries per gram) that alone would produce thesame acute dose to the whole body as the combinedactivities of noble gas nuclides Kr-85m, Kr-87, Kr-88,Xe-133m, Xe-1 33, Xe-1 35m, Xe-1 35, and Xe-1 38 actuallypresent. If a specific noble gas nuclide is not detected, itshould be assumed to be present at the minimum detectableactivity. The determination of DOSE EQUIVALENT XE-1 33shall be performed using effective dose conversion factorsfor air submersion listed in Table 111.1 of EPA FederalGuidance Report No. 12, 1993, "External Exposure toRadionuclides in Air, Water, and Soil."DOSE EQUIVALENT XE-1 33(continued)1.1-3 Unit I -Amendment No. 135, 155,156, 192,Unit 2 -Amendment No. 135, 155,156, 193,DIABLO CANYON -UNITS 1 & 2Rev 9 Page 3 of 24 RCS Specific Activity3.4.16SURVEILLANCE REQUIREMENTSSURVEILLANCE FREQUENCYSR 3.4.16.1 ---------------------NOTE -------------------------- In accordance withOnly required to be performed in MODE 1. the Surveillance------------------------------------------------------------..- Frequency C ontrolVerify reactor coolant DOSE EQUIVALENT XE-133 Programspecific activity < 690-0270.0 pCi/gm.SR 3.4.16.2 --------------------NOTE -----------------Only required to be performed in MODE 1.Verify reactor coolant DOSE EQUIVALENT 1-131 In accordance withspecific activity < 1.0 pCi/gm. the SurveillanceFrequency ControlProgramANDBetween 2 and 6.hours after aTHERMALPOWER change of> 15% RTP withina 1 hour period.DIABLO CANYON -UNITS 1 & 2Rev 10 Page 38 of 403.4-36 Unit I -Amendment No. 135,l-92,200,Unit 2 -Amendment No. 135,!93,201, Containment Isolation Valves3.6.33.6 CONTAINMENT SYSTEMS3.6.3 Containment Isolation ValvesLCO 3.6.3Each containment isolation valve shall be OPERABLE.APPLICABILITY:ACTIONSMODES 1, 2, 3, and 4.------------------------ N O TE S .1. Penetration flow path(s) except no mr.e than two of three flow paths for CoR.taincnt01 m/N 0, in!,- 14/ ,k-, 0+ Ain t-N,/ir/+ ;,il +11 IO InroO jrNrI F O+kO f+ fi/Nfi;;^r.. i il ......... Ln[JLi U L LJi i ........ ILJI48-inch purge valve flow paths, may be unisolated intermittently under administrativecontrols.2. Separate Condition entry is allowed for each penetration flow path.3. Enter applicable Conditions and Required Actions for systems made inoperable bycontainment isolation valves.4. Enter applicable Conditions and Required Actions of LCO 3.6.1, "Containment," whenisolation valve leakage results in exceeding the overall containment leakage rateacceptance criteria.CONDITION REQUIRED ACTION COMPLETION TIMEA. ---------NOTE--------- A.1 Isolate the affected 4 hoursOnly applicable to penetration flow path bypenetration flow paths with use of at least onetwo containment isolation closed and de-activatedvalves. automatic valve, closed------------------------------------ m anual valve, blindOne or more penetration flange, or check valveflow paths with one with flow through thecontainment isolation valve valve secured.inoperable except for acontainment purge supplyand exhaust valve orpressure/vacuum reliefvalve leakage not withinlimit. (continued)DIABLO CANYON -UNITS I & 2Rev 6 Page 5 of 183.6-5Unit 1 -Amendment No. -435,Unit 2 -Amendment No. 4-35, Containment Isolation Valves3.6.3SURVEILLANCE REQUIREMENTSSURVEILLANCE FREQUENCYSR 3.6.3.1 Net-usedVerify each 48 inch purge valve is In accordance with thesealed closed, except for one purge valve in a Surveillance Frequencypenetration flow path while in Condition D of this Control ProgramLCO.SR 3.6.3.2 Verify each 48 inch containment purge supply In accordance with theand exhaust and 12 inch vacuum/pressure relief Surveillance Frequencyvalve is closed, except when these valves are Control Programopen for pressure control, ALARA or air qualityconsiderations for personnel entry, or forSurveillances that require the valves to be open.SR 3.6.3.3 ----------------------------- NOTE ---------------------------Valves and blind flanges in high radiation areasmay be verified by use of administrative controls.Verify each containment isolation manual valve In accordance with theand blind flange that is located outside Surveillance Frequencycontainment and not locked, sealed or otherwise Control Programsecured and required to be closed duringaccident conditions is closed, except forcontainment isolation valves that are open underadministrative controls.SR 3.6.3.4 ----------------------------- NOTE --------------Valves and blind flanges in high radiation areasmay be verified by use of administrative means.Verify each containment isolation manual valve Prior to entering MODEand blind flange that is located inside 4 from MODE 5 if notcontainment and not locked, sealed or otherwise performed within thesecured and required to be closed during previous 92 daysaccident conditions is closed, except forcontainment isolation valves that are open underadministrative controls.SR 3.6.3.5 Verify the isolation time of each automatic power In accordance with theoperated containment isolation valve is within Inservice Testinglimits. ProgramSR 3.6.3.6 Not used(continued)DIABLO CANYON -UNITS 1 & 2Rev 6 Page 9 of 183.6-9 Unit 1 -Amendment. No. 135,200,Unit 2 -Amendment No. 135,20!,
+Containment Isolation Valves3.6.3SURVEILLANCE REQUIREMENTS (continued)SURVEILLANCEFREQUENCYSR 3.6.3.7----------------------------- NOTE --------------This surveillance is not required when thepenetration flow path is isolated by a leak testedblank flange.Perform leakage rate testing for containmentpurge supply and exhaust and vacuum/pressurerelief valves with resilient seals.In accordance with theSurveillance FrequencyControl ProgramANDPFo containment purFgesupply and exhaustvalves only, Within 92days opening theSR 3.6.3.8 Verify each automatic containment isolation valve In accordance with thethat is not locked, sealed or otherwise secured in Surveillance Frequencyposition, actuates to the isolation position on an Control Programactual or simulated actuation signal.SR 3.6.3.9 Not usedSR 3.6.3.10 Verify each 12 inch containment In accordance with thevacuum/pressure relief valve is blocked to restrict Surveillance Frequencythe valve from opening > 50'. Control ProgramSR 3.6.3.11 Not usedDIABLO CANYON -UNITS 1 & 2Rev 6 Page10 of 183.6-10Unit 1 -Amendment No. !35,!75,200 ,Unit 2 -Amendment No. 4,35477,20!,
+Programs and Manuals5.55.5 Programs and Manuals (continued)5.5.9 Steam Generator (SG) Tube Inspection ProgramA Steam Generator Program shall be established and implemented to ensure that SGtube integrity is maintained. In addition, the Steam Generator Program shall include thefollowing provisions:a. Provisions for condition monitoring assessments.Condition monitoring assessment means an evaluation of the "as found"condition of the tubing with respect to the performance criteria for structuralintegrity and accident induced leakage. The "as found" condition refers to thecondition of the tubing during an SG inspection outage, as determined from theinservice inspection results or by other means, prior to the plugging of tubes.Condition monitoring assessments shall be conducted during each outage duringwhich the SG tubes are inspected or plugged to confirm that the performancecriteria are being met.b. Performance criteria for SG tube integrity.SG tube integrity shall be maintained by meeting the performance criteria fortube structural integrity, accident induced leakage, and operational LEAKAGE.1. Structural integrity performance criterion: All in-service steam generatortubes shall retain structural integrity over the full range of normal operatingconditions (including startup, operation in the power range, hot standby,and cool down and all anticipated transients included in the designspecification) and design basis accidents. This includes retaining a safetyfactor of 3.0 against burst under normal steady state full power operationprimary-to-secondary pressure differential and a safety factor of 1.4against burst applied to the design basis accident primary-to-secondarypressure differentials. Apart from the above requirements, additionalloading conditions associated with the design basis accidents, orcombination of accidents in accordance with the design and licensingbasis, shall also be evaluated to determine if the associated loadscontribute significantly to burst or collapse. In the assessment of tubeintegrity, those loads that do significantly affect burst or collapse shall bedetermined and assessed in combination with the loads due to pressurewith a safety factor of 1.2 on the combined primary loads and 1.0 on axialsecondary loads.2. Accident induced leakage performance criterion: The primary to secondaryaccident induced leakage rate for any design basis accident, other than aSG tube rupture, shall not exceed the leakage rate assumed in theaccident analysis in terms of total leakage rate for all SGs and leakagerate for an indi-idual SG. Except during a SG tube rupture, leakage is alsonot to exceed 40.75 gallon per minute per-total for all four SGs.(continued)DIABLO CANYON -UNITS 1 & 2 5.0-10 Unit 1 -Amendment No. 4-98,Rev29 Page10of27 Unit 2 -Amendment No. 4-99, Programs and Manuals5.55.5 Programs and Manuals5.5.11 Ventilation Filter Testing Program (VFTP) (continued)c. Demonstrate for each of the ESF systems that a laboratory test of a sample ofthe charcoal absorber, when obtained as described in Regulatory Guide 1.52,Revision 2, shows the methyl iodide penetration less than the value specifiedbelow when tested in accordance with ASTM D3803-1989 at a temperature of300C and at the relative humidity specified below. Laboratory testing shall becompleted at least once per 24 months and after every 720 hours of charcoaloperation.ESF Ventilation System Penetration RHControl Room 2.5% 95%Auxiliary Building 45.0% 95%Fuel Handling Building 15.0% 95%d. Demonstrate for each of the ESF systems that the pressure drop across thecombined HEPA filters and the charcoal adsorbers is less than the valuespecified below when tested in accordance with ANSI N510-1980 at the systemflowrate specified below +/- 10% at least once per 24 months.ESF Ventilation System Delta P FlowrateControl Room 3.5 in. WG 2100 cfmAuxiliary Building 3.7 in. WG 73,500 cfmFuel Handling Building 4.1 in. WG 35,750 cfmThe provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring ProgramThis program provides controls for potentially explosive gas mixtures contained in theWaste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks,and the quantity of radioactivity contained in temporary unprotected outdoor liquidstorage tanks.The gaseous radioactivity quantities shall be determined following the methodology inRegulatory Guide 1.24 "Assumptions Used For Evaluating the Potential RadiologicalConsequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure."The liquid radwaste quantities shall be maintained such that 10 CFR Part 20 limits aremet.(continued)DIABLO CANYON -UNITS I & 2 5.0-13 Unit 1 -Amendment No. 135,142,163,-0,Rev 29 Page 13 of 27 Unit 2- Amendment No. 135,142,165,199, Programs and Manuals5:55.5 Programs and Manuals (continued)5.5.19 Control Room Envelope Habitability ProqramA Control Room Envelope (CRE) Habitability Program shall be established andimplemented to ensure that CRE habitability is maintained such that, with anOPERABLE Control Room Ventilation System (CRVS), CRE occupants can control thereactor safely under normal conditions and maintain it in a safe condition following aradiological event, hazardous chemical release, or a smoke challenge. The programshall ensure that adequate radiation protection is provided to permit access andoccupancy of the CRE under design basis accident (DBA) conditions without personnelreceiving radiation exposures in excess of 5 rem whole-body TEDE or its cquivalent toany part of the body for the duration of the accident. The program shall include thefollowing elements:a. The definition of the CRE and the CRE boundary.b. Requirements for maintaining the CRE boundary in its design condition, includingconfiguration control and preventive maintenance.c. Requirements for (i) determining the unfiltered air inleakage past the CREboundary into the CRE in accordance with the testing methods and at theFrequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197,"Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors,"Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequenciesspecified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.d. Measurement, at designated locations, of the CRE pressure relative to allexternal areas adjacent to the CRE boundary during the pressurization mode ofoperation by one train of the CRVS, operating at the flow rate required by theVFTP, at a Frequency of 24 months on a STAGGERED TEST BASIS. Theresults shall be trended and used as part of the 24 month assessment of theCRE boundary.e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shallbe stated in a manner to allow direct comparison to the unfiltered air inleakagemeasured by the testing described in paragraph c. The unfiltered air inleakagelimit for radiological challenges is the inleakage flow rate assumed in thelicensing basis analyses of DBA consequences. Unfiltered air inleakage limits forhazardous chemicals must ensure that exposure of CRE occupants to thesehazards will be within the assumptions in the licensing basis.f. The provisions of SR 3.0.2 are applicable to the Frequencies required byparagraphs c and d for determining CRE unfiltered inleakage and assessing CREhabitability, and measuring CRE pressure and assessing the CRE boundary.DIABLO CANYON -UNITS 1 & 2 5.0-17a Unit 1 -Amendment No. 2-14-,Rev 29 Page 18 of 27 Unit 2 -Amendment No. 2-2, EnclosureAttachment 2PG&E Letter DCL-15-069ATTACHMENT 2Proposed Technical Specification Changes(RETYPED)Remove Pages1.1-33.4-363.6-53.6-93.6-105.0-105.0-135.0-17aInsert Pages1.1-33.4-363.6-53.6-93.6-105.0-105.0-135.0-17a Definitions1.11.1 Definitions (continued)DOSE EQUIVALENT 1-131DOSE EQUIVALENT XE-1 33DOSE EQUIVALENT 1-131 shall be that concentration of1-131 (microcuries per gram) that alone would produce thesame dose when inhaled as the combined activities ofiodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actuallypresent. The determination of DOSE EQUIVALENT 1-131shall be performed using the committed thyroid doseconversion factors from Table 2.1 of EPA Federal GuidanceReport No. 11, 1988, "Limiting Values of RadionuclideIntake and Air Concentration and Dose Conversion Factorsfor Inhalation, Submersion, and Ingestion."DOSE EQUIVALENT XE-1 33 shall be that concentration ofXe-1 33 (microcuries per gram) that alone would produce thesame acute dose to the whole body as the combinedactivities of noble gas nuclides Kr-85m, Kr-87, Kr-88,Xe-1 33m, Xe-133, Xe-135m, Xe-1 35, and Xe-1 38 actuallypresent. If a specific noble gas nuclide is not detected, itshould be assumed to be present at the minimum detectableactivity. The determination of DOSE EQUIVALENT XE-133shall be performed using effective'dose conversion factorsfor air submersion listed in Table 111.1 of EPA FederalGuidance Report No. 12, 1993, "External Exposure toRadionuclides in Air, Water, and Soil."I(continued)1.1-3 Unit 1 -Amendment No. 135, 155,456, 192,Unit 2 -Amendment No. 135, 155,1546, 193,DIABLO CANYON -UNITS 1 & 2 RCS Specific Activity3.4.16SURVEILLANCE REQUIREMENTSSURVEILLANCE FREQUENCYSR 3.4.16.1 -------------------------------- NOTE -------------------------- In accordance withOnly required to be performed in MODE 1. the Surveillance------------------------------------------------------------..- F re quency C ontrolVerify reactor coolant DOSE EQUIVALENT XE-133 Programspecific activity -270.0 pCi/gm.SR 3.4.16.2 ----------------------------- NOTE -----------------Only required to be performed in MODE 1.Verify reactor coolant DOSE EQUIVALENT 1-131 In accordance withspecific activity -1.0 pCi/gm. the SurveillanceFrequency ControlProgramANDBetween 2 and 6.hours after aTHERMALPOWER change of> 15% RTP withina 1 hour period.IDIABLO CANYON -UNITS 1 & 23.4-36 Unit 1 -Amendment No. 135,4 92,20o,Unit 2 -Amendment No. 4-35,93,20,1 Containment Isolation Valves3.6.33.6 CONTAINMENT SYSTEMS3.6.3 Containment Isolation ValvesLCO 3.6.3Each containment isolation valve shall be OPERABLE.APPLICABILITY:ACTIONSMODES 1, 2, 3, and 4.--NOTES1. Penetration flow path(s) except for 48-inch purge valve flow paths, may be unisolatedintermittently under administrative controls.2. Separate Condition entry is allowed for each penetration flow path.3. Enter applicable Conditions and Required Actions for systems made inoperable bycontainment isolation valves.4. Enter applicable Conditions and Required Actions of LCO 3.6.1, "Containment," whenisolation valve leakage results in exceeding the overall containment leakage rateacceptance criteria.CONDITION REQUIRED ACTION COMPLETION TIMEA. -------------- NOTE--------- A.1 Isolate the affected 4 hoursOnly applicable to penetration flow path bypenetration flow paths with use of at least onetwo containment isolation closed and de-activatedvalves. automatic valve, closed------------------------------------ m anual valve, blindOne or more penetration flange, or check valveflow paths with one with flow through thecontainment isolation valve valve secured.inoperable except for acontainment purge supplyand exhaust valve orpressure/vacuum reliefvalve leakage not within (continued)limit. (continued)DIABLO CANYON -UNITS I & 23.6-5Unit 1 -Amendment No. 4-35,Unit 2 -Amendment No. 4-35, Containment Isolation Valves3.6.3SURVEILLANCE REQUIREMENTSSURVEILLANCE FREQUENCYSR 3.6.3.1 Verify each 48 inch purge valve is sealed closed, In accordance with theexcept for one purge valve in a penetration flow Surveillance Frequencypath while in Condition D of this LCO. Control ProgramSR 3.6.3.2 Verify each 12 inch vacuum/pressure relief valve In accordance with theis closed, except when these valves are open for Surveillance Frequencypressure control, ALARA or air quality Control Programconsiderations for personnel entry, or forSurveillances that require the valves to be open.SR 3.6.3.3 --------------------NOTE --------------Valves and blind flanges in high radiation areasmay be verified by use of administrative controls.Verify each containment isolation manual valve In accordance with theand blind flange that is located outside Surveillance Frequencycontainment and not locked, sealed or otherwise Control Programsecured and required to be closed duringaccident conditions is closed, except forcontainment isolation valves that are open underadministrative controls.SR 3.6.3.4 --------------------NOTE --------------Valves and blind flanges in high radiation areasmay be verified by use of administrative means.Verify each containment isolation manual valve Prior to entering MODEand blind flange that is located inside 4 from MODE 5 if notcontainment and not locked, sealed or otherwise performed within thesecured and required to be closed during previous 92 daysaccident conditions is closed, except forcontainment isolation valves that are open underadministrative controls.SR 3.6.3.5 Verify the isolation time of each automatic power In accordance with theoperated containment isolation valve is within Inservice Testinglimits. ProgramSR 3.6.3.6 Not used(continued)IDIABLO CANYON -UNITS 1 & 23.6-9 Unit 1 -Amendment No. !35,200,Unit 2 -Amendment No. 135,20!,
+Containment Isolation Valves3.6.3SURVEILLANCE REQUIREMENTS (continued)SURVEILLANCE FREQUENCYSR 3.6.3.7 --------------------NOTE --------------This surveillance is not required when thepenetration flow path is isolated by a leak testedblank flange.Perform leakage rate testing for containment In accordance with thepurge supply and exhaust and vacuum/pressure Surveillance Frequencyrelief valves with resilient seals. Control ProgramSR 3.6.3.8 Verify each automatic containment isolation valve In accordance with thethat is not locked, sealed or otherwise secured in Surveillance Frequencyposition, actuates to the isolation position on an Control Programactual or simulated actuation signal.SR 3.6.3.9 Not usedSR 3.6.3.10 Verify each 12 inch containment In accordance with thevacuum/pressure relief valve is blocked to restrict Surveillance Frequencythe valve from opening > 50'. Control ProgramSR 3.6.3.11 Not used-tDIABLO CANYON -UNITS 1 & 23.6-10Unit 1 -Amendment No. 135,!75,200,Unit 2 -Amendment No. 135,!77,201, Programs and Manuals5.55.5 Programs and Manuals (continued)5.5.9 Steam Generator (SG) Tube Inspection ProgramA Steam Generator Program shall be established and implemented to ensure that SGtube integrity is maintained. In addition, the Steam Generator Program shall include thefollowing provisions:a. Provisions for condition monitoring assessments.Condition monitoring assessment means an evaluation of the "as found"condition of the tubing with respect to the performance criteria for structuralintegrity and accident induced leakage. The "as found" condition refers to thecondition of the tubing during an SG inspection outage, as determined from theinservice inspection results or by other means, prior to the plugging of tubes.Condition monitoring assessments shall be conducted during each outage duringwhich the SG tubes are inspected or plugged to confirm that the performancecriteria are being met.b. Performance criteria for SG tube integrity.SG tube integrity shall be maintained by meeting the performance criteria fortube structural integrity, accident induced leakage, and operational LEAKAGE.1. Structural integrity performance criterion: All in-service steam generator.tubes shall retain structural integrity over the full range of normal operatingconditions (including startup, operation in the power range, hot standby,and cool down and all anticipated transients included in the designspecification) and design basis accidents. This includes retaining a safetyfactor of 3.0 against burst under normal steady state full power operationprimary-to-secondary pressure differential and a safety factor of 1.4against burst applied to the design basis accident primary-to-secondarypressure differentials. Apart from the above requirements, additionalloading conditions associated with the design basis accidents, orcombination of accidents in accordance with the design and licensingbasis, shall also be evaluated to determine if the associated loadscontribute significantly to burst or collapse. In the assessment of tubeintegrity, those loads that do significantly affect burst or collapse shall bedetermined and assessed in combination with the loads due to pressurewith a safety factor of 1.2 on the combined primary loads and 1.0 on axialsecondary loads.2. Accident induced leakage performance criterion: The primary to secondaryaccident induced leakage rate for any design basis accident, other than aSG tube rupture, shall not exceed the leakage rate assumed in theaccident analysis in terms of total leakage rate for all SGs. Except during aSG tube rupture, leakage is not to exceed 0.75 gallon per minute total forall four SGs.(continued)DIABLO CANYON -UNITS 1 & 2 5.0-10 Unit 1 -Amendment No. --98,Unit 2 -Amendment No. 99, Programs and Manuals5.55.5 Programs and Manuals5.5.11 Ventilation Filter Testing Program (VFTP) (continued)c. Demonstrate for each of the ESF systems that a laboratory test of a sample ofthe charcoal absorber, when obtained as described in Regulatory Guide 1.52,Revision 2, shows the methyl iodide penetration less than the value specifiedbelow when tested in accordance with ASTM D3803-1989 at a temperature of300C and at the relative humidity specified below. Laboratory testing shall becompleted at least once per 24 months and after every 720 hours of charcoaloperation.ESF Ventilation System Penetration RHControl Room 2.5% 95%Auxiliary Building 5.0% 95%Fuel Handling Building 15.0% 95%d. Demonstrate for each of the ESF systems that the pressure drop across thecombined HEPA filters and the charcoal adsorbers is less than the valuespecified below when tested in accordance with ANSI N510-1980 at the systemflowrate specified below +/- 10% at least once per 24 months.ESF Ventilation System Delta P FlowrateControl Room 3.5 in. WG 2100 cfmAuxiliary Building 3.7 in. WG 73,500 cfmFuel Handling Building 4.1 in. WG 35,750 cfmThe provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring ProgramThis program provides controls for potentially explosive gas mixtures contained in theWaste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks,and the quantity of radioactivity contained in temporary unprotected outdoor liquidstorage tanks.The gaseous radioactivity quantities shall be determined following the methodology inRegulatory Guide 1.24 "Assumptions Used For Evaluating the Potential RadiologicalConsequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure."The liquid radwaste quantities shall be maintained such that 10 CFR Part 20 limits aremet.(continued)DIABLO CANYON -UNITS 1 & 2 5.0-13 Unit 1 -Amendment No. 135,112,163,198,Unit 2 -Amendment No. 135,142,1654199, Programs and Manuals5.55.5 Programs and Manuals (continued)5.5.19 Control Room Envelope Habitability ProgramA Control Room Envelope (CRE) Habitability Program shall be established andimplemented to ensure that CRE habitability is maintained such that, with anOPERABLE Control Room Ventilation System (CRVS), CRE occupants can control thereactor safely under normal conditions and maintain it in a safe condition following aradiological event, hazardous chemical release, or a smoke challenge. The programshall ensure that adequate radiation protection is provided to permit access andoccupancy of the CRE under design basis accident (DBA) conditions without personnelreceiving radiation exposures in excess of 5 rem TEDE for the duration of the accident.The program shall include the following elements:a. The definition of the CRE and the CRE boundary.b. Requirements for maintaining the CRE boundary in its design condition, includingconfiguration control and preventive maintenance.c. Requirements for (i) determining the unfiltered air inleakage past the CREboundary into the CRE in accordance with the testing methods and at theFrequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197,"Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors,"'Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequenciesspecified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.d. Measurement, at designated locations, of the CRE pressure relative to allexternal areas adjacent to the CRE boundary during the pressurization mode ofoperation by one train of the CRVS, operating at the flow rate required by theVFTP, at a Frequency of 24 months on a STAGGERED TEST BASIS. Theresults shall be trended and used as part of the 24 month assessment of theCRE boundary.e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shallbe stated in a manner to allow direct comparison to the unfiltered air inleakagemeasured by the testing described in paragraph c. The unfiltered air inleakagelimit for radiological challenges is the inleakage flow rate assumed in thelicensing basis analyses of DBA consequences. Unfiltered air inleakage limits forhazardous chemicals must ensure that exposure of CRE occupants to thesehazards will be within the assumptions in the licensing basis.f. The provisions of SR 3.0.2 are applicable to the Frequencies required byparagraphs c and d for determining CRE unfiltered inleakage and assessing CREhabitability, and measuring CRE pressure and assessing the CRE boundary.DIABLO CANYON -UNITS I & 2 5.0-17a Unit I -Amendment No. 20-4-,Unit 2 -Amendment No. 2-02, EnclosureAttachment 3PG&E Letter DCL-15-069ATTACHMENT 3Technical Specification Bases Markup(For Information Only)
+RCS Pressure SLB 2.1.2B 2.0 SAFETY LIMITS (SLs)B 2.1.2 Reactor Coolant System (RCS) Pressure SLBASESBACKGROUNDThe SL on RCS pressure protects the integrity of the RCS againstoverpressurization. In the event of fuel cladding failure, fissionproducts are released into the reactor coolant. The RCS then servesas the primary barrier in preventing the release of fission products intothe atmosphere. By establishing an upper limit on RCS pressure, thecontinued integrity of the RCS is ensured. According to 10 CFR 50,Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," andGDC 15, "Reactor Coolant System Design" (Ref. 1), the reactorpressure coolant boundary (RCPB) design conditions are not to beexceeded during normal operation and anticipated operationaloccurrences (AOOs). Also, in accordance with GDC 28, "ReactivityLimits" (Ref. 1), reactivity accidents, including rod ejection, do notresult in damage to the RCPB greater than limited local yielding.The design pressure of the RCS is 2500 psia. During normal operationand AOOs, RCS pressure is limited from exceeding the designpressure by more than 10%, in accordance with Section III of theASME Code (Ref. 2). To ensure system integrity, all RCS componentswere hydrostatically tested at 125% of design pressure, according tothe ASME Code requirements prior to initial operation when there is nofuel in the core. Following inception of unit operation, RCScomponents shall be pressure tested, in accordance with therequirements of ASME Code, Section XI (Ref. 3).Overpressurization of the RCS could result in a breach of the RCPB. Ifsuch a breach occurs in conjunction with a fuel cladding failure, fissionproducts could enter the containment atmosphere, raising concernsrelative to limits on radioactive releases specified in 10 CFR 4-050.67,"Reactor Site CriteriaAccident Source Term" (Ref. 4).(continued)DIABLO CANYON -UNITS 1 & 2Rev 8 Page 4 of 6 RCS Pressure SLB 2.1.2BASES (continued)APPLICABILITYSL 2.1.2 applies in MODES 1, 2, 3, 4, and 5 because this SL could beapproached or exceeded in these MODES due to overpressurizationevents. The SL is not applicable in MODE 6 because the reactorvessel head closure bolts are not fully tightened, or the reactor vesselis sufficiently vented, making it unlikely that the RCS can bepressurized.SAFETY LIMITVIOLATIONSIf the RCS pressure SL is violated when the reactor is in MODE 1 or 2,the requirement is to restore compliance and be in MODE 3 withinI hour.Exceeding the RCS pressure SL may cause immediate RCS failureand create a potential for radioactive releases in excess of10 CFR 4-050.67, "Reactor Site CriteriaAccident Source Term" limits(Ref. 4).The allowable Completion Time of 1 hour recognizes the importance ofreducing power level to a MODE of operation where the potential forchallenges to safety systems is minimized.If the RCS pressure SL is exceeded in MODE 3, 4, or 5, RCS pressuremustbe restored to within the SL value within 5 minutes. Exceedingthe RCS pressure SL in MODE 3, 4, or 5 is more severe thanexceeding this SL in MODE I or 2, since the reactor vesseltemperature may be lower and the vessel material, consequently, lessductile. As such, pressure must be reduced to less than the SL within5 minutes. The action does not require reducing MODES, since thiswould require reducing temperature, which would compound theproblem by adding thermal gradient stresses to the existing pressurestress.REFERENCES1. 10 CFR 50, Appendix A, GDC 14 (associated with 1967 GDC 9per FSAR Appendix 3.1A), GDC 15 (no direct correlation to 1967GDC; however, intent of 1971 GDC is per met per FSARAppendix 3.1A), and GDC 28 (associated with 1967 GDC 30 perFSAR Appendix 3.1A).2. ASME, Boiler and Pressure Vessel Code, Section III, Summer1969.3. ASME, Boiler and Pressure Vessel Code, Section Xl.4. 10 CFR 4-050.67.5. FSAR, Section 7.2.6. DCM S-7, 3.4.1.DIABLO CANYON -UNITS 1 & 2Rev 8 Page 6 of 6 SDMB 3.1.1BASESAPPLICABLESAFETYANALYSIS(continued)In the boron dilution analysis, the required SDM defines the reactivitydifference between an initial subcritical boron concentration and thecorresponding critical boron concentration. These values, inconjunction with the configuration of the RCS and the assumed dilutionflow rate, directly affect the results of the analysis. This event is mostlimiting at the beginning of core life, when critical boron concentrationsare highest.Depending on the system initial conditions and reactivity insertion ratethe uncontrolled rod withdrawal transient is terminated by either a highpower level trip or a high pressurizer pressure trip. In all cases, powerlevel, RCS pressure, linear heat rate, and the DNBR do not exceedallowable limits.The ejection of a control rod rapidly adds reactivity to the reactor coreicausing both the core power level and heat flux to increase withcorresponding increases in reactor coolant temperatures and pressure.The ejection of a rod also produces a time dependent redistribution ofcore power.The startup of an inactive RCP in MODES 1 or 2 is precluded. InMODE 3, the startup of an inactive RCP cannot result in a "cold water"criticality, even if the maximum difference in temperature existsbetween the SG and the core. The maximum positive reactivityaddition that can occur due to an inadvertent start is less than half theminimum required SDM. Startup of an idle RCP cannot, therefore,produce a return to power from the hot standby condition.SDM satisfies Criterion 2 of IOCFR50.36(c)(2)(ii). Even though it is notdirectly observed from the control room, SDM is considered an initialcondition process variable because it is periodically monitored toensure that the unit is operating within the bounds of accident analysisassumptions.LCOSDM is a core design condition that can be ensured during operationthrough control rod positioning (control and shutdown banks) andthrough the soluble boron concentration.The MSLB (Ref. 2) and the boron dilution (Ref. 3) accidents are themost limiting analyses that establish the SDM value of the LCO. ForMSLB accidents, if the LCO is violated, there is a potential to exceedthe DNBR limit and to exceed 10 CFR 4-0050.67, "ReaGtOF-SiteGrt-e4aAccident Source Term," limits (Ref. 4). For the boron dilutionaccident, if the LCO is violated, the minimum required time assumedfor operator action to terminate dilution may no longer be sufficient.The required SDM is specified in the COLR.(contnued(continued)DIABLO CANYON -UNITS 1 & 2Rev 8A Page 3 of 45 SDMB 3.1.1BASESSURVEILLANCE SR 3.1.1.1 (continued)REQUIREMENTS Using the ITC accounts for Doppler reactivity in this calculationbecause the reactor is subcritical, and the fuel temperature will bechanging at the same rate as the RCS.The Surveillance Frequency is based on operating experience,equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.REFERENCES 1 10 CFR 50, Appendix A, GDC 26.2 FSAR, Chapter 15, Section 15.4.2.1.3 FSAR, Chapter 15, Section 15.2.4.4 10 CFR 40050.67.5 FSAR, Chapter 15, Section 15.4.6.1.6.DIABLO CANYON -UNITS 1 & 2Rev8A Page5of45 RTS InstrumentationB 3.3.1B 3.3 INSTRUMENTATIONB 3.3.1 Reactor Trip System (RTS) InstrumentationBASESBACKGROUND The RTS initiates a unit shutdown, based on the values of selected unitparameters, to protect against violating the core fuel design limits andReactor Coolant System (RCS) pressure boundary during anticipatedoperational occurrences (AOOs) and to assist the Engineered SafetyFeatures (ESF) Systems in mitigating accidents.The protection and monitoring systems have been designed to assuresafe operation of the reactor. This is achieved by specifying limitingsafety system settings (LSSS) in terms of parameters directlymonitored by the RTS, as well as specifying LCOs on other reactorsystem parameters and equipment performance.The LSSS, as defined in 10 CFR 50.36, are defined in this specificationas the Allowable Values, and in conjunction with the LCOs, establishthe threshold for protective system action to prevent exceedingacceptable limits during Design Basis Accidents (DBAs).During AOOs, which are those events expected to occur more thanonce during the unit life, the acceptable limits are:1. The Departure from Nucleate Boiling Ratio (DNBR) shall bemaintained above the Safety Limit (SL) value to prevent departurefrom nucleate boiling (DNB);2. Fuel centerline melt shall not occur; and3. The RCS pressure SL of 2735 psig shall not be exceeded.Operation within the SLs of Specification 2.0, "Safety Limits (SLs),"also maintains the above values and assures that offsite dose will bewithin the 10 CFR 50 and 10 CFR 4-0050.67 criteria during AOOs.Accidents are events that are analyzed even though they are notexpected to occur during the unit life. The acceptable limit duringaccidents is that offsite dose shall be maintained within an acceptablefraction of 10 CFR 050.67 limits. Different accident categories areallowed a different fraction of these limits, based on probability ofoccurrence. Meeting the acceptable dose limit for an accident category.is considered having acceptable consequences, for that event.(continued)DIABLO CANYON -UNITS I & 2Rev 8 Page I of 167 ESFAS InstrumentationB 3.3.2BASESBACKGROUND During AOOs, which are those events expected to occur one or more(continued) times during the unit life, the acceptable limits are:1. The Departure from Nucleate Boiling Ratio (DNBR) shall bemaintained above the Safety Limit (SL) value to prevent departurefrom nucleate boiling (DNB).2. Fuel centerline melt shall not occur, and3. The RCS pressure SL of 2750 psia shall not be exceeded.Operation within the SLs of Specification 2.0, "Safety Limits (SLs),"also maintains the above values and assures that offsite dose will bewithin the 10 CFR 50 and 10 CFR 4-1050.67 criteria during AOOs.Accidents are events that are analyzed even though they are notexpected to occur during the unit life. The acceptable limit duringaccidents is that offsite dose shall be maintained within an acceptablefraction of 10 CFR 40050.67 limits. Different accident categories areallowed a different fraction of these limits, based on probability ofoccurrence. Meeting the acceptable consequences for that event isconsidered having acceptable consequences for that event. However,these values and their associated NTSPs are not considered to beLSSS as defined in 10 CFR 50.36.The ESFAS instrumentation is segmented into three distinct butinterconnected modules as identified below:Field transmitters or process sensors and instrumentation: providea measurable electronic signal based on the physicalcharacteristics of the parameter being measured;Signal processing equipment including digital protection system,field contacts, and protection channel sets: provide signalconditioning, bistable setpoint comparison, process algorithmactuation, compatible electrical signal output to protection systemdevices, and control board/control room/miscellaneous indications;andSolid State Protection System (SSPS) including input, logic, andoutput bays: initiates the proper unit shutdown or engineeredsafety feature (ESF) actuation in accordance with the defined logicand based on the bistable outputs from the signal process controland protection system. The residual heat removal pump trip orrefueling water storage tank level-low signal is not processed by theSSPS. The associated relays are located in the residual heatremoval pumps control system.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8 Page 69 of 167 Containment Ventilation Isolation InstrumentationB 3.3.6BASESAPPLICABLE They are also the primary means for automatically isolatingSAFETY containment in the event of a fuel handling accident or any otherANALYSES source within containment during shutdown. Containment isolation in(continued) turn ensures meeting the containment leakage rate assumptions of thesafety analyses, and ensures that the calculated accidental offsiteradiological doses are below 10 CFR 41-050.67 (Ref. 1) limits. Due toradioactive decay, containment is only required to isolate during fuelhandling accidents involving handling recently irradiated fuel (i.e., fuelthat has occupied part of a critical reactor core within the previous* 4-072 hours.)The containment ventilation isolation instrumentation satisfiesCriterion 3 of 10 CFR 50.36(c)(2)(ii).LCO The LCO requirements ensure that the instrumentation necessary toinitiate Containment Ventilation Isolation, listed in Table 3.3.6-1, isOPERABLE.1. Manual Initiation -Not used2. Automatic Actuation Lociic and Actuation RelaysThe LCO requires two trains of Automatic Actuation Logic andActuation Relays OPERABLE to ensure that no single random failurecan prevent automatic actuation.Automatic Actuation Logic and Actuation Relays consist of the samefeatures and operate in the same manner as described for ESFASFunction 1.b, SI, and ESFAS Function 3.a, Containment Phase AIsolation. The applicable MODES and specified conditions for theContainment Ventilation Isolation portion of these Functions aredifferent and less restrictive than those for their Phase A isolation andSI roles. If one or more of the SI or Phase A isolation Functionsbecomes inoperable in such a manner that only the ContainmentVentilation Isolation Function is affected, the Conditions applicable totheir SI and Phase A isolation Functions need not be entered. The lessrestrictive Actions specified for inoperability of the ContainmentVentilation Isolation Functions specify Sufficient compensatorymeasures for this case.3. Containment RadiationThe LCO specifies two required channels of radiation monitors toensure that the radiation monitoring instrumentation necessary toinitiate Containment ventilation Isolation remains OPERABLE inMODES 1-4.The LCO only requires one monitor to be OPERABLE duringmovement of recently irradiated fuel assemblies in containment. Inorder to provide the CVI function under these conditions withoutplacing the entire SSPS in service, an alternate circuit is provided topower the output relays and provide logic actuation signalsindependent of the SSPS.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8 Page 151 of 167 Containment Ventilation Isolation Instrumentation.B 3.3.6BASESSURVEILLANCEREQUIREMENTSSR 3.3.6.7 (continued)The test verifies that the channel responds to a measured parameterwithin the necessary range and accuracy. The Surveillance Frequencyis based on operating experience, equipment reliability, and plant riskand is controlled under the Surveillance Frequency Control Program.SR 3.3.6.8This SR assures that the individual channel RESPONSE TIMES for theCVI from Containment Purge Radiation Gaseous and Particulatefunction are less than'or equal to the maximum values assumed in theaccident analysis. Response Time testing acceptance criteria areincluded in ECG 38.2. Individual component response times are notmodeled in the analyses. The analyses model the overall or elapsedtime, from the point at which the parameter exceeds the Trip Setpointvalue at the sensor, to the point at which the equipment in both trainsreaches the required functional state (e.g., valves in full closedposition). The response time may be measured by a series ofoverlapping tests such that the entire response time is measured.The Surveillance Frequency is based on operating experience,equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.REFERENCES 1. 10 CFR44-.0A450.67.2. NUREG-1366, December 1992.3. DCM No. T-16, Containment Function.4. WCAP-1 5376-P-A, Revision 1, "Risk-Informed Assessment of theRTS and ESFAS Surveillance Test Intervals and Reactor TripBreaker Test and Completion Times," March 2003.5. License Amendment 184/186, January 3, 2006.IDIABLO CANYON -UNITS I & 2Rev 8 Page 156 of 167 CRVS Actuation instrumentationB 3.3.7BASESBACKGROUND The CRVS has two additional manually selected emergency operating(continued) modes; smoke removal and recirculation. Neither of these modes arerequired for the CRVS to be OPERABLE, but they are useful for certainnon-DBA circumstances.APPLICABLE The control room must be kept habitable for the operators stationedSAFETY there during accident recovery and post accident operations.ANALYSES The CRVS acts to terminate the supply of unfiltered outside air to thecontrol room, initiate filtration, and pressurize the control room. Theseactions are necessary to ensure the control room is kept habitable forthe operators stationed there during accident recovery and postaccident operations by minimizing the radiation exposure of controlroom personnel.In MODES 1, 2, 3, and 4, the radiation monitor (located at the controlroom intakes) actuation of the CRVS is a backup for the Phase Asignal actuation. This ensures initiation of the CRVS during a loss ofcoolant accident, er--Steam generator tube rupture, control rod ejectionaccident and Main Steam Line Break involving a releasc of radioActi-.ve.materials.The radiation monitor actuation of the CRVS in MODES 5 and 6, duringmovement of recently irradiated fuel assemblies (i.e., fuel that hasoccupied part of a critical reactor core within the previous 4-G072hours), is the primary means to ensure control room habitability in theevent of a fuel handling or waste gas decay tank rupture accident.This actuation is credited in the FHA. The CRVS pressurization systemactuation instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).In MODES 1. 2. 3, 4. 5 and 6, credit is taken for the dual ventilationintake design of the CR Dressurization air intakes. Based on theavailability of redundant PG&E Design Class I radiation monitors ateach pressurization intake location, the DCPP design has the capabilityof initial selection of the cleaner intake, but does not have the capabilityof automatic selection of the clean intake throughout the event. Basedon the CRVS pressurization intake design, and the expectation that theoperator will manually make the proper intake selection throughout theevent, and oer RG 1.194, June 2003, Regulatory Position C.3.3.2.3,when the CRVS is in Mode 4, the X/Q values for the more favorableCR intake is reduced by a factor of 4 and utilized to estimate the doseconsequences.LCO The LCO requirements ensure that instrumentation necessary toinitiate the CRVS pressurization system is OPERABLE.1. Manual InitiationThe LCO requires two trains OPERABLE. The operator caninitiate the CRVS pressurization mode at any time by using eitherof two switches in the control room. This action will causeactuation of all components in the same manner as any of theDIABLO CANYON -UNITS 1 & 2Rev 8 Page 158 of 167 CRVS Actuation InstrumentationB 3.3.7automatic actuation signals.The LCO for Manual Initiation ensures the proper amount ofredundancy is maintained in the manual actuation circuitry toensure the operator has manual initiation capability.2. Automatic Actuation RelaysThe LCO requires two trains of Actuation Relays OPERABLE toensure that no single random failure can prevent automaticactuation of the pressurization system. Since each unit has onetrain of Actuation Relays consisting of two sets of actuation logic,each unit must have at least one logic set for both trains to beconsidered OPERABLE.(continued)(Spillover from previous page.)DIABLO CANYON -UNITS I & 2Rev 8 Page 158 of 167 CRVS Actuation InstrumentationB 3.3.7BASESSURVEILLANCEREQUIREMENTS(continued)SR 3.3.7.5SR 3.3.7.5 is the performance of a SLAVE RELAY TEST. This testenergizes the Slave Relays and verifies actuation of the equipment tothe pressurization mode. Although there are no "Slave Relays" as inthe SSPS, this surveillance was maintained to preserve the format ofthe standard specification. The surveillance is intended to ensure thatthe actuation relays, downstream of the logic, function to actuate thepressurization mode equipment. Since the radiation monitors directlyactuate the actuation relays, this test is performed as part of theperformance of SR 3.3.7.2.The Surveillance Frequency is based on operating experience,equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.SR 3.3.7.6SR 3.3.7.6 is the performance of a TADOT. This test is a check of theManual Actuation Functions. Each Manual Actuation Function is testedup to, and including, the master relay coils. In some instances, the testincludes actuation of the end device (i.e., pump starts, valve cycles,etc.).The Surveillance Frequency is based on operating experience,equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program. The SR is modified by aNote that excludes verification of setpoints during the TADOT. TheFunctions tested have no setpoints associated with them.SR 3.3.7.7CHANNEL CALIBRATION is a complete check of the instrument loop,including the sensor. The test verifies that the channel responds to ameasured parameter within the necessary range and accuracy.The Surveillance Frequency is based on operating experience,equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.REFERENCES 1. WCAP-13878, "Reliability of Potter & Brumfield MDR Relays",June 1994.2. WCAP-1 3900, "Extension of Slave Relay Surveillance TestIntervals", April 1994.3. License Amendment 184/186, January 3, 2006.4. RG 1.194, "Atmospheric Relative Concentrations for Control RoomRadiological Habitability Assessments at Nuclear Power Plants,"June 2003.DIABLO CANYON -UNITS I & 2Rev 8 Page 163 of 167 FBVS Actuation InstrumentationB 3.3.8B 3.3 INSTRUMENTATIONB 3.3.8 Fuel Building Ventilation System (FBVS) Actuation InstrumentationBASESBACKGROUNDThe FBVS ensures that radioactive materials in the fuel buildingatmosphere following a fuel handling accident involving handlingrecently irradiated fuel (i.e., fuel that has occupied part of a criticalreactor core within the previous 40072 hours) are filtered and adsorbedprior to exhausting to the environment. The system is described in theBases for LCO 3.7.13, "Fuel Handling Building Ventilation System."The system initiates filtered ventilation of the fuel building automaticallyfollowing receipt of a high radiation signal from the Spent Fuel PoolMonitor or from the New Fuel Storage Vault Monitor. Initiation mayalso be performed manually as needed from the main control room orfuel handling building.High radiation, from either of the two monitors, provides FBVSinitiation. These actions function to prevent exfiltration of contaminatedair by initiating filtered ventilation, which imposes a negative pressureon the fuel building.APPLICABLE The FBVS ensures that radioactive materials in the fuel buildingSAFETY atmosphere following a fuel handling accident involving handlingANALYSES recently irradiated fuel are filtered and adsorbed prior to beingexhausted to the environment. This action reduces the radioactivecontent in the fuel building exhaust following a fuel handling accidentso that offsite doses Fremain within the limits spccified in 10 CFR 100(Ref~.1)The FBVS actuation instrumentation satisfies Criterion 3 of 10 CFR50.36(c)(2)(ii).LCO The LCO requirements ensure that instrumentation necessary toinitiate the FBVS is OPERABLE.1. Manual InitiationThe LCO requires two channels OPERABLE. The operator caninitiate the FBVS at any time by using either of two switches, onein the control room and another in the fuel handling building. Thisaction will cause actuation of all components in the same manneras any of the automatic actuation signals.The LCO for Manual Initiation ensures the proper amount ofredundancy is maintained in the manual actuation circuitry toensure the operator has manual initiation capability.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8 Page 164 of 167 FBVS Actuation InstrumentationB 3.3.8BASESSURVEILLANCE SR 3.3.8.1 (continued)REQUIREMENTSThe CHANNEL CHECK supplements less formal, but more frequent,checks of channels during normal operational use of the displaysassociated with the LCO required channels. The SurveillanceFrequency is based on operating experience, equipment reliability, andplant risk and is controlled under the Surveillance Frequency ControlProgram.SR 3.3.8.2A CFT is performed on each required channel to ensure the entirechannel will perform the intended function. This test verifies thecapability of the instrumentation to provide the FBACS actuation. TheSurveillance Frequency is based on operating experience, equipmentreliability, and plant risk and is controlled under the SurveillanceFrequency Control Program.SR 3.3.8.3 -Not usedSR 3.3.8.4SR 3.3.8.4 is the performance of a TADOT. This test is a check of themanual actuation functions. Each manual actuation function is testedup to, and including, the master relay coils. In some instances, the testincludes actuation of the end device (e.g., pump starts, valve cycles,etc.). The Surveillance Frequency is based on operating experience,equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program. The SR is modified by aNote that excludes verification of setpoints during the TADOT. TheFunctions tested have no setpoints associated with them.SR 3.3.8.5CHANNEL CALIBRATION is a complete check of the instrument loop,including the sensor. The test verifies that the channel responds to ameasured parameter within the necessary range and accuracy. TheSurveillance Frequency is based on operating experience, equipmentreliability, and plant risk and is controlled under the SurveillanceFrequency Control Program.REFERENCES 1. 10CFR 100.!!Not used.2. License Amendment 184/186, January 3, 2006.3. PG&E Letter DCL-05-124DIABLO CANYON -UNITS 1 & 2Rev 8 Page 167 of 167 RCS Operational LEAKAGEB 3.4.13BASES (continued)APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses do notSAFETY address operational LEAKAGE. However, other operational LEAKAGEANALYSES is related to the safety analyses for LOCA; the amount of leakage canaffect the probability of such an event. Safety analyses for designbasis events that model primary to secondary LEAKAGE result insteam discharge to the atmosphere. The safety analysis for the S1LBevent assumes that prirnar; to secoendary LEAKAGE is 10.5 gpm (rootemperature conditions) from the faulted SG r incr.eases to 10.5 gp,,as a resiult of accident indced conditions, and 0.1 gpm (roomtemperature co.ditions) from eac. h intact SG. The safety analyses forevents resulting in steam discharge to the atmosphere, ethe-4 haRSGTR aRd SLB, assume that primary to secondary LEAKAGE from allSGs is 0.75 gpm (hot ......ie.sStandard Temperature and Pressure)under accident conditions. For conservatism, the SLB assumes that thetotal 0.75 gpm tube leakage is assigned to the faulted steam generatorand the SGTR assumes that the total 0.75 gpm tube leakage isassigned to the 3 intact steam generators. The LCO requirement tolimit primary to secondary LEAKAGE through any one SG to less thanor equal to 150 gallons per day is significantly less than the conditionsassumed in the SLB safety analysis for the faulted SG.Primary to secondary LEAKAGE is a factor in the dose releasesoutside containment resulting from a steam line break (SLB) accident.To a lesser extent, other accidents or transients involve secondarysteam release to the atmosphere, such as a steam generator tuberupture (SGTR). The leakage contaminates the secondary fluid.The SGTR (Ref. 3) is more limiting for radiological releases at the siteboundary. The radiological dose analysis assumes loss of off-sitepower at the time of reactor trip with no subsequent condenser coolingavailable. The SGTR assumes that the total 0.75 gpm tube leakage isassigned to the 3 intact steam generators. The steam generator (SG)PORV for the SG that has sustained the tube rupture is assumed to failopen for 30 minutes, at which time the operator closes the block valveto the PORV. The dose consequences resulting from the SGTRaccident are within the limits defined in 10 CFR 40050.67 (Ref. 6).'.The SLB is more limiting for site radiation releases for events otherthan SGTR. The safety a.alysis for the SLB accident assumes10.5 gpM primary to secondary LEAKA~GE is thro)ugh the faulted SG.The dose consequences resulting from the SLB accident are wellwithin the limits defined in10 CFR 400-5067- approvedlicensing basis (i.e., small fraction of these limits).The safety analysis for RCS main loop piping for GDC-4 (Ref. 1)assumes 1 gpm unidentified leakage and monitoring per RG 1.45(Ref. 2) are maintained (Ref. 4 and 5).The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).(continued)DIABLO CANYON -UNITS 1 & 2Rev 8A Page 71 of 105 RCS Operational LEAKAGEB 3.4.13BASES (continued)REFERENCES1.2.3.4.5.6.7.8.10 CFR 50, Appendix A, GDC 4 and 30.Regulatory Guide 1.45, May 1973.FSAR, Section 15.FSAR, Section 3.NUREG-1061, Volume 3, November, 1984.10 CFR -1050.67.NEI 97-06, "Steam Generator Program Guidelines."EPRI, "Pressurized Water Reactor Primary-to-Secondary LeakGuidelines."DIABLO CANYON -UNITS 1 & 2Rev 8A_ Page 77 of 105 RCS Specific ActivityB 3.4.16B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.16 RCS Specific ActivityBASESBACKGROUNDThe maximum dose to the whole body and the thyroid that an individualat the exclusion area boundary can receive for 2 hours following anaccident or at the low population zone outer boundary for theradiological release duration, is, specified in 10 CFR 410._-_50.67 (Ref.1). Doses to the control room operators must be limited per GDC 19.The limits on specific activity ensure that the doses are appropriatelylimited during analyzed transients and accidents.The RCS specific activity LCO limits the allowable concentration levelof radionuclides in the reactor coolant. The LCO limits are establishedto minimize the dose consequences in the event of a steam line break(SLB) or steam generator tube rupture (SGTR) accident.The LCO contains specific activity limits for both DOSE EQUIVALENT1-131 and DOSE EQUIVALENT XE-133. The allowable levels areintended to ensure that offsite and control room doses meet theappropriate acceptance criteria in the Standard Review Plan RG 1.183(Ref. 2).IAPPLICABLESAFETYANALYSESThe LCO limits on the specific activity of the reactor coolant ensuresthat the resulting offsite and control room doses meet the appropriateSRP acceptance criteria following a SLB or a SGTR accident. Thesafety analyses (Refs. 3 and 4) assume the specific activity of thereactor coolant is at or, more conseprative than the LCO limits, and anexisting reactor coolant steam generator (SG) tube leakage rate of40.75 gpm-exists. The safety analyses assume the specific activity ofthe secondary coolant is at its limit of 0.1 pCi/gm DOSE EQUIVALENT1-131 from LCO 3.7.18, "Secondary Specific Activity."IThe analysis for the SLB and SGTR accidents establish theacceptance limits for RCS specific activity. Reference to theseanalyses is used to assess changes to the unit that could affect RCSspecific activity, as they relate to the acceptance limits.The analyses consider two cases of reactor coolant specific activity.One case assumes specific activity at 1.0 pCi/gm DOSE EQUIVALENT1-131 with a concurrent large iodine spike that increases the rate ofrelease of iodine from the fuel rods containing cladding defects to theprimary coolant immediately after a SLB (by a factor of 500) or SGTR(by a factor of 335), respectively.(continued)DIABLO CANYON -UNITS I & 2Rev 8A Page 93 of 105 RCS Specific ActivityB 3.4.16BASESAPPLICABLESAFETYANALYSES(continued)The second case assumes the initial reactor coolant iodine activity at60.0 pCi/gm DOSE EQUIVALENT 1-131 due to a pre-accident iodinespike caused by an RCS transient. In both cases, the noble gasspecific activity is assumed to be 651-270 pCi/gm DOSE EQUIVALENTXE-133.The SGTR analysis also assumes a loss of offsite power at the sametime as the reactor trip. The SGTR causes a reduction in reactorcoolant inventory. The reduction initiates a reactor trip from a lowpressurizer pressure signal or an RCS overtemperature AT signal.The loss of offsite power causes the steam dump valves to close toprotect the condenser. The rise in pressure in the ruptured SGdischarges radioactively contaminated steam to the atmospherethrough the SG power operated relief valves and the main steam safetyvalves. The unaffected SGs remove core decay heat by venting steamto the atmosphere until the cooldown ends and the RHR system isplaced in service.Operation with iodine specific activity levels greater than the LCO limitis permissible, if the activity levels do not exceed 60.0 pCi/gm DOSEEQUIVALENT 1-131, for more than 48 hours.The limits on RCS specific activity are also used for establishingstandardization in radiation shielding and plant personnel radiationprotection practices.RCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).LCO 'The iodine specific activity in the reactor coolant is limited to1.0 pCi/gm DOSE EQUIVALENT 1-131, and the noble gas specificactivity in the reactor coolant is limited to 609.0270.0 pCi/gm DOSEEQUIVALENT XE-133, as contained in SR 3.4.16.2 and SR 3.4.16.1respectively. The limits on specific activity ensure that offsite andcontrol room doses will meet the appropriate SRR acceptance criteria(Refs. 1 and 2).(continued)IIDIABLO CANYON -UNITS I & 2Rev 8A Page 94 of 105 RCS Specific ActivityB 3.4.1,6BASESSURVEILLANCE The definition of DOSE EQUIVALENT XE-133 in Specification 1.1,REQUIREMENTS "Definitions," requires that the determination of DOSE EQUIVALENT(continued) XE-133 shall be performed using the effective dose conversion factorsfor air submersion listed in Table 111.1 of EPA Federal Guidance ReportNo. 12, 1993, "External Exposure to Radionuclides in Air, Water, andSoil." These dose conversion factors are consistent with the doseconversion factors used in the applicable dose consequence analyses.The Note modifies this SR to allow entry into and operation in MODE 4,MODE 3, and MODE 2 prior to performing the SR. This allows theSurveillance to be performed in those MODES, prior to enteringMODE 1.SR 3.4.16.2This Surveillance is performed to ensure iodine specific activity remainswithin the LCO limit during normal operation and following fast powerchanges when iodine spiking is more apt to occur. The SurveillanceFrequency is based on operating experience, equipment reliability, andplant risk and is controlled under the Surveillance Frequency Control "Program. The Frequency, between 2 and 6 hours after a power change-> 15% RTP within a 1 hour period, is established because the iodinelevels peak during this time following iodine spike initiation; samples atother times would provide inaccurate results.The definition of DOSE EQUIVALENT 1-131 in Specification 1.1,"Definitions," specifies the thyroid dose conversion factors which may beused to determine DOSE EQUIVALENT 1-131. The thyraid doseconversion factors used to determine DOSE EQUIVALENT 1-131 arethe committed thyroid dose conversion factors from Table 2.1 of EPAFederal Guidance Report No. 11, 1988, "Limiting Values ofRadionuclide Intake and Air Concentration and Dose ConversionFactors for Inhalation, Submersion and Ingestion." and arete-beconsistent with the dose conversion factors used in the applicable doseconsequence analyses, Or be .. .cnSative with reSPect to the doseconcrsonfactors used in the applicablc dose coensequence analyses..uch that a higher DOSE EQU/I'ALENT 1 131 is deteFrmined.The Note modifies this SR to allow entry into and operation in MODE 4,MODE 3, and MODE 2 prior to performing the SR. This allows theSurveillance to be performed in those MODES, prior to enteringMODE 1.(continued)DIABLO CANYON -UNITS 1 & 2Rev8A Page 97 of 105 RCS Specific ActivityB 3.4.16BASES (continued)REFERENCES 1. 10 CFR 100.41, 197350.67.2. Standard Review Plan (SRP), Section 6.4 (SLe and SGT-R controroomn dose limits), Section 4 5. 1.5Appendix A (SL=B offsite doselimits) and Section 5.6.3 (SGT-R offsite dose imits).RegulatoryGuide 1.183, July 2000.3. FSAR, Sections 15.4.3 and 15.5.20.4. FSAR Section 15.15.5.18.DIABLO CANYON -UNITS I & 2Rev 8A Page 98 of 105 Steam Generator (SG) Tube IntegrityB 3.4.17BASES (continued)APPLICABLESAFETYANALYSESThe steam generator tube rupture (SGTR) accident is the limitingdesign basis event for SG tubes and avoiding an SGTR is the basis forthis Specification. The analysis of a SGTR event assumes a teta-primary to secondary LEAKAGE rate of 40.75 gpm from the intact SGsplus the leakage rate associated with a double-ended rupture of asingle tube. The SGTR radiological dose analysis assumes loss of off-site power at the time of reactor trip with no subsequent condensercooling available. The SG PORV for the SG that has sustained thetube rupture is assumed to fail open for 30 minutes, at which time theoperator closes the block valve to the PORV. The SGTR radiologicaldose analysis assumes the contaminated secondary fluid is releasedbriefly to the atmosphere from all the PORVs following reactor trip, isreleased from the ruptured SG PORV for 30 minutes, is released fromthe intact SG PORVs during the cooldown, and is released from allPORVs following cooldown until termination of the event.The analysis for design basis accidents and transients other than aSGTR assume the SG tubes retain their structural integrity (i.e., theyare assumed not to rupture.) For the SLB event, the prim.ar; toSecondary L-E.AK1A.GE is 10.5 gpm from the faulted SGn o-r is assumedto increase to 10. 5 gp. .as a resuit of accident indu.ed conditions, and0.1 gpm from each intact SG. For other events, theThe steamdischarge to the atmosphere is based on the total primary to secondaryLEAKAGE from all SGs of 0.75 gpm under accident conditions. Foraccidents that do not involve fuel damage, the primary coolant activitylevel of DOSE EQUIVALENT 1-131 is assumed to be equal to the LCO3.4.16, "RCS Specific Activity," limits. For accidents that assume fueldamage, the primary coolant activity is a function of the amount ofactivity released from the damaged fuel. The dose consequences ofthese events are within the limits of GDC 19 (Ref. 2), 10 CFR 41-050.67(Ref. 3) or the NRC approved licensing basis (e.g., a small fraction atthese-limit and RG 1.183 (Ref. 7).Steam generator tube integrity satisfies Criterion 2 of10 CFR 50.36(c)(2)(ii).LCO The LCO requires that SG tube integrity be maintained. The LCO alsorequires that all SG tubes that satisfy the repair criteria be plugged inaccordance with the Steam Generator Program.During an SG inspection, any inspected tube that satisfies the SteamGenerator Program repair criteria is removed from service by plugging.If a tube was determined to satisfy the repair criteria but was notplugged, the tube may still have tube integrity.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8A Page 100 of 105 Steam Generator (SG) Tube IntegrityB 3.4.17BASESLCO Structural integrity requires that the primary membrane stress intensity(continued) in a tube not exceed the yield strength for all ASME Code, Section Ill,Service Level A (normal operating conditions) and Service Level B(upset or abnormal conditions) transients included in the designspecification. This includes safety factors and applicable design basisloads based on ASME Code, Section III, Subsection NB (Ref. 4) andDraft Regulatory Guide 1.121 (Ref. 5).The accident induced leakage performance criterion ensures (a) thatthe primary to secondary LEAKAGE caused by a design basisaccident, other than a SGTR, is within the accident analysisassumptions, and (b) that the primary to secondary LEAKAGE will notexceed 40.75 gpm total for all four peW SGs (except for specific types ofdegradation at specific locations where the NRC has approved greateraccident induced leakage) to ensure that the potential for inducedleakage during severe accidents will be maintained at a level that willnot increase risk. The accident analysis for the SLB event, the SGTRevent and other events resulting in steam release to the atmosphereassumes that accident induced leakage does not exceed 10 gpm in thefaulted SG and 0.1 gpmn in each intact SG. For the faulted SG in theSLB event, 10.5 gpm is the accid.nt induced leakage limit, of Which nomr..e thaRn 1 gpM can come from sour.es not specifically exempted bythe NRC fromn this I gpmn limit. The accident analyses for events otherthan SGTR aRd SLBI assume that leakage does not exceed 0.75 gpmtotal under accident conditions. The accident induced leakage rateincludes any primary to secondary LEAKAGE existing prior to theaccident in addition to primary to secondary LEAKAGE induced duringthe accident.The operational LEAKAGE performance criterion provides anobservable indication of SG tube conditions during plant operation.The limit on operational LEAKAGE is contained in LCO 3.4.13, "RCSOperational LEAKAGE," and limits primary to secondary LEAKAGEthrough any one SG to 150 gallons per day. This limit is based on theassumption that a single crack leaking this amount would notpropagate to a SGTR under the stress conditions of a LOCA or a mainsteam line break. If this amount of LEAKAGE is due to more than onecrack, the cracks are very small, and the above assumption isconservative.APPLICABILITY Steam generator tube integrity is challenged when the pressuredifferential across the tubes is large. Large differential pressuresacross SG tubes can only be experienced in MODE 1, 2, 3, or 4.RCS conditions are far less challenging in MODES 5 and 6 than duringMODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondarydifferential pressure is low, resulting in lower stresses and reducedpotential for LEAKAGE.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8A Page 102 of 105 Steam-Generator (SG) Tube IntegrityB 3.4.17BASESSURVEILLANCE In addition, Specification 5'.5.9.contains prescriptive requirementsREQUIREMENTS concerning inspection intervals to provide added assurance that the SG(continued) performance criteria will be met between scheduled inspections.SR 3.4.17.2During an SG inspection, any inspected tube that satisfies the SteamGenerator Program repair criteria is removed from service by plugging.The tube repair criteria delineated in Specification 5.5.9 are intended toensure that tubes accepted for continued service satisfy the SGperformance criteria with allowance for error in the flaw sizemeasurement and for future flaw growth. In addition, the tube repaircriteria, in conjunction with other elements of the Steam GeneratorProgram, ensure that the SG performance criteria will continue to bemet until the next inspection of the subject tube(s). Reference Iprovides guidance for performing operational assessments to verify thatthe tubes remaining in service will continue to meet the SG performancecriteria.The Frequency of prior to entering MODE 4 following a SG inspection.ensures that the Surveillance has been completed and all tubes meetingthe repair criteria are plugged prior to subjecting the SG tubes tosignificant primary to secondary pressure differential.REFERENCES 1. NEI 97-06, "Steam Generator Program Guidelines."2. 10 CFR 50 Appendix A, GDC 19 1_999.3. 10 CFR -I050.67.4. ASME Boiler and Pressure Vessel Code, Section III, SubsectionNB.5. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded SteamGenerator Tubes," August 1976.6. EPRI, "Pressurized Water Reactor Steam Generator ExaminationGuidelines."7. Regulatory Guide 1.183, July 2000.DIABLO CANYON -UNITS 1 & 2Rev8A Page105 of 105 Containment Isolation ValvesB 3.6.3BASESBACKGROUND(continued) ,Containment Puroe System (48 inch ourae valves)The Containment Purge System operates to supply outside air into thecontainment for ventilation and cooling or heating needed forprolonged containment access following.a shutdown and duringrefueling. The system may also be u.Sed to r-,ed-ucl e the concentration o,noble gases wfithinq containmnent priorto and duFrig personn~el accr-ess.The supply and exhaust lines each contain two isolation valves. The418 inch Containment Purge valves are qualified for: automnatic closurFefroM their open position under DBA c.nditions. The safety analysesassume that the 48-inch supply and exhaust line valves are closed atthe start of the DBA. Therefore, the 48 inch Containment Purge supplyand exhaust isolation valves are ...m.lly maintained sealed closed inMODES 1, 2, 3, and 4 to ensure the containment boundary ismaintained. The Purge Supply and Exhaust .selationvalves are supplied with an internalrblock which prevents opening thevalve beyond 80 degrees. This block was provided by the mnanufadtureto allow lim.iting the valve's opening. GCalcuations peormed duringqualification to B-ranc~h Technical Position CSB3 6 41 showed the block tobe unnecessary to assure clsu-re time within 2 seconds under DBAconditions (SSER 9, Juno 1980 and M 661). Adjustmntsof this block to values greater than or less than 80 degrees will nota.... Mnc valve s ability W. Gies., Tis des:gn ass .r.s that conEtanmentboundary is m.aintained. These valves may be opened as necessarya. Reduce noble gases within containment prior to and duringpersonnel access, andb. Mitigate the effects of controller leakage and other sources whichmay effect the habitabiity of the containment for personnelOperation in Modes 1, 2, 3, o .with the 48 inch purge valves or the12 incnh vacuu/pressure relief valves open providing a flow path islimnited t o Fnomoe than 200 hours per calendar year-.Containment Pressure/Vacuum Relief (12 inch isolation valves)rhThe Containment Pressure/Vacuum Relief valves are operated asnecessary to:a. Reduce the concentration of noble gases within containment priorto and during personnel access, andb. Equalize containment internal and external pressures.Since the 12 inch Containment Pressure/Vacuum Relief valves aredesigned to meet the requirements for automatic containment isolationwithin 5 seconds if mechanical blocks are installed to prevent openingmore then 500, these valves may be opened as needed in MODES 1,2, 3, and 4.(continued)DIABLO CANYON -UNITS I & 2Rev 8C Page 13 of 50 Containment Isolation ValvesB 3.6.3BASES (continued)APPLICABLE The containment isolation valve LCO was derived from theSAFETY assumptions related to minimizing the loss of reactor coolant inventoryANALYSES and establishing the containment boundary during major accidents. Aspart of the containment boundary, containment isolation valveOPERABILITY supports leak tightness of the containment. Therefore,the safety analyses of any event requiring isolation of containment isapplicable to this LCO.The DBA that results in a release of radioactive material withincontainment in MODES 1, 2, 3, or 4 is a loss of coolant accident(LOCA) (Ref. 1). In the analyses for this accident, it is assumed thatcontainment isolation valves are either closed or function to closewithin the required isolation time following event initiation. Thisensures that potential paths to the environment through containmentisolation valves (including the Containment Purge, and ContainmentVacuum/Pressure Relief valves) are minimized. The safety analysesassume that the 48 inch purge valves are closed at event initiation. Ifthe 48 inch ContaiRme t Purge supply and exhaust valVes close within2 seccnds and the 12 inch pressure/vacuum relief valves close within5 seconds after the DBA initiation, the safety analysis shows thatoffsite dose release will be less than 10 CFR-41O-50.67 guidelines.The DBA analysis assumes that containment isolation occurs andleakage is prevented except for the design leakage rate, La.The LOCA offsite dose analysis assumes leakage from thecontainment at a maximum leak rate of 0.10 percent of thecontainment volume per day for the first 24 hours, and at 0.05 percentof the containment volume per day for the duration of the accident.The single failure criterion required to be imposed in the conduct ofplant safety analyses was considered in the original design of the41 R inc Centainment Purge supply and exhaust and the 12 inchContainment Pressure/Vacuum Relief valves. Two valves in seriesprovide assurance that the flow paths can be isolated even if a singlefailure occurred. The inboard and outboard isolation valves areprovided with diverse power sources and are pneumatically operatedspring closed valves that will fail closed on the loss of power or air.The 48 inGh Purge supply and exhaust and 12 inchContainment PressureNacuum Relief valves are able to close in theenvironment following a LOCA. Therefore, each of the GGstaimentPurge supply and exhaust and Containment Vacuum/pressure Reliefvalves may be opened to provide a flow path. The 48il4ehContainment Purge supply and exhaust valves 12-inchvacuum/pressure relief valves may be open no more then 200 hoursper calendar year-while in MODES 1, 2, 3, and 4. Additionally, enly tweof the three flow paths (containment purge supply and exhaust, andcontainmnent vacuum/Wpressure relieD may be open at one time.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8C Page 14 of 50 Containment Isolation ValvesB 3.6.3BASESAPPLICABLESAFETYANALYSES(continued)The system is designed to preclude a single failure from compromisingthe containment boundary as long as the system is operated inaccordance with the subject LCO.The 48 inch Containment Purge supply and exhaust valves may beunable to close in the environment following a LOCA in sufficient timeto support DBA acceptance criteria. Therefore, each of the purgevalves is required to remain sealed closed during MODES 1. 2. 3. and4. In this case, the single failure criterion remains applicable to thecontainment purge valves due to failure in the control circuit associatedwith each valve. Again, the purge system valve design precludes asingle failure from compromising the containment boundary as long asthe system is operated in accordance with the subject LCO.The containment isolation valves satisfy Criterion 3 of1 OCFR50.36(c)(2)(ii).LCOContainment isolation valves form a part of the containment boundary.The containment isolation valves' safety function is related tominimizing the loss of reactor coolant inventory and establishing thecontainment boundary during a DBA. The automatic power operatedisolation valves are required to have isolation times within limits and toactuate on an automatic isolation signal. The 48 inch ContainmentPurge supply and exhaust valves aid 4he must be sealed closed duringMODES 1. 2, 3, and 4. The Pressure/Vacuum Relief valves must haveblocks installed to prevent full opening. These blocked valves alsoactuate on an automatic isolation signal. The valves covered by thisLCO are listed along with their associated stroke times in PlantProcedure AD13.DC1 (Ref. 5).Normally closed passive containment isolation valves/devices areconsidered OPERABLE when manual valves are closed, automaticvalves are de-activated and secured in their closed position, blindflanges are in place, and closed systems are intact. These passiveisolation valves/devices are those listed in Reference 5.Containment Purge supply and exhaust valves, and ContainmentPressure/Vacuum Relief valves with resilient seals must meetadditional leakage rate surveillance frequency requirements. The othercontainment isolation valve leakage rates are addressed by LCO 3.6.1,"Containment."This LCO provides assurance that the containment isolation valves andthe Containment Purge supply and exhaust, and ContainmentPressure/Vacuum Relief valves will perform their designed safetyfunction to minimize the loss of reactor coolant inventory and establishthe containment boundary during accidents.IAPPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactivematerial to containment. In MODES 5 and 6, the probability andconsequences of these events are reduced due to the pressure andtemperature limitations of these MODES. Therefore, the containmentDIABLO CANYON -UNITS 1 & 2Rev 8C Page 15 of 50 Containment Isolation ValvesB 3.6.3isolation valves are not required to be OPERABLE in MODE 5. Therequirements for containment isolation valves during MODE 6 areaddressed in LCO 3.9.4, "Containment Penetrations."(continued)(Spillover from previous page)DIABLO CANYON -UNITS 1 & 2Rev8C Page 15of 50 Containment Isolation ValvesB 3.6.3BASES (continued)ACTIONS The ACTIONS are modified by a Note allowing penetration flow paths,except 48-inch purge valve flow paths. to be unisolated intermittentlyunder administrative controls. These administrative controls consist ofstationing a person at the valve controls, who is in continuouscommunication with the control room. In this way, the penetration canbe rapidly isolated when a need for containment isolation is indicated.Due to the size of the containment purge line penetration and the factthat those penetrations exhaust directly from the containmentatmosphere to the environment, the penetration flow path containinqthese valves may not be opened under administrative controls. Asingle purge valve in a penetration flow path may be opened to effectrepairs to an inoperable valve, as allowed by SR 3.6.3.1.This Note alsolimits operation of the nrE)mally isolated Containment Supply andExhaust valves (2 penetration flow paths) and the VacUUM/PccsSUreRelief valves (I penetration flow path) to no more thanP 2 of 13penetrationR flow paths open at one time.A second Note has been added to provide clarification that, for thisLCO, separate Condition entry is allowed for each penetration flowpath. This is acceptable, since the Required Actions for eachCondition provide appropriate compensatory actions for eachinoperable containment isolation valve. Complying with the RequiredActions may allow for continued operation, and subsequent inoperablecontainment isolation valves are governed by subsequent Conditionentry and application of associated Required Actions.The ACTIONS are further modified by a third Note, which ensuresappropriate remedial actions are taken, if necessary, if the affectedsystems are rendered inoperable by an inoperable containmentisolation valve.In the event the containment isolation valve leakage results inexceeding the overall containment leakage rate acceptance criteria,Note 4 directs entry into the applicable Conditions and RequiredActions of LCO 3.6.1.Plant Procedure AD13.DC1 Attachment 7.7 (Ref. 5) provides theapplicable CONDITION to enter for each containment isolation valve ifthe valve is inoperable.A.1 and A.2In the event one containment isolation valve in one or more penetrationflow paths requiring isolation following a DBA is inoperable except forContainment Purge supply and exhaust, and ContainmentPressureNacuum Relief isolation valve leakage not within limit, theaffected penetration flow path must be isolated. The method ofisolation must include the use of at least one isolation barrier that(continued)DIABLO CANYON -UNITS I & 2Rev8C Page16of50 Containment Isolation ValvesB 3.6.3BASESACTIONSD.1, D.2, and D.3 (continued)condition, it is prudent to perform the SR more often. Therefore, aFrequency of once per 92 days was chosen and has been shown to beacceptable based on operating experience.Required Action D.2 is modified by two Notes. Note 1 applies to valvesand blind flanges located in high radiation areas and allows thesedevices to be verified closed by use of administrative means. Allowingverification by administrative means is considered acceptable, sinceaccess to these areas is typically restricted. Note 2 applies to isolationdevices that are locked, sealed, or otherwise secured in position andallows these devices to be verified closed by use of administrativemeans. Allowing verification by administrative means is consideredacceptable, since the function of locking, sealing, or securingcomponents is to ensure that these devices are not inadvertentlyrepositioned. Therefore, the probability of misalignment of thesevalves, once they have been verified to be in the proper position, issmall.E.1 and E.2If the Required Actions and associated Completion Times are not met,the plant must be brought to a MODE in which the LCO does not apply.To achieve this status, the plant must be brought to at least MODE 3within 6 hours and to MODE 5 within 36 hours. The allowedCompletion Times are reasonable, based on operating experience, toreach the required plant conditions from full power conditions in anorderly manner and without challenging plant systems.SURVEILLANCEREQUIREMENTSSR 3.6.3.1Ne4-Used Each 48 inch Containment Purge supply and exhaust valve isrequired to be verified sealed closed. This Surveillance is designed toensure that a gross breach of containment is not caused by aninadvertent or spurious opening of a Containment Purge valve. Thesevalves are assumed to be closed at the start of a DBA. Therefore.these valves are required to be in the sealed closed position duringMODES 1, 2, 3, and 4. A Containment Purge valve that is sealedclosed must have motive power to the valve operator removed. Thiscan be accomplished by de-energizing the source of electric power orby removing the air supply to the valve operator. In the event the purgevalve leakage requires entry into Condition D, the surveillance permitsopening one purge valve in a penetration flow path to perform repairs.The Surveillance Frequency is controlled under the SurveillanceFrequency Control Program.SR 3.6.3.2This SR ensures that the 48 inch Containment Purge supply dexhaust-and the 12 inch Containment PressureNacuum Relief valvesare closed as required or, if open, open for an allowable reason. If aDIABLO CANYON -UNITS 1 & 2Rev 8C Page 21 of 50 Containment Isolation ValvesB 3.6.3purge or pressure relief valve is open in violation of this SR, the valveis considered inoperable. If the inoperable valve is not otherwiseknown to have excessive leakage when closed, it is not considered tohave leakage outside of limits. The SR is not required to be met whenthe CoRtainment Purge supply and exhaust or Containment PressureRelief valves are open for the reasons stated. The valves may beopened for pressure control, ALARA or air quality considerations forpersonnel entry, or for Surveillances that require the valves to be open.The Containment Purge supply and exhaust or ContainmentPressureNacuum Relief valves are capable of closing in the(continued)(Spillover from previous page.)DIABLO CANYON -UNITS I & 2Rev 8C Page 21 of 50 Containment Spray and Cooling SystemsB 3.6.6BASESBACKGROUND Containment Spray System (continued)In the recirculation mode of operation, containment spray is supplied bymanual realignment of the residual heat removal (RHR) pumps afterthe RWST is empty.The Containment Spray System provides a spray of cold borated watermixed with sodium hydroxide (NaOH) from the spray additive tank intothe upper regions of containment to reduce the containment pressureand temperature, and to reduce fission products from the containmentatmosphere during a DBA. The RWST solution temperature is animportant factor in determining the heat removal capability of theContainment Spray System during the injection phase. In therecirculation mode of operation, heat is removed from the containmentsump water by the RHR heat exchangers. Each train of theContainment Spray System provides adequate spray coverage to meetthe system design requirements for containment atmospheric heatremoval.The Spray Additive System injects an NaOH solution into the spray.The resulting alkaline pH of the spray enhances the ability of the sprayto scavenge fission products from the containment atmosphere. TheNaOH added in the spray also ensures an alkaline pH for the solutionrecirculated in the containment sump. The alkaline pH of thecontainment sump water maximizes the retention of iodine andminimizes the occurrence of chloride and caustic stress corrosion onmechanical systems and components exposed to the fluid.The Containment Spray System is actuated either automatically by acontainment High-High pressure signal or manually. If an "S" signal ispresent, the High-High pressure signal automatically starts the twocontainment spray pumps, opens the containment spray pumpdischarge valves, opens the spray additive tank outlet valves, initiates aphase "B" isolation signal, and begins the injection phase. A manualactuation of the Containment Spray System will begin the samesequence and can be initiated by operator action from the main controlboard. The injection phase of containment spray continues until anRWST Low-Low level alarm is received. The Low-Low level alarm forthe RWST signals the operator to manually secure the system. Afterre-alignment of the RHR system to the containment recirculation sump,the associated RHR spray header isolation valve may-beis opened toallow continued spray operation of one train of spray utilizing the RHRpump to supply flow. The LOCA dose analysis takes credit for thismanual initiation of Containment Spray during recirculation to takeplace within 12 minutes following the termination of Containment Sprayduring the iniection phase.Containment Spray is fiat required to be actuated during therecirculation phase of a LOCA, but may be at the discretion oGthe Technical Support Center. Containment Spray operation (injectionplus recirculation) is credited until 6.25 hours following initiation of aDIABLO CANYON -UNITS 1 & 2Rev8C Page 34 of 50 Containment Spray and Cooling SystemsB 3.6.6LOCA. During the recirculation phase of a LOCA, the ContainmentSpray System must be capable of(continued)(Spillover from previous page.)DIABLO CANYON -UNITS I & 2Rev 8C Page 34 of 50 Containment Spray and Cooling SystemsB 3.6.6BASESAPPLICABLESAFETYANALYSES(continued)Analyses and evaluation show that containment spray is not requiredduring the recirculation phase of a LOCA for containment pressure andtemperature control (Ref. 7). However, for dose consequences.containment spray is required during the recirculation phase of a LOCAfor removing radioactive iodine and particulates from the containmentatmosphere.If only one RHR pump is available during the recirculation phase of aLOCA, it may not be possible to obtain significant containment spraywithout closing valves 8809A or B. If recirculation spray is used withonly one train of RHR in operation, ECCS flow to the reactor will bereduced, but analysis has shown that the flow to the reactor in thissituation is still in excess of that needed to supply the required corecooling.The effect of an inadvertent containment spray actuation has beenanalyzed. An inadvertent spray actuation results in a -1.8.0 psidcontainment pressure decrease and is based on a sudden coolingeffect of 70'F in the interior of the leak tight containment. Additionaldiscussion is provided in the Bases for LCO 3.6.4.The modeled Containment Spray System actuation from thecontainment analysis is based on a response time associated withexceeding the containment High-High pressure setpoint to achievingfull flow through the containment spray nozzles. The ContainmentSpray System total response time includes diesel generator (DG)startup (for loss of offsite power),(continued)DIABLO CANYON -UNITS I & 2Rev 8C Page 37 of 50 Containment Spray and Cooling SystemsB 3.6.6BASESAPPLICABLE sequenced loading of equipment, containment spray pump startup, andSAFETY spray line filling (Ref. 4). The CFCUs performance for post accidentANALYSES conditions is given in Reference 4. The result of the analysis is that(continued) each train (two CFCUs) combined with one train of containment spraycan provide 100% of the required peak cooling capacity during the postaccident condition.The modeled Containment Cooling System actuation from thecontainment analysis is based upon a response time associated withexceeding the containment High-High pressure setpoint to achievingfull Containment Cooling System air and safety grade cooling waterflow. The Containment Cooling System total response time includessignal delay, DG startup (for loss of offsite power), and componentcooling water pump startup times.The Containment Spray System and the Containment Cooling Systemsatisfies Criterion 3 of 10CFR50.36(c)(2)(ii).LCO During a DBA LOCA, a minimum of two CFCUs and one containmentspray train are required to maintain the containment peak pressure andtemperature below the design limits (Refs. 4). Additionally, onecontainment spray train is also required to remove radioactive iodineand particulates from the containment atmosphere and maintainconcentrations below those assumed in the safety analysis. To ensurethat these requirements are met, two containment spray trains and theCFCU system consisting of four CFCUs or thre6 CFCUs each suppliedby a different vital bus must be OPERABLE. Therefore, in the event ofan accident, at least one train of containment spray and two CFCUsoperate, assuming the worst case single active failure occurs. EachContainment Spray train typically includes a spray pump, sprayheaders, nozzles, valves, piping, instruments, and controls to ensurean OPERABLE flow path capable of taking suction from the RWSTupon an ESF actuation signal. Upon actuation of the RWST Low-Lowalarm, the containment spray pumps are secured. Containment sprayGeuldis then be-supplied as Fequired by an RHR pump taking suctionfrom the containment sump for a total spray operation (injection andrecirculation) of 6.25 hours.Each CFCU includes cooling coils, dampers, fans, instruments, andcontrols to ensure an OPERABLE flow path.APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactivematerial to containment and an increase in containment pressure andtemperature requiring the operation of the containment spray trainsand CFCUs.In MODES 5 and 6, the probability and consequences of these eventsare reduced due to the pressure and temperature limitations of theseMODES. Thus, the Containment Spray System and the ContainmentCooling System are not required to be OPERABLE in MODES 5 and 6.DIABLO CANYON -UNITS I & 2Rev 8C Page 38 of 50 Containment Spray and Cooling SystemsB 3.6.6(continued)DIABLO CANYON -UNITS 1 & 2Rev 8C Page 38 of 50 MSIVsB 3.7.2BASESAPPLICABLE c. A break downstream of the MSIVs will be isolated by the closure ofSAFETY the MSIVs.ANALYSES(continued) d. Following a steam generator tube rupture, closure of the MSIVsisolates the ruptured steam generator from the intact steamgenerators to minimize radiological releases.e. The MSIVs are also utilized during other events such as afeedwater line break. This event is less limiting so far as MSIVOPERABILITY is concerned.The MSIVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(i).LCO This LCO requires that four MSIVs in the steam lines be OPERABLE.The MSIVs are considered OPERABLE when the isolation times arewithin limits, and they close on an isolation actuation signal.This LCO provides assurance that the MSIVs will perform their designsafety function to mitigate the consequences of accidents that couldresult in offsite exposures comparable to the 10 CFR -0050.67 (Ref. 4)limits or the NRC staff approved licensing basis.APPLICABILITY The MSIVs must be OPERABLE in MODE 1, and in MODES 2 and 3except when closed and de-activated (vented or prevented fromopening), when there is significant mass and energy in the RCS andsteam generators. When the MSIVs are closed, they are alreadyperforming the safety function.In MODE 4, the steam generator energy is low, thus OPERABILITY inMODE 4 is not required.In MODE 5 or 6, the steam generators do not contain much energybecause their temperature is below the boiling point of water; therefore,the MSIVs are not required for isolation of potential high energysecondary system pipe breaks in these MODES.ACTIONS A.1With one MSIV inoperable in MODE 1, action must be taken to restoreOPERABLE status within 8 hours. Some repairs to the MSIV can bemade with the unit hot. The 8 hour Completion Time is reasonable,considering the low probability of an accident occurring during this timeperiod that would require a closure of the MSIVs.The 8 hour Completion Time is greater than that normally allowed forcontainment isolation valves because the MSIVs are valves that isolatea closed system penetrating containment. These valves differ fromother containment isolation valves in that the closed system providesan additional means for containment isolation.(continued)IDIABLO CANYON -UNITS I & 2Rev8D Page9of87 MSIVsB 3.7.2BASESSURVEILLANCEREQUIREMENTSSR 3.7.2.1 (continued)analyses. This Surveillance is normally performed upon returning theunit to operation following a refueling outage. The MSIVs should notbe tested at power, since even a part stroke exercise increases the riskof a valve closure when the unit is generating power.As the MSIVs are not tested at power, they are exempt from the ASMECode, Section XI (Ref. 5), requirements during operation in MODE Ior 2.The Frequency is in accordance with the Inservice Testing Program.This test may be conducted in MODE 3 with the unit at operatingtemperature and pressure. This SR is modified by a Note that allowsentry into and operation in MODE 3 prior to performing the SR.However, the test is normally conducted in MODE 5 as permitted bythe cold shutdown frequency justification provided in the InserviceTesting Program (IST) and as permitted by Reference 6,Subsection ISTC-3521 (c).SR 3.7.2.2This SR verifies that each MSIV can close on an actual or simulatedactuation signal. This Surveillance is normally performed uponreturning the plant to operation following a refueling outage. TheSurveillance Frequency is based on operating experience, equipmentreliability, and plant risk and is controlled under the SurveillanceFrequency Control Program.REFERENCES 1. FSAR, Section 10.3.2. FSAR, Section 6, Appendix 6.2 D.3. FSAR, Section 15.4.2.4. 10 CFR I4--.14,50.67.5. ASME, Boiler and Pressure Vessel Code, Section XI.6. ASME Code for Operation and Maintenance of Nuclear PowerPlants, 2001 Edition including 2002 and 2003 Addenda.DIABLO CANYON -UNITS I & 2Rev 8D Page 11 of 87 CRVSB 3.7.10BASESBACKGROUND Redundant supply and recirculation trains provide the required filtration(continued) should an excessive pressure drop develop across the other filter train.Normally open isolation dampers are arranged in series pairs so thatthe failure of one damper to shut will not result in a breach of isolation.The CRVS is designed in accordance with Seismic Category lrequirements.The CRVS is designed to maintain a habitable environment in the CREfor the duration of the most severe Design Basis Accident (DBA)without exceeding a 5 rem wheGe-bedyTEDE dose or its equivalent toany part of the body.APPLICABLE The CRVS components are arranged in redundant, safety relatedSAFETY ventilation trains. The location of components and ducting within theANALYSES CRE ensures an adequate supply of filtered air to all areas requiringaccess. The CRVS provides airborne radiological protection for theCRE occupants, as demonstrated by the CRE occupant dose analysesfor the most limiting design basis accident, fission product releasepresented in the FSAR, Chapter 15 (Ref. 2).There are no offsite or onsite hazardous chemicals that would pose acredible threat to control room habitability. Consequently, engineeredcontrols for the control room are not required to ensure habitabilityagainst a hazardous chemical threat. The amount Of CRE unfilteredinleakage is not incorporated into PG&E's hazardous chemicalassessment.The evaluation of -a smoke challenge demonstrated that smoke will notresult in the inability of the CRE occupants to control the reactor eitherfrom the control room or from the remote shutdown panels (Ref. 1).The assessment verified that a fire or smoke event anywhere within theplant would not simultaneously render the Hot Shutdown Panel (HSP)and the CRE uninhabitable, nor would it prevent access from the CREto the HSP in the event remote shutdown is required. No CRVSautomatic actuation is required for hazardous chemical releases orsmoke and no Surveillance Requirements are required to verifyoperability in cases of hazardous chemicals or smoke.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8D Page 54 of 87 CRVSB 3.7.10BASESAPPLICABLE The worst case single active failure of a component of the CRVS,SAFETY assuming a loss of offsite power, does not impair the ability of theANALYSES system to perform its design function.(continued) The CRVS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).LCO Two independent and redundant CRVS trains are required to beOPERABLE to ensure that at least one is available if a single activefailure disables the other train. The redundant train means a secondtrain from the other unit (Ref. 5). Total system failure, such as from aloss of both ventilation trains or from an inoperable CRE boundary,could result in exceeding a dose of 5 rem whTlebodyTEDE er-its,equivalent to any part of the body to the CRE occupants in the event ofa large radioactive release.Each CRVS train is considered OPERABLE when the individualcomponents necessary to limit CRE occupant exposure areOPERABLE. A CRVS train is OPERABLE when the associated:a. main supply fan (one), filter booster fan (one) and pressurizationfan (one) are OPERABLE;b. HEPA filters and charcoal adsorbers are not excessively restrictingflow, and are capable of performing their filtration functions; andc. Ductwork, valves, and dampers are OPERABLE, and aircirculation can be maintained.In order for the CRVS trains to be considered OPERABLE, the CREboundary must be maintained such that the CRE occupant dose from alarge radioactive release does not exceed the calculated dose in thelicensing basis consequence analyses for DBAs. In the event of aninoperable CRE boundary in MODES 1, 2, 3, or 4, mitigating actionsare required to ensure CRE occupants are protected from hazardouschemicals and smoke.DCPP does not have CRVS automatic actuation for hazardouschemicals or smoke. Current practices at DCPP do not utilizechemicals in sufficient quantity to present a chemical hazard to thecontrol room. Smoke is not considered in the DCPP safety analyses.Therefore, there are no specific limits at DCPP for hazardouschemicals or smoke.(continued)DIABLO CANYON -UNITS I & 2Rev 8D Page 55 of 87 CRVSB 3.7.10BASES (continued)APPLICABILITY In MODES 1, 2, 3, 4, 5, and 6, and during movement of recentlyirradiated fuel assemblies (i.e., fuel that has occupied part of a criticalreactor core within the previous 1-0072 hours) the CRVS must beOPERABLE to ensure that the CRE will remain habitable during andfollowing a DBA or the release from the rupture of an outside wastegas tank.During movement of recently irradiated fuel assemblies, the CRVSmust be OPERABLE to cope with the release from a fuel handlingaccident involving handling recently irradiated fuel.CRVS OPERABILITY requires that for MODE 5 and 6 and duringmovement of recently irradiated fuel assemblies in either unit, whenthere is only one OPERABLE train of CRVS, the OPERABLE CRVStrain must be capable of being powered from an OPERABLE dieselgenerator that is directly associated with the bus which is energizingthe OPERABLE CRVS train. This is an exception to LCO 3.0.6.ACTIONS The ACTIONS are modified by a NOTE that states that ACTIONSapply simultaneously to both units. The CRVS is common to bothunits.A.1When one CRVS train is inoperable for reasons other than aninoperable CRE boundary, action must be taken to restore OPERABLEstatus within 7 days. In this Condition, the remaining OPERABLECRVS train is adequate to perform the CRE occupant protectionfunction. However, the overall reliability is reduced because a singlefailure in the OPERABLE CRVS train could result in loss of CRVSfunction. The 7 day Completion Time is based on the low probability ofa DBA occurring during this time period, and ability of the remainingtrain to provide the required capability.B.1, 8.2, and B.3The CRE boundary is inoperable if unfiltered inleakage past the CREboundary can result in CRE occupant radiological dose greater thanthe calculated dose of the licensing basis analyses of DBAconsequences (allowed to be up to 5 rem whole body Or its equival ,nto any o-f the body"TEDE).In the event of an inoperable CRE boundary in MODES 1, 2, 3, or 4,action must be initiated to implement mitigating actions to lessen theeffect on CRE occupants from the potential hazards of a radiological orchemical event or a challenge from smoke. Actions must be takenwithin 24 hours to verify that in the event of a DBA, the mitigatingactions will ensure that CRE occupant radiological exposures will notexceed the calculated dose of the licensing basis analyses of DBAconsequences, and that CRE occupants are protected from potentialsmoke and chemical hazards.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8D Page 57 of 87 CRVSB 3.7.10BASESSURVEILLANCE SR 3.7.10.3REQUIREMENTS This SR verifies that the required CRVS testing is performed in(continued) accordance with the Ventilation Filter Testing Program (VFTP). TheCRVS filter tests are in accordance with ANSI N510-1980 (Ref. 3).The VFTP includes testing the performance of the HEPA filter,charcoal adsorber efficiency, minimum flow rate, and the physicalproperties of the activated charcoal. Specific test Frequencies andadditional information are discussed in detail in the VFTP.SR 3.7.10.4This 'SR verifies that each CRVS train automatically starts andoperates in the pressurization mode on an actual or simulatedactuation signal generated from a Phase "A" Isolation. TheSurveillance Frequency is based on operating experience, equipmentreliability, and plant risk and is controlled under the SurveillanceFrequency Control Program.SR 3.7.10.5This SR verifies the OPERABILITY of the CRE boundary by testing forunfiltered air inleakage past the CRE boundary and into the CRE. Thedetails of the testing are specified in the Control Room EnvelopeHabitability Program. Any changes to the most limiting configuration ofthe CRVS testing alignment for determining unfiltered air inleakagepast the CRE boundary into the CRE must be made using aconservative decision making process (References 11-13).The CRE is considered habitable when the radiological dose to CREoccupants calculated in the licensing basis analyses of DBAconsequences is no more than 5 rem whole body Or its equivai.. t toany of th b and the CRE occupants are protected fromhazardous chemicals and smoke. For DCPP, there is no CRVSautomatic actuation for hazardous chemical releases or smoke andthere are no CRVS Surveillance Requirements that verify operability incases of hazardous chemicals or smoke. This SR verifies that theunfiltered air inleakage into the CRE is no greater than the flow rateassumed in the licensing basis analyses of DBA consequences. Whenunfiltered air inleakage is greater than the assumed flow rate,Condition B must be entered. Required Action B.3 allows time torestore the CRE boundary to OPERABLE status provided mitigatingactions can ensure that the CRE remains within the licensing basishabitability limits for the occupants following an accident.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8D Page 60 of 87 ABVSB 3.7.12BASESBACKGROUND(continued)The ABVS is discussed in the FSAR, Sections 9.4 2, and 15.5 (Refs. 1,and 2, respectively) since it may be used for normal, as well as postaccident, ventilation and atmospheric cleanup functions. The primarypurpose of the single manually initiated heater is to maintain therelative humidity at an acceptable level, consistent with iodine removalefficiencies per ASTM D 3803-1989 (Ref. 3). There is no redundantheater since the failure of the charcoal adsorber and heater train wouldconstitute a second failure in addition to the RHR pump seal failureassumed in conjunction with a LOCA (Ref.7). The heaters are notrequired for ABVS operability.APPLICABLESAFETYANALYSESThe design basis of the ABVS is established by the large break LOCA.The system evaluation assumes a passive failure of the ECCS outsidecontainment, such as an RHR pump seal failure, during therecirculation mode. In such a case, the system limits radioactiverelease to within the 10 CFR 41-050.67 (Ref. 5) limits. The analysis ofthe effects and consequences of a large break LOCA is presented inReference 2. The ABVS also functions, following a LOCA, in thosecases where the ECCS goes into the recirculation mode of long termcooling, to clean up releases of smaller leaks, such as from valve stempacking.The ventilation flow is also required to maintain the temperatures of theoperating ECCS motors within allowable limits. The ventilation functionhas been designed for single failure and the system will continue tofunction to provide its ECCS motor cooling function.Two types of system failures are considered in the accident analysis forradiation release: complete loss of function of one train, and excessiveRHR pump seal LEAKAGE. Either type of failure may result in a lowerefficiency of removal for any gaseous and particulate activity releasedto the ECCS pump rooms following a LOCA.The ABVS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).LCOTwo trains of the ABVS are required to be OPERABLE to ensure thatat least one is available, assuming that a single failure disables theother train coincident with loss of offsite power. Total system failurecould result in the atmospheric release from the ECCS pump roomexceeding 10 CFR-40050.67 limits in the event of a Design BasisAccident (DBA).ABVS is considered OPERABLE when the individual componentsnecessary to maintain the ECCS pump room filtration and temperatureare OPERABLE in both trains.(continued)DIABLO CANYON -UNITS I & 2Rev8D Page 65 of 87 ABVSB 3.7.12BASESSURVEILLANCE SR 3.7.12.6REQUIREMENTS This SR verifies the leak tightness of dampers that isolate flow to the(continued) normally operating filter train. This SR assures that the flow from theauxiliary building passes through the HEPA filter and charcoal adsorberunit when the ABVS Buildings and Safeguards or Safeguards Onlymodes have been actuated coincident with an SI. The SurveillanceFrequency is based on operating experience, equipment reliability, andplant risk and is controlled under the Surveillance Frequency ControlProgram.REFERENCES 1. FSAR, Section 9.4.2.2. FSAR, Section 15.5.3. ASTM D 3803-19894. ANSI N510-19805. 10 CFR 40-.450.67.6. NUREG-0800, Section 6.5.1, Rev. 2, July 1981.7. DCM S-23B, "Main Auxiliary Building Heating and VentilationSystem".DIABLO CANYON -UNITS 1 & 2Rev 8D Page 69 of 87 FHBVSB 3.7.13BASESAPPLICABLESAFETYANALYSES(continued)FHBVS is only required to isolate during fuel handling accidentsinvolving the handling of recently irradiated fuel (i.e., fuel that hasoccupied part of a critical reactor core within the previous4--72 hours). In accordance with assumptions made in the fuelhandling accident analysis, loss of offsite power is not consideredconcurrent with a fuel handling accident. However. less of pe.w.eicnVcoped by the fuel handling accident aRalysis,. To maximize FHBVScapability to mitigate-the consequences of a fuel handling accident, atleast one of the FHBVS trains must be capable of being supplied froman operable emergency diesel generator at all times whenevermovement of recently irradiated fuel is taking place in the spent fuelpool. These assumptions and the analysis follow the guidanceprovided in Regulatory Guide 4-.2-51.183 (Ref. 3).The FHBVS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).ILCOTwo independent and redundant trains of the FHBVS are required tobe OPERABLE to ensure that at least one train is available, assuminga single failure that disables the other train. In accordance withassumptions made in the fuel handling accident analysis, loss of offsitepower is not considered concurrent with a fuel handling accident.However, loss of power is enveloped by the fuel handling accidentaalysi&s-This requires that when two trains of the FHBVS areOPERABLE, at least one train of the FHBVS must be capable of beingpowered from an OPERABLE diesel generator that is directlyassociated with the bus which energizes the FHBVS train. When onlyone train is OPERABLE, an OPERABLE diesel generator must bedirectly associated with the bus which energizes that one OPERABLEFHBVS train. Total system failure could result in the atmosphericrelease from the fuel handling building exceeding the 10 CFR 40050.67(Ref. 4) limits in the event of a fuel handling accident.The FHBVS is considered OPERABLE when the individualcomponents necessary to control releases from fuel handling buildingare OPERABLE in both trains. An FHBVS train is consideredOPERABLE when its associated:a. Exhaust fan is OPERABLE;b. HEPA filter and charcoal adsorber are not excessively restrictingflow, and are capable of performing their filtration function; andc. Ductwork, valves, and dampers are OPERABLE, and aircirculation can be maintained.(continued)DIABLO CANYON -UNITS I & 2Rev 8D Page 71 of 87 FHBVSB 3.7.13BASESSURVEILLANCEREQUIREMENTS(continued)SR 3.7.13.4This SR verifies the integrity of the fuel handling building enclosure.The ability of the fuel handling building to maintain negative pressurewith respect to potentially uncontaminated adjacent areas isperiodically tested to verify proper function of the FHBVS. During thepost accident mode of operation, the FHBVS is designed to maintain aslight negative pressure in the fuel handling building, to preventunfiltered LEAKAGE. The FHBVS is designed to maintain the buildingpressure < -0.125 inches water gauge with respect to atmosphericpressure. The Surveillance Frequency is based on operatingexperience, equipment reliability, and plant risk and is controlled underthe Surveillance Frequency Control Program.SR 3.7.13.5Operation of damper M-29 is necessary to ensure that the systemfunctions properly. The operability of damper M-29 is verified if it canbe closed. The Surveillance Frequency is based on operatingexperience, equipment reliability, and plant risk and is controlled underthe Surveillance Frequency Control Program.REFERENCES1. FSAR, Section 9.4.4.2. FSAR, Section 15.5.3.4.5.6.7.8.9.10.11.Regulatory Guide 4-251.183. July 2000.10 CFRI- 050.67.ASTM D 3802-1989ANSI N510-1980.NUREG-0800, Section 6.5.1, Rev. 2, July 1981.DCM S-23D, "Fuel handling Building HVAC System."Not usedLicense Amendment 184/186, January 3, 2006.PG&E Letter DCL-05-124DIABLO CANYON -UNITS 1 & 2Rev 8D Page 74 of 87 Spent Fuel Storage Pool Water LevelB 3.7.15B 3.7 PLANT SYSTEMSB 3.7.15 Spent Fuel Storage Pool Water LevelBASESBACKGROUNDThe minimum water level in the spent fuel pool meets the assumptionsof iodine decontamination factors following a fuel handling accident.The specified w~ater level shields and minimizes the general area dosewhen the storage racks are filled to their maximum capacity. Thewater also provides shielding during the movement of spent fuel.A general description of the spent fuel pool design is given in theFSAR, Section 9.1.2 (Ref. 1). A description of~the Spent Fuel PoolCooling and Cleanup System is given in the FSAR, Section 9.1.3(Ref. 2). The assumptions of the fuel handling accident are given inthe FSAR, Section 9.1.4.3.4, 15.4.5 and 15.5.22 (Ref. 3).APPLICABLESAFETYANALYSESThe minimum water level in the spent fuel pool meets the assumptionsof the fuel handling accident described in Regulatory Guide 4--251.183(Ref. 4). The resultant 2 hour thyroid dose per person at the exclusion.area boundary is a small fraction of the 10 CFR 4GQ50.67 (Ref. 5)limits.According to Reference 4, there is 23 ft of water between the top of thedamaged fuel rods and the fuel pool surface during a fuel handlingaccident. With 23 ft of water, the assumptions of Reference 4 can beused directly. Although there are other spent fuel pool elevationswhere fuel handling accidents can occur, the design basis fuel handlingaccident, which uses the conservative assumptions of RG 4-Z51.183, isexpected to be bounding. To add conservatism, the analysis assumesthat all fuel rods of the damaged fuel assembly fail.In practice, the water level maintained for fuel handling provides morethan 23 feet of water over the top of irradiated fuel assemblies seatedin the storage racks. FSAR Section 9.1.4.3.4 requires the water levelprovide a minimum of 8 feet of water shielding during fuel handling.This assures more than 24 feet 6 inches of water shielding over the topof the fuel assemblies in the racks and more than. 23 feet of watershielding over a fuel assembly lying horizontally on top of the racks.The spent fuel pool water level satisfies Criterion 2 of10 CFR 50.36(c)(2)(ii).ILCOThe spent fuel pool water level is required to be > 23 ft over the top ofirradiated fuel assemblies seated in the storage racks. The specifiedwater level preserves the assumptions of the fuel handling accidentanalysis (Ref. 3). As such, it is the minimum required for fuel storageand movement within the fuel storage pool.(continued)DIABLO CANYON -UNITS I & 2Rev 8D Page 76 of 87 Spent Fuel Storage Pool Water LevelB 3.7.15BASES (continued)APPLICABILITYThis LCO applies during movement of irradiated fuel assemblies in thespent fuel pool, since the potential for a release of fission productsexists.ACTIONSA.1IRequired Action A.1 is modified by a Note indicating that LCO 3.0.3does not apply.When the initial conditions for prevention of an accident cannot be met,steps should be taken to preclude the accident from occurring. Whenthe spent fuel pool water level is lower than the required level, themovement of irradiated fuel assembly in the spent fuel pool isimmediately suspended. This does not preclude movement of a fuelassembly to a safe position.If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3would not specify any action. If moving irradiated fuel assemblies whilein MODES 1, 2, 3, and 4, the fuel movement is independent of reactoroperations. Therefore, inability to suspend movement of irradiated fuelassemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCEREQUIREMENTSSR 3.7.15.1This SR is done during the movement of irradiated fuel assemblies asstated in the Applicability. This SR verifies sufficient fuel storage poolwater is available in the event of a fuel handling accident. The waterlevel in the spent fuel pool must be checked periodically. TheSurveillance Frequency is based on operating experience, equipmentreliability, and plant risk and is controlled under the SurveillanceFrequency Control Program.During refueling operations, the level in the spent fuel pool is inequilibrium with the refueling canal, and the level in the refueling canalis checked daily in accordance with SR 3.9.7.1.REFERENCES 1. FSAR, Section 9.1.2.2. FSAR, Section 9.1.3.3. FSAR, Section 9.1.4.3.4, 15.4.5 and 15.5.22.4. Regulatory Guide 41--5.183, July 2000.5. 10 CFR 4-1-.450.67.DIABLO CANYON -UNITS 1 & 2Rev8D Page 77 of 87 Secondary Specific ActivityB 3.7.18B 3.7 PLANT SYSTEMSB 3.7.18 Secondary Specific ActivityBASESBACKGROUNDActivity in the secondary coolant results from steam generator tubeoutleakage from the Reactor Coolant System (RCS). Under steadystate conditions, the activity is primarily iodines with relatively short halflives and, thus, indicates current conditions. During transients, 1-131spikes have been observed as well as increased releases of somenoble gases. Other fission product isotopes, as well as activatedcorrosion products in lesser amounts, may also be found in thesecondary coolant.A limit on secondary coolant specific activity during power operationminimizes releases to the environment because of normal operation,anticipated operational occurrences, and accidents.This limit is lower than the activity value that might be expected from a4--0.75 gpm tube leak (LCO 3.4.13, "RCS Operational LEAKAGE") ofprimary coolant at the limit of 1.0 pCi/gm (LCO 3.4.16, "RCS Specific'Activity"). The steam line failure is assumed to result in the release ofthe noble gas and iodine activity contained in the steam generatorinventory, the feedwater, and the reactor coolant LEAKAGE. Most ofthe iodine isotopes have short half lives, (i.e., < 20 hours). Operatingat or below 0.1 pCi/gm ensures that in the event of a DBA, offsitedoses will be less than 10 CFR 4-0050.67_requirements.IIAPPLICABLESAFETYANALYSESThe accident analysis of the main steam line break (MSLB), asdiscussed in the FSAR, Chapter 15 (Ref. 2) assumes the initialsecondary coolant specific activity to have a radioactive isotopeconcentration of 0.10 pCi/gm DOSE EQUIVALENT 1-131. Thisassumption is used in the analysis for determining the radiologicalconsequences of the postulated accident. The accident analysis,based on this and other assumptions, shows that the radiologicalconsequences of an MSLB do not exceed 10 CFR 40050.67 limits(Ref. 1) foF whole body and thyroid dose ra. Ft .With the loss of offsite power, the remaining steam generators areavailable for core decay heat dissipation by venting steam to theatmosphere through the MSSVs and steam generator atmosphericdump valves (ADVs). The Auxiliary Feedwater System supplies thenecessary makeup to the steam generators. Venting continues untilthe reactor coolant temperature and pressure have decreasedsufficiently for the Residual Heat Removal System to complete thecooldown.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8D Page 85 of 87 Secondary Specific ActivityB 3.7.18BASESACTIONS A.1 and A.2 (continued)least MODE 3 within 6 hours, and in MODE 5 within 36 hours. Theallowed Completion Times are reasonable, based on operatingexperience, to reach the required unit conditions from full powerconditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.18.1REQUIREMENTS This SR verifies that the secondary specific activity is within the limitsof the accident analysis. A gamma isotopic analysis of the secondarycoolant, which determines DOSE EQUIVALENT 1-131, confirms thevalidity of the safety analysis assumptions as to the source terms inpost accident releases. It also serves to identify and trend any unusualisotopic concentrations that might indicate changes in reactor coolantactivity or LEAKAGE. The Surveillance Frequency is based onoperating experience, equipment reliability, and plant risk and iscontrolled under the Surveillance Frequency Control Program.REFERENCES 1. 10 CFR 409.450.67.2. FSAR, Chapter 15..IDIABLO CANYON -UNITS I & 2Rev 8D Page 87 of 87 AC Sources -ShutdownB 3.8.2B 3.8 ELECTRICAL POWER SYSTEMSB 3.8.2 AC Sources-ShutdownBASESBACKGROUND A description of the AC sources is provided in the Bases for LCO 3.8.1,"AC Sources -Operating."APPLICABLE The OPERABILITY of the minimum AC sources during MODES 5 andSAFETY 6 and during movement of recently irradiated fuel assemblies ensuresANALYSES that:a. The unit can be maintained in the shutdown or refuelingcondition for extended periods;b. Sufficient instrumentation and control capability is available formonitoring and maintaining the unit status; andc. Adequate AC electrical power is provided to mitigate eventspostulated during shutdown, such as a fuel handling accidentinvolving handling recently irradiated fuel. Due to radioactivedecay, AC electrical power is only required to mitigate fuelhandling accidents involving recently irradiated fuel (i.e., fuelthat has occupied part of a critical reactor core within theprevious 40072 hours).In general, when the unit is shut down, the Technical Specificationsrequirements ensure that the unit has the capability to mitigate theconsequences of postulated accidents. However, assuming a singlefailure and concurrent loss of all offsite or all onsite power is notrequired. The rationale for this is based on the fact that many DesignBasis Accidents (DBAs) that are analyzed in MODES 1, 2, 3, and 4have no specific analyses in MODES 5 and 6. Worst case boundingevents are deemed not credible in MODES 5 and 6 because theenergy contained within the reactor pressure boundary, reactor coolanttemperature and pressure, and the corresponding stresses result in theprobabilities of occurrence being significantly reduced or eliminated,and in minimal consequences. These deviations from DBA analysisassumptions and designrequirements during shutdown conditions areallowed by the LCO for required systems.During MODES 1, 2, 3, and 4, various deviations from the analysisassumptions and design requirements are allowed within the RequiredActions. This allowance is in recognition that certain testing andmaintenance activities must be conducted, provided an acceptablelevel of risk is not exceeded.' During MODES 5 and 6, performance ofa significant number of required testing and maintenance activities isalso required. In MODES 5 and 6, the activities are generally plannedand administratively controlled. Relaxations from MODE 1, 2, 3, and 4LCO requirements are acceptable during shutdown modes based on:a. The fact that time in an outage is limited. This is a risk prudentgoal as well as a utility economic consideration.(continued)DIABLO CANYON -UNITS 1 & 2Rev 8B Page 34 of 90 DC Sources -ShutdownB 3.8.5B 3.8 ELECTRICAL POWER SYSTEMSB 3.8.5 DC Sources-ShutdownBASESBACKGROUNDA description of the DC sources is provided in the Bases for LCO 3.8.4,"DC Sources-Operating."APPLICABLESAFETYANALYSESThe initial conditions of Design Basis Accident and transient analysesin the FSAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assume thatEngineered Safety Feature systems are OPERABLE. The DCelectrical power system provides normal and emergency DC electricalpower for the diesel generators, emergency auxiliaries, and control andswitching during all MODES of operation.The OPERABILITY of the DC subsystems is consistent with the initialassumptions of the accident analyses and the requirements for thesupported systems' OPERABILITY.The OPERABILITY of the minimum DC electrical power sources duringMODES 5 and 6 and during movement of recently irradiated fuelassemblies ensures that:a. The unit can be maintained in the shutdown or refuelingcondition for extended periods;b. Sufficient instrumentation and control capability is available formonitoring and maintaining the unit status; andc. Adequate DC electrical power is provided to mitigate eventspostulated during shutdown, such as a fuel handling accidentinvolving handling recently irradiated fuel. Due to radioactivedecay, DC electrical power is only required to mitigate fuelhandling accidents involving recently irradiated fuel (i.e., fuelthat has occupied part of a critical reactor core within theprevious 4-072 hours).The DC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).LCO The DC electrical power subsystems, each subsystem consisting ofone battery, one battery charger per battery, and the correspondingcontrol equipment and interconnecting class 1 E cabling within thesubsystem, are required to be OPERABLE to support required trains ofthe distribution systems required OPERABLE by LCO 3.8.10,"Distribution Systems-Shutdown." An OPERABLE subsystem consistsof a DC bus connected to a battery with an OPERABLE battery chargerwhich is fed from an OPERABLE AC vital bus (Ref B.3.8.10).(continued)DIABLO CANYON -UNITS I & 2Rev 8B Page 60 of 90 Inverters -ShutdownB 3.8.8B 3.8 ELECTRICAL POWER SYSTEMSB 3.8.8 Inverters-ShutdownBASESBACKGROUNDA description of the inverters is provided in the Bases for LCO 3.8.7,"Inverters -Operating."APPLICABLESAFETYANALYSESThe initial conditions of Design Basis Accident (DBA) and transientanalyses in the FSAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2),assume Engineered Safety Feature systems are OPERABLE. TheClass 1 E UPS inverters are designed to provide the required capacity,capability, redundancy, and reliability to ensure the availability ofnecessary power to the Reactor Protective System and EngineeredSafety Features Actuation System instrumentation and controls so thatthe fuel, Reactor Coolant System, and containment design limits arenot exceeded.The OPERABILITY of the inverters is consistent with the initialassumptions of the accident analyses and the requirements for thesupported systems' OPERABILITY.The OPERABILITY of the minimum inverters to each 120 VAC vital busduring MODES 5 and 6 and during movement of recently irradiated fuelassemblies ensures that:a. The unit can be maintained in the shutdown or refuelingcondition for extended periods;b. Sufficient instrumentation and control capability is available formonitoring and maintaining the unit status; andc. Adequate power is available to mitigate events postulatedduring shutdown, such as a fuel handling accident involvinghandling recently irradiated fuel. Due to radioactive decay, ACand DC inverters are only required to mitigate fuel handlingaccidents involving recently irradiated fuel (i.e., fuel that hasoccupied part of a critical reactor core within the previous40072 hours).The inverters were previously identified as part of the distributionsystem and, as such, satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).LCOThis ensures the availability of sufficient inverter power sources tooperate the unit in a safe manner and to mitigate the consequences ofpostulated events during shutdown (e.g., fuel handling accidentsinvolving handling recently irradiated fuel).(continued)DIABLO CANYON -UNITS 1 & 2Rev 8B Page 75 of 90 Distribution Systems -ShutdownB 3.8.10B 3.8 ELECTRICAL POWER SYSTEMSB 3.8.10 Distribution Systems -ShutdownBASESBACKGROUNDA description of the Class 1E AC, DC, and 120 VAC vital bus electricalpower distribution systems is provided in the Bases for LCO 3.8.9,"Distribution Systems -Operating."APPLICABLESAFETYANALYSESThe initial conditions of Design Basis Accident and transient analysesin the FSAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assumeEngineered Safety Feature (ESF) systems are OPERABLE. The ClassI E AC, DC, and 120 VAC vital bus electrical power distribution systemsare designed to provide sufficient capacity, capability, redundancy, andreliability to ensure the availability of necessary power to ESF systemsso that the fuel, Reactor Coolant System, and containment designlimits are not exceeded.The OPERABILITY of the Class 1E AC, DC, and 120 VAC vital buselectrical power distribution system is consistent with the initialassumptions of the accident analyses and the requirements for thesupported systems' OPERABILITY.The OPERABILITY of the minimum Class IE AC, DC, and 120 VACvital bus electrical power distribution subsystems during MODES 5 and6, and during movement of recently irradiated fuel assemblies ensuresthat:a. The unit can be maintained in the shutdown or refuelingcondition for extended periods;b. Sufficient instrumentation and control capability is available formonitoring and maintaining the unit status; andc. Adequate power is provided to mitigate events postulatedduring shutdown, such as a fuel handling accident involvinghandling recently irradiated fuel. Due to radioactive decay, ACand DC electrical power is only required to mitigate fuelhandling accidents involving handling recently irradiated fuel(i.e., fuel that has occupied part of a critical reactor core withinthe previous 4-072 hours).The Class IE AC, DC, and 120 VAC electrical power distributionsystems satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).ILCOVarious combinations of subsystems, equipment, and components arerequired OPERABLE by other LCOs, depending on the specific plantcondition. An OPERABLE AC subsystem shall consist of a 4kV vitalbus powered from at least one energized offsite power source with thecapability of being powered from an OPERABLE DG. The DG may bethe DG associated with that bus or, with administrative controls inplace, a DG that can be cross-tied (via the startup cross-tie feederbreakers) to another bus. However, credit for this cross-tie capability(continued)DIABLO CANYON -UNITS I & 2Rev 8B Page 87 of 90 Containment PenetrationsB 3.9.4B 3.9 REFUELING OPERATIONSB 3.9.4 Containment PenetrationsBASESBACKGROUNDIn MODES 1, 2, 3, and 4, the containment serves to contain fissionproduct radioactivity that may be released from the reactor corefollowing an accident, such that offsite radiation exposures aremaintained well within the requirements of 10 CFR 4-050.67.Additionally, in all operating modes the containment provides radiationshielding from the fission products that may be present in thecontainment atmosphere following accident conditions. Howeverduring CORE ALTERATIONS or movement of irradiated fuelassemblies within containment, the potential for containmentpressurization as a result of an accident is not likely; therefore,requirements to maintain the pressure boundary can be less stringent.An analysis has been performed that shows by meeting the LCO,during CORE ALTERATION and movement of irradiated fuelassemblies in containment, the potential release as a result of a fuelhandling accident (FHA) will remain well-within the requirements of10 CFR 4-0050.67 limits.The containment equipment hatch, which is part of the containmentpressure boundary, provides a means for moving large equipment andcomponents into and out of containment. The LCO requires thatduring CORE ALTERATIONS or the movement of irradiated fuelassemblies the equipment hatch must be capable of being closed andheld in place by at least four bolts. Good engineering practice dictatesthat the bolts required by this LCO be approximately equally spaced.The containment Personnel Air Lock (PAL) and Emergency Air Lock(EAL), which are also part of the containment pressure boundary,provide a means for personnel and emergency access duringMODES 1, 2, 3, and 4 unit operation in accordance with LCO 3.6.2,"Containment Air Locks." Each of these air locks has a door at bothends. The doors are normally interlocked to prevent simultaneousopening when containment OPERABILITY is required. During periodsof unit shutdown when the PAL and EAL are not required to be closed,the door interlock mechanisms may be disabled, allowing both doors ofeach of the air locks to remain open for extended periods whenfrequent containment entry is necessary.(continued)IDIABLO CANYON -UNITS 1 & 2Rev 8A Page 10 of 26 Containment PenetrationsB 3.9.4BASESBACKGROUND Per the FHA inside containment analysis, there are no closure(continued) restrictions required to limit any release to well within the requirementsof 10 CFR 4GG50.67 limits for offsite dose as the result of a fuelhandling accident during refueling. The LCO requirements forcontainment penetration closure are not provided to meet regulatoryrequirements, but rather to reduce the potential volume of the releaseof fission product radioactivity within containment to the environment.The Containment Purge and Exhaust System includes twosubsystems. The normal subsystem includes a 48 inch purgepenetration and a 48 inch exhaust penetration in Which the flow path islimited to beiRg open 200 hour or less per calendar year. The secondsubsystem, a pressure equalization system provides a single 12 inchsupply and exhaust penetration. The three valves in the 12 inchpressure equalization penetration can be opened intermittently-. Eachof these systems are qualified to closed automatically by theEngineered Safety Features Actuation System .(ESFAS). Neither of-tsubsystems is subject to a Specification in MODE 5.In MODE 6, large air exchanges are necessary to conduct refuelingoperations. The normal 48 inch purge system is used for this purpose,and all four valves are closed by the ESFAS in accordance withLCO 3.3.6, "Containment Purge and Exhaust IsolationInstrumentation."The pressure equalization system is disassembled and used inMODE 6 for other outage functions.The other, containment penetrations that provide direct access fromcontainment atmosphere to outside atmosphere must be isolated on atleast one side if they are not opened under administrative controls.Isolation may be achieved by an OPERABLE automatic isolation valve,or by a manual isolation valve, blind flange, or equivalent. The fueltransfer tube is open but closure is provided by an equivalent isolationof a water loop seal. Equivalent isolation methods must be approvedand may include use of a material that can provide a temporary,ventilation barrier for the other containment penetrations during fuelmovements (Ref. 1).Although the historic severe weather patterns for DCPP do not requireconsideration of tornados as part of the design basis, severe weatherconditions might occur at the site that could necessitate closure ofopen penetrations with direct access to the outside atmosphere duringrefueling operations with core alterations or irradiated fuel movementinside containment. As a result, administrative procedures shallrequire that closure of these penetrations be initiated immediately ifsevere weather warnings are in effect. All fuel handling activities insidecontainment shall be suspended until closure of the equipment hatch iscompleted.(continued)DIABLO CANYON -UNITS 1 & 2Rev8A Pagellof26 Containment PenetrationsB 3.9.4BASES (continued)APPLICABLESAFETYANALYSISDuring CORE ALTERATIONS, or movement of irradiated fuelassemblies within containment, the most severe radiologicalconsequences result from a fuel handling accident. The fuel handlingaccident is a postulated event that involves damage to irradiated fuel(Ref. 2). Fuel handling accident inside the containment is based ondropping a single irradiated fuel assembly of which all 264 fuel rodsrupture. In addition the analysis assumes free and rapidcommunication of air from the containment to the outside environment;the accident occurs 40072 hours after reactor shutdown; almostinstantaneous release of the entire containment volume to the outsideatmosphere; thyroid dose con..rsion factors based on ICRP 30(Ref. 4); a radial peaking factor of 1.65 based on 105% full poweroperation; and the other guidance from RG 4-_251.183. (Ref 5)..The requirements of LCO 3.9.7, "Refueling Cavity Water Level," andthe minimum decay time of 1-4072 hours prior to CORE ALTERATIONSensure that the release of fission product radioactivity, subsequent to afuel handling accident, results in doses that are well within theguideline values specified in 10 CFR 100. Standard Review Pl4an-,Secti on 15.7.4,Rev. 1 (Ref. 3), defines 2wellwithin" 10 CFR 100 to be 259o or less bfthe 10 CFR 100 values. The acceptance limits for offsitc radiationexposure... .ill be 25% of 10 C FR 100 values. less than the accidentdose criteria specified in Table 6 of RG 1.183 (Ref. 5).Containment penetrations satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).(continued)DIABLO CANYON -UNITS 1 & 2Rev 8A Page 12 of 26 Containment PenetrationsB 3.9.4BASES (continued)REFERENCES 1. Design Criteria Memorandum T-16, Containment Functions.2. FSAR, Section 15.4.5 and 15.5.22.3. NUREG 0800, Section 15.7.4, Rev. 1, July 198!Not Used.4. international Comm~iSSion on Radiological Pro~tectioPublication 30, "Limits for intakes of Radionuclides byWorkers," 1 Not Used.5. RG --.2-51.183, July 2000.DIABLO CANYON -UNITS 1 & 2Rev8A Page 16 of 26 Refueling Cavity Water LevelB 3.9.7B 3.9 REFUELING OPERATIONSB 3.9.7 Refueling Cavity Water LevelBASESBACKGROUNDThe movement of irradiated fuel assemblies within containment requiresa minimum water level of 23 ft above the top of the reactor vesselflange. During refueling, this maintains sufficient water level in thecontainment, refueling canal, fuel transfer canal, refueling cavity, andspent fuel pool. Sufficient water is necessary to retain iodine fissionproduct activity in the water in the event of a fuel handling accident(Refs. 2-aad-61 and 2). Sufficient iodine activity would be retained tolimit offsite doses from the accident to < 25% of 10 CFR 100 limits, asprovided by the guidance of Reference 3the acceptance criteria of 10CFR 50.67 (Ref. 4) and RG 1.183 (Ref. 1).APPLICABLESAFETYANALYSISDuring CORE ALTERATIONS and movement of irradiated fuelassemblies, the water level in the refueling canal and the refuelingcavity is an initial condition design parameter in the analysis of a fuelhandling accident in containment, as postulated by Regulatory Guide4-.2-5 1.183 (Ref. 1). A minimum water level of 23 ft allows adecontamination factor of 200 (Appendix B (2) of Ref. 61 approved inRef. 7) to be used in the accident analysis for iodine. This relates to theassumption that 99.5% of the total iodine released from the pellet tocladding gap of all the dropped fuel assembly rods is retained by therefueling cavity water. The fuel pellet to cladding gap is assumed tocontain 41-012% of 1-131 and 10% of corethe tota- fuel rod iodineinventory of all other iodine isotopes (Ref. 42).I1The fuel handling accident analysis inside containment is described inReference 2. With a minimum water level of 23 ft and a minimum decaytime of 1-0072 hours prior to fuel handling, the analysis and testprograms demonstrate that the iodine release due to a postulated fuelhandling accident is adequately captured by the water and offsite dosesare maintained well within allowable limits (Refs. 1 and 4, and 5).Refueling cavity water level satisfies Criterion 2 of10 CFR 50.36(c)(2)(ii).IILCOA minimum refueling cavity water level of 23 ft above the reactor vesselflange is required to ensure that the radiological consequences of apostulated fuel handling accident inside containment are withinacceptable limits, as provided by the guidance of Reference 31.I(continued)DIABLO CANYON -UNITS 1 & 2Rev 8A Page 25 of 26 Refueling Cavity Water LevelB 3.9.7BASES (continued)APPLICABILITYLCO 3.9.7 is applicable during CORE ALTERATIONS, except duringlatching and unlatching of control rod drive shafts, and when movingirradiated fuel assemblies within containment. The LCO minimizes thepossibility of a fuel handling accident in containment that is beyond theassumptions of the safety analysis. If irradiated fuel assemblies arenot present in containment, there can be no significant radioactivityrelease as a result of a postulated fuel handling accident.Requirements for fuel handling accidents in the spent fuel pool arecovered by LCO 3.7.15, "Fuel Storage Pool Water Level."ACTIONS A..1With a water level of < 23 ft above the top of the reactor vessel flange,all operations involving movement of irradiated fuel assemblies withinthe, containment shall be suspended immediately to ensure that a fuelhandling accident cannot occur.The suspension of fuel movement shall not preclude completion ofmovement of a component to a safe position.SURVEILLANCE SR 3.9.7.1REQUIREMENTS Verification of a minimum water level of 23 ft above the top of thereactor vessel flange ensures that the design basis for the analysis ofthe postulated fuel handling accident during refueling operations ismet. Water at the required level above the top of the reactor vesselflange limits the consequences of damaged fuel rods that arepostulated to result from a fuel handling accident inside containment(Ref. 2).The Surveillance Frequency is based on operating experience,equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.REFERENCES 1. Regulatory Guide 1.25, March 23, 19721.183, July 2000.2. FSAR, Section 15.4.5 and 15.5.22.3. NUREG 0800, Section 15.7.4. Not Used4. 10 CFR 4004050.67.5. Malmn'e. .ki, D. D., Bell, M. j., Duhn, E., and Locante, J.,V\GAP_ 828, .O.sequencc. of a Fuel HandlingAccident, December 197-1.- Not Used6. Appendix B (2) of Regulator. , Guide i.183, july 2000Not Used.7. License Amendment 155/155, October 2\1, 2002DIABLO CANYON -UNITS I & 2Rev 8A Page 26 of 26}}

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