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| | number = ML15215A368 | | | number = ML15215A368 |
| | issue date = 07/27/2015 | | | issue date = 07/27/2015 |
| | title = Millstone Power Station Unit 3 - Response to Third Request for Additional Information Regarding License Amendment Request for Implementation of WCAP-14333 and WCAP-15376 Reactor Trip System Instrumentation and Engineered Safety Feature Actu | | | title = Response to Third Request for Additional Information Regarding License Amendment Request for Implementation of WCAP-14333 and WCAP-15376 Reactor Trip System Instrumentation and Engineered Safety Feature Actuation.. |
| | author name = Sartain M D | | | author name = Sartain M |
| | author affiliation = Dominion Nuclear Connecticut, Inc | | | author affiliation = Dominion Nuclear Connecticut, Inc |
| | addressee name = | | | addressee name = |
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| | page count = 7 | | | page count = 7 |
| | project = TAC:MF4131 | | | project = TAC:MF4131 |
| | stage = Other | | | stage = Request |
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| {{#Wiki_filter:5000 Dominion Boulevard, Glen Allen, VA 23060 Dominion"' Web Address: www.dom.com' July 27, 2015U.S. Nuclear Regulatory Commission Serial No. 15-344Attention: Document Control Desk NLOS/WDC R0Washington, DC 20555 Docket No. 50-423License No. NPF-49DOMINION NUCLEAR CONNECTICUT. INC.MILLSTONE POWER STATION UNIT 3RESPONSE TO THIRD REQUEST FOR ADDITIONAL INFORMATION REGARDINGLICENSE AMENDMENT REQUEST FOR IMPLEMENTATION OF WCAP-14333 ANDWCAP-15376. REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEEREDSAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TEST TIMES ANDCOMPLETION TIMES (TAC NO. MF41 31)By letter dated May 8, 2014 and supplemented by a letter dated August 14, 2014, DominionNuclear Connecticut, Inc. (DNC) submitted a license amendment request (LAR) for MillstonePower Station Unit 3 (MPS3). The proposed amendment would revise TS 3/4.3.1, "ReactorTrip System Instrumentation," and TS 3/4.3.2, "Engineered Safety Feature Actuation SystemInstrumentation." These proposed changes are based on Westinghouse Electric CompanyLLC topical reports WCAP-14333-P-A, Revision 1, "Probabilistic Risk Analysis of the RPS andESFAS Test Times and Completion Times," and WCAP-1 5376-P-A, Revision 1, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor TripBreaker Test and Completion Times." In an email dated September 22, 2014, the NuclearRegulatory Commission (NRC) transmitted a request for additional information (RAI) to DNCrelated to the LAR. DNC responded to the RAI on October 15, 2014. In an email dated March17, 2015, the NRC transmitted a second RAI. DNC responded to the RAI on May 18, 2015.In an email dated June 25, 2015, the NRC transmitted a third RAI as a follow up to theresponses in the May 18, 2015 response letter.The attachment to this letter provides DNC's response to the NRC's third RAI.If you have any questions regarding this submittal, please contact Wanda Craft at (804) 273-4687.Sincerely,wO A \~I' ONhY 1tIJtLIC! ~Commonwealth of Virgirtia MakD ati I Meg. #1 40542 -'Mark .Satainl My ommisionExpires May 31, 201!8Vice President -Nuclear Engineering -COMMONWE.ALTH OF VIRGINIA)COUNTY OF HENRICO)The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark 0. Sartain, who is VicePresident -Nuclear Engineering of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file theforegoing document in behalf of that Company, an that the s ~tements in the document are true to the best of his knowledge and belief.Acknowledged before me this 22._"ray of /"kl. 2015.My Commission Expires: S'- Nota.ry ________' _________ (~ | | {{#Wiki_filter:5000 Dominion Boulevard, Glen Allen, VA 23060 Dominion |
| Serial No. 15-344Docket No. 50-423Page 2 of 2Commitments made in this letter: NoneAttachment:Response to Third Request for Additional Information Regarding License AmendmentRequest for Implementation of WCAP-14333 and WCAP-15376, Reactor Trip SystemInstrumentation and Engineered Safety Feature Actuation System Instrumentation TestTimes and Completion Timescc: U.S. Nuclear Regulatory CommissionRegion I2100 Renaissance Blvd, Suite 100King of Prussia, PA 19406-2713R. V. GuzmanSenior Project ManagerU.S. Nuclear Regulatory CommissionOne White Flint North, Mail Stop 08-C 211555 Rockville PikeRockville, MD 20852-2738NRC Senior Resident InspectorMillstone Power StationDirector, Radiation DivisionDepartment of Energy and Environmental Protection79 Elm StreetHartford, CT 06106-5127 Serial No. 15-344Docket No. 50-423ATTACHMENTRESPONSE TO THIRD REQUEST FOR ADDITIONAL INFORMATIONREGARDING LICENSE AMENDMENT REQUEST FOR IMPLEMENTATION OFWCAP-1 4333 AND WCAP-15376. REACTOR TRIP SYSTEMINSTRUMENTATION AND ENGINEERED SAFETY FEATURE ACTUATIONSYSTEM INSTRUMENTATION TEST TIMES AND COMPLETION TIMESDOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 Serial No. 15-344Docket No. 50-423Attachment, Page 1 of 4By letter dated May 8, 2014 and supplemented by a letter dated August 8, 2014,Dominion Nuclear Connecticut, Inc. (DNC) submitted a license amendment request([AR) for Millstone Power Station Unit 3 (MPS3). The proposed amendment wouldrevise TS 3/4.3.1, "Reactor Trip System Instrumentation," and TS 3/4.3.2, "EngineeredSafety Feature Actuation System Instrumentation." These proposed changes are basedon Westinghouse Electric Company LLC topical reports WCAP-14333-P-A, Revision 1,"Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times,"and WCAP-15376-P-A, Revision 1, "Risk-Informed Assessment of the RTS and ESFASSurveillance Test Intervals and Reactor Trip Breaker Test and Completion Times." Inan email dated September 22, 2014, the Nuclear Regulatory Commission (NRC)transmitted a request for additional information (RAI) to DNC related to the LAR. DNCresponded to the RAI on October 16, 2014. In an email dated March 17, 2015, the NRCtransmitted a second RAI. DNC responded to the RAI on May 18, 2015. In an emaildated June 25, 2015, the NRC transmitted a third RAI as a follow up to the responses inthe May 18, 2015 response letter. This attachment provides DNC's response to theNRC's RAI.RAI 1.1 (follow up to PRA RAI 1)The response to PRA RAI 1 states that: "These statements do not apply to the Tier 2restrictions listed in Attachment 4 of the LAR," in reference to Tier 2 restrictions onpages 15 and 17 of the LAR. Please explain this statement.DNC ResponseThe statement was meant to convey that the Tier 2 restrictions listed in Attachment 4 ofthe [AR apply when a logic train or reactor trip breaker is tested in bypass.The wording of the Tier 2 restrictions listed in Attachment 4 of the [AR was revised andcontinues to apply when a logic train or reactor trip breaker are tested in bypass. Seethe response to RAI 4 in the May 18, 2015 response letter for the revised wording.RAI 5.1 (follow up to PRA RAI 5)PRA RAI 5 is based on information in the LAR Section 4.2.1, "WCAP-14333 Tier 2Restrictions," rather than WCAP-153 76. The LAR states that there are no Tier 2limitations when a slave relay, master relay, or analog channel is inoperable. Thisconclusion appears to be based on information provided in Tables Ql 1. and Q18.1Ifrom a letter dated December 20, 1996, transmitting a response to a request foradditional information regarding WCAP-14333. Please explain how this Tier 2conclusion was reached and whether the information in these tables, which includes anassessment of master or slave relay maintenance in Q18. 1, was considered. | | "' Web Address: www.dom.com |
| Serial No. 15-344Docket No. 50-423Attachment, Page 2 of 4DNC ResponseLAR Section 4.2.1 is a discussion of the Westinghouse evaluation of Tier 2 restrictions.DNC considered the information from the Westinghouse evaluation as documented inTables QII.l and Q18.1 when determining Tier 2 restrictions. The Westinghouseevaluation determined the system importances individually for plant configurations withno ongoing test and maintenance activities and for plant configurations with ongoingtest or maintenance activities on the analog channels, master relays, slave relays, andlogic trains. The importances were compared between the cases with individualcomponents unavailable and all components available. With respect to the analogchannels, master relays and slave relays, the Westinghouse evaluation determined theimportance rankings among the affected systems did not change. That is the basis forthe statement there are no Tier 2 restrictions when a slave relay, master relay or analogchannel is inoperable.RAI 6.1 (follow up to PRA RAI 6)The/licensee's response to PRA RAI 6 describes the process to determine if Tier 2 orTier 3 compensatory measures are needed for the LAR proposed changes with respectto fire-related risk. The process, according to the response, incorporates qualitativeinsights based on fire mitigation strategy as follows:* Identify components that, when removed from service, render the unit with nocore damage mitigation success paths* Remove those components that are Technical Specification limited (i.e., have anallowed outage time (AOQT) < 72 hours) AND require transitioning to mode 5The response appears to "or" these conditions together in that for components thatmeet either criterion, Technical Requirements Manual actions are established when afire risk significant component is removed from service for greater than 72 hours. Thesecond condition related to removing components that are Technical Specification (TS)limited does not consider potential risk significance. Based on the RAI response, itappears some components may not be qualitatively (or quantitatively) considered forTier 2 or Tier 3 during the proposed TS bypass times or completion times which are<- 72 hours. Therefore, describe an acceptable Tier 2 and Tier 3 process with respect tofire-related risk and the results of the Tier 2 assessment.DNC ResponseThe Tier 2 and Tier 3 process with respect to fire-related risk and the results of the Tier2 assessment is provided below. DNC used the guidance provided in NUMARC 93-01,Revision 4A Section 11.3.4.3 when assessing fire risk within the 10 CFR 50.65(a)(4) | | ' July 27, 2015 U.S. Nuclear Regulatory Commission Serial No. 15-344 Attention: Document Control Desk NLOS/WDC R0 Washington, DC 20555 Docket No. 50-423 License No. NPF-49 DOMINION NUCLEAR CONNECTICUT. INC. |
| Serial No. 15-344Docket No. 50-423Attachment, Page 3 of 4process. As a result of the clarification call associated with this RAI, a subset of thecriteria used to identify fire risk significant components within (a)(4) was applied to theequipment considered in the LAR. The criteria are:1. Incorporate quantitative PRA insights* Identify the components corresponding to random failure andtest/maintenance basic events with internal events core damage frequencyrisk achievement worth (CDF RAW) > 2.0.*Remove those components not listed on the Safe Shutdown Equipment List(SSEL).2. Incorporate qualitative insights based on fire mitigation strategy* Identify components that, when removed from service, render the unit with nocore damage mitigating success paths.The reactor protection system (RPS) components affected by the proposed amendmentare not fire risk significant and thus, any fire risk incurred would be due to the additionalequipment removed from service.Based on the fire risk significant criteria listed above, components only meeting criterion1 are less risk significant than components meeting criterion 2 since redundantequipment would be available to mitigate fire scenarios when criterion 1-onlycomponents are out of service. Furthermore, the risk of these configurations isadequately managed by Technical Requirements Manual (TRM) risk managementactions. As a result, criterion 1-only components do not warrant additional Tier 2 or Tier3 compensatory measures.Alternatively, criterion 2 components provide the only available core damage mitigationsuccess path for certain fire scenarios and therefore, have high fire risk significance dueto the lack of redundancy. Consequently, components meeting criterion 2 are deemedreasonable candidates for developing a Tier 2 or Tier 3 compensatory measure.The turbine driven AFW pump is the only component that meets criterion 2 and hasalready been included in the proposed Tier 2 restrictions (i.e., AFW system componentswill not be removed from service when a reactor trip breaker is inoperable formaintenance). As a result, no additional Tier 2 or Tier 3 compensatory measures arerecommended.RAI 12.1 (follow up to PRA RAI 12)The LAR Table 1 provides a comparison between WCAP-14333 analysis assumptionsand plant-specific parameters. PRA RAI 12 requested an explanation whether theWCAP-14333 Tier I analysis remained bounding for these plant-specific values. Theresponse to PRA RAI 12 states that the plant-specific values for MPS3 shown in Table Serial No. 15-344Docket No. 50-423Attachment, Page 4 of 4I of Attachment 3 of the LAR are consistent with those of Vogtle Electric GeneratingPlant (Vogtle) as shown on Table 1 on page 297 (Enclosure 5, page E5-2) of WCAP-14333-P-A, Revision 1, Supplement 1, dated September 2003. However, a review ofTSTF-418 shows that this reference is dated after the NRC staff's approval letter ofTSTF-418 (ADAMS Accession No. ML 030920633). Therefore, the referenceddocument does not appear to have been part of the NRC staff's review of TSTF-4 18,nor does it appear to be provided as a reference supporting the proposed TS changesin the LAR. Please provide an explanation as requested in PRA RAI 12, consistent withthe TSTF-418 traveler which the LAR is requesting to adopt.DNC ResponseGiven that less testing/maintenance is being performed on some slave relays and thereactor trip breakers, the unavailability of these components will be less than thatassumed in the WCAP. Since the MPS3 unavailability values are less than those usedin the WCAP, MPS3 is bounded by the WCAP risk analysis. | | MILLSTONE POWER STATION UNIT 3 RESPONSE TO THIRD REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST FOR IMPLEMENTATION OF WCAP-14333 AND WCAP-15376. REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TEST TIMES AND COMPLETION TIMES (TAC NO. MF41 31) |
| }} | | By letter dated May 8, 2014 and supplemented by a letter dated August 14, 2014, Dominion Nuclear Connecticut, Inc. (DNC) submitted a license amendment request (LAR) for Millstone Power Station Unit 3 (MPS3). The proposed amendment would revise TS 3/4.3.1, "Reactor Trip System Instrumentation," and TS 3/4.3.2, "Engineered Safety Feature Actuation System Instrumentation." These proposed changes are based on Westinghouse Electric Company LLC topical reports WCAP-14333-P-A, Revision 1, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times," and WCAP-1 5376-P-A, Revision 1, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times." In an email dated September 22, 2014, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) to DNC related to the LAR. DNC responded to the RAI on October 15, 2014. In an email dated March 17, 2015, the NRC transmitted a second RAI. DNC responded to the RAI on May 18, 2015. |
| | In an email dated June 25, 2015, the NRC transmitted a third RAI as a follow up to the responses in the May 18, 2015 response letter. |
| | The attachment to this letter provides DNC's response to the NRC's third RAI. |
| | If you have any questions regarding this submittal, please contact Wanda Craft at (804) 273-4687. |
| | Sincerely, wO A\~I' ONhY 1tIJtLIC |
| | ! ~Commonwealth of Virgirtia |
| | * MakD ati I Meg. #1 40542 -' |
| | Mark . Satainl My ommisionExpires May 31, 201!8 Vice President - Nuclear Engineering - |
| | COMMONWE.ALTH OF VIRGINIA) |
| | COUNTY OF HENRICO) |
| | The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark 0. Sartain, who is Vice President - Nuclear Engineering of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the |
| | ~tements |
| | (~ |
| | foregoing document in behalf of that Company, an that the s in the document are true to the best of his knowledge and belief. |
| | Acknowledged before me this 22._"ray of /"kl. 2015. |
| | My Commission Expires: S'- */-- Nota.ry |
| | * Serial No. 15-344 Docket No. 50-423 Page 2 of 2 Commitments made in this letter: None |
| | |
| | ==Attachment:== |
| | |
| | Response to Third Request for Additional Information Regarding License Amendment Request for Implementation of WCAP-14333 and WCAP-15376, Reactor Trip System Instrumentation and Engineered Safety Feature Actuation System Instrumentation Test Times and Completion Times cc: U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 R. V. Guzman Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08-C 2 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director, Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127 |
| | |
| | Serial No. 15-344 Docket No. 50-423 ATTACHMENT RESPONSE TO THIRD REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST FOR IMPLEMENTATION OF WCAP-1 4333 AND WCAP-15376. REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TEST TIMES AND COMPLETION TIMES DOMINION NUCLEAR CONNECTICUT, INC. |
| | MILLSTONE POWER STATION UNIT 3 |
| | |
| | Serial No. 15-344 Docket No. 50-423 Attachment, Page 1 of 4 By letter dated May 8, 2014 and supplemented by a letter dated August 8, 2014, Dominion Nuclear Connecticut, Inc. (DNC) submitted a license amendment request |
| | ([AR) for Millstone Power Station Unit 3 (MPS3). The proposed amendment would revise TS 3/4.3.1, "Reactor Trip System Instrumentation," and TS 3/4.3.2, "Engineered Safety Feature Actuation System Instrumentation." These proposed changes are based on Westinghouse Electric Company LLC topical reports WCAP-14333-P-A, Revision 1, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times," |
| | and WCAP-15376-P-A, Revision 1, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times." In an email dated September 22, 2014, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) to DNC related to the LAR. DNC responded to the RAI on October 16, 2014. In an email dated March 17, 2015, the NRC transmitted a second RAI. DNC responded to the RAI on May 18, 2015. In an email dated June 25, 2015, the NRC transmitted a third RAI as a follow up to the responses in the May 18, 2015 response letter. This attachment provides DNC's response to the NRC's RAI. |
| | RAI 1.1 (follow up to PRA RAI 1) |
| | The response to PRA RAI 1 states that: "These statements do not apply to the Tier 2 restrictions listed in Attachment 4 of the LAR," in reference to Tier 2 restrictions on pages 15 and 17 of the LAR. Please explain this statement. |
| | DNC Response The statement was meant to convey that the Tier 2 restrictions listed in Attachment 4 of the [AR apply when a logic train or reactor trip breaker is tested in bypass. |
| | The wording of the Tier 2 restrictions listed in Attachment 4 of the [AR was revised and continues to apply when a logic train or reactor trip breaker are tested in bypass. See the response to RAI 4 in the May 18, 2015 response letter for the revised wording. |
| | RAI 5.1 (follow up to PRA RAI 5) |
| | PRA RAI 5 is based on information in the LAR Section 4.2.1, "WCAP-14333 Tier 2 Restrictions," rather than WCAP-153 76. The LAR states that there are no Tier 2 limitations when a slave relay, master relay, or analog channel is inoperable. This conclusion appears to be based on information provided in Tables Ql 1. and Q18.1I from a letter dated December 20, 1996, transmitting a response to a request for additional information regarding WCAP-14333. Please explain how this Tier 2 conclusion was reached and whether the information in these tables, which includes an assessment of master or slave relay maintenance in Q18. 1, was considered. |
| | |
| | Serial No. 15-344 Docket No. 50-423 Attachment, Page 2 of 4 DNC Response LAR Section 4.2.1 is a discussion of the Westinghouse evaluation of Tier 2 restrictions. |
| | DNC considered the information from the Westinghouse evaluation as documented in Tables QII.l and Q18.1 when determining Tier 2 restrictions. The Westinghouse evaluation determined the system importances individually for plant configurations with no ongoing test and maintenance activities and for plant configurations with ongoing test or maintenance activities on the analog channels, master relays, slave relays, and logic trains. The importances were compared between the cases with individual components unavailable and all components available. With respect to the analog channels, master relays and slave relays, the Westinghouse evaluation determined the importance rankings among the affected systems did not change. That is the basis for the statement there are no Tier 2 restrictions when a slave relay, master relay or analog channel is inoperable. |
| | RAI 6.1 (follow up to PRA RAI 6) |
| | The/licensee's response to PRA RAI 6 describes the process to determine if Tier 2 or Tier 3 compensatory measures are needed for the LAR proposed changes with respect to fire-related risk. The process, according to the response, incorporates qualitative insights based on fire mitigation strategy as follows: |
| | * Identify components that, when removed from service, render the unit with no core damage mitigation success paths |
| | * Remove those components that are Technical Specification limited (i.e., have an allowed outage time (AOQT) < 72 hours) AND require transitioningto mode 5 The response appears to "or" these conditions together in that for components that meet either criterion, Technical Requirements Manual actions are established when a fire risk significant component is removed from service for greater than 72 hours. The second condition related to removing components that are Technical Specification (TS) limited does not consider potential risk significance. Based on the RAI response, it appears some components may not be qualitatively (or quantitatively) considered for Tier 2 or Tier 3 during the proposed TS bypass times or completion times which are |
| | <- 72 hours. Therefore, describe an acceptable Tier 2 and Tier 3 process with respect to fire-related risk and the results of the Tier 2 assessment. |
| | DNC Response The Tier 2 and Tier 3 process with respect to fire-related risk and the results of the Tier 2 assessment is provided below. DNC used the guidance provided in NUMARC 93-01, Revision 4A Section 11.3.4.3 when assessing fire risk within the 10 CFR 50.65(a)(4) |
| | |
| | Serial No. 15-344 Docket No. 50-423 Attachment, Page 3 of 4 process. As a result of the clarification call associated with this RAI, a subset of the criteria used to identify fire risk significant components within (a)(4) was applied to the equipment considered in the LAR. The criteria are: |
| | : 1. Incorporate quantitative PRA insights |
| | * Identify the components corresponding to random failure and test/maintenance basic events with internal events core damage frequency risk achievement worth (CDF RAW) > 2.0. |
| | *Remove those components not listed on the Safe Shutdown Equipment List (SSEL). |
| | : 2. Incorporate qualitative insights based on fire mitigation strategy |
| | * Identify components that, when removed from service, render the unit with no core damage mitigating success paths. |
| | The reactor protection system (RPS) components affected by the proposed amendment are not fire risk significant and thus, any fire risk incurred would be due to the additional equipment removed from service. |
| | Based on the fire risk significant criteria listed above, components only meeting criterion 1 are less risk significant than components meeting criterion 2 since redundant equipment would be available to mitigate fire scenarios when criterion 1-only components are out of service. Furthermore, the risk of these configurations is adequately managed by Technical Requirements Manual (TRM) risk management actions. As a result, criterion 1-only components do not warrant additional Tier 2 or Tier 3 compensatory measures. |
| | Alternatively, criterion 2 components provide the only available core damage mitigation success path for certain fire scenarios and therefore, have high fire risk significance due to the lack of redundancy. Consequently, components meeting criterion 2 are deemed reasonable candidates for developing a Tier 2 or Tier 3 compensatory measure. |
| | The turbine driven AFW pump is the only component that meets criterion 2 and has already been included in the proposed Tier 2 restrictions (i.e., AFW system components will not be removed from service when a reactor trip breaker is inoperable for maintenance). As a result, no additional Tier 2 or Tier 3 compensatory measures are recommended. |
| | RAI 12.1 (follow up to PRA RAI 12) |
| | The LAR Table 1 provides a comparison between WCAP-14333 analysis assumptions and plant-specific parameters. PRA RAI 12 requested an explanation whether the WCAP-14333 Tier I analysis remained bounding for these plant-specific values. The response to PRA RAI 12 states that the plant-specific values for MPS3 shown in Table |
| | |
| | Serial No. 15-344 Docket No. 50-423 Attachment, Page 4 of 4 I of Attachment 3 of the LAR are consistent with those of Vogtle Electric Generating Plant (Vogtle) as shown on Table 1 on page 297 (Enclosure 5, page E5-2) of WCAP-14333-P-A, Revision 1, Supplement 1, dated September 2003. However, a review of TSTF-418 shows that this reference is dated after the NRC staff's approval letter of TSTF-418 (ADAMS Accession No. ML030920633). Therefore, the referenced document does not appear to have been part of the NRC staff's review of TSTF-4 18, nor does it appear to be provided as a reference supporting the proposed TS changes in the LAR. Pleaseprovide an explanation as requested in PRA RAI 12, consistent with the TSTF-418 traveler which the LAR is requesting to adopt. |
| | DNC Response Given that less testing/maintenance is being performed on some slave relays and the reactor trip breakers, the unavailability of these components will be less than that assumed in the WCAP. Since the MPS3 unavailability values are less than those used in the WCAP, MPS3 is bounded by the WCAP risk analysis.}} |
Letter Sequence Request |
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TAC:MF4131, RPS and ESFAS Test Times and Completion Times - WCAP-14333, TSTF-41 (Open) |
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MONTHYEARML14234A0972014-08-14014 August 2014
Supplemental Information to License Amendment Request, Implementation of WCAP-14333 and WCAP-15376, Reactor Trip System Instrumentation and Engineered Safety Feature Actuation System Instrumentation Test Times..
Project stage: Supplement
ML15110A1542015-03-17017 March 2015
Email from M.Thadani to W.Craft Request for Additional Information, License Amendment Request - TSTF-411, TSTF-418
Project stage: RAI
ML15147A0182015-05-18018 May 2015
Response to Second Request for Additional Information Regarding License Amendment Request for Implementation of WCAP-14333 and WCAP-15376, Reactor Trip System Instrumentation and Engineered Safety Feature Actuation System..
Project stage: Request
ML15215A3682015-07-27027 July 2015
Response to Third Request for Additional Information Regarding License Amendment Request for Implementation of WCAP-14333 and WCAP-15376 Reactor Trip System Instrumentation and Engineered Safety Feature Actuation..
Project stage: Request
2015-03-17
[Table View] |
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Category:Letter
MONTHYEARIR 05000336/20240032024-11-0707 November 2024
Integrated Inspection Report 05000336/2024003 and 05000423/2024003 and Apparent Violation and Independent Spent Fuel Storage Installation Inspection Report 07200008/2024001
ML24289A0152024-10-21021 October 2024
Review of the Fall 2023 Steam Generator Tube Inspection Report
05000423/LER-2024-001, Loss of Safety Function and Condition Prohibited by Technical Specifications for Loss of Secondary Containment Boundary2024-10-14014 October 2024
Loss of Safety Function and Condition Prohibited by Technical Specifications for Loss of Secondary Containment Boundary
IR 05000336/20244022024-10-0808 October 2024
Security Baseline Inspection Report 05000336/2024402 and 05000423/2024402 (Cover Letter Only)
ML24281A1102024-10-0707 October 2024
Requalification Program Inspection
05000423/LER-2023-006-02, Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications2024-09-26026 September 2024
Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications
ML24240A1692024-09-18018 September 2024
Cy 2023 Summary of Decommissioning Trust Fund Status
ML24260A2192024-09-16016 September 2024
Decommissioning Trust Fund Disbursement - Revision to Previous Thirty-Day Written Notification
ML24260A1952024-09-16016 September 2024
Response to Request for Additional Information Regarding Proposed Amendment to Support Implementation of Framatome Gaia Fuel
ML24248A2272024-09-0404 September 2024
Operator Licensing Examination Approval
ML24240A1532024-09-0303 September 2024
Summary of Regulatory Audit Supporting the Review of License Amendment Request for Implementation of Framatome Gaia Fuel
IR 05000336/20240052024-08-29029 August 2024
Updated Inspection Plan for Millstone Power Station, Units 2 and 3 (Reports 05000336/2024005 and 05000423/2024005
IR 05000336/20240022024-08-13013 August 2024
Integrated Inspection Report 05000336/2024002 and 05000423/2024002
ML24221A2872024-08-0808 August 2024
Independent Spent Fuel Storage Installation (ISFSI) - Submittal of Cask Registration for Spent Fuel Storage
IR 05000336/20244412024-08-0606 August 2024
Supplemental Inspection Report 05000336/2024441 and 05000423/2024441 and Follow-Up Assessment Letter (Cover Letter Only)
ML24212A0742024-08-0505 August 2024
Request for Withholding Information from Public Disclosure - Millstone Power Station, Unit No. 3, Proposed Alternative Request IR-4-13 to Support Steam Generator Channel Head Drain Modification
ML24211A1712024-07-25025 July 2024
Associated Independent Spent Fuels Storage Installation, Revision to Emergency Plan - Report of Change
IR 05000336/20244032024-07-22022 July 2024
Information Request for the Cybersecurity Baseline Inspection, Notification to Perform Inspection 05000336/2024403 and 05000423/2024403
IR 05000336/20245012024-07-0101 July 2024
Emergency Preparedness Biennial Exercise Inspection Report 05000336/2024501 and 05000423/2024501
ML24180A0932024-06-28028 June 2024
Readiness for Additional Inspection: EA-23-144
IR 05000336/20240102024-06-26026 June 2024
Biennial Problem Identification and Resolution Inspection Report 05000336/2024010 and 05000423/2024010
ML24178A2422024-06-25025 June 2024
2023 Annual Report of Emergency Core Cooling System (ECCS) Model, Changes Pursuant to the Requirements of 10 CFR 50.46
IR 05000336/20244402024-06-24024 June 2024
Final Significance Determination for Security-Related Greater than Green Finding(S) with Assessment Follow-up; IR 05000336/2024440 and 05000423/2024440 and Notice of Violation(S), NRC Investigation Rpt 1-2024-001 (Cvr Ltr Only)
ML24176A1782024-06-20020 June 2024
Update to the Final Safety Analysis Report
ML24176A2622024-06-20020 June 2024
Update to the Final Safety Analysis Report, Revision 37
ML24177A2792024-06-20020 June 2024
Preparation and Scheduling of Operator Licensing Examinations
ML24280A0012024-06-20020 June 2024
Update to the Final Safety Analysis Report (Redacted Version)
ML24281A2072024-06-20020 June 2024
Update to the Final Safety Analysis Report, Revision 37 (Redacted Version)
05000336/LER-2024-001, Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications2024-06-10010 June 2024
Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications
ML24170B0532024-06-10010 June 2024
DOM-NAF-2-P/NP-A, Revision 0.5, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code
ML24165A1292024-06-0505 June 2024
ISFSI, 10 CFR 50.59 Annual Change Report for 2023 Annual Regulatory Commitment Change Report for 2023
ML24128A2772024-06-0404 June 2024
Issuance of Amendment No. 290 to Revise TSs for Reactor Core Safety Limits, Fuel Assemblies, and Core Operating Limits Report for Use of Framatome Gaia Fuel (EPID L-2023-LLA-0074) (Non-Proprietary)
ML24151A6482024-06-0303 June 2024
Changes in Reactor Decommissioning Branch Project Management Assignments for Some Decommissioning Facilities
ML24110A0562024-05-21021 May 2024
Exemption from the Requirements of 10 CFR Part 50, Section 50.46, and Appendix K Regarding Use of M5 Cladding Material (EPID L-2023-LLE-0013) (Letter)
ML24109A0032024-05-21021 May 2024
Issuance of Amendment No. 289 to Revise Technical Specifications to Use Framatome Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limit (EPID L-2023-LLA-0065) (Non-Proprietary)
ML24142A0952024-05-20020 May 2024
End of Cycle 22 Steam Generator Tube Inspection Report
05000423/LER-2023-006-01, Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications2024-05-20020 May 2024
Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications
ML24141A2432024-05-20020 May 2024
Response to Request for Additional Information Regarding Alloy 600 Aging Management Program Submittal Related to License Renewal Commitment No. 15
IR 05000336/20240012024-05-14014 May 2024
Integrated Inspection Report 05000336/2024001 and 05000423/2024001 and Apparent Violation
ML24123A2042024-05-0202 May 2024
Pre-Decisional Replay to EA-23-144
05000423/LER-2023-006, Pressurizer Power Operated Relief Valve Failed to Stroke Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications2024-05-0202 May 2024
Pressurizer Power Operated Relief Valve Failed to Stroke Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications
IR 05000336/20244012024-04-30030 April 2024
Security Baseline Inspection Report 05000336/2024401 and 05000423/2024401 (Cover Letter Only)
ML24123A1222024-04-30030 April 2024
Inservice Inspection Program - Owners Activity Report, Refueling Outage 22
ML24116A0452024-04-25025 April 2024
Special Inspection Follow-Up Report 05000336/2024440 and 05000423/2024440 and Preliminary Finding(S) of Greater than Very Low Significance and NRC Investigation Report No. 1-2024-001 (Cover Letter Only)
ML24116A1742024-04-24024 April 2024
Annual Radiological Environmental Operating Report
ML24114A2662024-04-24024 April 2024
Submittal of 2023 Annual Radioactive Effluent Release Report
ML24103A0202024-04-22022 April 2024
Summary of Regulatory Audit in Support of License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits
ML24106A2032024-04-15015 April 2024
2023 Annual Environmental Operating Report
ML24088A3302024-04-0404 April 2024
Regulatory Audit Plan in Support of License Amendment Request to Implement Framatome Gaia Fuel
ML24093A2162024-04-0101 April 2024
Response to Request for Additional Information Regarding License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits
2024-09-04
[Table view]
Category:Response to Request for Additional Information (RAI)
MONTHYEARML24260A1952024-09-16016 September 2024
Response to Request for Additional Information Regarding Proposed Amendment to Support Implementation of Framatome Gaia Fuel
ML24141A2432024-05-20020 May 2024
Response to Request for Additional Information Regarding Alloy 600 Aging Management Program Submittal Related to License Renewal Commitment No. 15
ML24093A2162024-04-0101 April 2024
Response to Request for Additional Information Regarding License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits
ML23361A0942023-12-21021 December 2023
Response to Request for Additional Information Regarding Proposed License Amendment Request to Revise Technical Specifications for Reactor Core Safety Limits, Fuel Assemblies and Core Operating Limits Report .
ML23248A2132023-08-30030 August 2023
Response to Request for Additional Information Regarding Proposed License Amendment Request to Revise the Applicability Term for Reactor Coolant System Heatup and Cooldown Pressure-Temperature.
ML23208A0922023-07-26026 July 2023
Request for Approval of Appendix F of Fleet Report DOM-NAF-2-P Qualification of Framatome ORFEO-GAIA and OORFE-NMGRID CHF Correlations in the Dominion Energy Vipre-D Computer Code Response
ML23124A3642023-04-20020 April 2023
Response to Request for Additional Information for Spring 2022 Steam Generator Tube Inspection Report
ML23096A2982023-04-0606 April 2023
Units 1 and 2 and Millstone Power Station, Units 2 and 3 - Request for Approval of Appendix F of Fleet Report DOM-NAF-2-P Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion
ML22312A4432022-11-0707 November 2022
NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Fleet Response to RAI
ML21259A0852021-09-15015 September 2021
North Ann, and Surry, Units 1 and 2, Millstone, Units 2 and 3, Request for Approval of Appendix E of Fleet Report DOM-NAF-2-A Qualification of the Framatome BWU-I CHF Correlation in the Vipre-D Computer Code Response to Request for Addition
ML21209A7622021-07-26026 July 2021
Response to Request for Additional Information for Alternative Request V-01 - Proposed Request for Alternative Frequency to Supplemental Valve Positionn Verification Testing Requirements
ML21153A4132021-06-0202 June 2021
Response to Request for Additional Information Regarding License Amendment Request for Measurement Uncertainty Recapture Power Uprate
ML21147A4772021-05-27027 May 2021
NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors - Final Supplemental Response
ML21140A2992021-05-20020 May 2021
Response to Request for Additional Information for Proposed License Amendment Request to Add an Analytical Methodology to the Core Operating Limits Report for a Large Break Loss of Coolant Accident
ML21133A2852021-05-13013 May 2021
Stations, Units 1 & 2 and Millstone Power Station, Units 2 and 3 - Request for Approval of Appendix E Fleet Report DOM-NAF-2-A Qualification of the Framatome Bwui CHF Correlation in the Dominion Energy VIPRE-D Computer Code
ML21105A4332021-04-15015 April 2021
Final Supplemental Response to NRC Genetic Letter 2004-02 on Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accident at Pressurized-Water Reactors
ML21105A4822021-04-15015 April 2021
Response to Request for Additional Information for Proposed License Amendment Request to Revise the Millstone, Unit 2 Technical Specification for Steam Generator Inspection Frequency
ML21081A1362021-03-19019 March 2021
Response to Request for Additional Information for Alternative Request RR-05-06 - Inspection Interval Extension for Steam Generator Pressure-Retaining Welds and Full-Penetration Welded Nozzles
ML20274A3462020-09-30030 September 2020
Response to Request for Additional Information for License Amendment Request to Revise Battery Survillance Requirements
ML20261H5982020-09-17017 September 2020
Response to Request for Additional Information for License Amendment Request to Revise Technical Specification Table 3.3-11, Accident Monitoring Instrumentation
ML20252A1912020-09-0404 September 2020
Response to Request for Additional Information Regarding License Amendment Request for a One-Time Deferral of the Millstone Unit 3 Steam Generator Inspections
ML20209A5362020-07-27027 July 2020
Response to Request for Additional Information Regarding Relief Request IR-3-33 for Limited Coverage Examinations Performed in the Second Period of the Third 10-Year Inspection Interval
ML20079K4242020-03-19019 March 2020
Response to Request for Additional Information for License and Request to Revise TS 3.8.1.1, A.C Sources - Operating, to Support Maintenance and Replacement of the Millstone Unit 3 'A' Reserve Station Service Transformer and 345 Kv South Bu
ML20076C8332020-03-16016 March 2020
Response to Request for Additional Information (E-mail Dated 3/16/2020) Alternative Request IR-4-03 for Use of Alternative Non-Code Methodology
ML20048A0192020-02-11011 February 2020
Response to Request for Additional Information for License Amendment Request to Revise TS 3.8.1.1, A.C. Sources - Operating, to Support Maintenance and Replacement of a Reserve Station Service Transformer and 345 Kv South Bus.
ML20042D9962020-02-10010 February 2020
Response to March 12, Request for Information Enclosure 2, Recommendation 2.1, Flooding Focused Evaluation/Integrated Assessment Submittal
ML19284A3972019-10-0303 October 2019
Response to NRC Request for Additional Information on License Amendment Request to Adopt 10 CFR 50.69
ML19249B7672019-08-29029 August 2019
Enclosure 1 - Millstone, Units 2 and 3 and ISFSI; North Anna, Units 1 and 2 and ISFSI; and Surry, Units 1 and 2 and ISFSI - Response to EAL Scheme Change RAIs
ML19092A3322019-03-27027 March 2019
Response to Request for Additional Information for Proposed Technical Specification Changes for Spent Fuel Pool Storage and New Fuel Storage
ML19011A1112018-12-18018 December 2018
Supplement to the Flooding Hazard Reevaluation Report in Response to 50.54(F) Information Request Regarding Near-Term Task Force Recommendation 2.1: Flooding
ML18340A0282018-11-29029 November 2018
Response to Request for Additional Information for Proposed Technical Specifications Changes for Spent Fuel Pool Storage and New Fuel Storage
ML18302A1202018-10-22022 October 2018
Response to Request for Additional Information for License Amendment Request to Revise the Technical Specification for Control Building Ventilation Inlet Instrumentation
ML18235A3212018-08-17017 August 2018
Response to Request for Additional Information for Proposed Alternative Request P-06 for 'C' Charging Pump
ML18225A0662018-08-0606 August 2018
Response to Request for Additional Information for Alternative Requests Associated with the In-Service Testing Program for Pumps, Valves, and Snubbers Fifth and Fourth 10-Year Interval Updates
ML18205A1762018-07-19019 July 2018
Response to Request for Additional Information for Alternative Requests Associated with the In-Service Testing Program for Pumps, Valves, and Snubbers Fifth and Fourth 10-Year Interval Updates for Units 2 and 3
ML18170A0932018-06-14014 June 2018
Response to Request for Additional Information Regarding License Amendment Request to Revise Integrated Leak Rate Test (Type a) and Type C Test Intervals
ML18151A4672018-05-24024 May 2018
Response to Request for Additional Information Regarding License Amendment Request to Revise Integrated Leak Rate Test (Typed a) and Type C Test Intervals
ML17338A0572017-11-22022 November 2017
Response to Request for Supplemental Information Regarding Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools
ML17300A2202017-10-24024 October 2017
Response to Information Need Request Regarding Mitigating Strategies Assessment (MSA) Report for Flooding
ML17053A1062017-02-16016 February 2017
Response to Request for Additional Information Regarding Proposed Alternative Requests RR-04-24 and IR-3-30 for Elimination of the Reactor Pressure Vessel Threads in Flange Examination
ML17038A0052017-01-31031 January 2017
Response to RAI Regarding End of Cycle 23 and End of Cycle 17 Steam Generator Tube Inspection Reports, CAC MF8507 & MF8506
ML16365A0362016-12-22022 December 2016
Response to March 12, 2012 Information Request High Frequency Sensitive Equipment Functional Confirmation for Recommendation 2.1
ML16365A0322016-12-21021 December 2016
Response to March 12, 2012 Information Request, Spent Fuel Pool Seismic Evaluation for Recommendation 2.1
ML16321A4542016-11-10010 November 2016
Connecticut and Virginia Electric & Power Company Response to Request for Additional Information Revision 22 of Quality Assurance Program Description Topical Report
ML16312A0642016-11-0101 November 2016
Units 1 & 2, Surry, Units 1 & 2, Response to Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools
ML16294A2702016-10-18018 October 2016
Response to Request for Additional Information for License Amendment Request Regarding Realistic Large Break Loss of Coolant Accident Analysis - RAI Questions 1 Through 3
ML16291A5082016-10-12012 October 2016
Response to Follow Up Request to Revise ECCS TS 3/4.5.2 and FSAR Chapter 14 to Remove Charging
ML16202A0402016-07-14014 July 2016
Response to Request for Additional Information Regarding Spent Fuel Pool Heat Load Analysis License Amendment Request
ML16188A1962016-06-30030 June 2016
NRC Regulatory Issue Summary 2016-09 Preparation and Scheduling of Operator Licensing Examinations
ML16182A0372016-06-27027 June 2016
Response to Request for Additional Information for License Amendment Request to Revise ECCS TS 3/4.5.2 and FSAR Chapter 14 to Remove Charging
2024-09-16
[Table view] |
Text
5000 Dominion Boulevard, Glen Allen, VA 23060 Dominion
"' Web Address: www.dom.com
' July 27, 2015 U.S. Nuclear Regulatory Commission Serial No.15-344 Attention: Document Control Desk NLOS/WDC R0 Washington, DC 20555 Docket No. 50-423 License No. NPF-49 DOMINION NUCLEAR CONNECTICUT. INC.
MILLSTONE POWER STATION UNIT 3 RESPONSE TO THIRD REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST FOR IMPLEMENTATION OF WCAP-14333 AND WCAP-15376. REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TEST TIMES AND COMPLETION TIMES (TAC NO. MF41 31)
By letter dated May 8, 2014 and supplemented by a letter dated August 14, 2014, Dominion Nuclear Connecticut, Inc. (DNC) submitted a license amendment request (LAR) for Millstone Power Station Unit 3 (MPS3). The proposed amendment would revise TS 3/4.3.1, "Reactor Trip System Instrumentation," and TS 3/4.3.2, "Engineered Safety Feature Actuation System Instrumentation." These proposed changes are based on Westinghouse Electric Company LLC topical reports WCAP-14333-P-A, Revision 1, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times," and WCAP-1 5376-P-A, Revision 1, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times." In an email dated September 22, 2014, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) to DNC related to the LAR. DNC responded to the RAI on October 15, 2014. In an email dated March 17, 2015, the NRC transmitted a second RAI. DNC responded to the RAI on May 18, 2015.
In an email dated June 25, 2015, the NRC transmitted a third RAI as a follow up to the responses in the May 18, 2015 response letter.
The attachment to this letter provides DNC's response to the NRC's third RAI.
If you have any questions regarding this submittal, please contact Wanda Craft at (804) 273-4687.
Sincerely, wO A\~I' ONhY 1tIJtLIC
! ~Commonwealth of Virgirtia
- MakD ati I Meg. #1 40542 -'
Mark . Satainl My ommisionExpires May 31, 201!8 Vice President - Nuclear Engineering -
COMMONWE.ALTH OF VIRGINIA)
COUNTY OF HENRICO)
The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark 0. Sartain, who is Vice President - Nuclear Engineering of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the
~tements
(~
foregoing document in behalf of that Company, an that the s in the document are true to the best of his knowledge and belief.
Acknowledged before me this 22._"ray of /"kl. 2015.
My Commission Expires: S'- */-- Nota.ry
- Serial No.15-344 Docket No. 50-423 Page 2 of 2 Commitments made in this letter: None
Attachment:
Response to Third Request for Additional Information Regarding License Amendment Request for Implementation of WCAP-14333 and WCAP-15376, Reactor Trip System Instrumentation and Engineered Safety Feature Actuation System Instrumentation Test Times and Completion Times cc: U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 R. V. Guzman Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08-C 2 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director, Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127
Serial No.15-344 Docket No. 50-423 ATTACHMENT RESPONSE TO THIRD REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST FOR IMPLEMENTATION OF WCAP-1 4333 AND WCAP-15376. REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TEST TIMES AND COMPLETION TIMES DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 3
Serial No.15-344 Docket No. 50-423 Attachment, Page 1 of 4 By letter dated May 8, 2014 and supplemented by a letter dated August 8, 2014, Dominion Nuclear Connecticut, Inc. (DNC) submitted a license amendment request
([AR) for Millstone Power Station Unit 3 (MPS3). The proposed amendment would revise TS 3/4.3.1, "Reactor Trip System Instrumentation," and TS 3/4.3.2, "Engineered Safety Feature Actuation System Instrumentation." These proposed changes are based on Westinghouse Electric Company LLC topical reports WCAP-14333-P-A, Revision 1, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times,"
and WCAP-15376-P-A, Revision 1, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times." In an email dated September 22, 2014, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) to DNC related to the LAR. DNC responded to the RAI on October 16, 2014. In an email dated March 17, 2015, the NRC transmitted a second RAI. DNC responded to the RAI on May 18, 2015. In an email dated June 25, 2015, the NRC transmitted a third RAI as a follow up to the responses in the May 18, 2015 response letter. This attachment provides DNC's response to the NRC's RAI.
RAI 1.1 (follow up to PRA RAI 1)
The response to PRA RAI 1 states that: "These statements do not apply to the Tier 2 restrictions listed in Attachment 4 of the LAR," in reference to Tier 2 restrictions on pages 15 and 17 of the LAR. Please explain this statement.
DNC Response The statement was meant to convey that the Tier 2 restrictions listed in Attachment 4 of the [AR apply when a logic train or reactor trip breaker is tested in bypass.
The wording of the Tier 2 restrictions listed in Attachment 4 of the [AR was revised and continues to apply when a logic train or reactor trip breaker are tested in bypass. See the response to RAI 4 in the May 18, 2015 response letter for the revised wording.
RAI 5.1 (follow up to PRA RAI 5)
PRA RAI 5 is based on information in the LAR Section 4.2.1, "WCAP-14333 Tier 2 Restrictions," rather than WCAP-153 76. The LAR states that there are no Tier 2 limitations when a slave relay, master relay, or analog channel is inoperable. This conclusion appears to be based on information provided in Tables Ql 1. and Q18.1I from a letter dated December 20, 1996, transmitting a response to a request for additional information regarding WCAP-14333. Please explain how this Tier 2 conclusion was reached and whether the information in these tables, which includes an assessment of master or slave relay maintenance in Q18. 1, was considered.
Serial No.15-344 Docket No. 50-423 Attachment, Page 2 of 4 DNC Response LAR Section 4.2.1 is a discussion of the Westinghouse evaluation of Tier 2 restrictions.
DNC considered the information from the Westinghouse evaluation as documented in Tables QII.l and Q18.1 when determining Tier 2 restrictions. The Westinghouse evaluation determined the system importances individually for plant configurations with no ongoing test and maintenance activities and for plant configurations with ongoing test or maintenance activities on the analog channels, master relays, slave relays, and logic trains. The importances were compared between the cases with individual components unavailable and all components available. With respect to the analog channels, master relays and slave relays, the Westinghouse evaluation determined the importance rankings among the affected systems did not change. That is the basis for the statement there are no Tier 2 restrictions when a slave relay, master relay or analog channel is inoperable.
RAI 6.1 (follow up to PRA RAI 6)
The/licensee's response to PRA RAI 6 describes the process to determine if Tier 2 or Tier 3 compensatory measures are needed for the LAR proposed changes with respect to fire-related risk. The process, according to the response, incorporates qualitative insights based on fire mitigation strategy as follows:
- Identify components that, when removed from service, render the unit with no core damage mitigation success paths
- Remove those components that are Technical Specification limited (i.e., have an allowed outage time (AOQT) < 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) AND require transitioningto mode 5 The response appears to "or" these conditions together in that for components that meet either criterion, Technical Requirements Manual actions are established when a fire risk significant component is removed from service for greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The second condition related to removing components that are Technical Specification (TS) limited does not consider potential risk significance. Based on the RAI response, it appears some components may not be qualitatively (or quantitatively) considered for Tier 2 or Tier 3 during the proposed TS bypass times or completion times which are
<- 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Therefore, describe an acceptable Tier 2 and Tier 3 process with respect to fire-related risk and the results of the Tier 2 assessment.
DNC Response The Tier 2 and Tier 3 process with respect to fire-related risk and the results of the Tier 2 assessment is provided below. DNC used the guidance provided in NUMARC 93-01, Revision 4A Section 11.3.4.3 when assessing fire risk within the 10 CFR 50.65(a)(4)
Serial No.15-344 Docket No. 50-423 Attachment, Page 3 of 4 process. As a result of the clarification call associated with this RAI, a subset of the criteria used to identify fire risk significant components within (a)(4) was applied to the equipment considered in the LAR. The criteria are:
- 1. Incorporate quantitative PRA insights
- Identify the components corresponding to random failure and test/maintenance basic events with internal events core damage frequency risk achievement worth (CDF RAW) > 2.0.
- Remove those components not listed on the Safe Shutdown Equipment List (SSEL).
- 2. Incorporate qualitative insights based on fire mitigation strategy
- Identify components that, when removed from service, render the unit with no core damage mitigating success paths.
The reactor protection system (RPS) components affected by the proposed amendment are not fire risk significant and thus, any fire risk incurred would be due to the additional equipment removed from service.
Based on the fire risk significant criteria listed above, components only meeting criterion 1 are less risk significant than components meeting criterion 2 since redundant equipment would be available to mitigate fire scenarios when criterion 1-only components are out of service. Furthermore, the risk of these configurations is adequately managed by Technical Requirements Manual (TRM) risk management actions. As a result, criterion 1-only components do not warrant additional Tier 2 or Tier 3 compensatory measures.
Alternatively, criterion 2 components provide the only available core damage mitigation success path for certain fire scenarios and therefore, have high fire risk significance due to the lack of redundancy. Consequently, components meeting criterion 2 are deemed reasonable candidates for developing a Tier 2 or Tier 3 compensatory measure.
The turbine driven AFW pump is the only component that meets criterion 2 and has already been included in the proposed Tier 2 restrictions (i.e., AFW system components will not be removed from service when a reactor trip breaker is inoperable for maintenance). As a result, no additional Tier 2 or Tier 3 compensatory measures are recommended.
RAI 12.1 (follow up to PRA RAI 12)
The LAR Table 1 provides a comparison between WCAP-14333 analysis assumptions and plant-specific parameters. PRA RAI 12 requested an explanation whether the WCAP-14333 Tier I analysis remained bounding for these plant-specific values. The response to PRA RAI 12 states that the plant-specific values for MPS3 shown in Table
Serial No.15-344 Docket No. 50-423 Attachment, Page 4 of 4 I of Attachment 3 of the LAR are consistent with those of Vogtle Electric Generating Plant (Vogtle) as shown on Table 1 on page 297 (Enclosure 5, page E5-2) of WCAP-14333-P-A, Revision 1, Supplement 1, dated September 2003. However, a review of TSTF-418 shows that this reference is dated after the NRC staff's approval letter of TSTF-418 (ADAMS Accession No. ML030920633). Therefore, the referenced document does not appear to have been part of the NRC staff's review of TSTF-4 18, nor does it appear to be provided as a reference supporting the proposed TS changes in the LAR. Pleaseprovide an explanation as requested in PRA RAI 12, consistent with the TSTF-418 traveler which the LAR is requesting to adopt.
DNC Response Given that less testing/maintenance is being performed on some slave relays and the reactor trip breakers, the unavailability of these components will be less than that assumed in the WCAP. Since the MPS3 unavailability values are less than those used in the WCAP, MPS3 is bounded by the WCAP risk analysis.
