ML15034A353

North Anna, Units 1 & 2 - Response to Request for Additional Information Proposed License Amendment Request Permanent Fifteen-Year Type a Test Interval

ML15034A353
Person / Time
Site:	North Anna
Issue date:	01/28/2015
From:	Mark D. Sartain Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
14-595
Download: ML15034A353 (40)

Text

VIRGINIA ELECTRIC AND POWER COMPANYRICHMOND, VIRGINIA 23261January 28, 2015U. S. Nuclear Regulatory Commission Serial No.: 14-595Attention:

Document Control Desk NLOS/ETS:

ROWashington, DC 20555-0001 Docket Nos.: 50-338/339 License Nos.: NPF-4/7VIRGINIA ELECTRIC AND POWER COMPANYNORTH ANNA POWER STATION UNITS 1 AND 2RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION PROPOSED LICENSE AMENDMENT REQUESTPERMANENT FIFTEEN-YEAR TYPE A TEST INTERVALIn a letter dated June 30, 2014 (Serial No.14-272),

Virginia Electric and PowerCompany (Dominion) requested license amendments in the form of changes to theTechnical Specifications, for facility Operating License Numbers NPF-4 and NPF-7 forNorth Anna Power Station Units 1 and 2, respectively.

The proposed amendments revise North Anna Power Station Units 1 and 2 Technical Specification (TS) 5.5.15,"Containment Leakage Rate Testing Program,"

by replacing the reference to Regulatory Guide (RG) 1.163 with a reference to Nuclear Energy Institute (NEI) topical report NEI94-01, Revision 3-A, as the implementation document used to develop the North Annaperformance-based leakage testing program in accordance with Option B of 10 CFR 50,Appendix J. On November 28, 2014, the NRC requested additional information associated with the proposed license amendment request.Attached is the response to the request for additional information (RAI) and revisedmarked-up and proposed (typed) Technical Specification pages, which address theNRC request to include the Conditions and Limitations for NEI 94-01, Revision 2A.In a telephone conference call January 12, 2015 with the NRC staff, the NRC ProjectManager concurred with Dominion providing the response to the RAI by January31, 2015 to permit providing the final leakage results for the North Anna Unit 2 Type Atest.The proposed revision to the amendment request does not affect the significant hazardsconsideration determination submitted with the original license amendment request norresult in any significant increase in the amount of effluents that may be released offsiteor any significant increase in individual or cumulative occupational radiation exposure.

The next Unit 1 ILRT is currently due no later than October 2017. Based on the currentoutage schedule for Unit 1, the current ten-year frequency would require the next Unit 1ILRT to be performed during the fall 2016 refueling outage. Due to lead time required toprocure the services and equipment to perform a Type A test, Dominion requestsapproval of the proposed change by December 31, 2015.

Serial No. 14-595Docket Nos. 50-338/339 Page 2 of 3Should you have any questions or require additional information, please contactMr.Jay Leberstien at (540) 894-2574.

Respectfully, Mark SartainVice President

-Nuclear Engineering Commitment contained in this letter: NoneAttachments:

1. Response to Request for Additional Information

2. Revised Marked-up Technical Specifications Page3. Revised Proposed Technical Specifications PageHikiL.Wll

.I NO'TARY PIlJIC pCommonwealth of Virginia4 R eg.o # 14 0 5 4 2 tCOMMONWEALTH OF VIRGINIA

) My Commission Expires May 31, 208 -COUNTY OF HENRICOThe foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, todayby Mr. Mark D. Sartain, who is Vice President

-Nuclear Engineering, of Virginia Electric and Power Company.

Hehas affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of thatcompany, and that the statements in the document are true to the best of his knowledge and belief.Acknowledged before me this a4 6 day of 2015.My Commission Expires:

0 *Notary Public Serial No. 14-595Docket Nos. 50-338/339 Page 3 of 3cc: U.S. Nuclear Regulatory Commission

-Region IIMarquis One Tower245 Peachtree Center Avenue, NE Suite 1200Atlanta, GA 30303-1257 State Health Commissioner Virginia Department of HealthJames Madison Building

-7th floor109 Governor StreetSuite 730Richmond, VA 23219Mr. J. E. Reasor, Jr.Old Dominion Electric Cooperative Innsbrook Corporate Center4201 Dominion Blvd.Suite 300Glen Allen, Virginia 23060Dr. V. Sreenivas NRC Project Manager North AnnaU.S. Nuclear Regulatory Commission One White Flint NorthMail Stop 08 G-9A11555 Rockville PikeRockville, MD 20852-2738 NRC Senior Resident Inspector North Anna Power Station Serial No. 14-595Docket Nos. 50-338/339 Attachment IResponse to Request for Additional Information Virginia Electric and Power Company(Dominion)

North Anna Station Units I and 2 Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 1 of 32REQUEST FOR ADDITIONAL INFORMATION TECHNICAL SPECIFICATION CHANGE PROPOSING PERMANENT FIFTEEN-YEAR TYPE A TEST INTERVALDOCKET NUMBERS 50-338 AND 50-339Backgqround By letter dated June 30, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14183B318),

Virginia Electric and Power Company(Dominion) requested an amendment to Operating License Numbers NPF-4 and NPF-7in the form of changes to the Technical Specifications (TSs) for North Anna PowerStation Units 1 and 2, respectively.

The license amendment request (LAR) proposes achange to TS 5.5.15, "Containment Leakage Rate Testing Program,"

by replacing thereference to Regulatory Guide (RG) 1.163 with a reference to Nuclear Energy Institute (NEI) topical report NEI 94-01, Revision 3-A, as the implementation document used todevelop the North Anna performance-based leakage testing program in accordance with Option B of 10 CFR 50, Appendix J. In order to complete its safety evaluation theNRC staff has the following comments and requests the following additional information.

In order to meet the due dates, please respond to these RAIs by January 16, 2015.In a telephone conference call January 12, 2015 with the NRC staff, the NRC ProjectManager concurred with Dominion providing the response to the RAI by January 31,2015 to permit providing the final leakage results for the North Anna Unit 2 Type A test.Mechanical and Civil Engineering Branch (EMCB):NRC EMCB RAI-1It is stated in Section 4.4.1 "IWE Examinations" of the LAR that the Interval 2, Period 2of the NAPS Unit 2 Containment IWE in-service inspection is scheduled to becompleted by October 2014. Please discuss the highlight of findings from this recentIWE inspection and any corrective actions taken to disposition the findings.

Dominion ResponseThe IWE Examination performed during the 2014 NAPS Unit 2 refueling outage wascompleted satisfactorily.

The original examination scope was expanded to include anew Examination

Category, E-A Containment

Surfaces, Item No. E1.30 -MoistureBarriers, in response to the NRC Information Notice 2014-07, "Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and ConcreteContainment Metallic Liner". Two deficiencies were noted and Condition ReportsCR558777 and CR558783 were submitted to document these discrepancies.

A missingtest port panel plug (CR558777) used to test the bottom liner welds during initialconstruction was replaced in accordance with work order 59102772082.

The Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 2 of 32replacement was completed after the scheduled Appendix J, Type A Test in order toconfirm liner integrity.

Work order 59102773080 was used to clean and inspect anothertest port panel plug which had unidentified deposits (CR558783).

Follow up generalvisual IWE examinations were completed satisfactorily for these two items.NRC EMCB RAI-2It is stated in the LAR and the current NAPS TS that the next Unit 2 integrated leak ratetest (ILRT), Type A test, is currently due no later than October 9, 2014.Please provide the following:

a. The results of as-found and as-left leak rate for the NAPS Unit 2 ILRT performed inOctober 2014 and a comparison with the corresponding leak rate acceptance limit.b. The results of visual inspection of containment concrete exterior surface areas,completed prior to the NAPS Unit 2 ILRT performed in October 2014.Dominion Responsea. Below are the results of the 2014 NAPS Unit 2 Integrated Leak Rate Test (ILRT).The maximum allowable containment leakage rate is 0.1% of containment air weightper day. The NAPS Unit 2 ILRT results are less than 40% of the Technical Specification limit.Unit 2 2014 Containment Integrated Leak Rate Test As Found As Left Performance (ILRT) ResultsCalculated Leak Rate 0.0242 %wt/day 0.0242 %wt/day 0.0242 %wt/dayUpper Confidence Limit 0.0291 %wt/day 0.0291 %wt/day 0.0291 %wt/dayLeakage Savings 0 %wt/day 0 %wt/day 0 %wt/dayNon-vented Penalities 0.0001 %wt/day 0 %wt/day 0 %wt/dayTotal Results 0.0292 %wt/day 0.0291 %wt/day 0.0291 %wt/dayb. The NAPS Unit 2 concrete examination was completed satisfactorily.

Two smallconcrete spalls were identified adjacent to instrumentation penetration 2-PE-EP-2F1.

The minor spalls were evaluated and accepted in accordance with Dominion andindustry standards (ACI 349.3R-14).

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 3 of 32NRC EMCB RAI-3Please provide a summary of any degradation identified during inspections of the NAPSUnit I and Unit 2 containment liner at the interface of containment floor slab and thecontainment wall liner, including moisture barrier (if any). Please describe thedegradation condition, corrective

actions, and additional monitoring programimplemented to manage the identified condition.

Dominion ResponseNorth Anna Units 1 and 2 do not have moisture barriers located at the interface.

TheNAPS Unit 1 IWE general visual examinations have not identified any degradation atthe floor slab and containment liner interface.

The initial NAPS Unit 2 IWE general visual examinations performed in 1999 identified some areas of minor surface corrosion at the interface.

Four of these areas wereevaluated and documented in Engineering Transmittal (ET)-MAT-99-0003, Containment Liner and Coating Evaluation North Anna Power Station, Unit 2 and ET-CEE-99-0007, Evaluation of Reduced Containment Liner Thickness North Anna Power Station, Unit 2.The apparent cause was due to water spraying during containment decontamination efforts during outages.The four inaccessible areas were excavated and volumetric examinations wereperformed.

The subsequent evaluation determined the corrosion was minimal andminor loss of liner thickness

(.035 in.) was evident.

Repairs, other than recoating, werenot required, but augmented examinations, in accordance with Category E-C(Item E4.12), were performed satisfactorily for the next three periods.

This information is documented in the Dominion letter dated April 3, 2008 (See RAI-5).NRC EMCB RAI-4It is stated in Section 4.0 of the LAR that although not a specific line item in the NorthAnna IWE program, accessible leak chase channel plugs and caps are inspected duringthe general visual examination completed in accordance with IWE program.Relative to the NRC Information Notice 2014-07, "Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and ConcreteContainment Metallic Liner," discuss the NAPS Units I and 2 operating experience, inspection results and any corrective actions taken.Dominion ResponseIn response to NRC Information Notice 2014-07, a new Examination

Category, E-AContainment

Surfaces, Item No. E1.30 -Moisture

Barriers, has been incorporated into Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 4 of 32the North Anna Containment Inservice Inspection IWE Plan and individual UnitImplementation Schedules.

For NAPS Unit 2, the additional general visual examination was completed satisfactorily.

Two deficiencies were noted and Condition ReportsCR558777 and CR558783 were submitted to document these discrepancies.

A missingtest port panel plug (CR558777) used to test the bottom liner welds during initialconstruction was replaced in accordance with work order 59102772082.

Thereplacement was completed after the scheduled Appendix J, Type A Test in order toconfirm liner integrity.

Work order 59102773080 was used to clean and inspect anothertest port panel plug which had unidentified deposits (CR558783).

Follow up generalvisual IWE examinations were completed satisfactory for these two items. The NAPSUnit 1 examination is scheduled for the upcoming 2015 refueling outage. However,most of these components (outside the In-core Instrumentation area) received a generalvisual inspection during the forced NAPS Unit 1 cold shutdown in December 2014. Nodeficiencies were noted.NRC EMCB RAI-5Please provide information relative to findings (if any) and actions taken whereexistence of or potential for degraded conditions in inaccessible areas of theNAPS Units I and 2 containment structure and steel liner were evaluated based onconditions found in accessible areas as required by IOCFR 50.55a(b)(2)(ix)(A) andI OCFR 50. 55a(b) (2) (viii) (E).Please note that in response to the NRC staff request for additional information, insupport of the one-time extension of the NAPS Unit 2 ILRT until 2014, Dominion, inits letter dated April 3, 2008, has already provided information relative to discovery ofa blister in the NAPS Unit 2 containment liner plate protective coating that promptedan examination of the liner plate which revealed the presence of a through-thickness hole. Therefore, information relative to this finding does not need to be resubmitted.

Dominion ResponseThe NAPS Unit 1 and 2 containment structures are in good material condition.

Nosignificant defects or concerns were observed during the last scheduled IWLexaminations.

NAPS Unit 1 IWL examinations were performed in July 2011 and theinspection findings are documented in Engineering Technical Evaluation (ETE)-NA-2011-0051.

The inspection findings can be summarized as follows:Five areas were identified exhibiting efflorescence or staining.

These areas weresounded with a hammer and sound concrete was found with no cracking or voids. Thestains appeared to be originating from abandoned inserts and tie bars and not fromprimary reinforcement.

These areas were evaluated as inconsequential requiring nofurther actions.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 5 of 32One area was identified with cracks measuring up to 0.05" in width on the top of thebarrel to the equipment hatch. These are shallow cracks that are relatively short inlength with no evidence of rust staining that would be indicative of degradation of thereinforcement.

This area was evaluated as inconsequential and requiring no furtheractions.Nineteen areas were identified as needing repair. These areas were characterized aspopouts, small spalls, small holes, rock pockets or embedded steel wire. However,these areas are considered non-structural as they do not extend beyond the face ofprimary reinforcement (4"). As such, the completed repairs did not rise to the level ofCode Repairs but were considered Cosmetic Repairs.One area extended beyond the concrete cover and had exposed primary reinforcement.

This area required an approved Repair/Replacement plan prior to making the ASMESection Xl, Subsection IWL code repair.Based on these inspection

findings, the Unit 1 Containment structure was found to be ingood material condition.

No significant defects or concerns were observed on theexterior concrete and for the most part, observed defects were due to originalconstruction flaws. Taken together or individually, the defects identified do notrepresent a structural concern.

The Containment structure continues to retain its abilityto perform as designed under all load cases including the design basis earthquake andpostulated strike from a tornado generated missile.

Required repairs were completed inaccordance with the work management process.The NAPS Unit 2 IWL examinations were performed in July 2011 and the inspection findings are documented in ETE-NA-2011-0052.

The inspection findings can besummarized as follows:Five areas are considered minor defects (efflorescence or staining, spall and small hole)and have been evaluated as inconsequential requiring no further actions.The remaining twenty (20) items required only cosmetic repairs as all of the excavations were considered non-structural as they do not extend beyond the face of primaryreinforcement (4").Based on these inspection

findings, no Code repairs were required and the Unit 2Containment structure was found to be in good material condition.

No significant defects or concerns were observed on the exterior concrete and observed defects weredue to original construction flaws. Taken together or individually, the defects identified do not represent a significant structural concern.

The Containment structure continues to retain its ability to perform as designed under all load cases including the designbasis earthquake and postulated strike from a tornado generated missile.

Requiredrepairs were completed in accordance with the work management process.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 6 of 32During the NAPS Unit 1 2009 refueling outage, general corrosion was discovered onexposed Recirculation Spray Sump test plug end caps. Additional examinations revealed several degraded grout hole test plates and plugs, pump suction well testconnections, and liner test plugs in the Recirculation Spray Sump, Containment Sump,and the In-core Instrument areas. Similar examination results were found in the nextNAPS Unit 2 refueling outage (2010). Repairs were conducted in accordance withDesign Changes NA-09-0123 (Unit 1) and NA-09-0133 (Unit 2). To prevent furtherdegradation of the Containment floor liner, the exposed carbon steel components of thetest connections were removed from boric acid exposure.

The repair approacheliminated carbon steel surfaces from Containment sumps that are exposed to boricacid, confirmed past and ensures future Containment Liner pressure retaining integrity, repaired any degraded sump pressure test connections and eliminated the potential forcorrosion of the Containment liner either from direct exposure to boric acid originating from sump contents or from indirect exposure from moisture migration from adjacentgrouting.

The discovery of the blister documented in Dominion letter dated April 3, 2008 led toadditional examinations of the liner and basement floor interface.

Four inaccessible areas were excavated, based on evidence of rust, and volumetric examinations wereperformed.

The subsequent evaluation determined the corrosion was minimal andminor loss of liner thickness (0.035 in.) was evident.

Repairs, other than recoating werenot required, but augmented examinations, in accordance with Category E-C(Item E4.12), were performed satisfactory for the next three inservice inspection periods.NRC EMCB RAI-6Section 4.0 of the LAR, Dominion response to limitation/condition 3 of NRC staff safetyevaluation (SE) for NEI 94-01, Revision 2, dated June 25, 2008, states that there are noprimary containment surface areas that require augmented examinations in accordance with American Society of Mechanical Engineers (ASME) Boiler and Pressure VesselCode (Code) Section Xl, IWE-1240.

Section 4.4.1 of the LAR states that for North AnnaUnit 2, Interval 2, Period 1, one area of the liner was observed to have exhibited someblistering and although no liner degradation was observed during the inspection prior torecoating, this area was considered as Category E-C (Item E4. 11) to be reexamined during the next Unit 2 refueling outage. Please provide further information regarding theabove condition and clarify whether there are any NAPS Unit 2 primary containment surface areas that require augmented examinations in accordance with ASMECode Section XI, IWE-1240.

Dominion ResponseAs stated, one area had a relevant condition, and six additional areas were observedwith blisters, and documented in CR373134.

These items were cleaned,

prepped, Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 7 of 32quality inspected, repairs made to the coating, and conservatively added to the NAPSUnit 2 IWE/IWL Implementation Schedule as a Category E-C (Item 4.11) to be re-inspected.

During the last NAPS Unit 2 refueling outage the required IWE detailedvisual, examinations were completed satisfactory.

No other augmented examinations are required.

NRC EMCB RAI-7Please provide the following information:

a. Percent of the total number of Type B tested components that are on 120-month extended performance-based test interval.

b. Percent of the total number of Type C tested components that are on 60-monthextended performance-based test interval.

Dominion Responsea. Type B tested Components NAPS Unit 1: With the exception of one, all electrical penetrations are on theextended test interval

(>99%).NAPS Unit 2: All electrical penetrations are on the extended test interval (100%).The air locks and the fuel transfer tube are tested every refueling outage.b. Type C Tested Components NAPS Unit 1: 92 % of all Type C valves are on the extended test interval.

NAPS Unit 2: 92 % of all Type C valves are on the extended test interval.

Although 92% of Type C valves perform well enough to be on an extended testinterval, only approximately 40% are tested at that frequency.

Because ofscheduled maintenance and testing methods, Type C penetrations are routinely tested more frequently than the 60-month interval.

bSerial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 8 of 32PRA Licensing Branch (APLA):NRC PRA RAI-1In the safety evaluation report for Electric Power Research Institute (EPRI) Technical Report (TR) 1009325, Revision 2, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals,"

the Nuclear Regulatory Commission (NRC) staff, in part,stated that for licensee requests for a permanent extension of the integrated leak ratetesting (ILRT) surveillance interval to 15 years "[clapability category I ofASME RA-Sa-2003 shall be applied as the standard, since approximate values of CDFand LERF and their distribution among release categories are sufficient for use in theEPRI methodology."

Section 4.6.2 of Attachment I to the license amendment request (LAR) states that the2013 Probabilistic Risk Assessment (PRA) full scope peer review found that 92 percentof the supporting requirements (SRs) were met with Capability Category I/Il or greater.Table B. I of Attachment 5 to the LAR provides the list of findings from the 2013 peerreview and provides an assessment of the impact on the ILRT extension application.

a. Provide a list of all SRs from Table B. I of Attachment 5 to the LAR which did notmeet Capability Category I requirements of the PRA Standard endorsed byRevision 2 of Regulatory Guide (RG) 1.200, "An Approach for Determining theTechnical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities."

Explain why not meeting Capability Category I has no impact on theILRT extension application.

b. Fact and observation (F&O) LE-GI-01 in Table B. I of Attachment 5 to the LARappears to indicate that the peer review team had difficulty in completely reviewing the Level 2/LERF analysis.

The finding states: "There is no adequate roadmapthat facilitates peer review of the Level 2/LERF documentation."

The findingfurther states that "[t]here are several dated self-assessment documents.

For LE,about one-third of the SRs do not have any discussion of how the SR is met andwhere the documentation can be found. Moreover, because of the conversion ofthe Volume numbers (e.g. LE.2 to LE. 1), there is additional confusion added forLE. Many of the referenced sections in the self-assessment (e.g., Section 5.4.1 ofLE. 1 (old LE.2)) appear to no longer exist. Finally, unlike the other technical elements that have completely revised the analysis, the Level 2 relies significantly on historical documents including the 20 year old IPE, SM-1243 and SM-1464."

If the LAR provides the summary of the peer review finding LE-GI-01, pleaseprovide the complete peer review feedback for F&O LE-GI-01.

c. For F&Os LE-GI-01, IE-C3-01, AS-B6-01, AS-Cl-01, DA-D8-01, DA-D8-02, andSY-CI-01 the impact assessment provided in the LAR states that there is noimpact on risk because "this is primarily a documentation enhancement."

Explain Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 9 of 32how it was determined that these gaps are limited only to documentation and haveno impact on the risk results, or alternatively explain why not meeting Capability Category I will have no impact on the ILRT extension application.

Dominion Responsea. Each Fact and Observation (F&O) and the associated SR from Table B.1 ofAttachment 5 to the LAR which did not meet Capability Category I requirements ofthe PRA Standard is listed below. The impact on the application is also includedfor each F&O explaining why not meeting Capability Category I has no impact onthe ILRT extension application.

This impact on the application discussed below issame justification provided in Table B.1 of Attachment 5 to the LAR with theexception of the F&Os specifically addressed by Part C of this RAI.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 10 of 32F&O Issue Unmet Impact on Application SRIE-A6-01Discussion:

Common cause and routine system alignments are generally appropriately considered for complicated safetysystem initiating event fault trees. However, for othersystems (notably, electrical systems) there is no discussion or evidence of a review for initiators due to common cause ofelectrical systems nor due to routine system alignments.

GARD NF-AA-PRA-101-204C identifies that transformers, battery chargers, and inverters are candidates for commoncause. These common cause failures are modeled in thecore damage mitigation fault trees. However, these commoncause failures are not considered as initiating events,particularly for RSST 4KV transformers, vital inverters, and125VDC battery chargers.

Also, for example, unavailability of a backup battery charger may drive a plant shutdowngiven loss of the normally operating charger.In addition, could not find a discussion of why common causeblockage of service water travelling screens was notconsidered.

Basis for Significance:

IE-A6 CAT II requires a systematic evaluation of initiating events, including events resulting frommultiple failures resulting from common cause or from routinesystem alignments.

Notebook IE.1 says that due to theindependency of busses, the loss of more than one bus at atime is assessed as negligible frequency, however thisstatement does not consider common cause. No evidence ofa systematic evaluation is evident.IE-A6(From LAR)Common cause initiating events are expected to haverelatively low frequencies, andtheir impact on the CDF andLERF are bounded by an orderof magnitude increase.

Thesensitivity study in Enclosure 1of the LAR demonstrates thatan order of magnitude increasein CDF or LERF does notimpact acceptability of theresults for this application.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 11 of 32F&O Issue Unmet Impact on Application SRIE-C3-01 Discussion:

Many recovery actions are credited in SSIE IE-C3 See response in RAI-1 Part C.fault trees. No discussion or analysis was found to justifythese credits.Basis for Significance:

SR IE-C3 requires justification forcredited recoveries in initiating events. These recoveries arealso used in the post-initiating event mitigation tree.AS-B6-01 Discussion:

No discussion could be identified in the AS AS-B6 See response in RAI-1 Part C.calculation and supporting information with respect to plantconfigurations and maintenance practices creatingdependencies among various system alignments.

Basis for Significance:

System alignments could have animpact on the risk profile if unique plant configurations ormaintenance practices are used.AS-Cl-01 Discussion:

Accident sequence analysis is a key element of AS-Cl See response in RAI-1 Part C.PRA to integrate many other elements of PRA, but accidentsequence notebook needs to improve for further application and update. For instance operator actions are generally described without specific governing procedures and basicevent name modeled in HRA. Observations in AS-C2provide more specific examples.

Observations in AS-Cl-02and AS-C2-01 and 02 provide more specific examples.

Basis for Significance:

This would facilitate emergent riskinformed applications using documents with bettertraceability.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 12 of 32F&O Issue Unmet Impact on Application SRSY-C1-01 Discussion:

The dependency matrix appears to address SY-C1 See response in RAI-1 Part C.dependency for front-line systems and mechanical supportsystems, but appears incomplete for electrical supportsystems.

For example, no dependency is listed for 125VDCpanel 2-BY-B-2-11 or MCC 2-EP-MCC-2A1-2.

In someinstances the support system gate is provided, in otherinstances only the system name is provided.

Basis for Significance:

This issue made it difficult to assessthe completeness of the dependency analysis and made itdifficult to assess the completeness of the identification of thesystems needed to provide or support the safety functions contained in the accident sequence analysis.

HR-G6-01 Discussion:

HR-G6 requires a check of the consistency of HR-G6 (From LAR)the post-initiator HEP quantifications.

The instructions are to A comparison between HFEsreview the HFEs and their final HEPs relative to each other to and their final HEPs for acheck their reasonableness given the scenario

context, plant reasonableness check washistory, procedures, operational practices, and experience, performed prior to release ofHR.2 states that an operator survey, which collects operator the NAPS-R07 model.response times, was performed to meet this requirement.

However, the documentation ofHowever, the surveys do not really check the consistency of the review requiresthe HEP quantifications.

enhancement.

There is little tono impact on CDF or LERF asBasis for Significance:

Confirm that quantifications are this is primarily areasonable.

documentation enhancement.

As a result, this gap has noimpact on the application.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 13 of 32F&O Issue Unmet Impact on Application SRHR-13-01 Discussion:

NAPS HR.1, HR.2, HR.3 section 2.3 and HR.4 HR-13 (From LAR)section 5 addresses assumptions and uncertainties.

Only There is little to no impact onsource of model uncertainty listed is lack of ERO credit which CDF or LERF as this isin reality can be accounted for using the recoveries available primarily a documentation in the HRA calculator.

NUREG/CR-1278 lists sources of enhancement.

As a result, thisuncertainty which could be referenced.

gap has no impact on theapplication.

Basis for Significance:

Need better documentation ofsources of uncertainty.

DA-B2-01 Discussion:

This SR instructs that outliers not be included in DA-B2 (From LAR)the definition of a data group. Looking at the NAPS Data Any change in CDF or LERFcalculation outliers with zero demands were included in resulting from addressing thisgroups with frequently tested components.

F&O is expected to be smalland bounded by an order ofBasis for Significance:

These data events could impact risk magnitude increase.

Theresults.

sensitivity study in Enclosure 1of the LAR demonstrates thatan order of magnitude increasein CDF or LERF does notimpact acceptability of theresults for this application.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 14 of 32F&O Issue Unmet Impact on Application SRDA-C14-01 Discussion:

Coincident maintenance events for intersystem DA-C14 (From LAR)events have not been looked at. Need to evaluate historical Coincident maintenance maymaintenance schedules to detect patterns of typical result in an increase in CDFmaintenance combinations and then add these identified and LERF, but the impact iscoincident maintenance events to the model. expected to be bounded by anorder of magnitude.

TheBasis for Significance:

These events could have an impact sensitivity study in Enclosure 1on the annual risk results.

Some plants have experienced a of the LAR demonstrates thatsignificant impact to their results form including such events an order of magnitude increasein the model. in CDF or LERF does notimpact acceptability of theresults for this application.

DA-D8-01 Discussion:

No discussion of evaluation of the impact of DA-D8 See response in RAI-1 Part C.plant modifications on the data could be found in any of thebelow:-GARD on Data (2061, 2063)-Data Calculation and Supporting Analyses-SY.3 System Notebooks Therefore this SR is considered to be Not MetBasis for Significance:

This item could change the resultsfrom the PRA.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 15 of 32F&O Issue Unmet Impact on Application SRDA-D8-02 Discussion:

No discussion of evaluation of the impact of See response in RAI-1 Part C.plant modifications could be found in any of the below:-GARD on Data (2061, 2063)-Data Calculation and Supporting Analyses-System Notebooks Basis for Significance:

Data could be impacted by a plantmod and effect risk results.IFPP-B3-01 Discussion:

No discussion is given in the various internal IFPP-B3 (From LAR)flooding notebooks with regard to the plant partitioning There is little to no impact onprocess or conclusions as what sources of uncertainty may CDF or LERF as this isbe present or may have been introduced as part of the primarily a documentation partitioning task. Assumptions are given in Section 2.3 of the enhancement.

As a result, thisIF.1B notebook related to flood area definitions, though no gap has no impact on thediscussion of their potential impacts to the analysis are given, application.

Sources of uncertainty related to the flooding initiating eventspipe mode are included in Section 6.0 of the IF.2 notebookand repeated in Section 2.0 of the QU.4 notebook (with noother internal flooding related uncertainties added in thisQU.4 notebook) while Section 5.0 of the IF.3 notebookindicates that sensitivities related to internal flooding arecontained in the QU notebooks, though only sensitivity casesrelated to HEP and CCF values were noted which contained the overall internal flooding events in the sensitivity casemodel quantifications.

Basis for Significance:

The SR was deemed 'not met' thusa finding level is appropriate.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 16 of 32F&O Issue Unmet Impact on Application SRIFSO-A5-01 Discussion:

The capacities of various sources are limited by IFSO-A5 (From LAR)an assumption that all flood isolations could be performed A sensitivity was performed towithin 60 minutes.

No basis is given for this assumption, and assess the impact of floodthe potential of all scenarios using a purely assumptive basis scenarios that screened outfor such inherent screening of potential impacts should also IFSN-A10 based on the 60 minutemodel non-isolated scenarios for the same pipe break timeframe, and the CDF impactsource. Also, the treatment is inconsistent with an IF HFE of those scenarios wasthat is evaluated past 60 minutes.

insignificant.

As a result, thisgap has no impact on theThis F&O applies to the following SRs: IFSO-B1, IFQU-A6, IFSN-A14 application.

IFQU-A5, IFSN-A9, IFSN-A15, IFSN-A16, IFSN-A1O, IFSN-A14, and IFSN-B2.Basis for Significance:

This assumption could have IFSN-A16significant impact to internal floods risk. REC-FLD-IRR hasavailable time of 84 minutes, yet still analyzed for failureprobability.

IFSO-B3-01 Discussion:

There is no uncertainty analysis related to flood IFSO-B3 (From LAR)sources.

There is little to no impact onCDF or LERF as this isBasis for Significance:

Missing uncertainty analysis.

SR primarily a documentation unmet. enhancement.

As a result, thisgap has no impact on theapplication.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Paae 17 of 32F&O Issue Unmet Impact on Application SRIFSN-A5-01 Discussion:

The critical height of all PRA-related SSCs isnot given in an easy to identify single location such as thetable listing of PRA-related SSCs within the various internalflood areas. In addition, the critical height is not alwaysdefined in the other sections of the internal floodingnotebooks such as walkdowns or area scenario discussions, only for the end-state important SSCs.Basis for Significance:

SR requires spatial location of SSCswhich was not consistently done.IFSN-A5(From LAR)There is little to no impact onCDF or LERF as this isprimarily a documentation enhancement.

As a result, thisgap has no impact on theapplication.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Paqe 18 of 32F&O Issue Unmet Impact on Application SRIFQU-A6-01 Discussion:

While the flooding-specific HFEs are developed with detailed assessments, several of the noted items in theSR were not accounted for.Items noted from review of SR IFQU-A6:(b) The impact of the flooding on cues that the control roomuses for a non-flooding HFEs is not discussed in thesupporting spreadsheet of the internal flooding HRAnotebook for internal events HFEs used in the floodinganalysis.

(a) The impact of the flooding on additional workload andstress in the control room uses for a non-flooding HFEs is notdiscussed in the supporting spreadsheet of the internalflooding HRA notebook for internal events HFEs used in theflooding analysis.

In addition, the stress levels for theflooding-specific events were evaluated at low stress levels,which is inconsistent with the intent of the SR.In addition, there appears to be inconsistent timings for theHEPs defined between the HRA calculator inputs and theNOTEBK-PRA-NAPS-IF.2 for time to perform the action(which is usually 1 minute less than the time to damage)being noted in the NOTEBK-PRA-NAPS-IF.2 notebook andthe time to damage being used in the HRA calculator.

Thisslight difference is not expected to cause significant changes,but should be reviewed for consistency and updated asneeded.Basis for Significance:

The SR was deemed 'not met' thusthe level of finding is appropriate.

IFQU-A6(From LAR)There is little to no impact onCDF or LERF as this isprimarily a documentation enhancement.

As a result, thisgap has no impact on theapplication.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Paae 19 of 32F&O Issue Unmet Impact on Application I SR IQU-B5-01Discussion:

Section 3.2 of fleet wide PRA procedure NF-AA-PRA-28 describes a method to break the circular logicappropriately and Table 3 in SY.2 attachment lists circularlogic break gates, but further review of the logic indicates thecircular logic is not handled properly.

A Gate 2-EP-CB-12A-LC "NO ELECTRIC POWER 125 V DCBUS 2-1 (U2 ESGR) (CIRC LOGIC BREAK)" is modeledunder EDG 2H. The 125V DC power supply with circularlogic break is supplied power only from battery under LOOPcondition which is required the EDG. However the batterypower is ANDed with battery charger failures as below:2-EP-CB-12A-PS-LC AND 2-BY-BC-2-I-FAIL 2-BY-BC-2C-l-FAIL 2-BY-B-2-1 Basis for Significance:

Improper breaking of circular logicswould result in improper accident sequence evaluation.

QU-B5(From LAR)There is uncertainty associated with the scope and impactassociated with this modelingissue. However, it is expectedthan the CDF and LERF impactresulting from correcting thecircular logic break modelingwould be bounded by an orderof magnitude increase.

Thesensitivity study in Enclosure 1of the LAR demonstrates thatan order of magnitude increasein CDF or LERF does notimpact acceptability of theresults for this application.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Paqe 20 of 329F&O Issue Unmet Impact on Application

__ __ SRLE-GI-01Discussion:

There is no adequate roadmap that facilitates peer review of the Level 2/LERF documentation.

This isexacerbated by the significant reliance on historical documents going back to the original IPE report.Basis for Significance:

There are several dated self-assessment documents.

For LE, about one-third of the SRsdo not have any discussion of how the SR is met and wherethe documentation can be found. Moreover, because of theconversion of the Volume numbers (e.g. LE.2 to LE.1), thereis additional confusion added for LE. Many of the referenced sections in the self-assessment (e.g., Section 5.4.1 of LE.1(old LE.2)) appear to no longer exist. Finally, unlike the othertechnical elements that have completely revised the analysis, the Level 2 relies significantly on historical documents including the 20 year old IPE, SM-1243 and SM-1464.LE-G1See response in RAI-1 Part C.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 21 of 32b. The LAR did not summarize the peer review finding LE-G1-01.

The peer reviewfeedback for F&O LE-G1-01 is provided below in addition to other instances inwhich the F&O is mentioned in the peer review report.High Level Requirement (HLR) Summary for LERF Analysis (LE)HLR Index HLR Summary of AssessedNumber Capability for PRAHLR-LE-G The documentation of LERF Documentation of LERF analysisanalysis shall be consistent with is consistent with the applicable the applicable supporting supporting requirements.

All SRsrequirements.

are met except LE-G1. OneFinding, F&O LE-G1-01, isassessed based on the difficulty oftracing the documents for ease ofpeer review. Additionally, asuggestion is provided regarding documentation of the capacityassessment for electrical penetration assemblies and othercontainment seals.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 22 of 32Supporti g Requirements Summary for LERF AnalysisIndex No.LE-G Capability Category I Capability Category II Capability Category IIILE-GI DOCUMENT the LERF analysis in a manner that facilitates PRAapplications,

upgrades, and peer review.F&O Assessment:

Not MetLE-G1-01 Basis: Reviewed LE.1, LE.2, and LE.3. The LERF analysis is in a mannerthat facilitates PRA applications and upgrades.

However, thedocumentation does not facilitate peer review and a Finding F&O isassessed.

LE-G5 IDENTIFY limitations in the LERF analysis that would impact applications.

F&O Assessment:

MetLE-G1-01 Basis: Reviewed LE.1, LE.2, LE.3, and Volume QU.4 Model, Assumptions and Uncertainties

Analysis, NOTEBK-PRA-NAPS-QU.4 Revision

0. Table2-1 of QU.4, items 15 through 22, provide discussion of impact onapplications.

However, these discussions should be documented in the LEnotebooks as well, or at least referenced from the LE notebooks.

Hence, alink is made to a previous F&O because of the difficulty in finding suchdiscussion during the peer review.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 23 of 32Facts and Observations No. F&O # Level Other IssueAffectedSRs64 LE-G1-01 Finding LE-G5 Discussion:

There is no adequate roadmapthat facilitates peer review of the Level2/LERF documentation.

This is exacerbated by the significant reliance on historical documents going back to the original IPEreport.Basis for Significance:

There are severaldated self-assessment documents.

For LE,about one-third of the SRs do not have anydiscussion of how the SR is met and wherethe documentation can be found. Moreover, because of the conversion of the Volumenumbers (e.g., LE.2 to LE.1), there isadditional confusion added for LE. Many ofthe referenced sections in the self-assessment (e.g., Section 5.4.1 of LE.1 (oldLE.2)) appear to no longer exist. Finally,unlike the other technical elements that havecompletely revised the analysis, the Level 2relies significantly on historical documents including the 20 year old IPE, SM-1243 andSM-1464.Possible Resolution:

In the LE.1 notebook, provide an SR-by-SR table of how each SR isaddressed and where the documentation canbe found.c. Each of the F&Os is listed below with an explanation of why these issues wereassessed as primarily documentation enhancements.

Since these issues arerelated to documentation and do not impact the model results, they do not impactthe ILRT extension application.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 24 of 32F&O Issue Unmet Impact on Application SRIE-C3-01 Discussion:

Many recovery actions are credited in IE-C3 A review of the recovery actionsSSIE fault trees. No discussion or analysis was found credited in the SSIE fault treesto justify these credits, indicated that the credit taken forthese actions is appropriate forBasis for Significance:

SR IE-C3 requires justification support system initiating eventsfor credited recoveries in initiating events. These based on the assumption, cues, andrecoveries are also used in the post-initiating event procedures.

However, justification formitigation tree. crediting these actions is notdocumented and needs to be addedto the model documentation.

As aresult, this is primarily adocumentation enhancement, andthere is no impact on the acceptability of the results for this application.

AS-B6-01 Discussion:

No discussion could be identified in the AS-B6 A review of the plant configurations AS calculation and supporting information with respect and maintenance practices hasto plant configurations and maintenance practices indicated that redundant equipment iscreating dependencies among various system rotated regularly to evenly wearalignments.

equipment and maximize reliability, and electrical loads are balancedBasis for Significance:

System alignments could have across both trains of power. No plantan impact on the risk profile if unique plant configurations or maintenance configurations or maintenance practices are used. practices were identified that wouldcreate a dependency among systemalignments that would impact themodel results.

As a result, this isprimarily a documentation enhancement, and there is no impacton the acceptability of the results forthis application.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 25 of 32F&O Issue Unmet Impact on Application SRAS-Cl-01 Discussion:

Accident sequence analysis is a key AS-Cl This F&O was assigned based on theelement of PRA to integrate many other elements of traceability of information within thePRA, but accident sequence notebook needs to model documentation.

However, theimprove for further application and update. For instance peer review and subsequent reviewoperator actions are generally described without of this issue by the SME did notspecific governing procedures and basic event name identify any related technical errors.modeled in HRA. Observations in AS-C2 provide more Of the referenced observations, AS-specific examples.

Observations in AS-Cl-02 and AS- C1-02 and AS-C2-02 areC2-01 and 02 provide more specific examples.

Suggestions, and AS-C2-01 is Met forCapability Category I. As a result,Basis for Significance:

This would facilitate emergent this is primarily a documentation risk informed applications using documents with better enhancement, and there is no impacttraceability.

on the acceptability of the results forthis application.

SY-Cl-01 Discussion:

The dependency matrix appears to SY-C1 This F&O was assigned based onaddress dependency for front-line systems and completeness of the dependency mechanical support systems, but appears incomplete matrix with respect to electrical for electrical support systems.

For example, no support systems.

Although thedependency is listed for 125VDC panel 2-BY-B-2-11 or dependency matrix needs to beMCC 2-EP-MCC-2A1-2.

In some instances the support enhanced to include electrical supportsystem gate is provided, in other instances only the systems, these dependencies aresystem name is provided.

captured in the PRA model and notechnical errors have been identified.

Basis for Significance:

This issue made it difficult to As a result, this is primarily aassess the completeness of the dependency analysis documentation enhancement, andand made it difficult to assess the completeness of the there is no impact on the acceptability identification of the systems needed to provide or of the results for this application.

support the safety functions contained in the accidentsequence analysis.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Paae 26 of 32F&O Issue Unmet Impact on Application I SRDA-D8-01Discussion:

No discussion of evaluation of the impactof plant modifications on the data could be found in anyof the below:-GARD on Data (2061, 2063)-Data Calculation and Supporting Analyses-SY.3 System Notebooks Therefore this SR is considered to be Not MetBasis for Significance:

This item could change theresults from the PRA.DA-D8This F&O was assigned based on thelack of guidance for including theimpact of plant modifications on thedata. Although the model updateguidance needs to be updated toprovide these specific instructions, plant modifications are reviewed as apart of the periodic model updateprocess, and their impacts on thedata are taken into consideration.

This is supported by the calculations in the data analysis documentation.

As a result, this is primarily adocumentation enhancement, andthere is no impact on the acceptability of the results for this application.

DA-D8-02Discussion:

No discussion of evaluation of the impactof plant modifications could be found in any of thebelow:-GARD on Data (2061, 2063)-Data Calculation and Supporting Analyses-System Notebooks Basis for Significance:

Data could be impacted by aplant mod and effect risk results.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 27 of 32F&O Issue Unmet Impact on Application SRLE-G1-01Discussion:

There is no adequate roadmap thatfacilitates peer review of the Level 2/LERFdocumentation.

This is exacerbated by the significant reliance on historical documents going back to theoriginal IPE report.Basis for Significance:

There are several dated self-assessment documents.

For LE, about one-third of theSRs do not have any discussion of how the SR is metand where the documentation can be found. Moreover, because of the conversion of the Volume numbers (e.g.LE.2 to LE.1), there is additional confusion added forLE. Many of the referenced sections in the self-assessment (e.g., Section 5.4.1 of LE.1 (old LE.2))appear to no longer exist. Finally, unlike the othertechnical elements that have completely revised theanalysis, the Level 2 relies significantly on historical documents including the 20 year old IPE, SM-1243 andSM-1464.LE-G1This F&O was assigned primarily dueto the lack of an adequate roadmapto facilitate the peer review byproviding discussion of how each SRwas met and the location ofsupporting documentation.

Inaddition, changes in the LE documentstructure have not been reflected inthe self-assessment documentreferences.

Although thesedocumentation issues need to beaddressed, no technical errors wereidentified that would impact the modelresults, and all other LE SRs wereconsidered met. Reliance on the IPEand other historical documents isacceptable since the information contained in these documents is stillrelevant.

As a result, this is primarily a documentation enhancement, andthere is no impact on the acceptability of the results for this application.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 28 of 32NRC PRA RAI-2Section 4.4 of the LAR uses the Calvert Cliffs Nuclear Power Plant methodology inevaluating the impact of liner corrosion on the extension of ILRT testing intervals.

Thisassessment was based on two observed corrosion events at North Anna Power Station,Unit 2 and Brunswick Steam Electric Plant, Unit 2. Additionally, the LAR references adata search performed by Peach Bottom Atomic Power Station in 2010, reviewing morerecent liner corrosion events. If there have been additional instances of liner corrosion that could be relevant to this assessment, provide an updated list of observed corrosion events relevant to North Anna containment, and an evaluation of the impact on riskresults when all relevant corrosion events are included in the risk assessment.

Dominion ResponseTwo corrosion events occurring over a 5.5-year period starting in September 1996 wereused to estimate the liner flaw probability in the Calvert Cliffs analysis.

Peach BottomAtomic Power Station documented a data search in 2010 which identified two linercorrosion events with through-wall holes occurring since the original data period.Dominion performed a data search to determine if more recent instances of linercorrosion with through-wall holes have occurred that could be relevant to thisassessment, and two were identified.

In October 2010, an area 4" by 32" was found tobe significantly

degraded, including through-wall damage, in the Turkey Point 3containment liner. In October 2013, a 0.40" by 0.28" through-wall hole was identified inthe Beaver Valley 1 containment liner. The liner flaw probability with six eventsoccurring over an 18-year period based on the inclusion of all relevant corrosion eventsis adequately represented by the two events occurring in the 5.5-year period of theCalvert Cliffs analysis.

As a result, the inclusion of all relevant corrosion events has noimpact on the steel liner corrosion analysis and the results of the risk assessment.

NRC PRA RAI-3Section 4.2.6 of EPRI TR-1009325, Revision 2-A states that "[pilants that rely oncontainment overpressure for net positive suction head (NPSH) for emergency corecooling system (ECCS) injection for certain accident sequences may experience anincrease in CDF", therefore requiring a risk assessment.

The fourth condition in thesafety evaluation report for EPRI TR-1009325, Revision 2 states that "[a] LAR isrequired in instances where containment over-pressure is relied upon for ECCSperformance."

Section 5.8 of Attachment 4 to the LAR states that the design basiscalculations credit containment overpressure to satisfy net positive suction head(NPSH) requirements for recirculation spray (RS) and low head safety injection (LHSI)pumps during loss of coolant accidents.

The MAAP analyses discussed in Attachment 4 of the LAR are intended to demonstrate that adequate NPSH is available assuming anincreased containment leak rate.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 29 of 32a. The MAAP calculations performed in the LAR analyzed the following break sizeLOCAs: 1 inch, 2 inch, 4 inch, 6 inch and 31 inch. Describe the basis for theselection of the break sizes analyzed and discuss how they compare to thedesign basis calculations crediting containment overpressure.

b. Discuss any key assumptions in the MAAP analysis that may be non-conservative and impacting the loss of NPSH assessment.

c. Attachment A, "MAAP Analyses,"

included in Attachment 4 of the LAR states thatthe presented MAAP results "did not include any loss of NPSH. This is welldemonstrated by Figures D- I through D-5." Explain how these figures indicatethat the available NPSH is sufficient for operation of the RS and LHSI pumps.Dominion Responsea. The break sizes were selected to cover the wide range of LOCAs including Small, Medium and Large Break LOCAs and demonstrate that there is alwayssufficient NPSH to the pumps regardless of the break size/LOCA type. UnlikePRA, the design basis analysis only models Large Break LOCA with double-ended rupture of hot leg (corresponds to 31" break MAAP case) for NPSHcalculation by providing justification that it is the bounding scenario.

b. No non-conservative assumptions were made.c. Investigation of the summary output files for all cases identified no indication ofloss of NPSH for any pumps, nor did it show any signs of core uncovery anddamage. The sump water level plots were provided to demonstrate theavailability and stability of the water level in the sump with the underlying conclusion that no core uncovery happens.Containment and Ventilation Branch (SCVB):NRC SCVB RAI-1Referring to Attachment I of letter dated June 30, 2014, Section 2.0, "Proposed Change",

states:North Anna TS 5.5.15 currently states: "A program shall establish the leakage ratetesting of the containment as required by 10 CFR 50.54(o) and 10 CFR 50,Appendix J, Option B, as modified by approved exemptions.

This program shall be inaccordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program,"

dated September 1995, modified by thefollowing exception:

NEI-94-01-1995, Section 9.2.3: The first Unit 2 Type A test performed after theOctober 9, 1999 Type A test shall be performed no later than October 9, 2014."

Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 30 of 32The licensee proposes to revise TS 5.5.15 as follows:"A program shall establish the leakage rate testing of the containment as required by10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approvedexemptions.

This program shall be in accordance with the guidelines contained inNEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July 2012."The proposed TS 5.5.15 language refers to Revision 3-A of NEI 94-01 which does notcontain the limitations and conditions for the extension of the Type A testing that arerequired in NEI 94-01 Revision 2-A. The current NRC staff position is as follows:* Licensees that plan to extend Type A test interval up to 15 years and do not plan toextend Type C test intervals beyond 60 months should reference NEI 94-01,Revision 2-A.* Licensees that plan to extend Type A test interval up to 15 years and/or acquire theoption to plan to extend Type C test intervals beyond 60 months up to 75 monthsshould reference NEI 94-01, Revision 3-A as well as the limitations and conditions required in the NRC staff Safety Evaluation Report for NEI 94-01 Revision 2-A.OR" Licensee that does not prefer to reference NEI 94-01 Revision 2-A in TS 5.5.15 shallinclude the following requirements in the TS 5.5.15:1. For calculating the Type A leakage rate, the licensee should use the definition inthe NEI TR 94-01, Revision 2, in lieu of that in ANSI/ANS-56.8-2002.

(Refer toSection 3.1.1.1 of NRC Safety Evaluation Report for NEI 94-01 Revision 2)2. The licensee submits a schedule of containment inspections to be performed prior to and between Type A tests. (Refer to Section 3.1.1.1 of NRC SafetyEvaluation Report for NEI 94-01 Revision 2)3. The licensee addresses the areas of the containment structure, potentially subjected to degradation.

(Refer to Section 3.1.1.1 of NRC Safety Evaluation Report for NEI 94-01 Revision 2)4. The licensee addresses any tests and inspections performed following majormodifications to the containment structure, as applicable.

(Refer to Section 3.1.4of NRC Safety Evaluation Report for NEI 94-01 Revision 2)5. The normal Type A test interval should be less than 15 years. If a licensee has toutilize the provision of Section 9.1 of NEI TR 94-01, Revision 2, related to Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 31 of 32extending the ILRT interval beyond 15 years, the licensee must demonstrate tothe NRC staff that it is an unforeseen emergent condition.

(Refer to Section3.1.1.2 of NRC Safety Evaluation Report for NEI 94-01 Revision 2)Explain and/or revise the license language to reflect the above for further review.Dominion ResponseSection 5.5.15 will be revised to include a discussion of the Limitations and Conditions contained in Section 4.1 of NEI TR 94-01, Revision 2 of the NRC Safety Evaluation Report in NEI 94-01 Revision 2A, dated October 2008. The revised TS page isattached in Attachment 2 to this letter.NRC SCVB RAI-2Refer to Attachment 1, Section 2.0, page 2 of 20, in the second from last line the words"currently states" appears to be an editorial error. Please provide furtherclarification/explanation.

Dominion ResponseThe "currently states" in the paragraph above is a typographical error. The paragraph below corrects the typographical error and includes the condition and limitations fromNEI 94-01 Revision 2A.TS 5.5.15, "Containment Leakage Rate Testing Program,":

"A program shallestablish the leakage rate testing of the containment as required by 10 CFR50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approvedexemptions.

This program shall be in accordance with the guidelines contained in NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July2012 and Section 4.1, "Limitations and Conditions for NEI TR 94-01,Revision 2" of the NRC Safety Evaluation Report in NEI 94-01 Revision 2A,dated October 2008.

4Serial No. 14-595Docket Nos. 50-338/339 Attachment 1Page 32 of 32References

[1] Industry Guideline for Implementing Performance-Based Option of 10 CFR Part50, Appendix J, NEI 94-01 Revision 2-A, October 2008.[2] Nuclear Energy Institute, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, NEI 94-01 Revision 3-A, July2012.[3] Letter from U.S. Nuclear Regulatory Commission to Biff Bradley (Director, Nuclear Energy Institute),

REQUEST REVISION TO TOPICAL REPORT NEI 94-01, REVISION 3-A, "lndustry Guideline for Implementing Performance-Based Option of 10 CFR PART 50, Appendix J", Accession Number ML13192A394, August 20, 2013.[4] Letter from Dominion to U.S. Nuclear Regulatory Commission, Virginia ElectricPower Company North Anna Power Station Units 1 and 2 Proposed LicenseAmendment Request Permanent Fifteen-Year Type A Test Interval, June 30,2014.

Serial No. 14-595Docket Nos. 50-338/339 Attachment 2Revised Marked-up Technical Specification PageVirginia Electric and Power Company(Dominion)

North Anna Station Units I and 2 Programs and Manuals5.55.5 Programs and Manuals5.5.14 Safety Function Determination Program (SFDP) (continued) analysis cannot be performed.

For the purpose of this program, aloss of safety function may exist when a support system isinoperable, and:a. A required system redundant to the system(s) supported by theinoperable support system is also inoperable; orb. A required system redundant to the system(s) in turn supported bythe inoperable supported system is also inoperable; orc. A required system redundant to the support system(s) for thesupported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If aloss of safety function is determined to exist by this program, theappropriate Conditions and Required Actions of the LCO in which theloss of safety function exists are required to be entered.

When aloss of safety function is caused by the inoperability of a singleTechnical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.5.5.15 Containment Leakage Rate Testing Programa. A program shall establish the leakage rate testing of thecontainment as required by 10 CFR 50.54(o) and 10 CFR 50,Appendix J, Option B, as modified by approved exemptions.

Thisprogram shall be in accordance with the guidelines contained inProgram,"

dated Septerbr19 as mno di-ýc -1b. The calculated peak containment iternal p ssure for the designdesign pressure is 45 psig.c. The maximum allowable containme tleakage rate, La, at Pa shallNEI 9dENE 9401 Reiso 3A"nust Guidel.in

.o Impleme ntn fPeF 111orý Mance-asd Option efeme of J,"dtoued uy21 ndScin41 Lmttosb.n CodThen calulte peak TR9-1,Rvsontainmenth NRC SaeyEaution deposign__NEI 94-01, Revision 2-A,"datd Octobry Guidlin fo mlmnigPefrac-North Anna Units 1 and 25.5-15Amendments 269ýý Serial No. 14-595Docket Nos. 50-338/339 Attachment 3Revised Proposed Technical Specification PageVirginia Electric and Power Company(Dominion)

North Anna Station Units 1 and 2 Programs and Manuals5.55.5 Programs and Manuals5.5.14 Safety Function Determination Program (SFDP) (continued) analysis cannot be performed.

For the purpose of this program, aloss of safety function may exist when a support system isinoperable, and:a. A required system redundant to the system(s) supported by theinoperable support system is also inoperable; orb. A required system redundant to the system(s) in turn supported bythe inoperable supported system is also inoperable; orc. A required system redundant to the support system(s) for thesupported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If aloss of safety function is determined to exist by this program, theappropriate Conditions and Required Actions of the LCO in which theloss of safety function exists are required to be entered.

When aloss of safety function is caused by the inoperability of a singleTechnical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.5.5.15 Containment Leakage Rate Testing Programa. A program shall establish the leakage rate testing of thecontainment as required by 10 CFR 50.54(o) and 10 CFR 50,Appendix J, Option B, as modified by approved exemptions.

Thisprogram shall be in accordance with the guidelines contained inNEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," datedJuly 2012 and Section 4.1 "Limitations and Conditions for NEITR 94-01, Revision 2" of the NRC Safety Evaluation Report in NEI94-01, Revision 2A, dated October 2008.b. The calculated peak containment internal pressure for the designbasis loss of coolant accident, Pas is 42.7 psig. The containment design pressure is 45 psig.c. The maximum allowable containment leakage rate, La, at Pas shallbe 0.1% of containment air weight per day.(continued)

North Anna Units I and 25.5-15Amendments