NL-14-0580, Pilot 10 CFR 50.69 License Amendment Request - Response to Request for Additional Information

Pilot 10 CFR 50.69 License Amendment Request - Response to Request for Additional Information

ML14122A364
Person / Time
Site:	Vogtle
Issue date:	05/02/2014
From:	Pierce C Southern Co, Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-14-0580
Download: ML14122A364 (50)

Text

Charles R. Pierce Southern Nuclear Regulatory Affairs Director Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Bim1ingham, AL 35242 SOUIHEIRN COMPANY May 2, 2014 Docket Nos.: 50-424 NL-14-0580 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant - Unit 1 and Unit 2 Pilot 10 CFR 50.69 License Amendment Request Response to Request for Additional Information Ladies and Gentlemen:

By letter dated August 31, 2012, Southern Nuclear Operating Company (SNC}

requested amendments to the Vogtle Electric Generating Plant (VEGP} Units 1 and 2 (TAC NOS ME9472 and ME9473}. The proposed amendments would revise the VEGP licensing basis to implement 10 CFR 50.69, risk informed categorization and treatment of structures, systems, and components for nuclear power plants.

By letter dated April 17, 2013, the NRC requested additional information. SNC responded to the request by letter dated May 17, 2013 and noted responses to requests for additional information (RAis} 19, 25, 26 and 27 would require additional time to develop and would be provided within 120 days from the date of the SNC letter. SNC provided the responses to RAis 19 and 27 by letter dated July 2, 2013. In the response to RA127, SNC identified that a Base-Case Sensitivity (BCS} model would be developed and used for the categorization process and to respond to RAls that request sensitivity analyses on NRC approved methods including RAis 25 and 26.

By letter dated September 13, 2013, SNC provided the response to RAis 25 and 26 and provided a summary of the BCS model results. By letter dated April 3, 2014, the NRC requested additional information. The enclosure to this letter provides responses to these RAis.

This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205} 992-7369.

U. S. Nuclear Regulatory Commission NL-14-0580 Page2 Mr. C. R. Pierce states he is Regulatory Affairs Director of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and, to the best of his knowledge and belief, the facts set forth in this letter are true.

Respectfully submitted, - . "--.

Cfi~

C. R. Pierce Regulatory Affairs Director Sworn to and subscribed before me this CLNo#Jlfr My commission expires: __ /-_2._-_Z._O_/ _.g_

Enclosure:

Response to Request for Additional Information

Attachment:

Revised Section 3.3.1.4 cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. T. E. Tynan, Vice President - Vogtle 1 & 2 Mr. B. L. lvey, Vice President- Regulatory Affairs Mr. D. R. Madison, Vice President- Fleet Operations Mr. B. J. Adams, Vice President - Engineering RType: CVC7000 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. R. E. Martin, NRR Senior Project Manager- Vogtle 1 & 2 Mr. L. M. Cain, Senior Resident Inspector- Vogtle 1 & 2 State of Georgia Mr. J. H. Turner, Environmental Director Protection Division

Vogtle Electric Generating Plant Request to Revise the Licensing Basis to Implement 10 CFR 50.69 Response to Request for Additional Information Regarding Pilot 10 CFR 50.69 License Amendment Request Enclosure Response to Request for Additional Information

Enclosure to NL-14*0580 Response to Request for Additional Information By letter dated August 31, 2012, and supplemental letters dated May 17, July 2, and September 13, 2013 (Agencywide Documents Access and Management System Accession Nos. ML12248A035, ML13137A480, ML13184A267 and ML13256A306, respectively), Southern Nuclear Operating Company, Inc. (SNC) submitted a license amendment request (LAR) to implement Title 10 of the Code of Federal Regulations, Section 50.69, ,"Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors." The latter three letters were in response to a request for additional information (RAI) dated April17, 2013 (ML13102A233) which contained 27 RAis. The numbering of the RAis below is consistent with that of the April 17, 2013, RAI, letter.

NRCRAI3 In response to RAI 3, SNC clarified that the May 2009 peer review was conducted against the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) Standard RA-Sb-2005 that was endorsed in Revision 1 of Regulatory Guide (RG) 1.200. Revision 2 of RG 1.200 endorses ASMEIANS RA-Sa-2009, a later version of the standard. Table 6 of the LAR identifies differences between Revisions 1 and 2 of RG 1.200, and relates those differences to the VEGP Probabilistic Risk Assessment (PRA). SNC intends to demonstrate compliance with Revision 2 of RG 1.200. However, some of the items listed in Table 6 of the LAR appear inconsistent with Section 3.3 of Nuclear Energy Institute (NEI) 05-04 "Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard", Revision 3, November 2009. NEI 05-041ists a number of supporting requirements that require or may require re-evaluation for updating PRAs to Revision 2 of RG 1.200. The majority of the supporting requirements listed in NEI 05-04 (for example, HR-G3, QU-A3, QU-85, QU-E3, IFPP-81) are not found in Table 6 of the LAR. Many of the items listed in Table 6 are not found in NEI 05-04. Please clarify these discrepancies and the basis for assuming compliance with Revision 2 of RG 1.200 for the Internal Events (including Internal Flooding) PRA.

SNC Response Table 6 in Section 3.3.1.4 of the LAR identifies the differences between Revisions 1 and 2 of RG 1.200 and relates those differences to the Plant Vogtle Probabilistic Risk Assessment (PRA). Table 6 has been revised to make it consistent with Section 3.3 of Nuclear Energy Institute (NEI) 05-04 "Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard", Revision 3, November 2009. The Attachment to this letter NL-14-0580 provides the revised Section 3.3.1.4 in its entirety.

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Enclosure to NL-14-0580 Response to Request for Additional Information NRC RAis 7.8 SNC's response to RAI #8 appears to indicate that the pumps and check valves of the containment spray (CS) system were categorized as low-safety-significant (LLS). In RAI

1. 7 the staff asked SNC to provide more details on how it answered questions 4 and 5 in Section 9.2.2 of NEI 00-04, 10 CFR 50.69 SSC [Structure, System and Component]

Categorization Guideline, Revision 0, July 2005, regarding the determination as to whether a function/SSe provides the "sole means" of accomplishing a specific mitigation function. The response to RAJ # 7 listed only one CS system function; the function related to containment pressure indication. Were other CS system functions considered?

Please clarify the categorization process as it was applied to the CS system pumps and check valves and the outcome of the licensee's trial categorization for these components.

SNC Response Other containment spray (CS) functions were also considered. The following information provides details on the categorization process as it was applied to the CS Pumps and Discharge Check Valves during the trial categorization. The process is described in the SNC procedures referenced herein.

A. The subject components were categorized as LSS by the Risk Hazards Analysis, in accordance with SNC Procedure NMP- ES-065-001, 10 CFR 50.69 Active Component Risk Significance Insights. The results are summarized below:

LSS LSS LSS LSS LSS B. The subject components were identified as supporting the following system functions:

EE 11206P6002 CS PUMP-TRAIN MV 11206U6015 [CHECK VALVE]

MV 11206U6016 [CHECK VALVE]

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Enclosure to NL-14-0580 Response to Request for Additional Information C. The above supported functions are described below along with the associated qualitative assessment. This assessment was performed in accordance with SNC Procedure NMP-ES-065-003, 10 CFR 50.69 Risk Informed Categorization for Systems, Structures, and Components.

1.1 SUPPLY BORATED LSS INIT No Failure of Function does not directly cause an initiating event, but rather WATER FROM THE is required to mitigate the effects of an initiating event which has already RWST TO THE CS occurred, i.e. a Loss of Coolant or a Main Steam Line Break.

RING HEADER RCPB No Failure of Function does not cause a loss of reactor coolant pressure DURING THE SHORT- boundary integrity. Limiting Containment pressure has no effect on the TERM INJECTION RCS barrier.

PHASE FBSF No The basic safety function of interest in this case is Containment Integrity.

SUBSEQUENT TO A Although failure of the containment spray function could allow LOCA OR STEAM containment pressure to rise above design limits, the ultimate BREAK TO LIMIT AND containment pressure capability is well above design limits. VEGP REDUCE employs a large, dry containment design. The VEGP Individual Plant CONTAINMENT Examination (IPE) for severe accident vulnerabilities evaluated the BUILDING containment response to such accidents. The IPE Report noted the PRESSURE AND following:

TEMPERATURE.

• The VEGP design includes two containment heat removal systems:

eight containment cooling units (CCUs) and two containment spray trains. However, only the CCU Sys. is designed to remove decay heat (long term) from the containment for design basis events.

Containment Spray can only be considered as a short term containment pressure reduction system.

• Only the CCUs need to be addressed for containment equip,ment survivability

• The VEGP overpressure evaluation concluded that the ultimate containment failure bound) is 127 psia.

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Enclosure to NL-14-0580 Response to Request for Additional Information Severe Accident Management Guideline (SAMG) Calculation X6CNA.11.8, page 141, identifies the containment pressure setpoint at which SGG-2, Depressurize Containment, should be entered. The SAMG calculation utilizes the data from the VEGP IPE study and conservatively sets the setpoint at 102 psig. Thus, it is clear that the ultimate containment pressure capability is well above the design limit.

In addition, a consequence analysis was performed to calculate the maximum post-accident containment pressure without containment spray (Ref. AMEC Report S0006/RP/001 R01, dated November 22, 2011 ). This analysis evaluated the following accident sequences: 1) double-ended Pump Suction LOCA with and without recirculation; and

2) double-ended Main Steam Line Break to one Steam Generator. The peak containment pressures for both LOCA and MSLB sequences were calculated to be below 50 psia, which is significantly lower than the 5%

containment failure value with margin of 116.7 psia [from the above SAMG calculation]. Thus, it can be concluded that loss of this function would not impact containment integrity. With regard to survivability of the CCUs, specifically the impact of increased containment pressure on the cooler tubes, a qualitative evaluation was performed which showed that the allowable external pressure on the tubes inside the cooler is nnrrwin"'<>tatu 545 osia. Thus, there is no adverse imoact on the CCUs.

EOPA No The EOP's have the operator verify or manually actuate containment spray when containment pressure exceeds the actuation setpoint.

However, Containment Spray is not the sole means for the successful performance of operator actions to mitigate the accident since the operation of the containment coolers (by EOP direction) also acts to reduce oressure and temoerature in the containment.

EOPC No The EOP's have the operator verify or manually actuate containment spray when containment pressure exceeds the actuation setpoint.

However, Containment Spray is not the sole means of achieving actions for assurina lana term containment intearitv since the ooeration of the E-4

Enclosure to NL-14-0580 Response to Request for Additional Information containment coolers (by EOP direction) also acts to reduce pressure in the containment. Although failure of the containment spray function could allow containment pressure to rise above design limits, the ultimate containment pressure capability is well above design limits (Reference Severe Accident Management Guideline Calculation X6CNA.11.3, page 141 ). The availability of the containment cooling units would ensure that, even with total loss of containment spray, containment integrity would be maintained. Refer to the FBSF basis for additional details.

SDMC No CS functions are required only for accident mitigation and are not applicable to mode changes or during shutdown conditions.

ORPA No The barriers to fission product release are fuel cladding, reactor coolant pressure boundary, and containment integrity. This function does not act as a barrier to fission product release during plant operation or during severe accidents.

1.2 RECIRCULATE LSS INIT No Failure of Function does not directly cause an initiating event, but rather BORATED WATER is required to mitigate the effects of an initiating event which has already WITH A TRISODIUM occurred, i.e. a Loss of Primary Coolant or a Main Steam Line Break.

PHOSPHATE (TSP) RCPB No Failure of Function does not cause a loss of reactor coolant pressure SOLUTION FROM boundary integrity. Limiting Containment pressure has no effect on the THE CONTAINMENT RCS barrier.

SUMP TO THE CS FBSF No The basic safety function of interest in this case is Containment Integrity.

RING HEADER Loss of this function would not impact containment integrity. Refer to the DURING THE LONG-basis for function 1.1 for details.

TERM RECIRCULATION EOPA No The EOP's have the operator swap CS suction to the containment PHASE sumps when the RWST empties. However, Containment Spray is not SUBSEQUENT TO A the sole means for the successful performance of operator actions to LOCA OR STEAM mitigate the accident since the operation of the containment coolers (by EOP direction) also acts to reduce pressure and temperature in the E-5

Enclosure to NL-14-0580 Response to Request for Additional Information BREAK TO LIMIT AND containment.

REDUCE EOPC No The EOP's have the operator swap CS suction to the containment CONTAINMENT sumps when the RWST empties. However, Containment Spray is not BUILDING the sole means of achieving actions for assuring long term containment PRESSURE AND integrity since the operation of the containment coolers (by EOP TEMPERATURE. direction) also acts to reduce pressure in the containment. Although failure of the containment spray function could allow containment pressure to rise above design limits, the ultimate containment pressure capability is well above design limits. Refer to the FBSF basis for function 1.1 for additional details.

SDMC No CS functions are required only for accident mitigation and are not to mode ordu shutdown conditions.

ORPA No The barriers to fission product release are fuel cladding, reactor coolant pressure boundary, and containment integrity. This function does not act as a barrier to fission product release during plant operation or during severe accidents.

ORPA No The barriers to fission product release are fuel cladding, reactor coolant pressure boundary, and containment integrity. This function does not act as a barrier to fission product release during plant operation or during severe accidents.

2.1 SUPPLY BORATED LSS INIT No Failure of Function does not directly cause an initiating event, but rather WATER FROM THE is required to mitigate the effects of an initiating event which has already RWST TO THE CS occurred, i.e. a Loss of Prim Coolant or a Main Steam Line Break.

RING HEADER RCPB No Failure of Function does not cause a loss of reactor coolant pressure DURING THE SHORT- boundary integrity. Reducing Iodine in containment mitigates the effects TERM INJECTION of a LOCA which has already occurred, but has no effect on the RCS PHASE barrier.

SUBSEQUENT TO A FBSF No Failure to reduce Iodine in the containment would lead to high LOCA OR STEAM containment radiation, but would not compromise containment integrity.

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Enclosure to NL-14-0580 Response to Request for Additional Information BREAK IN ORDER TO The other basic functions are not affected.

REDUCE THE EOPA No No EOP actions applicable to containment spray specifically address QUANTITY OF this function. Although high containment radiation is addressed in the IODINE AND FRP's, the guidance does not include this containment spray function.

PARTICULATE EOPC No No EOP actions applicable to containment spray specifically address FISSION PRODUCTS this function. Although high containment radiation is addressed in the IN THE FRP's, the does not include this containment function.

CONTAINMENT.

SDMC No CS functions are required only for accident mitigation and are not applicable to mode changes or during shutdown conditions.

ORPA No The barriers to fission product release are fuel cladding, reactor coolant pressure boundary, and containment integrity. This function does not act as a barrier to fission product release during plant operation or during severe accidents.

2.2 RECIRCULATE LSS INIT No Failure of Function does not directly cause an initiating event, but rather BORATED WATER is required to mitigate the effects of an initiating event which has already WITH A TRISODIUM occurred, i.e. a Loss of Prim Coolant or a Main Steam Line Break.

PHOSPHATE (TSP) RCPB No Failure of Function does not cause a loss of reactor coolant pressure SOLUTION FROM boundary integrity. Reducing Iodine in containment mitigates the effects CONTAINMENT of a LOCA, but has no effect on the RCS barrier.

SUMP TO CS RING FBSF No Failure to reduce Iodine in the containment would lead to high HEADER DURING containment radiation, but would not compromise containment integrity.

LONG-TERM RECIRC The other basic safety functions are not affected.

PHASE SUBSEQUENT TO A EOPA No No EOP actions applicable to containment spray specifically address LOCA OR STEAM this function. Although high containment radiation is addressed in the BREAK IN ORDER TO FRP's, the guidance does not include this containment function.

REDUCE QUANTITY EOPC No No EOP actions applicable to containment spray specifically address OF IODINE AND this function. Although high containment radiation is addressed in the PARTICULATE FRP's, the does not include this containment function.

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Enclosure to NL-14-0580 Response to Request for Additional Information FISSION PRODUCTS CS functions are required only for accident mitigation and are not IN CONTAINMENT. ao1:>11c:ao1e to mode or shutdown conditions.

ORPA No The barriers to fission product release are fuel cladding, reactor coolant pressure boundary, and containment integrity. This function does not act as a barrier to fission product release during plant operation or during severe accidents.

5.1 PROVIDE LSS INIT No Failure of Function does not directly cause an initiating event, but rather CONTAINMENT is required to mitigate the effects of an initiating event which has already ISOLATION. occurred, i.e. a Loss of Coolant or a Main Steam Line Break.

RCPB No Failure of Function would not affect the integrity of the RCS pressure FBSF No Although the containment spray system is open to the containment atmosphere on one side, it is a closed, filled system on the other side.

Therefore, failure of this function, Containment Isolation, would not result in a breach of the containment barrier basic safety function. No other basic function is affected by failure of containment isolation.

EOPA No The containment spray pump discharge valves and inside containment check valves provide a containment isolation function, but do not receive a Containment Isolation Actuation (CIA) signal and are not used in the EOP's to verify containment isolation. The pump discharge valves are specifically called out in the EOP's to verify CS actuation, but not for their isolation function.

EOPC No The containment spray pump discharge valves and inside containment check valves provide a containment isolation function, but do not receive a CIA signal and are not used in the EOP's to verify containment isolation. The pump discharge valves are specifically called out in the EOP's to CS actuation, but not for their isolation function.

SDMC No CS functions are required only for accident mitigation and are not applicable to mode or du shutdown conditions.

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Enclosure to NL-14-0580 Response to Request for Additional Information FUNCTION QUES ANSWER BASIS 10 RISK DESCRIPTION TION ORPA No This function does act as a barrier to fission product release. However, although the containment spray system is open to the containment atmosphere on one side, it is a closed, filled system on the other side.

Therefore, failure of this Function would not result in a breach of the containment barrier, and offsite protective actions would not be required.

7 MAINTAIN SYSTEM Refer to Passive Categorization Results PRESSURE BOUNDARY.

Legend: INIT =Does failure cause an initiating event? RCPB =Does failure cause aRC pressure boundary loss beyond normal makeup? FBSF =Does failure result in loss of a basic safety function? EOPA =Is function relied upon in EOPs as sole means of accident mitigation? EOPC = Is function relied upon in EOPs as sole means of assuring containment integrity, monitoring of post-accident conditions, or offsite emergency planning activities? SDMC = Does failure prevents plant from reaching safe shutdown; is function safety significant during mode changes? ORPA =Does failure of function that acts as barrier to fission product release result in implementation of offsite radiological protection actions?

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Enclosure to NL-14-0580 Response to Request for Additional Information D. The passive risk of the subject components was performed in accordance with SNC Procedure NMP-ES-065-002, 10 CFR 50.69 Passive Component Categorization. The results are summarized below.

LSS train B 8" line from CS pump B LSS MOV HV9001B LSS train A 8" middle spray ring LSS iddle part of ring header) and the ne from this ring to the lower spray LSS train B 8" middle spray ring LSS iddle part of ring header) and the ,

from this ring to the lower spray' E. The subject components were evaluated by the Defense-In-Depth (DID) Process in accordance with SNC Procedure NMP-ES-065-003. The subject check valves were categorized as HSS by the DID Process.

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Enclosure to NL-14-0580 Response to Request for Additional Information F. The results of the Risk Hazards Analyses, the Passive Classification, the qualitative risk based on the functions supported, and the DID review combined to show the overall final risk of the subject components, as shown below.

RISC-1 LSS LSS LSS HSS RISC-1 E- 11

Enclosure to NL-14-0580 Response to Request for Additional Information NRC RA128 The categorization process described in NEI 00-04 starts with component categorization based on risk (i.e., the PRA or defense-in-depth), followed by function categorization, followed by the final assessment of the Integrated Decision-making Panel (IDP). When a component's function is categorized as high-safety-significant (HSS) based on risk, the associated system or train function is considered HSS. System or train functions not addressed by the risk assessment are categorized based on the seven questions in section 9 of NEI 00-04. Once a function has been identified as HSS, then all components supporting the function are assigned as preliminary HSS and the IDP must intervene to assign any of these components to LSS. SNC's process is different from the categorization process described in NEI 00-04. Based on Clarification 1 in LAR Section 3.1.3, the licensee's categorization process performs function categorization based on the seven questions in Section 9 of NEI 00-04 (qualitative assessment), followed by component categorization based on risk, followed by the final assessment of the IDP. In SNC's process, it appears that a system or train function categorized as LSS based on the qualitative assessment would remain LSS even if the risk assessment identifies the associated component's function as HSS. If so, the risk assessment does not affect the preliminary LSS safety significance of other components associated with what appears should be a HSS function. Therefore, it appears that SNC's process could have more, perhaps many more, components categorized as preliminary LSS than the NEI method.

If this characterization is incorrect please clarify. Please also clarify the difference in the IDP final assessment between components that are preliminary HSS versus those that are preliminary LSS. Explain whether this process deviates from the NEI approach if components could follow different paths in the licensee's approach than they would in the NEI approach.

SNC Response SNC will revise the sequence of its process to match that of the NEI guidance regarding the identification of components that are preliminary HSS. Specifically, SNC will revise its process to ensure that if any sse is identified as safety significant from either the integrated PRA component safety significance assessment (Section 5 of NEI 00-04) or the defense-in-depth assessment (Section 6 of NEI 00-04), then the associated system function(s) will be identified as preliminary HSS. Once a system function is identified as HSS, then, as already described in the SNC process, all components that support this function are also preliminary HSS and the IDP must intervene to assign any of these components to LSS. Based on this, the IDP assessment process for components that are preliminary HSS and those that are preliminary LSS would be consistent with the NEI approach.

SNC considers that the process described in Clarification 1 in LAR Section 3.1.3 contains all of the risk evaluation elements but in a sequence that provides for a more structured and efficient approach that is applied the same way for all functions of a system and across different systems. However, SNC will revise the sequence of its process to match the NEI guidance.

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Enclosure to NL-14-0580 Response to Request for Additional Information NRC RAI29 Clarifications 2 and 3 in Section 3.1.3 of the LAR indicate that a passive component (i.e.,

a pressure retaining component) whose failure could fail an active HSS function can be assigned to LSS if the passive categorization process yields an LSS category. The passive categorization is driven by the consequence of failure and not the frequency so it is unclear how the pressure retaining function of a passive component whose failure would fail an HSS function can be found LSS in the passive categorization process.

Please explain and provide an example.

SNC Response The NEI 00-04 categorization methodology assigns risk at the component level. Per the methodology, a component gets assigned final risk if any of the following risks is HSS:

active risk, passive risk, or defense in depth. Active risk is determined using insights from the PRA and other qualitative considerations. Passive risk is determined using a passive component categorization methodology. Risk associated with defense in depth is determined using guidance provided in the NEI 00-04 categorization methodology.

The final risk of a component is the highest of these three risks. Then the critical attributes are identified for each HSS components to further understand the reason(s) for being HSS. For example, an HSS Motor Operated Valve (MOV) may have a critical attribute of fail to close because that is what made it HSS. However, the same valve may be LSS for passive risk (i.e., pressure boundary retention) assuming there is sufficient redundancy to respond to the event of interest and LSS from a defense in depth evaluation. Therefore,

• It is possible to have an LSS passive component that fails an HSS function if there is sufficient redundancy to respond to the event of interest. That is, there are other unaffected components/trains/systems available to fulfill the function.

• It is not possible to have an LSS passive component that fails an HSS function if there is not sufficient redundancy to respond to the event of interest.

As the RAI states, the passive categorization process is driven by the consequence of failure in that the process conservatively assumes that a failure occurs with a probability of 1.0. As such, some postulated passive failures will be categorized as HSS while from a pure risk perspective they may be low safety significant. As an example, postulated failures with conditional core damage probability (CCDP) values of 5 E-04 are HSS per the passive categorization process. However, many passive components have failure frequencies of 1E-08 and lower. Thus, if failure frequency were to be considered, they may be shown quantitatively to be low safety significant.

As shown in the figure below, the RCS heat removal function, MFW, EFW, AFW or HPSI can individually fulfill that function. For purposes of this example, it is assumed that each system is a two train system. From an active risk perspective, we will assume that the PRA risk insights or other qualitative considerations have determined the HPSI pumps to be HSS. However, if we were to postulate a pressure boundary failure that failed one HPSI pump, and no other impacts (e.g. spatial, loss of inventory), we see from the figure that not only is the second HPSI pump train available to fulfill the RCS heat removal function, but MFW, EFW and AFW are available to fulfill the RCS heat removal function; therefore, this passive failure of one HPSI pump would be considered LSS.

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Enclosure to NL-14-0580 Response to Request for Additional Information I RCS

'":1~

T RCS Heat Removal I \A,IfJUut Long erm Inventory Control/Heat Removal Transient Now from a different perspective, consider the following example in which a two train system is supplied from a tank. One of the system functions is to draw water from the tank and deliver downstream. Assume that the PRA risk insights or other considerations identify Pumps A and B as HSS. Therefore, all components that support the function are preliminary HSS.

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Enclosure to NL-14-0580 Response to Request for Additional Information Tank PointY Point X TrainA TrainB In order to determine risk of the passive components (i.e., pressure retention components), passive component categorization methodology is used. This methodology divides the entire system into small segments and postulates failure of one segment at a time. The following discussion outlines, at high level, the potential outcome from the passive component categorization perspective and compares with a component/function risk which was determined to be preliminary HSS from an active risk perspective.

Point X only:

A failure of a passive component (i.e., pipe) at point X would prevent Train A and Train B from performing its function. In contrast to the previous example, we assume there are no other unaffected trains/systems to fulfill this function, therefore, the passive component at Point X (i.e., pipe) would be candidate HSS (please see Table 2 of NMP-ES-065-002, 0.0 unaffected backup trains column). This would complement with the component/function risk from an active risk perspective.

Point Y only:

A failure of a passive component (i.e., pump) at pointY would prevent Train A from performing its function at the train level; however, Train B is not affected, so Train B would be able to perform the function at the train level, which in turn, will fulfill the function at system level. Hence, the passive component at PointY (i.e., pump) could be candidate LSS for pressure retention function although it is in the path of an HSS function. The ultimate passive categorization rank would be a function of frequency of challenge" and "exposure time to challenge" (please see Table 2 of NMP-ES-065-002, 1.0 unaffected backup trains column). It should be noted that the final risk will be higher E -15

Enclosure to NL-14-0580 Response to Request for Additional Information of active risk, passive risk, or defense in depth. So if the pump was determined to be HSS in any of these risk evaluations, then its final categorization would be HSS.

Point Z only:

A failure of passive component (i.e., pipe) at point Z would prevent Train B from performing its function; however, Train A is not affected, so Train A would be able to perform the system function. Hence, the passive component at Point Z (i.e., pipe) would be candidate LSS although it is in the path of an HSS function. Similar to pointY, the ultimate passive categorization rank would be a function of frequency of challenge" and "exposure time to challenge" (please see Table 2 of NMP-ES-065-002, 1.0 unaffected backup trains column).

As demonstrated above, it is possible to have an LSS passive component that fails an HSS component provided that there are a sufficient number of unaffected backup trains available to fulfill the function. However, postulated passive component failures that fail a function (e.g. zero redundancy) will be categorized as HSS.

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Enclosure to NL-14-0580 Response to Request for Additional Information NRC RAI30 The U.S. Nuclear Regulatory Commission staff is currently investigating methods available to perform a level3 PRA as described in Technical Analysis Approach Plan for Level 3 PRA Projecf', Rev Ob, October 2013 (ADAMS Accession No. ML13296A064). SNC is participating in the Level 3 PRA Project. As described in the Approach Plan, the Office of Nuclear Regulatory Research has notified the Office of Nuclear Reactor Regulation of three issues, as summarized below, regarding the VEGP PRA that have not been fully resolved and that might impact the results of SNC's pilot request to implement 50.69 for VEGP. Please summarize the impact of these three issues on your PRA models supporting your application to implement 10 CFR 50.69 and your proposed resolution of each issue with respect to this application.

i) Crediting offsite recovery within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> without considering the impact of earlier consequential failures that could lead to unrecoverable scenarios.

ii) Crediting battery DC power for a longer time than the design.

iii) Use of a human error probability estimation method without apparently fully exercising the attributes used to apply the method to time sensitive activities.

SNC Response i) Crediting offsite recovery within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> without considering the impact of earlier consequential failures that could lead to unrecoverable scenarios.

The impact of this issue on the PRA models which will be used to support SNC's application to implement 10CFR 50.69 is negligible due to the following reasons:

• The fire PRA model does not credit recovery of any fire-induced loss of offsite power events.

• The internal events (including internal flooding) PRA model will be modified to ensure that no credit is given to AC power recovery for the unrecoverable loss of power scenarios.

It should be noted that although SNC commits to refine the internal events (including internal flooding) PRA model, SNC believes that its current internal events PRA model correctly calculates the subject initiating event frequency and its recovery failure probability. The current approach uses plant specific data and physical configuration to calculate the loss of 4.16 KV power initiating event frequency and its probability of recovery within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. As a result, these values and their applications, which are based on the same set of plant specific data and configuration, are internally and inherently consistent and technically correct.

E -17

Enclosure to NL-14-0580 Response to Request for Additional Information ii) Crediting battery DC power for a longer time than the design.

This issue does not impact the PRA models that will be used to support implementation of 10CFR 50.69. This issue was identified by SNC during development of its fire PRA model and was addressed in the fire PRA model.

Although the impact on the internal events (including internal flooding) model is considered to be negligible, as part of the consistency between the two models (internal events (including internal flooding) model and fire PRA model), the identified issue will be addressed as part of the current internal events (including internal flooding) PRA model update. Plant Vogtle will perform all categorization using the updated internal events (including internal events) PRA model.

iii) Use of a human error probability estimation method without apparently fully exercising the attributes used to apply the method to time sensitive activities.

This issue does not impact the PRA models that will be used to support implementation of 10CFR 50.69. As stated in response to RAI 30 (ii), the internal events (including internal flooding) PRA model is being updated. As part of this model update, the Human Error Probability estimates (HEPs) have been updated using ERPI's latest state-of-the-practice methodology, which addresses time sensitive activities. The HEP estimates from this HEP update indicate that some HEP estimates will decrease, some HEP estimates essentially will stay the same, and some HEP estimates will increase.

Although the internal events PRA model update has not yet been completed, based on the pattern and the changes in the HEP estimates, it is judged that the overall impact on the regulatory figure of merits (Core Damage Frequency and Large Early Release Frequency) is negligible, and the PRA insights will remain unchanged.

The HRA associated with the Fire PRA will be updated prior to using Fire PRA for categorizing systems at Plant Vogtle.

E- 18

Vogtle Electric Generating Plant Request to Revise the Licensing Basis to Implement 10 CFR 50.69 Response to Request for Additional Information Regarding Pilot 10 CFR 50.69 License Amendment Request Attachment Revised Section 3.3.1.4

Attachment Revised Section 3.3.1.4 3.3.1.4 Comparison of RG 1.200, Revision 1 and Revision 2 Internal Events (including Internal Flooding) PRA Requirements The VEGP PRA model was reviewed against the 2007 version of the PRA Standard (Reference 26) as amended by RG 1.200, Revision 1 (Reference 27). The RG 1.200 Revision 2 (Reference 28) was issued in March 2009. So it would be prudent to review the VEGP PRA to the guidance of RG 1.200, Revision 2.

To ensure compliance with any new or changed RG 1.200 requirements, it is necessary to first identify the differences between the RG 1.200 Revision 1 and Revision 2 Capability Category 1/11, 11/111, and 1/11/111 requirements. A summary of the differences in these requirements is provided in the following Table 6 along with a response for each of the differences.

NEI 05-04, Revision 3 (Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard) provides guidance on performing a gap assessment between Addendum B of the ASME/ANS PRA Standard and RG 1.200 Revision 1, and RG 1.200 Revision 2. Section 3.3 of NEI 05-04 identifies those changes to SRs that would require a re-evaluation of the PRA against the PRA Standard requirements. These SRs are included in Table 6. Note that differences that were considered typographical, editorial, or provided additional descriptions of the SRs were not conidered technically significant and were excluded.

A-1

Attachment Revised Section 3.3.1.4 The VEGP internal events PRA provides mean values of HEPs CC-I/11/111: CC-1/11/111: with associated uncertainty PROVIDE an assessment of the CHARACTERIZE the uncertainty parameters.

uncertainty in the HEPs consistent in the estimates of the HEPs with the quantification approach. consistent with the quantification The revised SA does not require a USE mean values when providing approach, and PROVIDE mean revised or additional scope of point estimates of HEPs. values for use in the quantification work or change the conclusion of of the PRA results. the review for this SR.

A-2

Attachment Revised Section 3.3.1.4 The ng the 2009 The events PRA PRA Standard provides additional uses the HRA Calculator which CC-I: CC-I: description for the SR. takes into account plant-specific USE an approach that takes the USE an approach that takes the shaping factors related to the cues following into account: following into account: for event detection, and diagnosis, decision-making.

(a) the complexity of the response (a) the complexity of detection, diagnosis, decision-making and The revised SR does not require a The ASEP Approach is an executing the required response revised or additional scope of acceptable approach. work or change the conclusion of The ASEP Approach [2-6] is an the peer review for this SR.

acceptable approach.

CC-11/111: CC-11/111:

When estimating HEPs When estimating HEPs EVALUATE the impact of the EVALUATE the impact of the following plant-specific and following plant-specific and scenario-specific performance scenario-specific performance shaping factors: shaping factors:

(d) degree of clarity of the (d) degree of clarity of meaning of the cues/indications in supporting cues/indications. the detection, diagnosis, and decision-making give the plant-specific and scenario-specific context of the event.

A- 3

Attachment Revised Section 3.3.1.4 The additional SR was added as DA-DS from RG 1.200 Revision 1 part of the RG 1.200 Revision 1 was considered in the VEGP peer ee-1111/111: ee-1/11/111: clarifications to the 2007 standard review. The VEGP internal events For each SSe for which repair is For each SSe for which repair is and was subsequently PRA does not include any sses to be modeled, ESTIMATE, based to be modeled, ESTIMATE, based renumbered by RG 1.200 for which repair is modeled.

on the data collected in DA-e14, on the data collected in DA-C15, Revision 2.

the probability of failure to repair the probability of failure to repair the sse in time to prevent core the sse in time to prevent core damage as a function of the damage as a function of the accident sequence in which the accident sequence in which the sse failure appears. sse failure appears.

QU-A2a: QU-A2: The LERF requirement was added Section 10.3.2 of the internal by RG 1.200 Revision 2. events PRA calculation ee-1/11/111: ee-1/11/111: (Reference 14} presents estimates PROVIDE estimates of the PROVIDE estimates of the The updated SR explicitly requires for individual LERF sequence individual sequences in a manner individual sequences in a manner consideration of LERF. cutsets.

consistent with the estimation of consistent with the estimation of total eDF ... total eDF (and LERF} ...

A-4

Attachment Revised Section 3.3.1.4 The phrase, "from internal events", The peer review based on the was deleted from the 2009 version 2007 version of the PRA Standard CC-I: CC-I: of the PRA Standard. The LEAF (Reference 26) addressed these ESTIMATE the point estimate ESTIMATE the point estimate requirement was added by RG LEAF requirements. Section CDF from internal events. CDF (and LEAF). 1.200 Revision 2. 10.3.2 of the internal events PRA calculation (Reference 14)

CC-11: The SA explicitly requires presents the mean CDF and ESTIMATE the mean CDF from CC-11: consideration of LEAF. However, LEAF results.

internal events, accounting for the ESTIMATE the mean CDF (and per the note in 2007 SA LE-E4 "state-of-knowledge" correlation LEAF), accounting for the state-of- and LE-F3, LEAF was addressed between event probabilities [Note knowledge correlation between in applicable requirements of (1 )]. event probabilities [Note (1)]. Table 4.5.8, which includes all QU SRs. Thus, the peer review using CC-III: the 2007 version of the PRA CALCULATE the mean CDF from CC-III: Standard addressed these LEAF internal events by propagating the CALCULATE the mean CDF (and requirements.

uncertainty distributions, ensuring LEAF) by propagating the that the "state-of-knowledge" uncertainty distributions, ensuring correlation between event that the state-of-knowledge probabilities is taken into account. correlation between event lities is taken into account.

A- 5

Attachment Revised Section 3.3.1.4 The revised wording in the 2009 The VEGP internal events PRA PRA Standard provides less logic loops were broken according CC-1/11/111: CC-1/11/111: ambiguous wording for the SR. to the guidance provided without Fault tree linking and some other Fault tree linking and some other the introduction of unnecessary modeling approaches may result modeling approaches may result conservatisms or non-in circular logic that must be in circular logic that must be conservatisms.

broken before the model is solved. broken before the model is solved.

BREAK the circular logic BREAK the circular logic The revised SR does not require a appropriately. Guidance for appropriately. Guidance for revised or additional scope of breaking logic loops is provided in breaking logic loops is provided in work or change the conclusion of NUREG/CR-2728 [Note (1)]. NUREG/CR-2728 [2-13]. When the peer review for this SR.

When resolving circular logic, resolving circular logic, DO NOT AVOID introducing unnecessary introduce unnecessary conservatisms or non- conservatisms or non-conservatisms. conservatisms.

The LERF requirement was added The peer review based on the by RG 1.200 Revision 2. 2007 version of the PRA Standard CC-1/11/111: CC-1/11/111: (Reference 26) addressed these ACCOUNT for system successes ACCOUNT for system successes The SR explicitly requires LERF requirements. The Level 2 in addition to system failures in the in addition to system failures in the consideration of LERF. However, PRA event trees presented in evaluation of accident sequences evaluation of accident sequences per the note in 2007 SR LE-E4 Section 9.2 of the internal events to the extent needed for realistic to the extent needed for realistic and LE-F3, LERF was addressed PRA calculation (Reference 14) estimation of CDF. This estimation of CDF or LERF. This in applicable requirements of explicitly account for system accounting may be accomplished accounting may be accomplished Table 4.5.8, which includes all OU successes.

by using numerical quantification by using numerical quantification SRs. Thus, the peer review using of success probability, of success probability, the 2007 version of the PRA complementary logic, or a delete complementary logic, or a delete Standard addressed these LERF term approximation and includes term approximation and includes requirements.

the treatment of transfers among the treatment of transfers among event trees where the "successes" event trees where the "successes" may not be transferred between may not be transferred between event trees. event trees.

A-6

Attachment Revised Section 3.3.1.4 t;:;;~~~~~~M;;~ The peer review based on the 2007 version of the PRA Standard CC-I: CC-I: (Reference 26) addressed these ESTIMATE the uncertainty interval ESTIMATE the uncertainty interval The SR explicitly requires LERF requirements. Section 10.4 of the COF results. Provide a of the COF (and LERF) results. consideration of LERF. However, of the internal events PRA basis for the estimate consistent Provide a basis for the estimate per the Note in 2007 SR LE-E4 calculation (Reference 14) with the characterization consistent with the and LE-F3, LERF was addressed presents the uncertainty intervals parameter uncertainties (OA-03, characterization parameter in applicable requirements of for both COF and LERF, with HR-06, HR-GS, IE-C15). uncertainties (OA-03, HR-06, HA- Table 4.5.S, which includes all QU consideration of the state-of-GS, IE-C15). SRs. Thus, the peer review using knowledge correlation.

CC-11: the 2007 version of the PRA ESTIMATE the uncertainty interval CC-11: Standard addressed these LERF of the COF results. ESTIMATE the ESTIMATE the uncertainty interval requirements.

uncertainty intervals associated of the COF (and LEAF) results.

with parameter uncertainties (OA- ESTIMATE the uncertainty 03, HR-06, HR-GS, IE-C15), intervals associated with taking into account the state-of- parameter uncertainties (OA-03, knowledge correlation. HR-06, HR-GS, IE-C15), taking into account the state-of-CC-III: knowledge correlation.

PROPAGATE parameter uncertainties (OA-03, HR-06, HA- CC-III:

GS, IE-C15) .... (no change) PROPAGATE parameter uncertainties (OA-03, HR-06, HR-IE-C1 A-7

Attachment Revised Section 3.3.1.4 met 2007 and Ill were collapsed into a single version of the PRA Standard CC-I: CC-1/11/111: requirement for CC-1/11/111 in the (Reference 26).

PROVIDE an assessment of the For each source of model 2009 version of the PRA impact of the model uncertainties uncertainty and related Standard. The reference to Note and assumptions on the results of assumption identified in QU-E1 1 was deleted by RG 1.200 the PRA. and QU-E2, respectively, Revision 2.

IDENTIFY how the PRA model is CC-11: affected (e.g., introduction of a The updated SR assigns the same EVALUATE the sensitivity of the new basic event, changes to basic requirement to all three CCs.

results to model uncertainties and event probabilities, change in Meeting CC-II in the 2007 version key assumptions using sensitivity success criterion, introduction of a of the PRA Standard assures that analyses [Note (1)]. new initiating event). the new SR is met.

CC-III:

EVALUATE the sensitivity of the results to uncertain model boundary conditions and other assumptions using sensitivity analyses except where such sources of uncertainty have been adequately treated in the quantitative uncertainty analysis A- 8

Attachment Revised Section 3.3.1.4 2009 PRA the SR for documentation that CC-1/111111: CC-1/11/111: facilitates applications, upgrades, DOCUMENT the internal flooding DOCUMENT the internal flood and review into separate SRs for analysis in a manner that plant partitioning in a manner that each of the internal flooding facilitates PRA applications, facilitates PRA applications, elements (plant partitioning, flood The revised SR does not require a upgrades, and peer review. upgrades, and peer review. sources, flood scenarios, initiating revised or additional scope of events, and quantification). work or change the conclusion of IFS0-81: the peer review for this SR.

CC-1/11/111:

DOCUMENT the internal flood sources in a manner that facilitates PRA applications, upgrades, and peer review.

IFSN-81:

CC-1/11/111:

DOCUMENT the internal flood scenarios in a manner that facilitates PRA applications, upgrades, and peer review.

IFEV-81:

CC-1111/111:

DOCUMENT the internal flood-induced initiating events in a manner that facilitates PRA applications, upgrades, and peer review.

A-9

Attachment Revised Section 3.3.1.4 CC-1/11/111: CC-1/11/111:

DOCUMENT the internal flooding DOCUMENT the internal flood analysis in a manner that accident sequences and facilitates PRA applications, quantification in a manner that upgrades, and peer review. facilitates PRA applications, and review.

IF-F2: The rement to document 5 and the walkdowns performed in support internal flooding PRA (Reference CC-1/11/111: CC-1/11/111: of plant partitioning was added to 13) document the walkdowns DOCUMENT the process used to DOCUMENT the process used to the 2009 version of the PRA performed to validate information identify ... flood areas, . . . For identify flood areas. For example, Standard. related to flood areas, flood example, this documentation this documentation typically sources, SSCs, mitigation and typically includes includes The updated SR cites examples of other flood related features in the acceptable documentation of the flood areas.

(b) flood areas used in the (a) flood areas used in the process to identify flood sources.

analysis and the reason for analysis and the reason for eliminating areas from further eliminating areas from further Since documentation of analysis. analysis. walkdowns was not in the 2007 version of the PRA Standard, it (b) any walkdowns performed in was not reviewed as part of the support of the plant partitioning. peer review conducted using that version of the PRA Standard.

A-10

Attachment Revised Section 3.3.1.4 The requirement to document The internal flooding PRA walkdowns performed in support documents the walkdowns CC-1/111111: CC-1/11/111: of the identification or screening of performed to validate information DOCUMENT the process used to DOCUMENT the process used to flood sources was added to 2009 related to flood areas, flood identify applicable flood sources. identify applicable flood sources. version of the PRA Standard. sources, SSCs, mitigation and For example, this documentation For example, this documentation other flood related features in the typically includes typically includes The updated SR cites examples of flood areas.

acceptable documentation of the (a) flood sources identified in the (a) flood sources identified in the process to identify flood sources.

analysis, rules used to screen analysis, rules used to screen out these sources, and the out these sources, and the Since documentation of resulting list of sources to be resulting list of sources to be walkdowns was not in the 2007 further examined. further examined version of the PRA Standard, it was not reviewed as part of the (f) screening criteria used in the (b)screening criteria used in the peer review conducted using that analysis. analysis version of the PRA Standard.

(j) calculations or other analyses (c) calculations or other analyses used to support or refine the used to support or refine the flooding evaluation. flooding evaluation (d) any walkdowns performed in support of the identification or

.,.,,..,.,n..,,n of flood sources.

A-ll

Attachment Revised Section 3.3.1.4 The requirement to document The internal flooding PRA walkdowns performed in support documents the walkdowns CC-1/11/111: CC-1111/111: of the identification or screening of performed to validate information DOCUMENT the process used to DOCUMENT the process used to flood scenarios was added to related to flood areas, flood identify applicable flood scenarios. identify applicable flood scenarios. 2009 version of the PRA sources, SSCs, mitigation and For example, this documentation For example, this documentation Standard. other flood related features in the typically includes typically includes flood areas.

(a) propagation pathways ... The updated SA cites examples of (c) propagation pathways ... acceptable documentation of the (b) accident mitigating features process to identify flood scenarios.

(d) accident mitigating features and barriers credited ...

and barriers credited ... Since documentation of (c) assumptions or calculations walkdowns was not in the 2007 (e) assumptions or calculations used in the determination of ... version of the PRA Standard, it used in the determination of ... flood-induced effects on was not reviewed as part of the flood-induced effects on equipment operability. peer review conducted using that equipment operability. version of the PRA Standard.

(d) screening criteria used in the (f) screening criteria used in the analysis.

analysis.

(e) flooding scenarios considered, (g) flooding scenarios considered, screened, and retained.

screened, and retained.

(f) description of how the internal (h) description of how the internal event analysis models were event analysis models were modified ...

modified ...

(g) calculations or other analyses (j) calculations or other analyses used to support or refine the used to support or refine the flooding evaluation.

flooding evaluation.

(h) any walkdowns performed in support of the identification or screeni of flood os.

A-12

Attachment Revised Section 3.3.1.4 The 2009 PRA Standard divides The VEGP internal flooding report the SR for typical items included in adequately documents each of the CC-1/11/111: CC-1/11/111: the internal flooding analysis elements of the internal flooding DOCUMENT the process used ... DOCUMENT the process used to process documentation into analysis.

internal flood model development. identify applicable flood-induced separate SRs for each of the For example, this documentation initiating events. For example, this internal flooding elements (plant The revised SR does not require a typically includes documentation typically includes partitioning, flood sources, flood revised or additional scope of scenarios, initiating events, and work or change the conclusion of (f) screening criteria used in the (a) flood frequencies, component quantification). While most the peer review for this SR.

analysis unreliabilities/unavailabilities, elements (PP, SO, SN, QU) and HEPs used in the analysis included additional requirements (i) flood frequencies, component (i.e., the data values unique to to document walkdowns, the EV unreliabilities/unavailabilities, the flooding analysis) element included no change in and HEPs used in the analysis requirements.

(i.e., the data values unique to (b) calculations or other analyses the flooding analysis) used to support or refine the flooding evaluation (j) calculations or other analyses used to support or refine the (c) screening criteria used in the flooding evaluation analysis A-13

Attachment Revised Section 3.3.1.4 The requirement to document internal flooding walkdowns performed in support documents the walkdowns CC-1/111111: CC-1/11/111: of internal flood accident performed to validate information DOCUMENT the process used to DOCUMENT the process used to sequence quantification was related to flood areas, flood define the applicable internal flood define the applicable internal flood added in 2009 version of the PRA sources, SSCs, mitigation and accident sequences and their accident sequences and their Standard. other flood related features in the associated quantification. For associated quantification. For flood areas that are considered in example, this documentation example, this documentation The updated SR cites examples of flood sequence definition.

typically includes: typically includes: acceptable documentation of the process to identify flood related (j) calculations or other analyses (a) calculations or other analyses features considered in flood used to support or refine the used to support or refine the sequence quantification.

flooding evaluation flooding evaluation Since documentation of (f) screening criteria used in the (b) screening criteria used in the walkdowns was not in the 2007 analysis analysis version of the PRA Standard, it was not reviewed as part of the (i) flooding scenarios considered, (c) flooding scenarios considered, peer review conducted using that screened, and retained screened, and retained version of the PRA Standard.

(k) results of the internal flood (d) results of the internal flood analysis, consistent with the analysis, consistent with the quantification requirements quantification requirements provided in HLR-QU-D provided in HLR-QU-D (e) any walkdowns performed in support of internal flood accident sequence A-14

Attachment Revised Section 3.3.1.4 The 2009 T~ re~rt the SR for documentation of did not include documentation of CC-1/11/111: CC-1/11/111: assumptions and sources of sources of model uncertainty for DOCUMENT the assumptions and DOCUMENT sources of model uncertainty into separate SRs for the internal flooding analysis.

sources of uncertainty associated uncertainty and related each of the internal flooding with the internal flooding analysis. assumptions (as identified in QU- elements {plant partitioning, flood However, as part of this Ucense E1 and QU-E2) associated with sources, flood scenarios, initiating Amendment Request, the internal flood plant partitioning. events, and quantification). identification and documentation of the sources of model IFS0-63: uncertainty in all elements of the VEGP internal events Level 1 and CC-1/11/111: Level 2 PRA (including internal DOCUMENT sources of model flooding) was performed.

uncertainty and related assumptions (as identified in QU-E1 and QU-E2) associated with the internal flood sources.

IFSN-63:

CC-1/11/111:

DOCUMENT sources of model uncertainty and related assumptions (as identified in QU-E1 and QU-E2) associated with the internal flood scenarios.

A-15

Attachment Revised Section 3.3.1.4 The 2009 PRA Standard divides The VEGP internal flooding report the SR for documentation of did not include documentation of CC-1/11/111: CC-1/11/111: assumptions and sources of sources of model uncertainty for DOCUMENT the assumptions and Document sources of model uncertainty into separate SRs for the internal flooding analysis.

sources of uncertainty associated uncertainty and related each of the internal flooding with the internal flooding analysis. assumptions (as identified in QU- elements (plant partitioning, flood However, as part of this License E1 and QU-E2) associated with sources, flood scenarios, initiating Amendment Request, the internal flood-induced initiating events, and quantification). identification and documentation events. of the sources of model uncertainty in all elements of the IFQU-83: VEGP internal events Level 1 and Level 2 PRA (including internal CC-1/11/111: flooding) was performed.

DOCUMENT sources of model uncertainty and related assumptions (as identified in QU-E1 and QU-E2) associated with the internal flood accident and uantification.

A-16

Attachment Revised Section 3.3.1.4 The clarifications to the 2007 PRA The VEGP internal flooding report Standard provided in RG 1.200, met the combined CC-IIII CC-I: CC-I: Revision 1 included the separation describing the mechanisms For the SSCs identified in IF-C2c, For the SSCs identified in IFSN- of CC-I and C-11. The revised considered, including

... (no change) A5, ... (no change) wording in the 2009 PRA submergence and spray; thereby Standard provides additional meeting CC-11. However, there CC-II: CC-II: description for the SR. was not sufficient consideration of INCLUDE failure by submergence For the SSCs identified in IFSN- humidity, condensation, and spray in the identification A5, IDENTIFY the susceptibility of temperature, etc. to merit CC-III.

process. each sse in a flood area to flood-ASSESS qualitatively the impact induced failure mechanisms.

of flood-induced mechanisms that INCLUDE failure by submergence are not formally addressed (e.g., and spray in the identification using the mechanisms listed process.

under Capability Category Ill of ASSESS qualitatively the impact this requirement), by using of flood-induced mechanisms that conservative assumptions. are not formally addressed (e.g.,

using the mechanisms listed CC-III: under Capability Category Ill of For the SSCs identified in I F-C2c, this requirement), by using

.. .(no change) conservative assumptions .

CC-III:

For the SSCs identified in IFSN-A5, ... (no change)

A-17

Attachment Revised Section 3.3.1.4 The NRC recommended wording The Recovery Analysis and changes on "key assumptions Uncertainty Analysis for the VEGP and sources of uncertainty were Level 1 and Level 2 Model CC-IIII/III: CC-1/11/111: reflected in the 2009 PRA (Reference 14, Chapter 10)

DOCUMENT the key assumptions DOCUMENT the sources of model Standard. included documentation on the and key sources uncertainty uncertainty and related sources of modeling uncertainty.

associated with the initiating event assumptions (as identified in QU- Since the wording changes were analysis (also accident sequence E1 and QU-E2 or LE-F3) not in the 2007 version of the PRA In addition, as part of this License analysis, development of success associated with the initiating event standard, the wording changes Amendment Request, an criteria, systems analysis, human analysis (also accident sequence were not directly considered assessment and characterization reliability analysis, data analysis, analysis, development of success during the peer review conducted of the sources of model and LERF analysis, including results criteria, systems analysis, human using the 2007 PRA Standard. uncertainty in the VEGP internal and important insights from reliability analysis, data analysis, events Level 1 and Level 2 PRA sensitivity studies.) and LERF analysis, including results was performed according to the and important insights from guidance in NUREG-1855 and EPRI TR-1016737.

QU-E1: The NRC recommended wording The Recovery Analysis and changes on "key assumptions Uncertainty Analysis for the VEGP cc 1/11/11: cc 1/11/11: and sources of uncertainty were Level 1 and Level 2 Model IDENTIFY key sources of model IDENTIFY sources of model reflected in the 2009 PRA (Reference 14, Chapter 10) uncertainty. uncertainty. Standard. included documentation on the sources of modeling uncertainty.

Since the wording changes were not in the 2007 version of the PRA In addition, as part of this License standard, the wording changes Amendment Request, an were not directly considered assessment and characterization during the peer review conducted of the sources of model using the 2007 PRA Standard. uncertainty in the VEGP internal events Level 1 and Level 2 PRA was performed according to the guidance in NUREG-1855 and EPRI TR-1016737.

A-18

Attachment Revised Section 3.3.1.4 The NRC recommended wording The Recovery Analysis and changes on "key assumptions Uncertainty Analysis for the VEGP cc 1/11/11: cc 1/11/11: and sources of uncertainty were Level 1 and Level 2 Model IDENTIFY key assumptions made IDENTIFY assumptions made in reflected in the 2009 PRA (Reference 14, Chapter 10) in the development of the PRA the development of the PRA Standard. included documentation on the model. model. sources of modeling uncertainty.

Since the wording changes were not in the 2007 version of the PRA In addition, as part of this License standard, the wording changes Amendment Request, an were not directly considered assessment and characterization during the peer review conducted of the sources of model using the 2007 PRA Standard. uncertainty in the VEGP internal events Level 1 and Level 2 PRA was performed according to the guidance in NUREG-1855 and EPRI TR-1016737.

QU-F4: QU-F4: The NRC recommended wording The Recovery Analysis and changes on "key assumptions Uncertainty Analysis for the VEGP cc 1/11/11: cc 1/11/11: and sources of uncertainty were Level 1 and Level 2 Model DOCUMENT key assumptions DOCUMENT the characterization reflected in the 2009 PRA (Reference 14, Chapter 10) and key sources of uncertainty, of the sources of model Standard. included documentation on the such as: possible optimistic or uncertainty and related sources of modeling uncertainty.

conservative success criteria, assumptions (as identified in QU- Since the wording changes were suitability of the reliability data, E4). not in the 2007 version of the PRA In addition, as part of this License possible modeling uncertainties standard, the wording changes Amendment Request, an (modeling limitations due to the were not directly considered assessment and characterization method selected), degree of during the peer review conducted of the sources of model completeness in the selection of using the 2007 PRA Standard. uncertainty in the VEGP internal initiating events, possible spatial events Level 1 and Level 2 PRA dependencies, etc. was performed according to the guidance in NUREG-1855 and EPRI TR-1016737.

A-19

Attachment Revised Section 3.3.1.4 wording and numbering The peer review based on the changes were incorporated in the 2007 version of the PRA Standard

.QJ.;_ Cl: 2009 version of the PRA (Reference 26) addressed these PERFORM containment isolation PERFORM containment isolation Standard. LERF requirements and is not analysis in a conservative analysis in a conservative affected by the wording change.

manner ... (unchanged) manner... (unchanged) The difference is considered editorial only.

C II: C II:

PERFORM containment isolation PERFORM containment isolation analysis in a realistic manner for analysis in a realistic manner for the significant accident the significant accident progression sequences resulting progression sequences resulting in a large early release. USE in a large early release. USE conservative or a combination of conservative or a combination of conservative or realistic treatment conservative or realistic treatment for the non-significant accident for the nonsignificant accident progression sequences progression sequences

... (unchanged) ... (unchanged)

CC Ill:

PERFORM containment isolation PERFORM containment isolation analysis in a realistic analysis in a realistic manner .. manner ...

IE-A4a IE-A6 The wording change is a The revised wording is a clarification/editorial correction of clarification and did not result in CC II and CC Ill: CC II and CC Ill: the requirements, and does not any revised or additional scope of When performing the systematic When performing the systematic need to be re-assessed unless the work relative to the VEGP CC evaluation required in IE-A4, evaluation required in IE-A5, SR was not met in the peer systematic evaluation, or change INCLUDE initiating events INCLUDE initiating events review. the conclusion of the peer review resulting from multiple failures, if resulting from multiple failures, if for this SR. This SR was met.

the equipment failures result from the equipment failures result from a common cause, and from a common cause, or from system system alignments resulting from alignments resulting from preventive and corrective preventive and corrective maintenance. maintenance.

A-20

Attachment Revised Section 3.3.1.4 internal events PRA Standard provides less grouping of initiating events was Cl: Cl: ambiguous wording for the SR. found to be appropriate.

GROUP initiating events only GROUP initiating events only when the following is true: when the following can be The revised SR does not require a (items (a) and (b) unchanged) ensured: revised or additional scope of (items (a) and (b) unchanged) work or change the conclusion of C II: the peer review for this SR.

GROUP initiating events only C II:

when the following is true: GROUP initiating events only (items (a) and (b) unchanged) when the following can be DO NOT SUBSUME events into a ensured:

group unless (items (a) and (b) unchanged)

(items (1) and (2) unchanged) DO NOT SUBSUME scenarios into a group unless C Ill: (items (1) and (2) unchanged)

GROUP initiating events only when the following is true: C Ill:

(items (a) and (b) unchanged) GROUP initiating events only when the following can be ensured:

(items (a) and (b) unchanged)

A-21

Attachment Revised Section 3.3.1.4 The clarifications to the 2007 PRA The internal events PRA Standard were provided in RG accident sequence construction CCII: CCII: 1.200, Revision 1 and were was assigned CC Ill by the peer In constructing the accident In constructing the accident subsequently also included in the review team for this SR.

sequence models, INCLUDE, for sequence models, INCLUDE, for 2009 PRA Standard.

each modeled initiating event, each modeled initiating event, The revised SR does not require a sufficient detail that differences in sufficient detail that differences in revised or additional scope of requirements on systems and requirements on systems and work or change the conclusion of required operator interactions required operator interactions the peer review for this SR.

(e.g., systems initiations or valve (e.g., systems initiations or valve alignments) are captured. Where alignment) are captured. Where diverse systems and/ or operator diverse systems and/ or operator actions provide a similar function, actions provide a similar function, if choosing one over another if choosing one over another changes the requirements for changes the requirements for operator intervention or the need operator intervention or the need for other systems, MODEL each for other systems, MODEL each The clarifications to the 2007 PRA The VEGP human reliability Standard were provided in RG analysis includes both the cc 1/11/111: cc 1/11/111: 1.200, Revision 1 and were cognitive (diagnosis) and IDENTIFY IDENTIFY those actions subsequently also included in the execution portions of control room (a) those actions required to (a) required to initiate ... (not 2009 PRA Standard to remove actions.

initiate ... (not changed) changed) ambiguity.

(b) those actions performed by the {b) performed by the control room The revised SR does not require a control room staff either in staff either in response to revised or additional scope of response to procedural procedural direction or as skill-of- work or change the conclusion of direction or as skill-of-the-craft the-craft to diagnose and then the peer review for this SR.

to diagnose and then recover ... (not changed).

recover ..

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Attachment Revised Section 3.3.1.4 The clarifications to the 2007 PRA The VEGP internal events PRA Standard were provided in RG does not include any SSCs for c 1/11/111: c 1/11/111: 1.200, Revision 1 and were which repair is modeled.

For each SSC for which repair is For each SSC for which repair is subsequently also included in the to be modeled (see SY-A22), to be modeled (see SY-A22), renumbered SR in RG 1.200 The SR is judged to be not IDENTIFY instances of plant- IDENTIFY instances of plant- Revision 2. applicable.

specific experience and, when that specific experience and, when that is insufficient to estimate failure to is insufficient to estimate failure to repair consistent with DA- repair consistent with DA-DS ... 09 ...

The clarifications to the 2007 PRA Parameter estimates most Standard were provided in RG failure data in the VEGP internal CC II and Ill: CC II and Ill: 1.200, Revision 1 and were events PRA were based on a

.. .USE a Bayes update process . ... USE a Bayes update process . subsequently also included in the Bayesian update process. A few CHOOSE... CHOOSE... 2009 PRA Standard to remove estimates were based on ambiguity. published generic data only.

The revised SR does not change the conclusion of the peer review for this SR.

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Attachment Revised Section 3.3.1.4 The requirement to include the fire Potential flood sources identified protection system in Item (a) as a in Section 5 of the internal flooding CC-1/11/111: CC-1/11/111: potential flooding source was PRA reviewed as part of 2009 For each flood area, IDENTIFY For each flood area, IDENTIFY added by RG 1.200 Revision 1. peer review against 2007 version the potential sources of flooding the potential sources of flooding This requirement was addressed of the PRA standard amended by

[Note (1 )]. INCLUDE: [Note (1)]. INCLUDE in the peer review, which used the RG 1.200, Revision 1 (Reference 2007 version of the PRA Standard 27) include RCS-connected (a) equipment (e.g., piping, valves, (a) equipment (e.g., piping, valves, amended by RG 1.200 Revision 1. systems - chemical and volume pumps) located in the area that pumps) located in the area that control system (CVCS),

are connected to fluid systems are connected to fluid systems The requirement to include the containment spray (CS), residual (e.g., circulating water system, (e.g., circulating water system, reactor coolant system in Item (a) heat removal (RHR), reactor service water system, fire service water system, fire as a potential flooding source was coolant system drain tank protection system, component protection system, component added to the 2009 version of the (RCSDT), safety injection (SI),

cooling water system, cooling water system, PRA Standard. Thus, it was not and reactor water makeup system feedwater system, condensate feedwater system, condensate reviewed as part of the peer (RMWS). As outlined in the Plant and steam systems) and steam systems, and review conducted using that Vogtle Internal Flooding notebook, reactor coolant system) version of the PRA Standard. the Containment Building (and RCS components therein) is not included in the scope of the internal IF-B3 IFSO-A5 The clarifications to the 2007 PRA Flood calculations for the VEGP Standard were provided in RG internal flooding analysis consider cc 1/11/111: cc 1/111111: 1.200, Revision 1 and were a range of flow rates.

For each source and its identified For each source and its identified subsequently also included in the failure mechanism, IDENTIFY the failure mechanism, IDENTIFY the re-numbered SR in the 2009 PRA The revised SR does not change characteristic of release and the characteristic of release and the Standard. the conclusion of the peer review capacity of the source. INCLUDE: capacity of the source. INCLUDE: for this SR.

(b) range of flow rates (b) range of flow rates A-24

Attachment Revised Section 3.3.1.4 r*r*r*~Tit"\n~ to the 2007 PRA Flood calculations for the VEGP Standard were provided in RG internal flooding analysis consider C II: C II: 1.200, Revision 1 and were drain lines and other pathways not IDENTIFY inter-area propagation IDENTIFY inter-area propagation subsequently removed from CC II directly related to room openings.

through the normal flow path from through the normal flow path from for the re-numbered SR in the one area to another via drain one area to another via drain 2009 PRA Standard. The revised SR requires a lines; and areas connected via lines; and areas connected via reduced scope of word and does back flow through drain lines backflow through drain lines not change the conclusion of the involving failed check valves, pipe involving failed check valves, pipe peer review for this SR.

and cable penetrations (including and cable penetrations (including cable trays), doors, stairwells, cable trays), doors, stairwells, hatchways, and HVAC ducts. hatchways, and HVAC ducts.

INCLUDE potential for structural INCLUDE potential for structural failure (e.g., of doors or walls) due failure (e.g., of doors or walls) due to flooding loads, and the potential to flooding loads.

for barrier unavailability, including maintenance ~,.,.",., *.,.~

IF-D3: IFEV-A2 The revised wording in the 2009 The VEGP internal flooding PRA Standard provides less analysis grouping (and some CCII: ambiguous wording for the SR. consequential subsuming) was AVOID subsuming scenarios into found to be appropriate.

a group unless DO NOT SUBSUME scenarios into a group unless The revised SR does not require a revised or additional scope of work or change the conclusion of the review for this SR.

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Attachment Revised Section 3.3.1.4 The VEGP internal flooding phrase "due to causes quantification was performed cc 1/11/111: cc 1/11/111: independent of the flooding" which using the same basic events for INCLUDE, in the quantification, INCLUDE, in the quantification, did not change the intent of the other non-flooding failures as used the combined effects of failures the combined effects of failures requirement. in the internal events model.

caused by flooding and those caused by flooding and those coincident with the flooding due to coincident with the flooding due to The revised SA does not require a causes independent of the independent causes including revised or additional scope of flooding including unavailability equipment failures, unavailability work or change the conclusion of due to maintenance, common- due to maintenance, common- the peer review for this SR.

cause failures, and other credible cause failures, and other credible causes.

IE-C10: The sentences were clarifications provided in RG 1.200 Revision 1 CC-1/11/111: CC-1/111111: and Revision 2, respectively.

An example of an acceptable The updated SR cites a more generic data sources is recent example of an acceptable Per NEI 05-0, Revision 3, this SR NUREG/CR-5750 n,..,,.ril" data source. does not re-evaluation.

DA-C1: Reference NUREG-1715 was R-6928 is used as the added by RG 1.200 Revision 1; source for generic data priors in CC-1/111111: CC-1/11/111: References NUREG-1715 and Revision 4 of the VEGP internal NUREG/CR-6928 were included events PRA.

Examples of parameter estimates Examples of parameter estimates in the 2009 version of the PRA and associated sources include: and associated sources include Standard. Per NEI 05-0, Revision 3, this SR (a) component failure rates and (a) component failure rates and does not require re-evaluation.

probabilities: NUREG/CR-4639 probabilities: NUREG/CR-4639 The updated SA cites more recent

[Note (1 )], NUREG/CR-4550 [2-7], NUREG/CR-4550 [2-3], examples of acceptable generic

[Note (2)], NUREG-1715 [Note NUREG-1715 [2-21], data sources.

NUREG/CR-6928 A-26

Attachment Revised Section 3.3.1.4 LE-F2: LE-F3: Separate requirements for CC-I, II, No action, CC-II met for 2007 and Ill were collapsed into a single version of the PRA Standard CC-I: CC-1/11/111 : requirement for CC-1/11/111 in the (Reference 26).

PROVIDE a qualitative IDENTIFY and CHARACTERIZE 2009 version of the PRA assessment of the key sources of the LEAF sources of model Standard. The difference was not uncertainty. uncertainty and related included in NEI 05-04, Revision 3.

Examples: assumptions, in a manner (a) Identify bounding assumptions. consistent with the applicable (b) Identify conservative treatment requirements of Tables 2-2.7-2(d) The updated SR assigns the same of phenomena. and 2-2.7-2(e). requirement to all three CCs.

Meeting CC-II in the 2007 version CC-11: of the PRA Standard assures that PROVIDE uncertainty analysis the new SR is met.

that identifies the key sources of uncertainty and includes sensitivity studies for the significant contributors to LEAF.

CC-III:

PROVIDE uncertainty analysis that identifies the key sources of uncertainty and includes sensiti studies.

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Attachment Revised Section 3.3.1.4 RG 1.200 Revision 2 removed the No action, This SR was met for phrase "including uncertainty and 2007 version of the PRA Standard C-1/11/111: C-1/11/111: sensitivity analyses, as (Reference 26).

appropriate for the level of detail DOCUMENT the process used to DOCUMENT the process used to of the analysis", which is a identify plant damage states and identify plant damage states and reduction in the requirement.

accident progression contributors, accident progression contributors, define accident progression define accident progression sequences, evaluate accident sequences, evaluate accident progression analyses of progression analyses of containment capability, and containment capability, and quantify and review the LERF quantify and review the LERF results. For example, this results. For example, this documentation typically includes documentation typically includes (h) The model integration process (h) The model integration process including the results of the including the results of the quantification including quantification.

uncertainty and sensitivity analyses, as appropriate for the level of detail the ,.n~'"""'

SY-815: The sentences were provided in As noted in Table 9.2-1 of the the 2007 and 2009 versions of the internal events PRA calculation CC-1/11/111: CC-1/111111: PRA Standard, respectively. The (Reference 14), failure of the difference was not included in NEI containment boundary due to (h) harsh environments induced (h) harsh environments induced 05-04, Revision 3. venting is not applicable to the by containment venting, or by containment venting, failure VEGP large, dry, subatmospheric failure that may occur prior to of the containment venting The updated SR explicitly requires containment.

the onset of core damage. ducts, or failure of the consideration of containment containment boundary that may venting ducts and failure of the occur prior to the onset of core containment boundary prior to damage core damage.

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