NL-12-0932, Pilot 10 CFR 50.69 License Amendment Request

Pilot 10 CFR 50.69 License Amendment Request

ML12248A035
Person / Time
Site:	Vogtle
Issue date:	08/31/2012
From:	Ajluni M Southern Nuclear Operating Co, Southern Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NL-12-0932
Download: ML12248A035 (160)

Text

Mark J. Ajluni, P.E. Southern Nuclear Nuclear Licensing Director Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birm ingham. Alabama 35201 Tel 2059927673 Fax 205.9927885 SOUTHERN ' \

August 31 , 2012 COMPANY Docket Nos. : 50-424 NL-12-0932 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington , D. C. 20555-0001 Vogtle Electric Generating Plant Pilot 10 CFR 50.69 License Amendment Request Ladies and Gentlemen :

Pursuant to 10 CFR 50.69(b)(2) and 10 CFR 50.90, Southern Nuclear Operating Company (SNC) hereby requests amendments to the Vogtle Electric Generating Plant (VEGP) Units 1 and 2 Renewed License Numbers NPF-68 and NPF 81.

The proposed amendments would revise the VEGP licensing basis to implement 10 CFR 50.69, risk-informed categorization and treatment of structures, system ,

and components (SSCs) for nuclear power plants. The Nuclear Regulatory Commission (NRC), by letter dated June 17, 2011, in response to SNC 's letter dated December 6, 2010, granted pilot status for the VEGP 10 CFR 50.69 license amendment request (LAR).

Implementation of 10 CFR 50.69 allows for application of a risk-informed categorization process per 10 CFR 50.69(c) to modify the scope of SSCs subject to special treatment requirements . Alternative treatments per 10 CFR 50.69(b)(1) and 10 CFR 50.69(d)(2) can then be applied consistent with the categorization of the SSCs . Implementation of 10 CFR 50.69 will allow the licensee and the NRC to better focus attention and resources on SSCs that have safety significance, resulting in improved plant safety.

The VEGP 10 CFR 50.69 LAR conforms to the scope requirements of 10 CFR 50.69(b)(2). The categorization process described in the LAR conforms to the guidance in Regulatory Guide 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," Revision 1 dated May 2006. The categorization process also conforms to the guidance in NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline, " Revision 0 dated July 2005. SNC has determined that the proposed amendments do not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c). provides the basis for the proposed change to the VEGP Units 1 and 2 Operating Licenses. Enclosure 2 provides the VEGP Operating Licenses marked-up pages showing the proposed change . Enclosure 3 provides

U. S. Nuclear Regulatory Commission NL-12-0932 Page 2 the VEGP Operating Licenses clean typed pages showing the proposed change.

SNC requests approval of the proposed license amendments by August 31, 2013.

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

Mr. M. J. Ajluni states he is the Nuclear Licensing 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.

and SUbscrir before me this 3 Ij.>-"day of ~.j- . 2012.

~~~.YtrJ-Notary Public My commission expires: I {- 0 Z - 20 I 3 MJAlCLTflac Respectfully submitted, M. J. Ajluni Nuclear Licensing Director

Enclosures:

1. Basis for Proposed Change

2. Operating Licenses Marked-up Pages

3. Operating Licenses Clean Typed Pages 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 Mr. B. L. Ivey, Vice President - Regulatory Affairs Mr. B. J. Adams, Vice President - Fleet Operations RType: CVC7000 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. R. E. Martin, NRR Senior Project Manager - Vogtle Mr. L. M. Cain , Senior Resident Inspector - Vogtle Mr. M. O. Miller, Senior Project Engineer, NRC Region II State of Georgia Mr. J. H. Turner, Environmental Director Protection Division

Vogtle Electric Generating Plant Pilot 10 CFR 50.69 License Amendment Request Enclosure 1 Basis for Proposed Change to NL-1L-V,JUL for Proposed Change Table of Contents 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION TECHNICAL EVALUATION 1 Categorization Description (10 CFR 50.69(b)(2)(i>>

3.1.1 Categorization Overview 3.1 Categorization Key Steps 3.1 Conformance to NEI 00-04 Technical Adequacy Evaluation (10 CFR 50.69(b}(2)(ii>>

1 internal with Internal Flooding Probabilistic Risk Assessment (PRA) Model 3.2.2 PRA Model 3.2.3 Other Techniques 3.3 Review Process Results (10 50. 69(b )(2)(iii))

1 Internal (including Internal Flooding) Model 3.3.2 PRA Model 3.4 Risk Evaluations (10 50.69(b)(2)(iv>>

1 Rule Requirements 3.4.2 Explanation of Process 3.4.3 Conclusion

4.0 REGULATORY EVALUATION

4.1 Significant Hazards Consideration Applicable Regulatory Requirements/Criteria Precedent 4.4 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

6.0 APPENDIX A: Regulatory Guide 1 Conformance Matrix E1 to I\lL-12-0932 Basis for Proposed Change 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend the Vogtle Electric Generating Plant (VEGP) Units 1 and 2 Renewed Facility Operating License (OL) Numbers NPF-68 and NPF-81. Specifically, the VEGP Units 1 and 2 OLs will be amended to add proposed OL license conditions (LCs) which will allow for the voluntary implementation of 10 CFR 50.69.

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 (SSCs) for nuclear power plants (Reference 1). The Nuclear Regulatory Commission (NRC), by letter dated June 17,2011 (Reference 2), in response to Southern Nuclear Operating Company's (SNC) letter dated December 6,2010, granted pilot status for the VEGP 10 CFR 50.69 license amendment request (LAR). Implementation of 10 CFR 50.69 will allow the licensee and the NRC to better focus attention and resources on SSCs that have safety significance, thereby resulting in improved plant safety.

SNC requests approval of the proposed license amendments by August 31, 2013. Initial 10 CFR 50.69 implementation will consist of SNC issuing the VEGP procedures documenting the 10 CFR 50.69 SSC categorization process within 90 days of issuance of the amendments. The resultant VEGP procedures will reflect the 10 CFR 50.69 processes as approved by the NRC and as documented in the NRC safety evaluation report associated with the subject license amendments.

2.0 DETAILED DESCRIPTION The proposed VEGP Units 1 and 2 OL LCs will allow for the voluntary implementation of 10 CFR 50.69. Implementation of 10 CFR 50.69 allows for application of a risk-informed categorization process per 10 CFR 50.69(c) to modify the scope of SSCs subject to special treatment requirements. Alternative treatments per 10 CFR 50.69(b)(1) and 10 CFR 50.69(d)(2) can then be applied consistent with the categorization of the SSCs. Implementation of 10 CFR 50.69 allows the licensee and NRC to better focus attention and resources on SSCs that have safety significance resulting in improved plant safety.

The VEGP 10 CFR 50.69 LAR conforms to scope requirements of 10 CFR 50.69(b)(2). The categorization process described in the LAR conforms to the guidance in NRC Regulatory Guide (RG) 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," Revision 1 dated May 2006 (Reference 3). The categorization process also conforms to the guidance in I\lEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0 dated July 2005 (Reference 4). SNC has determined that the proposed amendments do not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c).

E1-3 to NL-12-0932 Basis for Proposed Change

3.0 TECHNICAL EVALUATION

3.1 Categorization Process Description (10 CFR 50.69(b)(2)(i>>

3.1.1 Categorization Process Overview The process that SNC will use for categorization of SSCs into the four risk informed safety classes (RISC) defined in 10 CFR 50.69, specifically RISC-1, RISC-2, RISC-3, and RISC-4, will be specified in SI\JC procedures. The SNC categorization process conforms to the categorization process guidance provided in [\lEI 00-04, as endorsed by RG 1.201. Conformance of the SNC categorization process to RG 1.201 is documented in Appendix A of this enclosure. As indicated in Appendix A, no exceptions are taken to RG 1.201.

Per 10 CFR 50.69, once a system is selected for categorization, all the components associated with this system shall be categorized. A system has active and passive components. Consequently, NEI 00-04 provides guidance on categorizing active and passive components, which includes the involvement of a group of experienced and plant-knowledgeable personnel known as the Integrated Decision-making Panel (lOP).

For active components, the SNC procedures implement the active function categorization process defined in NEI 00-04 with clarifications as outlined in section 3.1.3 of this enclosure. The process includes the various quantitative and qualitative evaluations and sensitivity studies intended to provide reasonable confidence that the initial categorization is valid and that validity of the process is maintained.

Passive components are defined as SSCs having only a pressure retaining function. Active components can also have a passive or pressure retaining function. Therefore, the term "passive component" also refers to the passive function of active components, if applicable. NEI 00-04 states that passive component categorization should be performed using the guidance of ASME Code Case N-660 Revision 0, "Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities," (Reference 5) or subsequent versions of this guidance approved by ASME. RG 1.201 states that alternatives to this code case may be submitted for NRC review and approval as part of a specific 10 CFR 50.69 application. SNC has elected to use an alternate method for categorization of passive components. The alternate method was developed after NEI 00-04 was approved but is based on the EPRI risk-informed in-service inspection methodology EPRI TR-112657 Revision B-A (Reference 6) that is cited as an acceptable approach in NEI 00-04. This alternative method is conservative but still provides sufficiently realistic insights with regard to categorization of passive components.

A comparison was conducted of the SNC process for 10 CFR 50.69 passive component categorization and WCAP-16308-NP-A, "Pressurized Water Reactor Owners Group 10 CFR 50.69 Pilot Program - Categorization Process - Wolf Creek Generating Station," (Reference 7). The review included the main body of the WCAP and Appendices, responses to related NRC requests for additional E1-4 to NL-12-0932 Basis for Proposed Change information (RAls), and the I\JRC's associated safety evaluation report (SER)

(Reference 8). The results of this review are presented in the following two tables.

Table 1A compares Table 1 of the NRC SER for WCAP-16308-NP-A, titled "NRC Staff Position on Proposed Changes in ASME Code Case N-660 in TR WCAP 16308," and the SNC process for 10 CFR 50.69 passive component categorization . Table 1B presents a comparison of the guidance contained in Appendix A of WCAP-16308-NP-A, specifically pages A-18 through A-30, to the SNC process for 10 CFR 50.69 passive component categorization.

WCAP-16308-NP-A pages A-18 through A-30 provide the ASME Code Case N-660 Revision 0, as approved in 2002, marked up to reflect the NRC approved passive categorization process.

Additionally, some clarifications of the SI\JC process for 10 CFR 50.69 passive component categorization are provided below:

• Consistent with EPRI TR-112657 Revision 8-A, the SNC process requires that all safety functions supported by a system be completely evaluated as part of that system's categorization, while WCAP-16308-NP-A would allow an

'interim' categorization.

• Operator actions in the SNC process, when credited, need to meet the requirements of NRC approved EPRI TR-112657 Revision 8-A.

• A spectrum of break sizes needs to be evaluated in the SNC process and the one with the highest consequence rank used.

• The SI\JC process currently limits the application to Class 2 and 3 components (i.e., Class 1 is always high-safety-significant (HSS) for passive categorization) .

• The SNC process requires that all relevant configurations be assessed as part of the categorization process (e.g., Table 3-2 of EPRI TR-112657, Revision 8-A).

The overall categorization process guidance includes the involvement of a group of experienced and plant-knowledgeable professionals known as the lOP. The lOP is responsible for ensuring that an integrated decision-making process is employed that systematically considers the quantitative and qualitative information available regarding the various modes of plant operation and initiating events, including PRA quantitative risk results and insights (e.g., core damage frequency (CDF), large early release frequency (LERF), and importance measures), deterministic engineering insights (e.g., defense-in-depth, safety margins, and containment integrity); and other pertinent information (e.g., industry and plant-specific operational and performance experience, feedback, and corrective actions program) in the categorization of SSCs. The lOP is responsible for approving the final categorizations.

SNC procedures define the duties and responsibilities of the lOP members, the lOP evaluation, and the lOP overall assessment process as defined in NEI-00-04 without exception. When holding lOP meetings, the SNC procedure requires a quorum for the lOP. A quorum consists of at least five qualified persons, collectively having site specific expertise in the following functional areas: plant E1-5 to NL-12-0932 Basis for Proposed Change operations (senior reactor operator (SRO) qualified), safety analysis, design engineering, systems engineering, and PRA. The training and qualification of each lOP member is documented.

E1-6 to NL-12-0932 Basis for Proposed Change Table 1A: 10 CFR 50.69 Passive Component Categorization Process Comparison of NRC Staff Positions on WCAP-16308-NP-A Methods (Table 1 of NRC SER for WCAP-16308-NP-A)

And the SNC Passive Component Categorization Process

{N-660 Revision 0 Section} NRC Position NRC Mark-up of SNC Passive Component

[WCAP-16308 Section] WCAP-16308 Proposed Changes to Categorization Process ASME Code Case N-660 Revision 0 (additions in bold, deletions in strikeout, comments in italics)

{1-2.0} Objection The owner shall define the boundaries Conforms to NRC Mark-up.

[1-2.0] requiring included in the scope of the RISC qualification evaluation process subject to the SNC limits the application to Class 2 constraints in paragraph 50.59(c )(1 lev) and 3 only. This is conservative.

that the categorization must be performed for entire systems. Items optionally classified to Class 1 and Class 1 items connected to the reactor coolant pressure boundary, as defined in paragraphs 10 CFR 50.55a (c)(2)(i) and (c)(2)(ii), are within the scope of the RISC evaluation process. All other Class.1 items shall be classified High Safety Significant (HSS) and the provisions of the RISC evaluation shall not apply.

{1-3.0} Objection Additionally, information considered Conforms to NRC Mark-up. I

[1-3.0] requiring relevant to the classification shall be I I

qualification collected for each piping segment e.g., SNC procedure requires all initiating information regarding design basis events and operating modes to be I accidents, at-power risk, shutdown risk, evaluated. This is consistent with the I E1-7 to NL-12-0932 Basis for Proposed Change Table 1A: 10 CFR 50.69 Passive Component Categorization Process Comparison of NRC Staff Positions on WCAP-16308-NP-A Methods (Table 1 of NRC SER for WCAP-16308-NP-A)

And the SNC Passive Component Categorization Process

{N-660 Revision 0 Section} NRC Position NRC Mark-up of SNC Passive Component

[WCAP-16308 Section] WCAP-16308 Proposed Changes to Categorization Process ASME Code Case N-660 Revision 0 (additions in bold, deletions in strikeout, comments in italics) containment isolation, flooding, fires, EPRI TR-112657 SEA.

seismic conditions, etc.). Consistent with 50.69(c)(1)(ii), the classification must address all initiating events and plant operating modes.

{ } Objection Entire proposed section should be Conforms to NRC Mark-up. I

[1-3.1.1 (a)(4)] requiring deleted I

qualification The applicable SNC procedure I

requires that breaks from small to large are postulated and the break size with the highest consequence rank be used. This is consistent with the EPRI TR-112657 SEA.

{1-3.1.2} Objection " .in accordance with (a) through (d) Clarification to NRC Mark-up.

[1-3.1.2] requiring below. In assessing the appropriate qualification consequence category, risk information Consistent with the EPRI TR-112657 for all initiating events, including fire and SER, the applicable SNC procedure seismic, should be considered. To requires all initiating events, capture the risk importance from operating modes and external E1-8

Enclosure 1 to NL-12-0932 Basis for Proposed Change Table 1A: 10 CFR 50.69 Passive Component Categorization Process Comparison of NRC Staff Positions on WCAP-16308-NP-A Methods (Table 1 of NRC SER for WCAP-16308-NP-A)

And the SNC Passive Component Categorization Process

{N-660 Revision 0 Section} NRC Position NRC Mark-up of SNC Passive Component

[WCAP-16308 Section] WCAP-16308 Proposed Changes to Categorization Process ASME Code Case N-660 Revision 0 (additions in bold, deletions in strikeout, comments in italics) initiating events for which no hazards to be evaluated for each quantitative PRA is available, any individual system application.

piping segment supporting a safe Experience has shown that these shutdown pathway would be classified evaluations, which complete the risk as HSS. assessment process, identify portions of the respective system that are not high from an at-power perspective, but can be important from a shutdown perspective or their support of other systems (e.g., Emergency Core Cooling Systems during internal events or external hazards).

{1-3.1.2(b)(3)} No objection In lieu of Table 1-2, quantitative indices Clarification to NRC Mark-up.

[J-3.1.2(b)] may be used to assign consequence I categories in accordance with Table 1-5. Per the applicable SNC procedure, quantitative indices may be used to assign consequence categories in accordance with Table 5 in lieu of Table 2 provided the quantitative basis of Table 2 (e.g., one full train E1-9 to NL-12-0932 Basis for Proposed Change Table 1A: 10 CFR 50.69 Passive Component Categorization Process Comparison of NRC Staff Positions on WCAP-16308-NP-A Methods (Table 1 of NRC SER for WCAP-16308-NP-A)

And the SNC Passive Component Categorization Process

{N-660 Revision 0 Section} NRC Position NRC Mark-up of SNC Passive Component

[WCAP-16308 Section] WCAP-16308 Proposed Changes to Categorization Process ASME Code Case N-660 Revision 0 (additions in bold, deletions in strikeout, comment~ in italics) unavailability approximately 10"2, exposure time) is consistent with the failure scenario being evaluated.

Differences in the consequence rank between the use of Table 2 and 5 shall be reviewed, justified and documented or the higher consequence rank assigned. This is further clarification of the technical intent of the methodology and is consistent with or more conservative than the NRC pOSition.

E1-10 to NL-12-0932 Basis for Proposed Change Table 1A: 10 CFR 50.69 Passive Component Categorization Process I

Comparison of NRC Staff Positions on WCAP-16308-NP-A Methods (Table 1 of NRC SER for WCAP-16308-NP-A)

And the SNC Passive Component Categorization Process

{N-660 Revision 0 Section} NRC Position NRC Mark-up of SNC Passive Component

[WCAP-16308 Section] WCAP-16308 Proposed Changes to Categorization Process ASME Code Case N-660 Revision 0 I (additions in bold, deletions in I strikeout, comments in italics)

{1-3.1.3(a)(3)} Objection Even when taking credit for plant features Clarification to NRC Mark-up.

[1-3.2.2(b)(1 )] requiring and operator actions, failure of the piping qualification segment will not directly fail anothor high Consistent with the EPRI TR-112657 safety-significant function. SER, the applicable SNC procedure uses the term "basic safety function" (e.g., reactivity control, core cooling, heat sink, RCS inventory) instead of "high safety significant function" .

This term is more complete and/or conservative.

{1-3.1.3(a)(4)} Objection Failure of the piping segment will not Clarification to NRC Mark-up.

[1-3.2.2(b)(2)] requiring result in failure of anothor high safety qualification significant piping segment, e.g. through Consistent with the EPRI TR-112657 indirect effects SER, the applicable SNC procedure uses the term "basic safety function" (e.g. reactivity control, core cooling, heat sink, RCS inventory) instead of "high safety significant function".

This term is more complete and/or conservative.

E1-11 to NL-12-0932 Basis for Proposed Change

- - ~

Table 1A: 10 CFR 50.69 Passive Component Categorization Process Comparison of NRC Staff Positions on WCAP-16308-NP-A Methods (Table 1 of NRC SER for WCAP-16308-NP-A)

And the SNC Passive Component Categorization Process

{N-660 Revision 0 Section} NRC Position NRC Mark-up of SNC Passive Component

[WCAP-16308 Section] WCAP-16308 Proposed Changes to Categorization Process ASME Code Case N-660 Revision 0 (additions in bold, deletions in strikeout, comments in italics)

{1-3.1.3(a)(5)} Objection Consideration changed and moved to Clarification to NRC Mark-up.

[1-3.2.2(b)(3)] requiring new section 1-3.2.2(b)(3), eVeR !NAeR qualification takiRg credit for plaRt features aRd Operator actions can be credited operator actioRs, failure of the piping provided they are evaluated in segment will not prevent or adversely accordance with EPRI TR-112657 atfeGt the plant's capability to reacA OF Revision B-A. This is consistent with maiRtaiR reaching or maintaining safe N-660 Revision 0 and the EPRI shutdown conditions. TR-112657 SER.

{1-3.1.3(b)(3)} Objection e\!eR !NAeR takiRg credit for plaRt features Clarification to NRC Mark-up.

[1-3.2.2(b)(6)] requiring aRd operator actioRs, f Failure of the qualification piping segment will not result in releases Operator actions can be credited of radioactive material that would result in provided they are evaluated in the implementation of off-site emergency accordance with EPRI TR-112657 response and protective actions. Revision B-A. This is consistent with N-660 Revision 0 and the EPRI TR-112657 SEA.

{1-3.2.2(b)} No objection The plant condition monitoring would Clarification to NRC Mark-up.

[1-3.2.2(b)(5)) identify any known active degradation mechanisms in the pipe segment prior to While condition monitoring programs E1-12

Enclosure 1 to NL-12-0932 Basis for Proposed Change Table 1 A: 10 CFR 50.69 Passive Component Categorization Process I Comparison of NRC Staff Positions on WCAP-16308-NP-A Methods (Table 1 of NRC SER for WCAP-16308-NP-A)

And the SNC Passive Component Categorization Process

{N-660 Revision 0 Section} NRC Position NRC Mark-up of SNC Passive Component I

[WCAP-16308 Section] WCAP-16308 Proposed Changes to Categorization Process ASME Code Case N-660 Revision 0 J (additions in bold, deletions in strikeout, comments in italics) its failure in test or an actual demand certainly exist, "monitoring" is not an event (e.g., flow accelerated corrosion element of the categorization process program). and need not be addressed in categorization guidance. This wording was originally added to various draft revisions of N-660 to allow for the postulation of small or medium break sizes instead of large breaks. The applicable SNC procedure requires that the full spectrum of break sizes be evaluated and the one with the highest consequence be used.

E1-13

Enclosure 1 to NL-12-0932 Basis for Proposed Change Table 1B: 10 CFR 50.69 Passive Component Categorization Process Comparison of the NRC Approved Passive Categorization Process (Mark-up of ASME Code Case N-660 Revision 0)

And the SNC Passive Component Categorization Process WCAP-16308-NP-A Section Topic SNC Passive Component Categorization Page Process A-18 1100 Scope WCAP-16308-NP-A allows process to be SNC limits the application to Class 2 and 3 applied to Class 1 components only. This is conservative.

A-18 1320 Required WCAP-16308-NP-A has requirements for lOP SNC has a standalone procedure for the lOP.

Disciplines A-19 9000 Glossary WCAP-16308-NP-A has a Glossary The applicable SNC procedure has similar information.

A-21 1-2.0 Scope WCAP-16308-NP-A has wording as follows: SNC limits the application to Class 2 and 3 Identification only. This is conservative.

The owner shall define the boundaries included in the scope of the RISC evaluation process subject to the constraints in paragraph SO.S9(c){1){v) that the categorization must be performed for entire systems. Items optionally classified to Class 1 and Class 1 items connected to the reactor coolant pressure boundary, as defined in paragraphs 10 CFR SO.SSa (c)(2)(i) and I (c){2)(ii), are within the scope of the RISC evaluation process. All other Class.1 items shall be classified High Safety Significant E1-14

Enclosure 1 to NL-12-0932 Basis for Proposed Change J

Table 1B: 10 CFR 50.69 Passive Component Categorization Process Comparison of the NRC Approved Passive Categorization Process (Mark-up of ASME Code Case N-660 Revision 0)

And the SNC Passive Component Categorization Process

. WCAP-16308-NP-A Section Topic SNC Passive Component Categorization Page Process (HSS) and the provisions of the RISC evaluation shall not apply A-22 1-3.1 .1(a) WCAP-16308-NP-A allows leak before break SNC does not which is conservative.

to support the small break assumption A-22 1-3.1.1 (b) and WCAP-16308-NP-A allows operator actions to Operator actions can only be credited (e) be credited but no criteria or requirements are provided they are evaluated in accordance I

identified with EPRI TR-112657 Revision B-A. This is consistent with the EPRI TR-112657 SER. I I

I A-22 1-3.1.1 WCAP-16308-NP-A does not require that Based upon lessons learned from application :

success criteria diagrams be developed for of the methodology, SNC's applicable relevant initiating events procedure requires that success criteria diagrams be developed for relevant initiating events.

A-22 1-3.1.1 WCAP-16308-NP-A does not require that Based upon lessons learned from application each of the relevant operating configurations of the methodology, SNC's applicable identified in EPRI TR-112657, Rev B-A, procedure requires that this be done. This (Chapter 3.3.3) be evaluated provides for a more comprehensive and comQlete assessment. - -

E1-15 to NL-12-0932 Basis for Proposed Change Table 18: 10 CFR 50.69 Passive Component Categorization Process Comparison of the NRC Approved Passive Categorization Process (Mark-up of ASME Code Case N-660 Revision 0)

And the SNC Passive Component Categorization Process WCAP-16308-NP-A Section Topic SNC Passive Component Categorization Pag_e Process A-22 1-3.1.2 WCAP-16308-NP-A wording: SNC applicable procedure has requirements consistent with the EPRI TR-112657 To capture risk importance from initiating Revision B-A SER to evaluate the impact of events for which no quantitative PRA is shutdown events and external hazards.

available, any piping segment supporting a safe shutdown pathway would be classified as HSS.

A-22 1-3.1.2(a) WCAP-16308-NP-A has no requirement to SNC does which provides for a more investigate differences in the consequence comprehensive and complete assessment.

rank in the use of Table 1-1 and 1-5.

A-23 1-3.1.2(b) WCAP-16308-NP-A has no requirement to SNC does which provides for a more investigate differences in the consequence comprehensive and complete assessment.

rank in the use of Table 1-2 and 1-5.

A-23 1-3.1.2(b) WCAP-16308-N P-A has no requirement that SNC does which provides for a more robust for defense in depth purposes, all postulated consequence ranking.

failures leading to "zero defense" (i.e., no backup trains) shall be assigned a high consequence rank.

A-23 1-3.1.2(c) WCAP-16308-NP-A has no requirement to SNC does which provides for a more investigate differences in the consequence comprehensive and complete assessment.

E1-16 to NL-12-0932 Basis for Proposed Change Table 18: 10 CFR 50.69 -Passive Component Categorization Process Comparison of the NRC Approved Passive Categorization Process (Mark-up of ASME Code Case N-660 Revision 0)

And the SNC Passive Component Categorization Process WCAP-16308-NP-A Section Topic SNC Passive Component Categorization Page Process rank in the use of Table 1-3 and 1-5.

A-24 1-3.2.2(b) WCAP-16308-NP-A allows operator actions to Operator actions can only be credited be credited but no criteria or requirements are provided they are evaluated in accordance identified. with EPRI TR-112657 Revision B-A. This is consistent with the EPRI TR-112657 SER.

A-24/A-25 1-3.2.2(b) and "Additional Considerations" - wording is While condition monitoring programs certainly (c) generally consistent except for editorial exist, "monitoring" is not an element of the changes. WCAP-16308-NP-A also contains categorization process and need not be the following: addressed in categorization guidance. This wording was originally added to various draft The plant condition monitoring program would revisions of N-660 to allow for the postulation identify any known active degradation of small or medium break sizes instead of mechanisms in the pipe segment prior to its large breaks. The applicable SNC procedure failure in test or an actual demand event (e.g., requires that the full spectrum of break sizes flow accelerated corrosion program). be evaluated and the one with the highest consequence rank be used.

A-30 Table 1-5 Use of the equal sign is inconsistent with EPRI SNC applicable procedure consistent with the TR-112657 Revision B-A. EPRI TR-112657 Revision B-A SER N/A N/A WCAP-16308-NP-A does not contain a table The SNC applicable procedure has such a identifying relevant operating configurations table. This provides for a more for postulating when a piping segment fails. comprehensive and complete assessment.

E1-17 to NL-12-0932 Basis for Proposed Change 3.1.2 Categorization Process Key Steps The key steps of the SNC 10 CFR 50.69 categorization process for active and passive components, including lOP deliberation, are provided below.

• The SNC 50.69 categorization process utilizes a combination of deterministic engineering (e.g., defense-in-depth, safety margins, and containment integrity), quantitative PRA, and qualitative risk insights to determine if a SSC performs one or more HSS functions and identifies those functions. The term "HSS function" is synonymous with the term "safety-significant function", as defined in 10 CFR 50.69.

• The process is applied only to those plant systems or structures (herein referred to simply as system) that VEGP chooses to categorize.

• Once a plant system is selected for categorization, all components within that system are categorized into one of the following categories:

• RISC-1: safety-related SSCs that perform HSS functions

• RISC-2: nonsafety-related SSCs that perform HSS functions

• RISC-3: safety-related SSCs that perform low safety-significant (LSS) functions

• RISC-4: nonsafety-related SSCs that perform LSS functions

• The identification of components within a system is based on the VEGP Plant Data Management System (POMS) which is a design-controlled database that reflects the current plant configuration.

• The functions performed by the system are identified, including but not necessarily limited to, design basis functions, maintenance rule functions, and functions credited for mitigation and prevention of severe accidents.

• Plant and industry operating experience with the system being categorized is obtained and evaluated for applicability and insights.

• System functions are qualitatively categorized using the set of deterministic questions identified in NEI 00-04.

• For each component in the system, the function(s) that the component supports are identified.

• Component functional importance is determined through an integrated, systematic process that considers all of the following factors:

• Internal Events (including Internal Flooding) At-Power - the VEGP Internal Events (including Internal Flooding) PRA models severe accident scenarios resulting from internal initiating events, including internal flooding, occurring at full power operation. Importance measures related to COF and LERF are used to identify SSCs that are HSS with respect to internal events.

E1-18 to NL-12-0932 Basis for Proposed Change

• Fire Hazards - the VEGP Fire PRA models severe accident scenarios resulting from fire-initiated events, occurring at-power operation.

Importance measures related to CDF and LERF are used to identify SSCs that are HSS with respect to fire events.

• Seismic hazards - Seismic hazards are assessed using the VEGP Seismic Margins Analysis (SMA) developed for the Individual Plant Examination of External Events (I PEEE) (References 9 and 10). As stated in NEI 00-04, components that are credited as part of the seismic safe shutdown path are considered HSS with respect to seismic risk.

• Other External hazards - Other external hazards, such as high winds, external flooding, and accidents from transportation routes and nearby industrial facilities, are assessed using the results of the VEGP IPEEE.

The other external hazards evaluation was updated in 2011 using guidance provided in ASMEIANS RA-Sa-2009 (Reference 11). The evaluation included the impact of construction activities at Vogtle Units 3 and 4. The evaluation concluded that the other external hazards screened out. Therefore, components are LSS with respect to risk from other external events.

• Low Power/Shutdown Risk -In the absence of low power or shutdown PRA risk models, the low power risk is evaluated using at-power PRA for those cases where the at-power PRA model is valid for evaluating risk for low power conditions. In addition, in those cases where at-power PRA risk model is not appropriate for use at low power conditions, the low power risk will be evaluated by qualitatively assessing the function(s) performed by the system as outlined in NEI 00-04 section 9.2.2 review of risk information item 6. A component is assigned a preliminary risk, w~lich is higher of these two risks where applicable. For shutdown risk, VEGP has performed and maintains a Shutdown Defense-In-Depth Evaluation, in accordance with the NUMARC 91-06 Program (Reference 12).

Components that playa primary or alternate role in accomplishing key shutdown safety functions or whose failure would initiate a shutdown event (e.g., loss of shutdown cooling, drain down, etc.) are identified as HSS with respect to shutdown risk.

• Sensitivity Studies - For PRA-modeled components, several sensitivity studies are performed as outlined in NEI 00-04 to ensure that assumptions in the PRA are not masking the importance of an SSC.

E1-19 to NL-12-0932 Basis for Proposed Change

• Passive Component Categorization - For the purpose of 10 CFR 50.69 categorization, passive components are those components that have a pressure retaining function. Passive components and the passive function of active components are evaluated through a process that utilizes the guidance in EPRI TR-112657, Revision B-A, with the following additional constraints:

o Component failure is assumed and only the consequence evaluation is performed .

o Additional deterministic considerations (e.g., defense in depth, safety margins) are applied.

o ASME Class 1 components are, by default, categorized as HSS with respect to passive risk.

o Component supports are assigned the same safety significance as the highest passively ranked component within the bounds of the associated analytical pipe stress model.

• Active components are assigned a qualitative risk based on the system function(s) they support. If a component supports one or more HSS functions, it is considered to be HSS initially. A component may be assigned a qualitative risk of LSS, even it supports an HSS function , if a credible failure of the component would not preclude the fulfillment of the HSS function (e.g., locked open valve).

• An active component is assigned an overall categorization of HSS if one or more of the following evaluations identify that component as HSS:

integrated results of Internal Events (including Internal Flooding) PRA and Fire PRA assessment, seismic hazards, other external hazards, shutdown risk, and qualitative risk based on the system function(s). A passive component is categorized as HSS per processes and criteria discussed herein. In addition, an active component having an HSS passive categorization is categorized as HSS.

• Components that are still LSS are then evaluated for their role in providing defense-in-depth and, if appropriate, upgraded to HSS.

• Overall Risk Sensitivity Study - For PRA-modeled components, an overall risk sensitivity study is used to confirm that the categorization process results in acceptably small increases to CDF and LERF. The overall risk sensitivity study is performed using guidance outlined in NEI 00-04. SNC will use a factor of 3 to increase unreliability and unavailability of RISC-3 LSS components modeled in PRAs.

• For components that are HSS or that are LSS but support HSS functions, the associated critical attributes are identified.

• The above categorization results are presented to the lOP for review and approval. The VEGP lOP is staffed with expert, plant-knowledgeable, and process-trained members whose collective expertise includes, at a minimum, E1-20 to NL-12-0932 Basis for Proposed Change PRA, safety analysis, plant operations (SRO qualified), design engineering, and system engineering. The lOP will review the categorization results and make the final determination on the safety significance of system functions and components, in accordance with the following requirements:

• The basis for the adequacy of the PRA results; the results of non-modeled hazards; and the system functions and the basis for their categorization are reviewed to ensure adequacy in supporting the process.

• The detailed categorization results, including results of sensitivity studies, and defense-in-depth considerations are reviewed for completeness and adequacy.

• Components that are HSS from one of the following evaluations cannot be categorized as LSS, although the lOP may request that further clarification and/or analysis be performed and brought back to the lOP.

o Internal Events PRA o Non-PRA evaluations of seismic, other external events, or shutdown risk hazards.

o Passive categorization

• Components that are HSS only from the Fire PRA may be categorized as LSS if the integrated assessment of component risk importances over all PRA models shows the integrated risk importance measures meet the LSS criteria.

• The lOP may change the categorization of a component 'from LSS to HSS based on its assessment and use of conservative decision-making.

Conversely, as outlined in NEI 00-04, the lOP may re-categorize components from HSS to LSS if a credible failure of the component would not preclude the fulfillment of the HSS function and if the component did not meet one of the previously stated criteria precluding LSS categorization .

• Once the safety significance of the components in the selected system has been approved by the lOP, the components will be categorized into one of the four Rise categories as detailed at the beginning of this section 3.1.2.

• Periodic reviews are conducted at least once every two Unit 1 refueling outages to ensure continued validity and performance monitoring for those SSCs that have been categorized.

• The categorization results, supporting bases, lOP meeting minutes, and results of periodic reviews are documented as Quality Assurance records and maintained for the life of the plant.

E1-21 to NL-12-0932 Basis for Proposed Change 3.1.3 SNC Conformance to NEI 00-04 In summary, the SNC 10 CFR 50.69 process to be implemented at VEGP addresses all aspects of the guidance provided in NEI 00-04 as endorsed by RG 1.201, with the following clarifications.

Clarification 1: Approach Used to Risk Rank System Functions:

The NEI 00-04 process uses component risk significance (determined using the methodology outlined in sections 5.0 and 6.0) to determine function risk significance (section 7.0). For example, if a component identified by PRA as HSS is mapped to a function, then the function is considered HSS. Subsequently, all the components mapped to this function are preliminary categorized as HSS.

NEI-00-04 section 9.2 states that "The lOP approval of the categorization of system functions, based on the coarse mapping of components to system functions, would be used to define the safety significance of each SSC as described in Section 10. Thus, if a system function is found to be safety-significant by the lOP, then all components required to support that function would initially be considered safety-significant. If a more detailed categorization of the SSCs associated with a safety-significant function is performed after the initial lOP, then the basis for that re-categorization must be considered in a follow-up lOP session. In this follow-up session, the lOP would be expected to review the basis for the re-categorization and to assess the impact of this re categorization on the risk importance and defense in depth implications using the same criteria as in the original lOP session for candidate low safety significant SSCs."

NEI 00-04 section 10.2 provides guidance on performing detailed categorization of the SSCs and the criteria for assignment of low safety significance for an SSC supporting a safety-significant function. Consistent with section 9.2, section 10.2 further states that "For SSCs that are re-categorized to a lower classification (e.g., components in a safety-significant function that are determined to be LSS based on the above considerations), the new categorization and its basis should be presented to another session of the lOP to be re-categorized using the same rigor as described in Section 9. If the SSCs being considered for re categorization to a lower classification are modeled in the PRA, then the risk sensitivity described in Section 5 would need to be completed prior to presentation to the lOP."

The following is a clarification of the way SNC has applied the above mentioned steps when performing a trial categorization of three systems.

SNC evaluates each function using the seven questions outlined in section 9.2.2 of NEI 00-04. If any of the seven questions is answered "yes" for a function, the function is considered HSS. The components mapped to the functions are considered preliminary HSS. This is called "qualitative assessment".

E1-22 to NL-12-0932 Basis for Proposed Change The component safety significance assessment is performed using guidance provided in section 5.0 of NEI 00-04. This portion of the assessment is called "Active Risk Assessment".

The passive risk for applicable components is determined through the process outlined in this LAR. This portion of the assessment is called "Passive Risk Assessment" .

A component is assigned a preliminary risk, which is the highest of the following three risks - qualitative assessment, active risk assessment, and passive risk assessment. Then the defense-in-depth and safety margin analysis are performed using the guidance provided in section 6.0 of NEI 00-04. The preliminary risk of a component is adjusted, as needed, to reflect insights gained from defense-in-depth and safety margin analysis. The overall risk sensitivity study is performed per section 8.0 of NEI 00-04.

The categorization results are provided to the lOP for their review per section 9.0 of NEI 00-04. The lOP uses information in section 9.0 and 10.2 to evaluate the categorization results and disposition them appropriately.

The SNC process meets the guidance of I\lEI 00-04 in the manner described herein. SNC utilizes all the sections of the NEI 00-04 methodology. The sequence in which the sections (sections 5.0, 6.0, 9.2.1, 9.2.2, 9.2.3, and 10.2) are applied varies without compromising the fidelity of the methodology. The SNC process provides a more logical and intuitive top-down approach for the qualitative portion of the assessment. System functions are categorized first, followed by component-to-function mapping, and ending with component categorization. The SNC process improves the overall process efficiency while applying all the elements of the NEI 00-04 methodology.

Clarification 2: Mapping of Truly Passive Components to Each Function:

In section 4.0, under "Identification of System Functions", NEI 00-04 states that the classification of SSCs having only a pressure retaining function (also referred to as passive components) or the passive function of active components should be performed using passive component categorization methodology.

In the same section 4.0, under "Coarse Mapping of Components to Functions",

I\lEI 00-04 states that The assignment of SSCs to each of the functions is necessary at this step to ensure that every SSC with a tag identifier for the system being considered is represented in at least one of the functions." SNC complies with this step with the following clarification. The insights obtained from the trial categorization of three systems indicate that truly passive components like pipes and thermowells should be assigned to a pressure retention function only. These types of components do not perform an active function. During the trial categorization, SNC observed that when these components were mapped to various functions (as currently stated in NEI 00-04), some of these components were determined to be candidate HSS only because they were in the flow path of safety significant function(s) . However, when the lOP review was performed per E1-23 to NL-12-0932 Basis for Proposed Change section 9.0 and 10.2 of NEI 00-04, their safety significance was ultimately determined using the results of the passive component categorization methodology. The current verbiage (quoted above) in NEI 00-04 does not allow this exclusion.

SNC takes the position that by mapping truly passive components only to the passive function, the methodology becomes more efficient without compromising the fidelity of the process.

Clarification 3: Components in the Flow Path vs. Components Required for a Function(s):

On page 28, NEI 00-04 states that "Define the pathway associated with each function and then define the components associated with that pathway." SI\JC complies with this step with the following clarification.

The current wordings in NEI 00-04 imply that each and every component in the pathway of a function should be mapped to the function. However, the insights obtained from the trial categorization indicate that there will be instances when select components will be in the function pathway, but these components are not required to perform the function. Consider the following two examples.

• When categorizing a system, functions performed by the system are identified. One of the functions performed by the Chemical and Volume Control System (CVCS) is to maintain primary coolant inventory during normal operations, startup, and shutdown (includes operation in support of accident response when restoring CVCS inventory control). When a pathway is traced on piping and instrumentation diagrams (P&ID) for this function (i.e.,

maintain primary coolant inventory), the pathway includes components associated with demineralizer beds. However, these components are not required to perform this function other than a passive role of pressure retention to maintain a flow path. When categorizing a system, the risk associated with the pressure retention function is captured in a separate function (e.g., maintain CVCS pressure boundary during normal operations, startup, and shutdown). Therefore, it is not necessary to map the components associated with demineralizer beds to this function (i.e., maintain primary coolant inventory). Note that the components associated with demineralizer beds are required for other functions (such as control primary coolant pH during normal operations, startup, and shutdown; limit primary coolant radioactivity levels during normal operations, startup, and shutdown; etc.); therefore, they are mapped to these functions.

• Virtually all mechanical systems have drain, test, and vent valves. These components are there to support a specific maintenance or test function.

However, they will show up in the pathway for other functions performed by a system being categorized. The failure of these components will not impact other system functions. Assuming flow diversion does not affect specific function, it is not necessary to map these components to the other functions identified for the system being categorized.

E1-24 to NL-12-0932 Basis for Proposed Change Therefore, SNC takes the position that only components that are required to perform an active function will be mapped to that function. This makes the methodology more efficient without compromising the fidelity of the process.

3.2 Technical Adequacy Evaluation (10 CFR 50.69{b){2){ii>>

SNC employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for VEGP. This approach includes both a proceduralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews.

The NRC-endorsed !\lEI 00-04 methodology also allows for categorization of SSCs to be performed in the absence of quantifiable PRA models for the following other risks - seismic hazards, other external events hazards (e.g.,

tornados, high winds, transportation, etc.), and low power/shutdown hazards.

The following sections describe the approach used to meet the requirements of 10 CFR 50.69(b)(2)(ii).

3.2.1 Internal Events with Internal Flooding PRA Model VEGP has two operating units of Westinghouse 4-loop pressurized light water reactors with a core thermal power rating of 3625 MWt. The VEGP baseline PRA models CDF and LERF due to internal events, including internal flooding, at power. There is one model for VEGP because each unit is a mirror image of the other from a PRA perspective. The model has been significantly upgraded since the IPE PRA model and reflects the as-built, as-operated plant. The baseline PRA is a large fault tree methodology model, using the EPRI R&R Workstation (CAFTA) software and FTREX software for quantification. The baseline CDF and LERF values are 2.25E-05/yr and 7.37E-08/yr, respectively.

3.2.1.1 PRA Maintenance/Updates and Application to 10 CFR 50.69 Categorization Process The administrative controls applicable to the PRA models used to support the 10 CFR 50.69 program ensure that these models reflect the as-built, as-operated plant. Plant changes, including physical modifications and procedure or operating practice changes, are reviewed prior to implementation to determine if they could impact the PRA models. If so, the process then determines the quantitative significance of the change and, if appropriate, implements the PRA model change concurrently with the plant change. Otherwise, the PRA model change is prioritized for implementation at a routine model update. Such pending changes are considered when evaluating other changes until they are fully implemented into the PRA models. Routine updates are performed as a minimum every two refueling cycles.

E1-25 to NL-12-0932 Basis for Proposed Change Internal Events (including Internal Flooding) PRA Maintenance and Update The SNC risk management process ensures that the applicable PRA model reflects the as-built and as-operated plant for each of the VEGP units. The process delineates the responsibilities and guidelines for updating the full power internal events and internal fire PRA models at all operating SI\lC sites, and includes both regularly scheduled and interim PRA model updates. The process includes provisions for monitoring potential impact areas affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operational experience), for assessing the risk impact of unincorporated changes, and for controlling the model and associated computer files.

Review of Plant Changes for Incorporation into the PRA Model- Refueling and Non-Refueling Outage Period (1) Plant changes (including both physical modifications to the facility and changes to procedures or operating practices) are reviewed as follows:

a. Refueling Outage Cycle Implementation: Six months prior to each refueling outage, changes slated for implementation during the refueling outage are reviewed to determine if any plant changes planned for the refueling outage could have an impact on the PRA models used to support 10 CFR 50.69 categorization.

b. Non-Refueling Outage Cycle Implementation: Changes slated for implementation during non-refueling outage periods are monitored concurrently with the planning phase to determine if changes planned for implementation prior to the next refueling outage could have an impact on the PRA models used to support 10 CFR 50.69 program risk calculations.

(2) When considered along with other changes, plant changes are categorized as "Required" if they are considered to have a quantitative significant impact.

"Required" plant changes are incorporated in the model to coincide with the time that the changes are implemented at the plant (refueling outage or non refueling outage period). Plant changes that do not meet the criteria for "Required" are assigned to other categories (i.e., "High Priority", "Moderate Priority" , and "Low Priority") and are incorporated at the next regular update, or in a future update, based on the priority category assigned. Accordingly, these changes are placed into a "Pending" file, which may be maintained in a database, spreadsheet, or other appropriate tool. When new changes are being evaluated (as discussed above) these pending changes are also considered with the newly proposed changes for their cumulative effect in order to determine the appropriate schedule and priority of PRA model changes.

(3) PRA updates, including updates to reliability data, failure data, initiating events frequency data, human reliability data, and other PRA inputs, are performed at least once every two Unit 1 refueling cycles.

E1-26 to NL-12-0932 Basis for Proposed Change (4) If PRA model errors are discovered, they are reviewed within the SNC Corrective Action Program to determine the quantitative impact on PRA results. A similar prioritization process to the plant change evaluation process is then applied to determine the appropriate priority and schedule for correcting the error, based on the significance of the quantitative impact.

(5) When a PRA model change is required but cannot be immediately implemented for a significant plant change or model error, either:

a. Alternative analyses to conservatively bound the expected risk impacts of the changes are performed. In such a case, these alternative analyses become part of the 10 CFR 50.69 categorization process until the plant changes are incorporated into the PRA model during the next update.

b. Appropriate administrative restrictions, if any, on the use of the categorization results are put in place until the model changes are completed.

3.2.1.2 Identification and Disposition of 10 CFR 50.69 Application Specific Sources of Uncertainty for Internal Events (including Internal Flooding)

PRAModel The purpose of this section is to identify and disposition the impact of epistemic uncertainty from the internal events (including internal flooding) PRA model on the 10 CFR 50.69 program. The baseline internal events (including internal flooding) PRA model documents assumptions and sources of uncertainty. These have been reviewed during the model peer review. Therefore, in order to disposition the impact of epistemic uncertainty from the internal events (including internal flooding) PRA model on the 10 CFR 50.69 program, the approach taken is to review these documents to identify those items which may be directly relevant to the 10 CFR 50.69 program, to perform sensitivity analyses where appropriate, to discuss the results, and to provide a disposition for the 10 CFR 50.69 program.

Background:

NEI 00-04 outlines a set of sensitivity studies that must be performed when using the PRA models for different risk hazards. NEI 00-04 requires the following sensitivity studies that evaluate the potential impact on the categorization results of several defined areas of uncertainty when using the internal events (including internal flooding) PRA.

• Increase all human error basic events to their 95th percentile value

• Decrease all human error basic events to their 5th percentile value

• Increase all component common cause events to their 95th percentile value

• Decrease all component common cause events to their 5th percentile value

• Set all maintenance unavailability terms to 0.0

• Perform any applicable sensitivity studies identified in the characterization of PRA adequacy E1-27 to NL-12-0932 Basis for Proposed Change The first five sensitivity studies are self-explanatory. Therefore, they are not discussed further in this section. The discussion in this section is focused on the requirement to perform: "Any applicable sensitivity studies identified in the characterization of PRA adequacy". In order to determine which, if any, such studies are needed, it is first necessary to evaluate the model and application specific sources of uncertainty.

Methodology for Identifying 10 CFR 50.69 Application Speci'fic Sources of Uncertainty for Internal Events (including Internal Flooding) PRA Model:

The baseline PRA model uncertainty report was developed using guidance provided in NUREG-1855 (Reference 13). As described in NUREG-1855, sources of uncertainty include "parametric" uncertainties, "modeling" uncertainties, and "completeness (or scope and level of detail)" uncertainties.

Parametric uncertainty was addressed as part of the VEGP aleatory uncertainty analysis as documented in Reference 14.

Assumptions are made during the PRA development as a way to address a particular modeling uncertainty because there is not a single definitive approach.

The assumptions are defined consistent with the definition provided in NUREG 1855. Plant-specific assumptions made for each of the VEGP internal events PRA technical elements are noted in the individual PRA notebooks. These assumptions were collected from each notebook and were evaluated to determine if they were related to a source of modeling uncertainty. A source of uncertainty having a potential impact on the 10 CFR 50.69 application was retained for further evaluation.

EPRI TR-1 016737 <<(Reference 15, Table A-3) compiled a listing of generic sources of modeling uncertainty to be considered for each PRA technical element. An evaluation of each generic source of modeling uncertainty was performed.

Completeness uncertainty addresses scope and level of detail. Uncertainties associated with scope and level of detail are documented in the PRA but are only considered for their impact on a specific application.

Disposition of 10 CFR 50.69 Application Specific Sources of Uncertainty for Internal Events (including Internal Flooding) PRA Model:

From the characterization of potential sources of uncertainty in the baseline PRA model and of supplementary issues from EPRI TR-1 016737 (Reference 15 Table A-3), the following items were identified as having a potentially important impact on the internal events (including internal flooding) PRA results.

1. Pressure-Induced SGTR

2. Seasonal Impacts on Initiating Events

3. Basis for human error probabilities (HEPs)

E1-28 to NL-1L.-u,ClUL.

Basis for Proposed The potential impacts of each of the above three sources of uncertainty on the 10 CFR 50.69 application were evaluated to determine jf any of them should be retained as part of standard sensitivity studies performed for the internal events (including flooding) PRA model.

2 performed to assess of these sou rces of model uncertainty on 10 CFR 50.69 application. of evaluation indicated that epistemic sources of would not an impact on the 10 application.

to NL-12-0932 Basis for Proposed Change Table 2: VEGP Internal Events (including Internal Flooding) PRA Model Sources of Uncertainty and Impact on 10 CFR 50.69 Program Source of Epistemic Related Assumptions Sensitivity Case Sensitivity Disposition of Uncertainty Conclusions Conclusions Pressure-Induced SGTR:

High reactor coolant system Scenarios with significant Sensitivity cases were Several scenarios (RCP No explicit (RCS) pressure impacts the reactor coolant pump (RCP) performed by reclass ifying seal leak, SORV, etc.) sensitivity potential for induced SGTR. seal leakage, a stuck open several scenarios (RCP were reclassified from evaluation is pressurizer valve, or a seal leak, stuck-open relief high pressure to low needed for 50.69 Only the Large loss of coolant pressurizer power-operated valve (SORV), etc.) from pressure and LERF re impact.

accident (LOCA), Reactor relief valve (PORV) open for high pressure to low evaluated. There was no vessel rupture, and the feed and bleed are pressure to determ ine change in LERF.

Medium LOCA scenarios are conservatively considered impact on large early Therefore, this treated as low RCS pressure RCS high pressure scenarios release frequency (LERF). uncertainty would not scenarios. have an impact on the 10 CFR 50.69 application, as Model may over-estimate no change in LERF will contribution of pressure- also result in no change induced SGTR (PI-SGTR) to in components safety LERF. significance. I

---

- I E1-30 to NL-12-0932 Basis for Proposed Change Table 2: VEGP Internal Events (including Internal Flooding) PRA Model Sources of Uncertainty and Im~act on 10 CFR 50.69 Program Source of Epistemic Related Assumptions Sensitivity Case Sensitivity Disposition of Uncertainty -

Conclusions Conclusions Seasonal Impacts on Initiating Events:

Certain initiating events can The generic industry frequency N/A Not a source of No explicit be affected by seasonal for the LOSP event developed uncertainty because the sensitivity impacts (e.g., LOSP, loss of in NUREG/CR-6890 is VEGP PRA model evaluation is SW, etc.) . applicable to the VEGP site. reflects average needed for 50.69 conditions (e.g. , overall impact.

Although the LOSP IE The NSCW cooling towers are fraction of time NSCW includes weather-related not needed in cold weather. eTs in bypass). So for events, any seasonal an application such as variation is not addressed. 50.69 which uses the average PRA model, this The fraction of time the is not a source of NSCW towers are in bypass uncertainty.

is modeled rather than specific season or weather conditions.

LOSP and LON sew frequency may be a potential source of uncertainty.

Applications pertaining to or affected by specific LOSP or LONSCW configurations should further evaluate seasonal impacts potential source of uncertainty. - - _ .. - "-

E1-31 to NL-12-0932 Basis for Proposed Change Table 2: VEGP Internal Events (including Internal Flooding) PRA Model Sources of Uncertainty and Impact on 10 CFR 50.69 Program Source of Epistemic Related Assumptions Sensitivity Case Sensitivity Disposition of Uncenainty Conclusions Conclusions Basis for HEPs: Detailed evaluations of HEPs The overall modeling Since there are no Since the VEGP are performed for the risk uncertainty associated with specific HFEs affected by PRA model uses The method of calculation of significant, pre- and post- the general basis for HEPs the 10 CFR 50.69 industry consensus human error probabilities initiator human failure events is addressed by the application, no additional modeling (HEPs) for the Human (HFEs) using industry standard HEP sensitivity explicit sensitivity approaches for its Reliability Analysis (HRA) consensus methods. The cases (required by NEI 00- evaluation is needed for HEP calculations, may introduce uncertainty THERP method is applied for 04 (Reference 4) 10 CFR 50.69 impact this is not based on the particular pre-initiator HFEs. The process). considered a methodology applied. CBDTM is used for cognitive significant source errors and THERP for of epistem ic I execution errors for post- uncertainty. I initiator HFEs. Therefore, no I additional explicit I sensitivity evaluation is needed for 50.69 impact.

th Note that 5 and th 95 percentiles sensitivity cases are part of the NEI 00-04 process.

E1-32 to NL-12-0932 Basis for Proposed Change 3.2.2 Fire PRA Model A state-of-the-art VEGP Fire PRA model was developed using the guidance provided in the ASMEIANS PRA standard (RA-Sa-2009) (Reference 11) in early 2012. A peer review was conducted during the week of February 13, 2012. The results of the peer review are provided in section 3.3.2 of this LAR.

VEGP has two operating units of Westinghouse 4-loop pressurized light water reactors with a core thermal power rating of 3625 MWt. The VEGP baseline Fire PRA models core damage frequency and large early release frequency due to internal fire, at power. There are two models, one for each unit, because each unit is not a mirror image of the other from a Fire PRA perspective. The models reflect the as-built, as-operated plant. Because the models have been developed recently, they have not gone through major upgrades or revision. However, SNC anticipates that they will continue to be refined over the coming years, similar to the internal events (including internal flooding) model. The Fire baseline PRA is a large fault tree methodology model, using the EPRI R&R Workstation (CAFTA) software and FTREX software for quantification. The baseline CDF and LERF values are 5.22E-05/yr and 2.1 OE-06lyr, respectively for U 1 and 5.19E-05/yr and 2.38E-06/yr, respectively for U2.

3.2.2.1 Fire PRA Maintenance/Updates and Application to 10 CFR 50.69 Categorization Process The Fire PRA models would be maintained and updated in the same manner as the internal events (including internal flooding) PRA model. This process is described in section 3.2.1.1 of this LAR.

3.2.2.2 Identification and Disposition of 10 CFR 50.69 Application Specific Sources of Uncertainty for Fire PRA Model The purpose of this section is to describe the evaluation of epistemic uncertainty in the Fire PRA (FPRA) model for impact on the implementation of 10 CFR 50.69.

Background:

The Vogtle FPRA model includes various sources of uncertainty that occur because there is inherent randomness in the elements that comprise the FPRA and because the state of knowledge in these elements continues to evolve.

NEI 00-04 outlines standard sensitivity studies that must be performed when using a PRA for different risk hazards. NEI 00-04 requires the following sensitivity studies that exercise key areas of uncertainty when using the Fire PRA.

• Increase all human error basic events to their 95th percentile value

• Decrease all human error basic events to their 5th percentile value

• Increase all component common cause events to their 95th percentile value E1-33 to NL-12-0932 Basis for Proposed Change

• Decrease all component common cause events to their 5th percentile value

• Set all maintenance unavailability terms to 0.0

• Take no credit for manual suppression

• Perform any applicable sensitivity studies identified in the characterization of PRA adequacy The first five sensitivity studies are self-explanatory. Therefore, they are not discussed further in this section. The discussion in this section is focused on the last two sensitivity studies - "No credit for manual suppression" and "Any applicable sensitivity studies identified in the characterization of PRA adequacy."

Take No Credit for Manual Suppression:

In performing this sensitivity study for the 10 CFR 50.69 categorization process, VEGP will set all manual suppression credit in the fire PRA model to zero (Le.,

manual suppression fails) with the exception of scenarios involving main control room abandonment. Manual suppression will be credited for scenarios involving main control room abandonment because if a fire is potentially necessitating control room abandonment, it is not reasonable to assume that manual suppression will not be attempted. Treating this sensitivity in this manner makes it more consistent with the treatment of the human error and common cause failure sensitivities listed above. With this approach, the sensitivity of the categorization considers the impacts of application of reasonable ranges of issue treatments, rather than preemptively removing all credit. In the case of main control room abandonment, plant procedures and operator training make it highly unlikely that manual suppression would not be attempted.

Any Applicable Sensitivity Studies Identified in the Characterization of PRA Adeguacy:

The discussion in this section is focused on the requirement to perform: "Any applicable sensitivity studies identified in the characterization of PRA adequacy".

In order to determine which, if any, such studies are needed, it is first necessary to evaluate the model and application-specific sources of uncertainty.

Methodology for Identifying 10 CFR 50.69 Application Specific Sources of Uncertainty for FPRA:

The development of the Vogtle FPRA was guided by NUREG/CR-6850 (Reference 16). The VEGP FPRA model used either the consensus model described in NUREG/CR-6850 or, in one instance, an alternate consensus model that was found to be adequate to meet the fire PRA standard. The following discussion provides information on one instance in which an alternate consensus approach was used. The approach dealt with assigning severity factors, which, per the peer review team, was based on an unreviewed analysis method. Consequently, an F&O was assigned. The F&O related to an alternate approach stated:

"The severity factors approach is based on an unreviewed analysis method. Currently in the FPRA there appears to be no credit for E1-34 to NL-12-0932 Basis for Proposed Change suppression activities when the severity factor is applied. Therefore, the scenario quantification does not limit the severity of the fire from the perspective of suppression credit. This assessment is limited to the treatment of severity factors and non suppression probabilities at the time of the peer review."

The F&O resolution that reflects the alternate consensus approach states:

"The application of the method to the FPRA is consistent with the results of the recent industry review completed subsequent to the Peer Review.

This method involves the application of a modified fire factor developed through a review of the individual industry fire events that form the basis for the generic fire frequency. However, the FPRA has applied a conservative value - higher than the value proposed in the method reviewed."

VEG P has used guidance provided in NUREG -1855 (Reference 13) to address uncertainties associated with Fire PRA for 10 CFR 50.69 application.

As stated in Section 1.5 of NUREG -1855:

"Although the guidance does not currently address all sources of uncertainty, the guidance provided on the process for their identification and characterization and for how to factor the results into the decision making is generic and is independent of the specific source.

Consequently, the process is applicable for other sources such as internal fire, external events, and low power and shutdown."

NU REG-1855 (Reference 13) describes an approach for addressing sources of model uncertainty and related assumptions. It defines "a source of model uncertainty is one that is related to an issue in which no consensus approach or model exists and where the choice of approach or model is known to have an effect on the PRA (e.g., introduction of a new basic event, changes to basic event probabilities, change in success criterion, introduction of a new initiating event}." NUREG-1855 defines a consensus model as "a model that has a publicly available published basis and has been peer reviewed and widely adopted by an appropriate stakeholder group. In addition, widely accepted PRA practices may be regarded as consensus models. Examples of the latter include the use of the constant probability of failure on demand model for standby components and the Poisson model for initiating events.

For risk-informed regulatory decisions, the consensus model approach is one that NRC has utilized or accepted for the specific risk-informed application for which it is proposed."

Disposition of 10 CFR 50.69 Application Specific Sources of Uncertainty for FPRA:

The potential sources of model uncertainty in the Fire PRA model were characterized for the 16 tasks identi'fied by NUREG/CR-6850 (Reference 16).

This framework was used to organize the assessment of baseline FPRA E1-35 to NL-12-0932 Basis for Proposed Change epistemic uncertainty and evaluate the impact of this uncertainty on 10 CFR 50.69.

Table 3 outlines sources of uncertainties by task and their disposition. As noted in the table, the VEGP FPRA was developed using methods outlined in the NUREG/CR-6850 (Reference 16), which is considered a consensus methodology. Furthermore, the use of unreviewed analysis method has been determined to be consensus approach. Therefore, consistent with NUREG 1855 (Reference 13), FPRA modeling does not introduce epistemic uncertainties to be addressed in the 10 CFR 50.69 application.

E1-36

Enclosure 1 to NL-12-0932 Basis for Proposed Change

[

Table 3: Fire PRA Uncertainty and Sensitivity Matrix I

Task Description Sources of Uncertainty Disposition Number 1 Analysis boundary This task establishes the overall spatial scope of Based on the discussion of sources of and partitioning the analysis and provides a framework for uncertainly, it is concluded that the methodology organizing the data for the analysis. The for the Analysis Boundary and Partitioning task I I

partitioning features credited are required to satisfy does not introduce any epistemic uncertainties I established industry standards. that would require sensitivity treatment.

Therefore, this task does not have an impact on I 10 CFR 50.69 aQplication. I 2 Component This task involves the selection of components to In the context of the FPRA, the uncertainty that is Selection be treated in the analysis in the context of initiating unique to the analYSis is related to initiating event I events and mitigation. The potential sources of identification. However, that impact is minimized uncertainty include those inherent in the internal though use of the PWROG Generic Multiple events PRA model as that model provides the Spurious Operation (MSO) list and the process foundation for the FPRA. used to identify and assess potential MSOs.

Based on the discussion of sources of uncertainty and the discussion above, it is concluded that the methodology for the Component Selection task does not introduce any epistemic uncertainties that would require sensitivity treatment. Therefore, this task does not have an impact on 10 CFR 50.69 application.

3 Cable Selection The selection of cables to be considered in the Based on the discussion of sources of analysis is identified using industry guidance uncertainty it is concluded that the methodology documents. The overall process is essentially the for the Cable Selection task does not introduce same as that used to perform the analyses to any epistemic uncertainties that would require demonstrate compliance with 10 CFR 50.48. sensitivity treatment. Therefore, this task does not have an impact on 10 CFR 50.69 application.

E1-37

Enclosure 1 to NL-12-0932 Basis for Proposed Change Table 3: Fire PRA Uncertainty and Sensitivity Matrix I Task Description Sources of Uncertainty Disposition Number 4 Qualitative Qualitative screening was performed; however, In the event a structure (location) which could Screening some structures (locations) were eliminated from result in a plant trip was incorrectly excluded, its the global analysis boundary and ignition sources contribution to CDF would be small (with a CCDP deemed to have no impact on the fire PRA (based commensurate with base risk). Such a location on industry guidance and criteria) were excluded would have a negligible risk contribution to the from the quantification based on qualitative overall FPRA.

screening criteria. The only criterion subject to uncertainty is the potential for plant trip. However, Based on the discussion of sources of such locations would not contain any features uncertainty and the discussion above, it is (equipment or cables identified in the prior two concluded that the methodology for the tasks) and consequently are expected to have low Qualitative Screening task does not introduce risk contribution. any epistemic uncertainties that would require sensitivity treatment. Therefore, this task does not have an impact on 10 CFR 50.69 application.

5 Fire-Induced Risk The internal events PRA model was updated to The identified source of uncertainty could result Model add fire specific initiating event structure as well as in the over-estimation of fire risk. In general, the additional system logic. The methodology FPRA development process would have processes used are consistent with that used for reviewed all significant fire initiating events and the internal events PRA model development as performed supplemental assessments to address was subjected to industry Peer Review. this possible source of uncertainty. I I

The developed model is applied in such a fashion Based on the discussion of sources of that all postulated fires are assumed to generate a uncertainty and the discussion above, it is plant trip. This represents a source of uncertainty, concluded that the methodology for the Fire-as it is not necessarily clear that fires would result Induced Risk Model task does not introduce any in a trip. In the event the fire results in damage to epistemic uncertainties that would require cables and/or equipment identified in Task 2, the sensitivity treatment. Therefore, this task does E1-38 to NL-12-0932 Basis for Proposed Change Table 3: Fire PRA Uncertainty and Sensitivity Matrix Task Description Sources of Uncertainty Disposition Number PRA model includes structure to translate them not have an impact on 10 CFR 50.69 application.

into the appropriate induced initiator.

6 Fire Ignition Fire ignition frequency is an area with inherent Industry generic frequency values were used Frequency uncertainty. Part of this uncertainty arises due to when developing VEGP FPRA. Based on the the counting and related partitioning methodology. discussion of sources of uncertainty, it is However, the resulting frequency is not particularly concluded that the methodology for the Fire sensitive to changes in ignition source counts. The Ignition Frequency task does not introduce any primary source of uncertainty for this task is epistemic uncertainties that would require associated with the industry generic frequency sensitivity treatment. Therefore, this task does values used for the FPRA. This is because there is not have an impact on 10 CFR 50.69 application.

no specific treatment for variability among plants along with some significant conservatism in defining the frequencies, and their associated heat release rates. The applied fire frequency values are believed to be over-estimated.

7 Quantitative Other than screening out potentially risk significant The Vogtle FPRA development did not screen Screening scenarios (ignition sources), this task is not a out any fire initiating events based on low source of uncertainty. CDF/LERF contribution. Screening of individual fire ignition sources occurred only if it involved a discrete component and the consequences of the associated fire did not involve failure of any other plant component or feature.

Based on the discussion of sources of uncertainty and the discussion above, it is concluded that the methodolog~ for the E1-39 to NL-12-0932 Basis for Proposed Change Table 3: Fire PRA Uncertainty and Sensitivity Matrix Task Description Sources of Uncertainty Disposition Number Quantitative Screening task does not introduce any epistemic uncertainties that would require sensitivity treatment. Therefore , this task does not have an impact on 10 CFR 50.69 application.

8 Scoping Fire The framework of NUREG/CR-6850 includes two See Task 11 discussion.

Modeling tasks related to fire scenario development. These two tasks are 8 and 11. The discussion of uncertainty for both tasks is provided in the discussion for Task 11.

9 Detailed Circuit The circuit analysis is performed using standard Circuit analysis was performed as part of the Failure Analysis electrical engineering principles. However, the deterministic post fire safe shutdown analysis.

behavior of electrical insulation properties and the Refinements in the application of the circuit response of electrical circuits to fire induced analysis results to the FPRA were performed on I

failures is a potential source of uncertainty. This a case-by-case basis where the scenario risk uncertainty is associated with the dynamics of fire quantification was large enough to warrant further i and the inability to ascertain the relative timing of detailed analysis. The uncertainty (conservatism) circuit failures. The analysis methodology which may remain in the FPRA is associated with I assumes failures would occur in the worst possible scenarios that do not contribute significantly to configuration, or if multiple circuits are involved, at the overall fire risk.

whatever relative timing is required to cause a bounding worst-case outcome. This results in a Based on the discussion of sources of skewing of the risk estimates such that they are uncertainty and the discussion above, it is over-estimated . concluded that the methodology for the Detailed Circuit Failure Analysis task does not introduce any epistemic uncertainties that would require sensitivity treatment. Therefore, this task does E1-40 to NL-12-0932 Basis for Proposed Change Table 3: Fire PRA Uncertainty and Sensitivity Matrix Task Description Sources of Uncertainty Disposition Number not have an impact on 10 CFR 50.69 application.

10 Circuit Failure One of the failure modes for a circuit (cable) given Uncertainty in the circuit failure mode likelihood Mode Likelihood fire induced failure is a hot short. A conditional analysis could lead to assumed failures of related Analysis probability is assigned using industry guidance components and related system functions. This such as that published in NUREG/CR-6850. The would generate conservative results and that uncertainty associated with the applied conditional would typically be acceptable for most failure probabilities poses competing applications. Furthermore, consensus modeling considerations. On the one hand, a failure approach is used for Circuit Failure Mode probability for spurious operation could be applied Likelihood Analysis. I I

based solely on cable scope without consideration of less direct fire affects (e.g., a 0.3 failure Based on the discussion of sources of I

likelihood applied to the spurious operation of an uncertainty and the discussion above, it is I I

MOV without consideration of the fire-induced concluded that the methodology for the Circuit generation of spurious signal to close or open the Failure Mode Likelihood Analysis task does not I MOV). The analysis has biased the treatment such introduce any epistemic uncertainties that would that it is assumed the spurious signal will always require sensitivity treatment. Therefore, this task drive the valve in the unsafe direction. In addition, does not have an impact on 10 CFR 50.69 for those valves that might have multiple desired application.

functions - consideration of spurious closure and consideration of failure to open on demand, the non-spurious failure state is treated with a logical TRUE rather the complement of the spurious probability. For those valves that only have an active function, the potential for a spurious Signal to drive the valve in the desired direction is ignored.

The treatment results in skewing of the results E1-41

Enclosure 1 to NL-12-0932 Basis for Proposed Change Table 3: Fire PRA Uncertainty and Sensitivity Matrix I Task Description Sources of Uncertainty Disposition Number such that the resulting risk is over-estimated.

11 Detailed Fire The application of fire modeling technology is used in Consensus modeling approach is used for the Modeling the FPRA to translate a fire initiating event into a set of Detailed Fire Modeling.

consequences (fire induced failures). The performance of the analysis requires a number of key input Based on the discussion of sources of parameters. These input parameters include the heat uncertainty and the discussion above, it is release rate (HRR) for the fire, the growth rate, the damage threshold for the targets, and the response of concluded that the methodology for the Detailed plant staff (detection, fire control, and fire suppression). Fire Modeling task does not introduce any epistemic uncertainties that would require The fire modeling methodology itself is largely empirical sensitivity treatment. Therefore, this task does in some respects and consequently is another source of not have an impact on 10 CFR 50.69 application.

uncertainty. For a given set of input parameters, the fire modeling results (temperatures as a function of distance from the fire) are characterized as having some distribution (aleatory uncertainty). The epistemic uncertainty arises from the selection of the input parameters (specifically the HRR and growth rate) and how the parameters are related to the fire initiating I event. While industry guidance is available, that I I

guidance is derived from laboratory tests and may not necessarily be representative of randomly occurring I events.

The fire modeling results using these input parameters are used to identify a zone of influence (ZOI) for the fire and cables/equipment within that ZOI are assumed to be damaged. In general, the guidance provided for the treatment of fires is conservative and the application of E1-42 to NL-12-0932 Basis for Proposed Change Table 3: Fire PRA Uncertainty and Sensitivity Matrix Task Description Sources of Uncertainty Disposition I

Number that guidance retains that conservatism. The resulting risk estimates are also conservative.

12 Post-Fire Human The human error probabilities used in the FPRA The human error probabilities were obtained Reliability Analysis were adjusted to consider the additional challenges using the EPRI HRAC and included the that may be present given a fire. The human error consideration of degradation or loss of necessary probabilities were obtained using the EPRI HRAC cues due to fire. The impact of any remaining I

and included the consideration of degradation or uncertainties is expected to be small.

loss of necessary cues due to fire. Given the methodology used, the impact of any remaining Based on the discussion of sources of uncertainties is expected to be small. uncertainty and the discussion above, it is concluded that the methodology for the Post-Fire Human Reliability Analysis task does not introduce any epistemic uncertainties that would require sensitivity treatment. Therefore, this task does not have an impact on 10 CFR 50.69 application.

13 Seismic-Fire Since this is a qualitative evaluation, there is no The qualitative assessment of seismic induced Interactions quantitative impact with respect to the uncertainty fires should not be a source of model uncertainty Assessment of this task. as it is not expected to provide changes to the quantified fire PRA model.

Based on the discussion of sources of uncertainty and the discussion above, it is concluded that the methodology for the Seismic-Fire Interactions Assessment task does not introduce any epistemic uncertainties that would require sensitivity treatment. Therefore, this task E1-43 to NL-12-0932 Basis for Proposed Change Table 3: Fire PRA Uncertainty and Sensitivity Matrix Task Description Sources of Uncertainty Disposition Number does not have an impact on 10 CFR 50.69 I I

application.

14 Fire Risk As the culmination of other tasks, most of the The selected truncation was confirmed to be Quantification uncertainty associated with quantification has consistent with the requirements of the PRA already been addressed. The other source of Standard.

uncertainty is the selection of the truncation limit.

However, the selected truncation was confirmed to Based on the discussion of sources of be consistent with the requirements of the PRA uncertainty and the discussion above, it is Standard. concluded that the methodology for the Fire Risk Quantification task does not introduce any epistemic uncertainties that would require sensitivity treatment Therefore, this task does not have an impact on 10 CFR 50.69 application.

15 Uncertainty and This task does not introduce any new uncertainties. This task does not introduce any new Sensitivity This task is intended to address how the fire risk uncertainties. This task is intended to address Analyses assessment could be impacted by the various how the fire risk assessment could be impacted sources of uncertainty. by the various sources of uncertainty.

Based on the discussion of sources of uncertainty and the discussion above, it is concluded that the methodology for the Uncertainty and Sensitivity Analyses task does not introduce any epistemic uncertainties that would require sensitivity treatment. Therefore, this task does not have an impact on 10 CFR 50.69 application.

16 Fire PRA This task does not introduce any new uncertainties This task does not introduce any new E1-44 to NL-12-0932 Basis for Proposed Change Table 3: Fire PRA Uncertainty and Sensitivity Matrix Task Description Sources of Uncertainty Disposition Number Documentation to the fire risk. uncertainties to the fire risk as it outlines documentation requirements.

Based on the discussion of sources of uncertainty and the discussion above, it is concluded that the methodology for the Fire PRA Documentation task does not introduce any epistemic uncertainties that would require sensitivity treatment. Therefore, this task does not have an impact on 10 CFR 50.69 application.

E1-45 to NL-12-0932 Basis for Proposed Change 3.2.3 Other Techniques The NRC endorsed NEI-OO-04 methodology allows for categorization of SSCs to be performed in the absence of quantifiable PRA models for other risk hazards, specifically seismic hazards, other external events hazards (e.g., tornados, high winds, transportation, etc.), and shutdown hazards. The following information provides a historic perspective of how these risks were analyzed in the past and how the results will be used when categorizing SSCs per NEI 00-04.

External hazards were evaluated in the VEGP Individual Plant Examination of External Events (IPEEE) submitted in response to the NRC IPEEE Generic Letter 88-20, Supplement 4 (Reference 9). The IPEEE was a one-time review of external hazard risk and was limited in its purpose to the identification of potential plant vulnerabilities and the understanding of associated severe accident risks.

The results of the VEGP IPEEE study are documented in the VEGP IPEEE main report.

In the VEGP IPEEE, the seismic risk evaluation was performed in accordance with the EPRI Seismic Margins Analysis (SMA) methodology (References 17, 18, and 19). The SMA Review Level Earthquake (RLE) for VEGP was a 0.3g Peak Ground Acceleration (PGA) NUREG/CR-0098 (Reference 20) spectrum. VEGP structures and equipment were designed for a safe shutdown earthquake (SSE) defined by a Regulatory Guide 1.60 spectrum tied to a PGA of 0.2 g. However, due to conservatism applied to the demand and/or evaluation techniques, most of the seismic Category I structures and equipment were designed and qualified for a 0.3g PGA capacity.

Because the SMA approach was used, there are no CDF and LERF models or values available from the seismic analysis in the VEGP IPEEE. One of the insights from the VEGP SMA was that VEGP is one of the most seismically rugged nuclear power plants. A conclusion from the SMA was that VEGP has a high-confidence-Iow-probability-of-failure (HCLPF) capacity of at least 0.3g PGA.

The VEGP IPEEE analysiS of High winds, Floods, and Other (HFO) external hazards was accomplished by using a progressive screening approach described in NUREG-1407 (Reference 21). The VEGP IPEEE concluded that the existing VEGP design was in conformance with the 1975 Standard Review Plan (SRP),

NUREG 75-087 (Reference 22), criteria, in all reviewed areas and no potential vulnerabilities were identified. HFO events were screened out by compliance with the SRP. As such these hazards were determined to be negligible contributors to the overall plant risk.

As stated earlier, the NEI 00-04 methodology allows for categorization of SSCs to be performed in the absence of quantifiable PRA models for the following other risk hazards - seismic hazards, other external events hazards (e.g., tornados, high winds, transportation, etc.), and low power/shutdown hazards. The following information provides how each of these risks will be evaluated when categorizing SSCs per NEI 00-04.

E1-46 to NL-12-0932 Basis for Proposed Change 3.2.3.1 Updated Seismic Margin Analysis SSEl Currently, a state-of-the-art VEGP Seismic PRA model is being developed that will address the requirements of the ASMEIANS PRA standard (Reference 11). It is anticipated that the model will be available for use to categorize SSCs in early 2015. As such, the use of a seismic PRA model is not proposed as part of the VEGP 10 CFR 50.69 process at this time.

In the absence of a seismic PRA model that meets the seismic PRA standard, the

!\lEI 00-04 methodology allows the use of the seismic margin analyses, whicrl is a screening approach to evaluating seismic hazards. It does not generate core damage values; rather, it functions to identify potential seismic susceptibilities and vulnerabilities.

To respond to the NRC IPEEE Generic Letter 88-20, Supplement 4, VEGP chose to address IPEEE-Seismic by performing an SMA per the EPRI SMA methodology (Reference 17). In NUREG-1407 (Reference 10), the NRC states that the EPRI SMA methodology is an acceptable methodology for resolution of IPEEE-Seismic. Per the EPRI methodology, one preferred and one alternate path capable of achieving and maintaining a safe shutdown condition for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a seismic margin earthquake (SME) were selected for each unit.

The SMA included the development of seismic Safe Shutdown Equipment Lists (SSELs) and the performance of seismic capability walkdowns and evaluations of the SSEL components. In 2011 and 2012, SNC reviewed the IPEEE SSELs against the current plant configuration, updated SSELs accordingly, and performed seismic capability walkdowns and evaluations on accessible new and modified SSEL components in accordance with EPRI NP-6041-SL (Reference 17). The primary remaining work consists of evaluating inaccessible new and modified SSEL components and resolving open items.

In the updated seismic SSEL, the review level earthquake (RLE) for VEGP was the same as that was used in IPEEE, which was a 0.3g peak ground acceleration (PGA) NUREG/CR-0098 (reference 20) spectrum. VEGP structures and equipment were designed for a safe shutdown earthquake (SSE) defined by a RG 1.60 (Reference 18) spectrum tied to a PGA of 0.2g. However, due to conservatism applied to the demand and/or evaluation techniques, most of the seismic Category I structures and equipment were designed and qualified for a 0.3g PGA capacity. One of the insights from the VEGP IPEEE SMA was that VEGP is one of the most seismically rugged nuclear power plants. A conclusion from the SMA was that VEGP has a high-confidence-Iow-probability-of-failure (HCLPF) capacity of at least 0.3g PGA. These insight and conclusion also apply to the SMA containing updated seismic SSEL.

The SMA identified the seismic design basis and severe accident functions of components. A determination was made to find out if a component would be needed during and after the seismic event. A seismic SSEL was developed containing components that would be needed during and after the seismic event.

When categorizing SSCs per 10 CFR 50.69, a component that is credited as part E1-47 NL-12-0932 Basis for Proposed Change of a seismic-margins-evaluated shutdown path will be considered safety-significant from a seismic risk perspective. The attributes, which yielded that conclusion, will identified. If the component does not participate in the seismic safe shutdown path, then it is considered a candidate low safety-significant with seismic 3.2.3.2 Updated IPEEE Screening (other external events)

When the I analysis was performed, the ASMEIANS standards did not exist. However, ASMEIANS standards have since been developed for analyzing various hazards. The current ASMEIANS Combined PRA standard (RA-Sa 2009) (Reference 11) guidance for other external As cited in RA-Sa 2009, other external events include external events other than events.

In 2011, SNC evaluated other external risks, including the impact of construction activities at Vogtle Units 3 and 4, using the criteria cited in RA-Sa-2009.

conclusion of evaluation is all other out.

Therefore, per RA-Sa-2009, SNC is not to develop VEGP PRA models for other external The results of the screening indicate all SSCs will be treated as candidate low safety significant with respect to other external events risk.

3.2.3.3 Low Power/Shutdown Currently, a consensus Low Power (LP) PRA standard is not available.

Consequently, VEGP does not have a LP model. A typical LP PRA methodology would extensively on the extension of at-power analysis methods to LP conditions. Accordingly, the at-power methodology employed for higher power operations are typically adapted or extended to address conditions. The at-power PRA model is generally valid for evaluating LP conditions. When accounting for failure P not differentiate between at-power conditions versus low power conditions. Therefore, in order to account risks, insights can be obtained using the at-power PRA model for those cases where at-power PRA model is valid for evaluating low power conditions. In addition, in those cases where at-power PRA model is not appropriate for use at low power conditions, the low power risk will be evaluated by qualitatively the function(s) performed by the system as outlined in NEI 00-04 section 9.2.2 review of risk information item 6. A component is a preliminary risk, which is higher of these two risks where applicable.

Currently, a consensus Shutdown (SO) PRA standard is not available, and consequently, not have a SO PRA model. In of SO PRA, NEI 00-04 methodology permits use of qualitative analysis. It is recognized the categorization process using a qualitative approach is more conservative (Le., deSigned to identify more as safetY-Significant) than using a plant-specific SO PRA. is due to fact that the qualitative approach provides safety function defense-in-depth without regard to the likelihood demand or reliability of the functions credited. 00-04 approach identifies all to support primary shutdown safety systems as safety-significant. This

-48 to NL-12-0932 Basis for Proposed Change measure of safety significance assures that the SSCs that were required to maintain low shutdown risk are retained as safety-significant. A typical SO PRA would credit all of the same SSCs in a probabilistic framework, so some may avoid being identified as safety-significant using the PRA, but the shutdown safety management plan approach identifies them as safety-significant regardless of the frequency of challenge or level of functional diversity.

The NEI 00-04 methodology uses a shutdown safety assessment process per NUMARC 91-06. VEGP updated its Shutdown Safety Program developed per NUMARC 91-06 (Reference 12) in early 2011. As a result, VEGP has a qualitative defense-in-depth (DID) shutdown model for shutdown configuration risk management (CRM). The development of the model is based on the framework for and guidance for DID provided in NUMARC 91-06, which provides guidance for assessing and enhancing safety during shutdown operations. Other documents considered in this model development include INPO 06-08 (Reference 23), SOER 09-01 (Reference 24), and NRC Inspection Manual Chapter 0609, Appendix G (Reference 25). This qualitative model will be used as follows. It is recognized that the use of shutdown DID model introduces an additional deterministic conservatism.

1. A component can be identified as safety-significant for shutdown conditions for either of the following reasons :

• NUMARC 91-06 specifies that a defense in depth approach should be used with respect to each defined shutdown key safety function. This is accomplished by designating a running and alternative system/train to accomplish the given key safety function. When multiple systems/trains are available to satisfy the key safety function, only SSCs that support the primary and first alternative methods to satisfy the key safety function are considered to be the "primary shutdown safety system" and are thus candidate safety-significant.

• Its failure would initiate a shutdown event (e.g., loss of shutdown cooling, drain down, etc.).

2. If the component does not participate in either of these manners, then it is considered a candidate as low safety significance with respect to shutdown safety.

Note that in this assessment, a primary shutdown safety system refers to a system that has the following attributes:

• It has a technical basis for its ability to perform the function.

• It has margin to fulfill the safety function.

• It does not require extensive manual manipulation to fulfill its safety function.

E1-49 to NL-12-0932 for Proposed Change 3.3 Review Process Results (10 CFR 50.69(b)(2)(iii>>

NEI 00-04 (Reference 4) requires that the internal events (including internal flooding) and PRA models be reviewed to the guidance of Regulatory Guide 1.200 to ensure that PRA meet Capability Category (CC) II for supporting requirements of the ASME PRA standard. NEI 00-04 also requires that deviations from CC II relative to the 10 CFR program be justified documented. This section provides PRA review results for the internal events (including flooding) PRA and Fire PRA models.

3.3.1 Internal Events (including Internal Flooding) PRA Model The PRA been subjected to a number of reviews and self-assessments, including one performed in accordance with the 2007 version the PRA Standard (Reference 26) as clarified by RG 1.200, Revision 1 (Reference The information provided in section demonstrates that the internal PRA model (including flooding) meets requirements of RG 1.200, Revision 2 (Reference 28).

Section .1 summarizes prior peer reviews and self-assessments for the PRA model.

Section 3.3.1 describes the overall results of the 1.200 review (Capability Category and Findings) performed in 2009.

1.3 summarizes the resolution of Findings Observations (F&Os) identified in the RG 1.200 peer review.

In May 2009, the VEGP was reviewed against the 2007 version of PRA Standard as amended by RG 1.200, 1 (Reference 27) not the version of the PRA Standard (Reference 11) issued in 2009 as clarified by 1.200, Revision 2 (Reference 28). Therefore, Section 1 provides a list of additional requirements associated with the PRA Standard. This section also describes the VEGP PRA model model documentation revisions that assure consistency with the version of the PRA Standard RG 1 3.3.1.1 Previous Review and VEGP PRA Model In addition to independent internal and external review during each VEGP PRA model development and update, several assessments of technical capability were prior to the Pressurized Water Owners Group (PWROG) peer review against the ASME Standard and RG 1.200, Revision 1 in May 2009. Listed below are previous assessments of the PRA.

.. An independent PRA review was conducted under auspices of the Westinghouse Owners Group (WOG) in December 2001, following Industry PRA Review process (Reference 29). This peer review included an assessment of the PRA model maintenance and update process.

-50

assessment did not identity any uN' F&Os. from the 2001 Industry PRA Review for were raC","an in VEGP PRA model Revision 3.

• PRA model were performed in support of mitigating systems performance indicator (MSPI) Results cross-comparison are presented in WCAP-16464, Westinghouse Owner's Mitigating Systems Performance Index Cross Comparison. The Comparison Candidate Outlier Status was in section MSPI document. Noted in this document was the fact that, allowing for plant-specific features, were no MSPI cross-comparison for VEGP

• In 2006, a gap analysis was performed against the versions ASME PRA Standard (Reference 30) and Regulatory Guide 1.200, Revision 0 (2003 trial version). Documentation (especially system notebooks), '.:::I,",'

to the internal flooding PRA, and treatment of uncertainty correlations were identified as major gaps. The identified gaps were resolved in P model 4 in 2009.

• In 2008, the PRA model Revision 4) was benchmarked with three Westinghouse PWRs (Comanche Peak, Callaway, Wolf Creek) as a part of an MSPI margin study. benchmarking concluded that were no significant in the VEGP model which would impact MSPI calculations.

3.3.1 Industry VEGP Internal ",l:IOn1rc. (including Internal Flooding)

VEGP PRA model for internal events (including internal flooding) at-power was updated to 4 early in 2009 to close the from the 2006 to ASME PRA supporting requirements (including NRC clarifications as stated in RG 1 Revision 1), and to represent as-built as-operated plant.

In May 2009, the PRA internal events model 4 (including internal flooding) was reviewed against the requirements of 2007 version of the PRA Standard (Reference 26). as amended by RG 1.200, Revision 1 (Reference 27).

A summary of this review is below:

1. 2007 of ASME Standard contains a total of numbered supporting requirements (SRs) in nine technical elements and configuration control element. of the SRs represent deleted requirements (I IE-A9, SY-A9, HR-G8, IF-A2, IF D2, IF-E2, and QU-D2) and 20 were determined by the peer review to not applicable to the VEGP Thus, a total of 296 SRs were applicable.

-51 to NL-12-0932 Basis for Proposed Change

2. Among 296 applicable SRs, 99% of SRs met Capability Category II or higher as follows:

Table 4: Summary of VEGP Internal Event (including Internal Flooding) PRA Cap_ability Categories Number of  % of Total Applicable Capability Category Met SRs SRs CC-IIII/Ill (or SR 210 70.9%

Met)

CCI 0 0%

CC II 38 12.8%

CC III 7 2.4%

CC 1111 14 4.7%

CC II/III 24 8.1%

SR Not Met 3 1.0%

SR (CC-IIII1I1I) Met 296 100

3. Three SRs were judged to be not met. These are HR-G6, aU-03, and LE-G5.

• SR HR-G6 was not met because the reasonableness check of Human Reliability Analyses (HRA) was done for the previous revision of the PRA model and not the latest revision.

• SR aU-03 was not met because the SR requires the PRA results to be compared with those from similar plants. The VEGP PRA report cites the MSPI benchmark report as evidence of meeting this requirement, but the peer reviewers viewed this as an outdated comparison.

• SR LE-G5 was characterized as "Not Met" because the limitation of the LERF calculations that could impact risk-informed applications was not identified.

4. The peer review generated 11 Findings. These Findings and their resolutions are described in section 3.3.1.3. Resolution of Findings HR-G6-01, aU-03 01, and LE-G5-01 resulted in SRs HR-G6, aU-03, and LE-G5 being met to a Capability Category 1111/111. Thus, the VEGP internal events PRA (including flood) model meets the requirements of RG 1.200.

3.3.1.3 Resolution of Findings from VEGP PRA Peer Review Table 5 documents the VEGP Internal Events (including Internal Flooding) PRA model peer review F&Os and their resolutions. As shown in Table 5, the three "Not Met" SRs have since been resolved.

Table 5 reflects two types of F&O - those that are already resolved and those, as indicated in the table, that would be resolved prior to implementation of the 10 CFR 50.69 program. As indicated in Table 5, the unresolved F&O do not E1-52 to NL-12-0932 Basis for Proposed Change involve any Not Met SRs and mostly are related to enhancing documentation and are not expected to impact 10 CFR 50.69 categorization results.

E1-53 to NL-12-0932 Basis for Proposed Change Table 5: VEGP Internal Events (including Internal Flooding)

Resolution of the Peer Review Findings and Observations (F&Os)

F&O Review Capability Finding Description Resolution Number Element Category (CC)

IE-A4-01 IE-A4 CC-II The SR requires a systematic evaluation of each This F&O will be resolved prior to the Met system to assess the possibility of an initiating implementation of the 10 CFR 50.69 event occurring due to failure of the system. The Program when the Vogtle internal events reviewers could not find documentation of such a PRA notebook documentation is enhanced systematic review. to describe the systematic evaluation performed for identifying special initiating Additional notes made by review team in response events.

to SNC's comments: When the reviewers asked for the Initiating Events (IE) notebook (NB), they A systematic evaluation for initiating events were told that Chapter 2 of the main report is the was performed during the Vogtle IPE and IE NB. Chapter 2 does not contain any evidence the baseline Vogtle internal events PRA that a systematic evaluation of each system was model credits this process. The internal performed. Nor does Chapter 2 contain a system event model documentation provides a block failure modes and effects analysis (FMEA) as diagram describing the IE identification task ..

required by the Standard which would have been an acceptable alternate. The fact that a systematic evaluation was performed during the Individual Plant Examination (lPE), in of itself, is not sufficient. The evaluation performed for the IPE should have been reviewed and a statement to that extent should have been presented in the Chapter 2. In absence of such evidence, the review comment stays.

As noted elsewhere in the report, it is very important to have good documentation.

E1-54 to NL-12-0932 Basis for Proposed Change Table 5: VEGP Internal Events (including Internal Flooding)

Resolution of the Peer Review Findings and Observations (F&Os)

F&O Review Capability Finding Description Resolution Number Element Category (CC)

IE-01-01 IE-01 CC-I/IIIIII The lack of a central place for all the information This F&O will be resolved when SNC Met related to initiating events made it difficult for the develops a separate IE notebook prior to the review team to review this topic. Most plants have implementation of 10 CFR 50.69 program.

all this information stored in a separate IE NB.

The review team recommends that VEGP do the same.

Additional notes in response to SNC's comments:

The review team disagrees with SNC's comments.

The Standard requires that the work be documented in a manner that facilitates PRA applications, upgrades, and peer review. The review team does not believe that the work was documented a manner to facilitate peer review.

One could almost make a case for 'not met' categorization for this element as the documentation is the weakest link in this whole effort. The F&O stays as written. - - -

E1-55 to NL-12-0932 Basis for Proposed Change I Table 5: VEGP Internal Events (including Internal Flooding)

Resolution of the Peer Review Findings and Observations (F&Os)

F&O Review Capability Finding Description Resolution Number Element Category (CC)

AS-A 11-01 AS-A11 CC-I/II/III Dependencies are not preserved for This F&O will be resolved prior to the Met consequential anticipated transient without implementation of 10 CFR 50.69 when SNC SCRAM (ATWS) for the small loss of coolant modifies the PRA Calculation to note that the accident (SLOCA) initiator and the steam A TWS event tree was developed to include generator tube rupture (SGTR) initiator. The consideration on loss of RCS inventory existing A TWS trees, based on a LOFW initiator, (LOCA) and SGTR. This is accomplished as were developed for transients that do not include documented in the internal events PRA a loss of reactor coolant system (RCS) inventory calculation (Reference 14) and by or operator actions to mitigate a SGTR. incorporating logic gates into the system logic models to address primary integrity and Note: The review team decided to leave the F&O system failures caused by SLOCA- or as is after reviewing SNC's comments. SGTR-initiated ATWS events.

E1-56 to NL-12-0932 Basis for Proposed Change

- Table 5: VEGP Internal Events (including Internal Flooding)

Resolution of the Peer Review Findings and Observations (F&Os)

F&O Review Capability Finding Description Resolution Number Element Category (CC)

SY-B3-01 SY-B3 CC-IIIIIIII The treatment of main or frontline system and This F&O will be resolved prior to the Met supporting or mitigating system Common Cause implementation of 10 CFR 50.69 when SNC Failure (CCF) event groupings do not appear to develops updated CCF documentation consistent in the current PRA documents. It reflecting that all potential CCF groups have appears that for systems considered non-risk been considered and documented.

significant, reviews for CCF groups may not have been undertaken since the IPE modeling. The The systems that may be lacking CCF updated standards require a systematic treatment grouping are non-risk Significant systems; of all systems, not just the main systems therefore, resolution of this F&O is not contributing to core damage. New CCF groups expected to impact the 10 CFR 50.69 may be required or updated documentation as to categorization results.

why these groups are not required is needed.

Note: The review team decided to leave the F&O as is after reviewing SNC's comments.

HR-G6-01 HR-G6 CC-I/IIIIII Check of consistency and review for This F&O is resolved.

Not Met reasonableness is missing in the Revision 4 updated HRA draft and the prior revision Reasonableness check for all HRAs for document information related to these items is not Revision 4 model was re-performed. All appropriate to use in light of the updates HRAs have been determined to be performed and changes to the results. Section 8 reasonable or have been appropriately includes a table of human failure events (HFEs) revised. The reasonableness check is I and human error probabilities (HEPs) but does not documented in Section 8.2.2 of the internal I include HEP reasonableness check, as is events PRA calculation (Reference 14).

documented in Section 8.3 of the November 2005 HRA update for Revision 3. I E1-57 to NL-12-0932 Basis for Proposed Change Table 5: VEGP Internal Events (including Internal Flooding)

Resolution of the Peer Review Findings and Observations (F&Os)

F&O Review Capability Finding Description Resolution Number Element Category (CCt DA-C2-01 DA-C2 CC-1I111111 Generic data alone was used for the probability This F&O will be resolved prior to the Met that a power-operated relief valve (PORV) is implementation of 10 CFR 50.69 program blocked -- refer to Table 6.3.9. Since PORV when SNC revises Table 6.3-9 of the PRA availability is a critical plant feature with respect to Calculation (Reference 14) to remove the ATWS pressure control, the use of generic data entry for PORV leakage (POLRC) since a for this parameter is deemed a weakness. plant-specific value has been used.

IF-C2a-01 IF-C2a CC-I/II/III Because of a lack of well documented analysis, a This F&O will be resolved prior to the Met lot of information had to be obtained by talking to implementation of 10 CFR 50.69 program to the analyst who performed the analysis, which is clarify documentation related to flooding the basis for the F&O. analysis.

Original F&O: From a more detailed review of the The VEGP internal events flooding analysis IPE flood calculations (which are the main input to conservatively assumes equipment in a defining the flood events and consequences), it is room is damaged due to flood when a pipe noted that successful operator mitigation of ALL break occurs in that room. The analysis does flood events is assumed to occur 30 minutes into not credit operator actions for flood any flood scenario and fully terminate the flood isolation/mitigation (that is, screening HEP flow, and it appears to be based on assumptions values used were equal to 1.0). Screening only, as no detailed discussion of the actual ability HEP values in human induced flooding of operators to perform such actions is given. events do not make use of the results of the This appears to be in direct conflict with the HFEs design related calculations which assumes a included (but not modeled in the PRA model) in 30 minutes flow termination time. As a result the flooding report (assumed perfect response vs. resolution of this F&O has no impact on the HFE calculation). Also , the report lists hundreds 10 CFR 50.69 categorization.

of pages of a detailed analysis approach using screening criteria, flow calculations, etc. and only E1-58 to NL-12-0932 Basis for Proposed Change Table 5: VEGP Internal Events (including Internal Flooding)

Resolution of the Peer Review Findi~s and Observations (F&Os)

F&O I Review I Capability I Finding Description Resolution Number Element Category (CC) by locating very specific statements.

There appear to be conflicts of the inputs to the flooding PRA and the subsequent discussions of operator mitigation as well as using the information from the IPE calculations for propagation assessments. This is more than an editorial finding and impacts the entire basis of using the older calculated results in the current analysis.

Additional notes made in response to SNC's comments: The lengthy flooding methodology outlined in the report is not used in the current VEGP flooding results as mentioned in the original F&O. The previous IPE flooding analysis is used as inputs to the flooding targets and propagations for a bounding case estimation, and the more thorough analysis outlined in the report is not undertaken but is in place for use (as was explained by the SNC analyst in charge of the flooding project during the peer review and very briefly mentioned in the flooding report).

Each and every IPE flooding calculation reviewed during the peer review contained the assumption E1-59 to NL-12-0932 Basis for Proposed Change Table 5: VEGP Internal Events (including Internal Flooding)

Resolution of the Peer Review Findings and Observations (F&Os)

F&O Review Capability Finding Description Resolution Number Element Category (CC) that the flood water flow was successfully isolated at 30 minutes, and all calculations for flood volumes, propagations, etc. were done with the amount of water generated in this 30 minutes with considerations for system characteristics. If any IPE flooding calculations are done, which do not contain this assumption, they were not seen during the peer review.

I No change to the F&O is warranted. The problem with this scenario of using the IPE flooding calculations for inputs to the described methodology is the following: If the flooding analysis was performed in accordance to the methodology outlined in the current report, new flooding volumes and propagation assessments would be required that did not take into account successful isolation at 30 minutes (as the IPE calculations do) since another operator isolation assessment is outlined in the flooding report methodology for normal HFE calculations for flow isolation.

QU-03-01 QU-03 CC-IIIIIIII Reviewer asked the VEGP staff to provide This F&O is resolved.

Not Met evidence of comparison of the VEGP results to those from similar plants. The VEGP staff A new comparison study was performed by presented the benchmark report for MSPI as comparing VEGP PRA results with two PWR E1-60 to NL-12-0932 Basis for Proposed Change Table 5: VEGP Internal Events (including Internal Flooding)

Resolution of the Peer Review Findings and Observations jF&Os)

F&b Review Capability Finding Description Resolution Number Element Category (CC) evidence of comparison. Reviewers concluded PRAs (Callaway and Wolf Creek), which are that report is not sufficient evidence for considered relatively similar to VEGP. In demonstrating compliance to this SR. addition to the comparison of PRA reports, a plant visit to Callaway was performed to identify more details of Callaway systems and PRA modeling.

The comparison showed that all three plants have loss of offsite power (LOSP)/station blackout (SBO) as the most dominant contributors which indicated that the VEGP PRA results are not an outlier, as compared ,

to similar PWRs. Differences in dominant CDF contributors were investigated, and it I was found that those differences are due to differences in details of system I configuration/operation and physical barriers for internal flooding and in the sources for generic initiating event frequency data (VEGP PRA used the latest generic initiating frequency and failure data along with VEGP specific experience data for its data update).

E1-61 to NL-12-0932 Basis for Proposed Change Table 5: VEGP Internal Events (including Internal Flooding)

Resolution of the Peer Review Findings and Observations (F&Os)

F&O Review Capability Finding Description Resolution Number Element Category (CC)

QU-F5-01 QU-F5 CC-I/IIIIII In Chapter 10, there is insufficient documentation This F&O will be resolved prior to the Met for the quantification process, which would impact implementation of 10 CFR 50.69 program application (only EOOS). Reviews conclude that when SNC revises Chapter 10 to provide the documentation currently in Chapter 10 is not additional documentation of the sufficient to meet this SR fully. quantification process, including additional limitations when using the PRA model for EOOS and other applications, and verifies that all quantification supporting requirements are met and documented.

Section 10.3 of the internal events PRA calculation (Reference 14) discusses a few limitations in the quantification process that would impact applications, including the use of an average configuration, average test and maintenance unavailabilities, and point estimate mean data values.

E1-62 to NL-12-0932 Basis for Proposed Change Table 5: VEGP Internal Events (including Internal Flooding)

Resolution of the Peer Review Findin~s and Observations (F&Os)

F&O Review Capability Finding Description Resolution Number Element Category (CC)

LE-G5-01 LE-G5 CC-I/IIIIII Limitations in the LERF analysiS that would impact This F&O is resolved.

Not Met applications are not identified. The LERF analysis documentation is incomplete because limitations A comparison of Vogtle LERF scenarios with in the LERF analysis that would impact those in Table 4.5.9.3 of the ASME PRA applications, as required by SR LE-G5, are not standard revealed that the Vogtle PRA I identified. included more potential LERF scenarios than as required for a large dry containment plant in ASME PRA standard.

The LERF scenarios modeled in VEGP PRA include containment bypass core damage scenarios (steam generator tube rupture and Interfacing systems LOCA), thermally or pressure induced steam generator tube rupture after core damage, containment isolation failure with core damage, and various early_containment failure modes.

E1-63 to NL-12-0932 Basis for Proposed Change Table 5: VEGP Internal Events (including Internal Flooding)

Resolution of the Peer Review Findings and Observations IF&Os)

F&O Review Capability Finding Description Resolution Number Element Category (CC)

MU -B4-01 MU-B4 CC-I/II/III The VEGP plant procedures do not specifically This F&O is resolved.

Met call for a peer review after a PRA upgrade has been completed. But the plant has had this peer Procedure(s) outlining requirements dealing review and other peer reviews in the past. This with PRA configuration control, as change is required by the SA. referenced in ASMEIANS RA-Sa-2009, Section 1-5 (Reference 11), have been developed to comply with requirements of 10 CFR 50.69. The revision includes a reference for peer review update after a PRA upgrade has been completed. The PRA model update process is discussed in Section 3.2.1 .1 of this licensing submittal.

E1 -64 to NL-12-0932 Basis for Proposed Change 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 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 1111, 111111, 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. Note that differences that were considered typographical, editorial, or provided additional descriptions of the SRs were not considered technically significant and were excluded from Table 6.

Among these is the exclusion of adjectives, such as the term "key", because no model document can contain all assumptions, only those that are considered significant enough to merit mention.

The review concluded that the VEG P internal events (including internal flooding)

PRA model meets RG 1.200, Revision 2.

E1-65 to NL-12-0932 Basis for Proposed Change Table 6: Comparison of RG 1.200 Revision 1 and Revision 2 SRs Applicable to CC-I/II, CC-IIIIII, and CC-1I111111 SR in 2007 PRA Standard as SR in 2009 PRA Standard as Amended by RG 1.200, Amended by RG 1.200, Description of Change Resolution Revision 1 Revision 2 IE-C10: IE-C12: The sentences were NUREG/CR-6928 is used as the clarifications provided in RG source for generic data priors in CC-I/IIIIII: CC-I/IIIIII: 1.200 Revision 1 and Revision Revision 4 of the VEGP internal I

... ... 2, respectively . events PRA.

An example of an acceptable An example of an acceptable generic data sources is generic data sources is The updated SR cites a more NUREG/CR-5750 [Note 1]. NUREG/CR-6928 [Note 1]. recent example of an acceptable generic data source.

SY-B15: SY-B14: The sentences were As noted in Table 9.2-1 of the clarifications provided in RG internal events PRA calculation CC-IIIIIIII: CC-I/II/III: 1.200 Revision 1 and Revision (Reference 14), failure of the

... ... 2, respectively. containment boundary due to (h) harsh environments induced (h) harsh environments induced venting is not an applicable to by containment venting, or by containment venting, failure The updated SR explicitly the VEGP large, dry, failure that may occur prior to of the containment venting requires consideration of subatmospheric containment.

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

damage E1-66 to NL-12-0932 Basis for Proposed Change Table 6: Comparison of RG 1.200 Revision 1 and Revision 2 SRs Applicable to CC-IIII, CC-IVIII, and CC-VII/III SR in 2007 PRA Standard as SR in 2009 PRA Standard as Amended by RG 1.200, Amended by RG 1.200, Description of Change Resolution Revision 1 Revision 2 DA-C1: DA-C1: Reference NUREG-1715 was NUREG/CR-6928 is used as the added by RG 1.200 Revision 1; source for generic data priors in CC-IIII1I1I: CC-1I111111: References NUREG-1715 and Revision 4 of the VEGP internal

... ... NUREG/CR-6928 were included events PRA.

Examples of parameter Examples of parameter in the 2009 version of the PRA estimates and associated estimates and associated Standard.

sources include: sources include (a) component failure rates and (a) component failure rates and The updated SR cites more probabilities: NU REG/CR-4639 probabilities: NUREG/CR-4639 recent examples of acceptable

[Note (1)], NUREG/CR-4550 [2-7], NUREG/CR-4550 [2-3], generic data sources.

[Note (2)], NUREG-1715 [Note NUREG-1715 [2-21],

7] NUREG/CR-6928 [2-20]

QU-A2a: QU-A2: The LERF requirement was Section 10.3.2 of the internal added by RG 1.200 Revision 2. events PRA calculation CC-I/II/III: CC-I/II/III: (Reference 14) presents PROVIDE estimates of the PROVIDE estimates of the The updated SR explicitly estimates for individual LERF individual sequences in a individual sequences in a requires consideration of LERF. sequence cutsets.

manner consistent with the manner consistent with the estimation of total CDF ... estimation of total CDF (and LERF) ...

E1-67 to NL-12-0932 Basis for Proposed Change Table 6: Comparison of RG 1.200 Revision 1 and Revision 2 SRs Applicable to CC-IJII, CC-IIIIII, and CC-1I111111 SR in 2007 PRA Standard as SR in 2009 PRA Standard as Amended by RG 1.200, Amended by RG 1.200, Description of Change Resolution Revision 1 - Revision 2 aU-A2b: aU-A3: The phrase, "from internal The peer review based on the events", was deleted from the 2007 version of the PRA CC-I: CC-I: 2009 version of the PRA Standard (Reference 26)

ESTIMATE the point estimate ESTIMATE the point estimate Standard. The LERF addressed these LERF CDF from internal events. CDF (and LERF). requirement was added by RG requirements. Section 10.3.2 of 1.200 Revision 2. the internal events PRA I

CC-II: calculation (Reference 14) I ESTI MATE the mean CDF from CC-II: The SR explicitly requires presents the mean CDF LERF internal events, accounting ESTIMATE the mean CDF (and consideration of LERF. results. I for the "state-of-knowledge" LERF), accounting for the However, per the note in 2007 I correlation between event state-of-knowledg e SR LE-E4 and LE-F3, LERF I probabilities [Note (1 )]. correlation between event was addressed in applicable I probabilities [Note (1 )]. requirements of Table 4.5.8, CC-III : which includes all au SRs.

CALCULATE the mean CDF Thus, the peer review using the from internal events by CC-III: 2007 version of the PRA propagating the uncertainty CALCULATE the mean CDF Standard was addressed these distributions, ensuring that the (and LERF) by propagating the LERF requirements.

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

E1-68

Enclosure 1 to NL-12-0932 Basis for Proposed Change Table 6: Comparison of RG 1.200 Revision 1 and Revision 2 SRs Applicable to CC-IIII, CC-IUIII, and CC-UUIIII SR in 2007 PRA Standard as SR in 2009 PRA Standard as Amended by RG 1.200, Amended by RG 1.200, Description of Change Resolution I Revision 1 Revision 2 QU-B6: aU-B6: The LERF requirement was The peer review based on the added by RG 1.200 Revision 2. 2007 version of the PRA CC-I/II/III: CC-I/II/III: Standard (Reference 26)

ACCOUNT for system ACCOUNT for system The SR explicitly requires addressed these LERF successes in addition to system successes in addition to system consideration of LERF. requirements. The Level 2 PRA failures in the evaluation of failures in the evaluation of However, per the note in 2007 event trees presented in Section I accident sequences to the accident sequences to the SR LE-E4 and LE-F3, LERF 9.2 of the internal events PRA extent needed for realistic extent needed for realistic was addressed in applicable calculation (Reference 14)

I estimation of CDF. This estimation of CDF or LERF. requirements of Table 4.5.8, explicitly account for system I

accounting may be This accounting may be which includes all au SRs. successes.

accomplished by using accomplished by using Thus, the peer review using the I numerical quantification of numerical quantification of 2007 version of the PRA I success probability, success probability, Standard was addressed these complementary logic, or a complementary logic, or a LERF requirements.

I delete term approximation and delete term approximation and includes the treatment of includes the treatment of I transfers among event trees transfers among event trees I where the "successes" may not where the "successes" may not be transferred between event be transferred between event

~es. trees.

E1-69 to NL-12-0932 Basis for Proposed Change Table 6: Comparison of RG 1.200 Revision 1 and Revision 2 SRs Applicable to CC-IIII, CC-IIIIII, and CC-1I11/111 SR in 2007 PRA Standard as SR in 2009 PRA Standard as Amended by RG 1.200, Amended by RG 1.200, Description of Change Resolution Revision 1 Revision 2 aU-E3: aU-E3: The LERF requirement was The peer review based on the added by RG 1.200 Revision 2. 2007 version of the PRA CC-I : CC-I: Standard (Reference 26)

ESTIMATE the uncertainty ESTIMATE the uncertainty The SR explicitly requires addressed these LERF interval of the CDF results. interval of the CDF (and LERF) consideration of LERF. requirements. Section 10.4 of Provide a basis for the results. Provide a basis for the However, per the Note in 2007 the internal events PRA estimate consistent with the estimate consistent with the SR LE-E4 and LE-F3, LERF calculation (Reference 14) characterization parameter characterization parameter was addressed in applicable presents the uncertainty uncertainties (DA-D3, HR-D6, uncertainties (DA-D3, HR-D6, requirements of Table 4.5.8, intervals for both CDF and HR-G8,IE-C1S). HR-G8, IE-C1S). which includes all au SRs. LERF, with consideration of the Thus, the peer review using the state-of-knowledge correlation.

CC-II: CC-II: 2007 version of the PRA ESTI MATE the uncertainty ESTIMATE the uncertainty Standard was addressed these interval of the CDF results. interval of the CDF (and LERF) LERF requirements.

ESTIMATE the uncertainty results. ESTIMATE the intervals associated with uncertainty intervals parameter uncertainties (DA- associated with parameter I D3, HR-D6, HR-G8, IE-C1S), uncertainties (DA-D3, HR-D6, I taking into account the state- HR-G8, IE-C1S), taking into of-knowledge correlation. account the state-of knowledge correlation.

CC-III:

PROPAGATE parameter CC-III:

uncertainties (DA-D3, HR-D6, PROPAGATE parameter HR-G8, IE-C1S) .... (no change) uncertainties (DA-D3, HR-D6, HR-G8, IE-C1S) .... (no change)

E1-70

Enclosure 1 to NL-12-0932 Basis for Proposed Change Table 6: Comparison of RG 1.200 Revision 1 and Revision 2 SRs Applicable to CC-IIII, CC';'II/III, and CC-1I11/111 I SR in 2007 PRA Standard as SR in 2009 PRA Standard as Amended by RG 1.200, Amended by RG 1.200, Description of Change Resolution Revision 1 Revision 2 QU-E4: QU-E4: Separate requirements for CC-I, No action, CC-II met for 2007 II, and III were collapsed into a version of the PRA Standard CC-I: CC-1I111111 : single requirement for CC-1I111111 (Reference 14).

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

IDENTIFY how the PRA model CC-II: is affected (e.g., introduction of The updated SR assigns the EVALUATE the sensitivity of the a new basic event, changes to same requirement to all three results to model uncertainties basic event probabilities, CCs. Meeting CC-II: in the and key assumptions using change in success criterion, 2007 version of the PRA ,

sensitivity analyses [Note (1)]. introduction of a new initiating Standard assures that the new event). SR is met. I CC-III: I EVALUATE the sensitivity of the I results to uncertain model I boundary conditions and other I

assumptions using sensitivity analyses except where such sources of uncertainty have been adequately treated in the quantitative uncertainty analYSis

[Note (1)].

E1-71 to NL-12-0932 Basis for Proposed Change Table 6: Comparison of RG 1.200 Revision 1 and Revision 2 SRs Applicable to CC-IIII, CC-Ilflll, and CC-I/IIIIII SR in 2007 PRA Standard as SR in 2009 PRA Standard as Amended by RG 1.200, Amended by RG 1.200, Description of Change Resolution Revision 1 Revision 2 LE-F2: LE-F3: Separate requirements for CC-I, No action, CC-II met for 2007 II, and III were collapsed into a version of the PRA Standard CC-I: CC-I/IIIIII: Single requirement for CC-1/1I/111 (Reference 14).

PROVIDE a qualitative IDENTIFY and in the 2009 version of the PRA assessment of the key sources CHARACTERIZE the LERF Standard.

of uncertainty. sources of model uncertainty Examples: and related assumptions, in a The updated SR assigns the (a) Identify bounding manner consistent with the same requirement to all three assumptions. applicable requirements of CCs. Meeting CC-II: in the (b) Identify conservative Tables 2-2.7-2(d) and 2-2.7 2007 version of the PRA treatment of phenomena. 2(e). Standard assures that the new SR is met.

CC-II:

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

CC-III:

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

E1-72 to NL-12-0932 Basis for Proposed Change Table 6: Comparison of RG 1.200 Revision 1 and Revision 2 SRs Applicable to CC-lnI, CC-IVlII, and CC-VIVIII SR in 2007 PRA Standard as SR in 2009 PRA Standard as Amended by RG 1.200, Amended by RG 1.200, Description of Change Resolution Revision 1 Revision 2 IF-F2: IFPP-B2: The requirement to document Section 5 and Appendix A of the walkdowns performed in support internal flooding PRA CC-IIII/III : CC-I/I 1111 I: of plant partitioning was added (Reference 13) document the DOCUMENT the process used DOCUMENT the process used to the 2009 version of the PRA walkdowns performed to to identify ... flood areas, .. , For to identify flood areas. For Standard. validate information related to example, this documentation example, this documentation flood areas, flood sources, typically includes typically includes The updated SR cites examples SSCs, mitigation and other flood

... of acceptable documentation of related features in the flood (b) flood areas used in the (a) flood areas used in the the process to identify flood areas.

analysis and the reason for analysis and the reason for sources.

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

E1-73 to NL-12-0932 Basis for Proposed Change Table 6: Comparison of RG 1.200 Revision 1 and Revision 2 SRs Applicable to CC-IIIJ, CC-IIIIIJ, and CC-1I11111J SR in 2007 PRA Standard as SR in 2009 PRA Standard as Amended by RG 1.200, Amended by RG 1.200, Description of Change Resolution Revision 1 Revision 2 IF-B1: IFSO-A1 : The requirement to include the Potential flood sources identified fire protection system in Item (a) in Section 5 of the internal CC-I/II/III: CC-I/II/III: as a potential flooding source flooding PRA reviewed as part For each flood area, IDENTIFY For each flood area, IDENTIFY was added by RG 1.200 of 2009 peer review against the potential sources of flooding the potential sources of flooding Revision 1. This requirement 2007 version of the PRA

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

fluid systems (e.g., circulating fluid systems (e.g., circulating containment spray (CS),

water system, service water water system, service water The requirement to include the residual heat removal (RHR),

system, fire protection system, fire protection reactor coolant system in Item reactor coolant system drain system, component cooling system, component cooling (a) as a potential flooding tank (RCSDT), safety injection water system, feedwater water system, feedwater source was added to the 2009 (SI), and reactor water makeup system, condensate and system, condensate and version of the PRA Standard. system (RMWS). As outlined in steam systems) steam systems, and reactor Thus, it was not reviewed as the Plant Vogtle Internal coolant system) ... part of the peer review Flooding notebook, the conducted using that version of Containment Building (and RCS the PRA Standard. components therein) is not included in the scope of the internal flooding anal},sis.

E1-74 to NL-12-0932 Basis for Proposed Change Table 6: Comparison of RG 1.200 Revision 1 and Revision 2 SRs Applicable to CC-IIII, CC-IIIIII, and CC-1I111111 SR in 2007 PRA Standard as SR in 2009 PRA Standard as Amended by RG 1.200, Amended by RG 1.200, Description of Change Resolution Revision 1 Revision 2 IF-F2: IFSO-B2: The requirement to document The internal flooding PRA walkdowns performed in support documents the walkdowns CC-I/II/III : CC-IIIIIIII: of the identification or screening performed to validate DOCUMENT the process used DOCUMENT the process used of flood sources was added to information related to flood to identify applicable flood to identify applicable flood 2009 version of the PRA areas, flood sources, SSCs, sources. For example, this sources. For example, this Standard. mitigation and other flood documentation typically documentation typically related features in the flood includes: includes: The updated SR cites examples areas.

of acceptable documentation of (a) flood sources identified in (a) flood sources identified in the process to identify flood the analysis, rules used to the analysis, rules used to sources.

screen out these sources, screen out these sources, and the resulting list of and the resulting list of Since documentation of sources to be further sources to be further walkdowns was not in the 2007 examined 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 analysis analysis that version of the PRA

.. . Standard .

OJ calculations or other analyses (c) calculations or other used to support or refine the analyses used to support or flooding evaluation refine the flooding evaluation (d) any walkdowns performed in I support of the identification or screeninq of flood sources I E1-75

Enclosure 1 to NL-12-0932 Basis for Proposed Change I Table 6: Comparison of RG 1.200 Revision 1 and Revision 2 SRs Applicable to CC-IIII, CC-IIIIII, and CC-IIII1111 SR in 2007 PRA Standard as SR in 2009 PRA Standard as Amended by RG 1.200, Amended by RG 1.200, Description of Change Resolution Revision 1 Revision 2 IF-F2: IFSN-B2: The requirement to document The internal flooding PRA walkdowns performed in support documents the walkdowns CC-I/II/III: CC-I/II/III: of the identification or screening performed to validate DOCUMENT the process used DOCUMENT the process used of flood scenarios was added to information related to flood to identify applicable flood to identify applicable flood 2009 version of the PRA areas, flood sources, SSCs, scenarios. For example, this scenarios. For example, this Standard. mitigation and other flood documentation typically documentation typically related features in the flood includes: includes The updated SR cites examples areas.

... of acceptable documentation of (c) propagation pathways ... (a) propagation pathways ... the process to identify flood

.. . scenarios.

(d) accident mitigating features (b) accident mitigating features and barriers credited ... and barriers credited ... Since documentation of

... walkdowns was not in the 2007 (e) assumptions or calculations (c) assumptions or calculations version of the PRA Standard, it used in the determination of used in the determination of was not reviewed as part of the

... flood-induced effects on ... flood-induced effects on peer review conducted using equipment operability equipment operability that version of the PRA

... Standard .

(f) screening criteria used in the (d) screening criteria used in the analysis analysis (g) flooding scenarios (e) flooding scenarios considered, screened, and considered, screened, and retained retained E1-76 to NL-12-0932 Basis for Proposed Change Table 6: Comparison of RG 1.200 Revision 1 and Revision 2 SRs Applicable to CC-IIII, CC-IIIIII, and CC-1I11/111 SR in 2007 PRA Standard as SR in 2009 PRA Standard as Amended by RG 1.200, Amended by RG 1.200, Description of Change Resolution I Revision 1 Revision 2 (h) description of how the (f) description of how the internal event analysis internal event analysis models were modified ... models were modified ...

(j) calculations or other analyses (g) calculations or other used to support or refine the analyses used to support or flooding evaluation refine the flooding evaluation (h) any walkdowns performed in support of the identification or screening of flood scenarios IF-F2: IFQU-B2: The requirement to document The internal flooding PRA walkdowns performed in support documents the walkdowns CC-IIIIIIII: CC-1/1II111 : of internal flood accident performed to validate DOCUMENT the process used DOCUMENT the process used sequence quantification was information related to flood to define the applicable internal to define the applicable internal added in 2009 version of the areas, flood sources, SSCs, flood accident sequences and flood accident sequences and PRA Standard. mitigation and other flood their associated quantification. their associated quantification. related features in the flood For example, this For example, this The updated SR cites examples areas that are considered in documentation typically documentation typically of acceptable documentation of flood sequence definition.

includes: includes: the process to identify flood

... related features considered in (j) calculations or other analyses (a) calculations or other flood sequence quantification.

used to support or refine the analyses used to support or flooding evaluation refine the flooding evaluation Since documentation of

~

walkdowns was not in the 2007 E1-77 to NL-12-0932 Basis for Proposed Change Table 6: Comparison of RG 1.200 Revision 1 and Revision 2 SRs Applicable to CC-IIII, CC-IIIIII, and CC-1I111111 SR in 2007 PRA Standard as SR in 2009 PRA Standard as Amended by RG 1.200, Amended by RG 1.200, Description of Change Resolution Revision 1 Revision 2 (f) screening criteria used in the (b) screening criteria used in the version of the PRA Standard, it analysis analysis was not reviewed as part of the

... peer review conducted using (i) flooding scenarios (c) flooding scenarios that version of the PRA considered, screened, and considered, screened, and Standard.

retained retained (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 quantification - - -- - - - -- -- -- - I E1-78 to NL-12-0932 Basis for Proposed Change 3.3.1.5 General Conclusions Regarding PRA Capability The information provided in this section demonstrates that the VEGP at-power internal events (including internal flooding) PRA model conforms to the standard at CC-II which satisfies the guidance of RG 1.200, Revision 2 (Reference 28). In addition, the VEGP PRA model complies with all requirements for technical adequacy of the baseline PRA as defined in NEI 00-04.

The VEGP internal events (including flooding) PRA model technical capability evaluations described above provide a robust basis for concluding that the PRA model is suitable for use in supporting the implementation of 10 CFR 50.69.

3.3.2 Fire PRA Model The VEGP Fire PRA (FPRA) was reviewed in 2012 by the PWROG against Combined ASMEIANS RA-SA-2009 Standards (Reference 11), RG 1.200, Revision 2 (Reference 28), and NEI 07-12 (Reference 31). A self assessment of the VEGP FPRA against the ASMEIANS RA-SA-2009 was performed by SNC in house staff prior to the PWROG peer review. The overall results of the PWROG peer review (Capability Category and Findings) are described in section 3.3.2.1.

Section 3.3.2.2 summarizes the resolution of F&Os identified in the PWROG peer review.

The PWROG peer review exit meeting concluded that the Vogtle Unit 1 FPRA was complete, that the CDF was within an acceptable range, and that the supporting documents were complete. Some refinements to address F&Os and to address documentation issues were identified. The PWROG peer review team stated that:

• FPRA overall quality was very good

• FPRA team was qualified and capable

• FPRA was found to be of high technical quality

• The ignition frequency assumptions, sources, conditions and corresponding actions were explicitly addressed. The depth and completeness of the uncertainty related discussion in the documents was outstanding

• Documentation was extensive

• HRA was very detailed and evaluation of fire procedures was addressed

• Documentation of Internal Events F&Os resolution against fire PRA requirements was good The information provided in section 3.3.2 demonstrates that the VEGP FPRA model meets the requirements of RG 1.200, Revision 2.

E1-79 to NL-12-0932 Basis for Proposed Change 3.3.2.1 Previous Peer Review and Self Assessment for VEGP FPRA Model Before performing the PWROG peer review of FPRA against the Combined ASME/ANS RA-SA-2009 Standards (Reference 11), RG 1.200, Revision 2 (Reference 28), and NEI 07-12 (Reference 31) in 2012, a self-assessment of the VEGP FPRA model against the ASMEIANS RA-SA-2009 was performed by SNC in-house staff. The scope of the self-assessment included a review of the following FPRA tasks:

TASK 1 & 6: Vogtle Fire PRA Plant Partitioning and Fire Ignition Frequency TASK 2: Component and Cable Selection Task 3 & 9: Cable Selection and Detailed Circuit Failure Analysis TASK 5: Fire Model Development TASK 10: Circuit Failure Mode and Likelihood Analysis TASK 8 & 11: Fire Scenario Selection TASK 12: Human Reliability Analysis TASK 13: Seismic-Fire Interactions Assessment TASK 14 & 15: Fire Risk Quantification The results of the self assessment concluded that when compared against the SRs of the ASMEIANS RA-SA-2009 FPRA standard, all the SRs were classified as Met. This favorable outcome provided an assurance that the FPRA was ready for submittal to the PWROG peer review team.

3.3.2.2 Industry PRA Peer Review for VEGP Fire PRA Model The ASMEIANS RA-SA-2009 version of the PRA Standard (Reference 11) contains a total of 173 numbered SRs in 13 technical elements. The configuration control element has 10 additional SRs. Thus, a total of 183 SRs were assessed.

Among 183 SRs, 25 were determined to be Not Applicable, resulting in a total of 158 SRs that were assessed. 13 of the 25 not applicable SRs were associated with the QLS and QNS Technical Elements, which were assessed as Not Reviewed. After the PWROG Peer Review, a focused-scope peer review was conducted for the QLS and QNS elements that were marked as Not Reviewed by the peer review team. The focused-scope peer review dispositioned the 7 SRs for QLS as Met and the 6 SRs for QNS as NA (not applicable). The focused scoped peer review did not identify any new Findings.

Of the 158 total applicable SRs, approximately 80% met Capability Category II or higher, as shown in Table 7. It is noted that the results of the focused scope peer review are not included in the Table 7 statistics.

E1-80 to NL-12-0932 Basis for Proposed Change Table 7: Summary of VEGP Fire Events Capability Categories Capability  % of Total Category Met Number of SRs  % of Total SRs Applicable SRs Met 95 51.9% 60.1%

Not Met 31 16.9% 19.6%

CCI 5 2.7% 3.2%

CC II 7 3.8% 4.4%

CCIII 7 3.8% 4.4%

CC I1II 5 2.7% 3.2%

CC 11/111 8 4.4% 5.1%

NA 25 13.7%

NR 0 -

Total 183 100.0% 100%

The peer review generated 50 Findings out of which 31 SRs were judged to be not met. These Findings and their resolutions are described in Section 3.3.2.3. It is noted that several Findings resulted in more than one SR being judged as not met. The resolution of the Findings results in two SRs, SC-C2-02 and PRM-B13-01, being not met. Resolution of these two Findings relates to enhancing documentation and has no impact on any technical element of the analysis. These SRs will be resolved prior to implementation of the VEGP 10 CFR 50.69 Program. All SRs (including the two not met when resolved) are being met at CC II or better, with the exception of FSS-E3 and PP-B5. In the case of FSS-E3 and PP-B5, the treatment of these items in the FPRA satisfies the requirements of CC I, which is judged to be sufficient for this application.

Thus, the VEGP Fire PRA meets the requirements of RG 1.200, Revision 2.

3.3.2.3 Resolution of Findings from RG 1.200 Fire PRA Peer Review Table 8 shows the details of the 50 Findings and the associated resolutions developed after the peer review.

E1-81 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number PP-A1-01 PP-A1, NOT MET This F&O has been resolved, and a focused PP-C2 This SR states: INCLUDE within the global analysis scope Peer Review for the QLS element found boundary all fire areas, fire compartments, or all associated SRs are MET with no FINDINGS.

locations within the licensee-controlled area where a fire could adversely affect any equipment or cable The existing analysis includes a number of plant item to be credited in the Fire PRA plant response site locations where a postulated fire is either model including those locations of a sister unit that not reasonably expected to occur (natural draft contain shared equipment credited in the Fire PRA. cooling tower) or where a fire is not anticipated A review of PRA-BC-V-12-004, Version 0, identified to cause or require a plant shutdown several structures within the Protected Area that (demineralizer water). These locations were were identified within the global analysis boundary, originally screened from the analysis. All plant but were then screened from further analysis, but the locations within licensee-controlled area were justification for screening appears to be inadequate reassessed and the scope of locations included or incorrect. in the analysis was expanded. Those locations that were screened were addressed in a Table 2.1-1 of PRA-BC-V-12-004, Version 0, lists the focused-scope Peer Review for the QLS plant structures that were considered when element which found all associated SRs MET determining the scope of the Fire PRA. This table with no FINDINGS.

also "screens" some of the structures based on varying criteria. Since a screening is performed The overall FPRA analysis and related without performing any quantitative evaluations, the documentation was updated to reflect the screening needs to meet the requirements of Section changes in the global analysis boundary.

4.2-4 (QLS) of the ASME/ANS Standard. For a number of the structures screened, this criterion appears to not be met - and the structures should have been retained for further evaluation in the I analysis J E1-82 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number A number of structures are screened stating: "A walk down and a review of equipment layout and electrical one-line drawings confirmed it contains no equipment that would impact plant operations and is not susceptible to fire." This does not address the potential for cables that have the potential to impact plant operations to be in the facility. Any PAU that has cables whose failure could cause spurious operation of any equipment, system, function, or Operator action credited in the Fire PRA must be retained in the analysis. IF there are any cables in the facility that are unknown, the facility cannot be verified to not have cables that meet this criterion, and must be retained for further analysis."

Re-evaluate the facilities listed in Table 2.1-1 and ensure any that are "screened" are documented to meet the criteria specified in Section 4-2.4 (QLS) of the ASMEIANS standard. Any that do not meet the criteria specified should be retained in the analysis.

PP-A1-02 PP-A1 NOT MET This SR states: INCLUDE within the global analysis This F&O has been resolved.

boundary all fire areas, fire compartments, or locations within the licensee-controlled area where a The F&O details include items that overlap with fire could adversely affect any equipment or cable and duplicate F&O PP-A 1-01 . The discussion of item to be credited in the Fire PRA plant response the resolution of this F&O focuses only on those model including those locations of a sister unit that details not already addressed and resolved as contain shared equ~ment credited in the Fire PRA. described in the resolution for PP-A 1-01.

E1-83 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings I Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number A review of section 2.1 of PRA-BC-V-12-004, I

Version 0, indicates that the development of the fire The scope of the plant locations included in the PRA global analysis boundary may have missed analysis was reviewed and confirmed to have several key fire areas. included the locations identified in the F&O. All of the plant Fire Areas as well as the specific Section 2.1.2 describes the process that was used to locations identified in the F&O are included in develop the global analysis boundary, and Table 2.1 the analysis scope - none of these locations 1 is supposed to be a summary of all fire areas that were screened.

were considered for the global analysis boundary, and justification for what was excluded. A review of Those locations that were screened were this table indicates that some potentially significant addressed in a focused-scope Peer Review for fire areas were not identified as being within scope of the QLS element which found all associated the global analysis boundary. SRs MET with no FINDINGS.

Based on the fact that there are several major fire areas that are not identified as being with the Global Boundary, it is not apparent that all potential risk significant fire areas were identified as part of the global boundary analysis definition.

Develop a complete list of all structures, above ground and below ground, within the Owner Controlled Area for the Vogtle site. Once the list is developed, ensure that ALL fire areas are addressed within the analysis. If appropriate, identified fire areas can be qualitatively screened - but only if the criteria in Section 4.2-4 is fullv met.

PP-B2-01 PP-B2, NOT MET Calculation PRA-BC-V-12-004 was reviewed for This F&O has been resolved.

E1-84 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number PP-B3 treatment of non-rated barriers credited in defining PAUs. Table 2.1 addresses the credited non-rated Additional walkdowns were performed to confirm barriers and documents the results of a walkdown of the adequacy of the credited non-rated barriers the barriers. This table appears to be incomplete. for both units. The criteria that was used For example, the floor/ceiling between zone 49 and required barrier construction using non-zone 149 in fire area 1-AB-LJ-B is not addressed combustible materials (concrete or masonry (note that others may exist, this example was the block) with no visible openings or confirmation first example one looked at). This barrier is not a that there were no combustible targets within the rated per drawing AXYDR801. In addition, no fire zone of influence when projected beyond barriers for Unit 2 are included, and per discussions any opening within a barrier. In all instances, no walkdowns of Unit 2 were performed. Inspection the results of the effort confirmed that the of Unit 1 non-rated barriers for justification of Unit 2 existing treatment was consistent with the non-rated barriers is likely not sufficient. characteristics and configuration of the feature.

The documentation was expanded and Additionally, the walkdown evaluated spatial enhanced to include additional technical separations that were credited in definition of PAUs information to justify the basis for all credited and concluded they were adequate given there were non-rated fire barriers.

no ignition sources in the spatial separation or no propagation path in the zone of influence of postulated transients. This walkdown did not consider spatial separations in unit 2.

The analysis performed during the walkdown I appears to have supports that the barriers would "substantially contain the damaging effects of fires",

I but the documentation of the walkdowns is lacking I detail to call it a justification. For example, in some cases the description is "Closed Wall". Additional I E1-85 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number I details would add to credibility to the justification of the barriers."

Inappropriate credit of non-rated barriers could lead I to inappropriate definition of PAUs. I Review all non-rated barriers. As a minimum, review all Fire Areas that have multiple fire zones and I confirm that the zones are not separated by non rated barriers. Suggest that additional detail be added to table 2.1 to solidify the justification of using the barriers.

Walkdown unit 2 sj:)atial seRarations and document.

PP-BS-01 PP-BS NOT MET The first assumption in PRA-BC-V-12-004 states that This F&O has been resolved.

active fire barriers were credited as part of plant partitioning. No information is provided that indicates The design of the plant includes the use of fire where these barriers are used and does not justify dampers and active fire doors. All of the active the active fire barriers except to say that they are features credited in the FPRA were confirmed to utilized in the Fire Hazard Analysis (FHA). be the same as those used and controlled by the plant Fire Protection Program. The analysis Failure to identify the active fire barriers may hamper documentation was updated to add this model maintenance and no justification is provided. technical information. The FPRA does not credit any other active features.

Provide a list of the active barriers credited and the justification for credit. The associated SR is met at CC I which is judged to be adequate for this application and not crediting other active features is E1-86 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number I- ....

conservative for the calculation of fire risk contributions. The standard would require the analysis to credit additional active features that are not used and controlled by the plant Fire Protection ProQram to met CC 11/111.

PP-B7-01 PP-B7 MET Attachment D of the Plant Partitioning document This F&O has been resolved.

CC 1111/111 shows that walkdowns were performed but does not really indicate any information associated with the Documentation was expanded to extract details I fire barriers. Table 2.2-1 shows that walkdowns from the Fire Hazards Analysis and replicate in were performed on the Unit 1 non-rated barriers. the FPRA documentation. Other documentation Based on discussions with the plant team, the rated enhancements were also incorporated as noted barriers are controlled under the Fire Protection in the F&O.

Visual Inspection procedures 29144-1 and 29144-C.

These procedures should be referenced as part of Section 2.2 in PRA-BC-V-12-004.

Rated fire barriers are discussed in Attachment D and not clearly addressed in the current document.

Add reference to the FP visual inspections to PRA BC-V-12-004.

ES-A1-01 ES-A1 MET This SR States: IDENTIFY equipment whose failure, This F&O has been resolved.

CC 1/11/111 including spurious operation (see ES-A4), caused by an initiating fire would contribute to or otherwise The FPRA assumes at a minimum an induced cause an automatic trip, a manual trip per procedure plant trip for all locations within the global direction, or would invoke a limiting condition of analysis boundary. This treatment bounds the operation (LCO) that would necessitate a shutdown instances in which a manual shutdown may be where required given a Tech Spec LCO.

E1-87

Enclosure 1 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings i

Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element I (CC)

Number ,

(a) shutdo~wn is likely to be required before the fire is ,

extinguished, An operator reviewed each plant location to I (b) a potentially significant effect on safe shutdown determine if a fire at the location would result in capability is caused by the affected equipment, or a plant trip or manual shutdown. These (c) the shutdown will be modeled as a plant trip locations were included in the analysis. A rather than a slow, controlled shutdown of the plant focused-scope Peer Review was performed for based on the current modeling practice in the the QLS element which found all associated Internal Events PRA. SRs MET with no FINDINGS. Therefore, it was confirmed that a fire at a location in the global A Tech Spec LCO review could not be found. analysis boundary that may lead to a reactor trip given a Tech Spec LCO was retained in the The SR requires a systematic review of Tech Specs analysis.

to identify any equipment whose failure could invoke a LCO that would require a shutdown (no time frame Based on this existing treatment, no technical specified) WHERE other conditions/plant impacts change in the FPRA is needed to resolve this could also exist. No documentation of any LCO F&O. The analysis documentation will be review could be found. updated to clarify this element of the analysis.

Perform, and document a systematic review of Technical SpeCification, and any other governing documents that could require a plant shutdown, and ensure that any equipment that meets the criteria specified in this SR are added into the scope of the Fire PRA.

ES-C2-01 ES-C2 NOT MET Each category of this SR starts with: IDENTIFY This F&O has been resolved.

instrumentation associated with each operator action to be addressed, based on the following: a) fire- The Peer Review found that when the E1-88

Enclosure 1 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution i (F&O) Element (CC)

Number induced failure of any single instrument whereby one documentation for 5 of the 6 new operator of the modes of failure to be considered is spurious actions that were added to the FPRA was operation of the instrument. This was found to not reviewed it did not identify associated have occurred for several credited operator actions. instrumentation. A review of all credited operator actions was performed to confirm that All of the Operator actions from the internal events all associated instrumentation was identified and PRA model do identify at least one instrument/cue treated appropriately in the FPRA. That review that triggers the Operator to perform the action. determined that no technical change to the However, only 1 of the 6 Fire specific Operator FPRA model was required. An update to the actions identified at least one instrument/cue that documentation has been completed to add the would trigger the Operator to perform the action. information that was inadvertently omitted.

Since this SR states that "each operator action be addressed," and only one of the new operator actions meets this requirement, potentially significant instruments may not have been identified .

Identify at least one instrument/cue for each credited Operator action, and ensure that the instrument is included in the scope of the Fire PRA.

ES-D1-01 ES-D1 MET This SR is associated with the documentation of the This F&O has been resolved.

CC 11111111 Equipment selected and the process for identifying the equipment. The details in the F&O refer to the use of screening type codes in Table 4.1-3 of the Table 4.1-3 of the Component and Cable Selection analysis report. The report creates a source of report uses Type Codes that are not reflective of confusion as the type code that is referred to is current Vogtle Type Codes and a correlation not the type code used in the basic events E1-89 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number between the "N1" categorized Type Codes and the naming convention. The report will be updated Vogtle Type Codes does not exist. to clarify this usage.

Table 4.1-3 should be updated/revised to reflect the current Internal Events PRA Type Code naming scheme, or the type codes could be removed and only component types and failure modes used in the table.

CS-C2-01 CS-C2 MET The list of components, cables, and function codes This F&O has been resolved.

CC 1/11/1" (Attachment 3 of PRA-BC-V-12-005) did not contain all components modeled in the Fire PRA and The version of the database output that was contained a number of function codes not used in included in the report did not include the use of a the Fire PRA model. To facilitate review, upgrade, filter. As a result, records that are retained in and application, the list of function codes, the database for historical purposes were components, cables, etc. should be limited to only inadvertently included. The report has been those actually used in the Fire PRA model. updated to correct this issue. Accordingly, the circuit analysis packages from Attachment 3 Example: Condensate storage tank level now correspond to the functional states modeled transmitters 1LT5101 and 2LT51 01 are included in in the FPRA.

the Fire PRA model but were not listed in Attachment 3 of PRA-BC-V-12-005 (Data Query Circuit Report: Function Code Sort.) These transmitters are in listed in the ARC database along with the required power, instrumentation, and indication cables along with their cable routing.

I Missing information makes the document I E1-90 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number incomplete; extraneous information makes it difficult I to confirm which equipment was modeled in the Fire I PRA and unnecessarily complicates review and update.

I Consistently mark records as significant to FPRA based on the function codes used in the Fire PRA model and produce a report containing only FPRA-significant information.

CS-C2-02 CS-C2 MET A summary of the fire zone nomenclature (e.g. used CC 1111/111 in cable routing) and table associating fire zones This F&O will be resolved in a future with physical analysis units and referring to documentation update appropriate plant drawings and site maps would simplify review . This F&O refers to a documentation enhancement. The resolution of this F&O has Information is available but scattered, complicating no impact on any technical element of the review. analysis.

Condense the information from the FSAR Chapter 9A (Fire Hazards Analysis) into a table. Add nomenclature description and appropriate plant drawings and site maps.

PRM-A1-01 PRM-A1 MET This SR states: CONSTRUCT the Fire PRA plant This F&O has been resolved.

CC 1111/111 response model so that it is capable of determining fire-initiated conditional core damage probabilities The entire scope of mutually exclusive gates (CCDPs) and conditional large early release was reviewed and no other instances beyond probabilities (CLERPs) for various fire scenarios. the three occurrences identified in the F&O were Although most of the model appears to be found (MUTEX46, MUTEX-ISINJ-HE1, and E1-91 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number appropriate, there were several Mutex gates that do MUTEX529). All of these instances have been not appear to be technically valid and should be corrected. The results of these changes did not revised/removed to ensure that the Fire-induced alter the reported FPRA results for CDF or CCDPs and CLERP are modeled appropriately. LERF.

The following 3 Mutex gates appear to be invalid, and should be revisedlremoved from the Fire PRA Plant Response Model. The other Mutex gates modeled in the Fire PRM appear to be valid for the Fire PRA.

MUTEX46 - This configuration is allowed by Tech Specs and is not mutually exclusive so the logic should be removed (note that this is also true for the Internal Events model and logic should be removed from that model as well).

MUTEX-ISINJ-HE1 - This logic appears to be valid for an Internal Events scenario, but may not be valid for a Fire-induced initiator since the timing of the events for Internal Events is easy to follow, but the timing of events for the Fire is not as straight forward, and the combination of events may be phenomenologically possible. Suggest adding the

%ISINJ tag into the logic to retain the logic for Internal Events and eliminate it from the Fire-induced scenarios.

E1-92 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings I Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number _.

MUTEX529 - Discussions with SNOC personnel indicate that this combination is not expected to show up. Since it is unclear what the technical basis is behind this logic, it is recommended that this logic be removed from the model."

Remove/revise the 3 MUTEX gates as discussed above.

PRM-B13-01 PRM-B13 NOT MET This SR states: For any item identified per PRM- This F&O will be resolved prior to B12, PERFORM the data analysis portion of the Fire implementation of the 10 CFR 50.69 program in PRA plant response model in accordance with HLR- a future model update. This issue has no DA-A, HLR-DA-B, HLR-DA-C, and HLR-DA-D and technical impact on the quantification.

their SRs in Section 2 with the following clarifications: a) All the SRs under HLR-DA-A, HLR- This F&O refers to the naming convention used DA-B, HLR-DA-C, and HLR-DA-D in Section 2 are to in the FPRA for fire specific FLAG events. The be addressed in the context of both random events naming convention used was the same as that I as well as fire events causing damage to eqUipment used for random failure basic events instead of and associated cabling, and b) DEVELOP a defined the convention used for FLAGs. This creates a basis to support the claim of no applicability of any of source of confusion as basic events appear in these requirements in Section 2. the FPRA model with a 'zero' probability.

Based on the PRM document and the fault tree These events were added to the model for model, a significant number of new basic events potential fire impact on MSO and operator associated with equipment failures were added into actions from instrumentation impacts. These the Fire PRM. Each of the newly added basic events events are treated for fire induced failure only currently have a random failure probability of "0" and are set to "TRUE" or the circuit failure assigned to them. This convention means they likelihood probability when potentially impacted never randomly fail. If the intent of these added by a postulated fire. This treatment is consistent E1-93 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number failures was to function as "Tag/flag" events for use with SR SY-A1S. Documentation has been in the fire PRA only, which is an acceptable updated to clarify treatment of events added for approach, then the analysis should use the Vogtle fire induced failure concerns.

Tag/Flag naming convention instead of a basic event naming scheme that links to pre-defined type codes. This issue has no technical impact on the If the intent of these added failures is to function as quantification.

basic events, they need to be revisited, and data assigned to them in accordance with the requirements of this SA.

If the intent of these added failures was to function as "Tag/flag" events for use in the fire PRA only, then they should use the Vogtle Tag/Flag naming convention instead of a basic event naming scheme that links to pre-defined type codes. If the intent of these added failures is to function as basic events, they need to be revisited, and data assigned to them in accordance with the requirements of this SR. In either case, the documentation needs to be updated to identify all the added failures, and how they are intended to be used in the model.

PRM-C1-01 PRM-C1 NOT MET This SR is associated with the quality of the This F&O has been resolved.

documentation of the PRM model.

The information that was provided in the Although some portions of the PRM model analysis documentation included a listing of the documentation was easy to follow and understand, basic events from the CAFT A database. This there was a lot of extraneous/no longer used printout included extraneous events that are not E1-94 to N L-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number modeling included in the report that resulted in used in the model but exist in the database.

confusion as to what was and was not included in the current Fire PRA model. Additionally a listing of For the fire PRA, the database was purged to modified/added basic events was not provided and clearly identify basic events used in the fire could not be easily obtained from the basic event file. PRA. Additionally, the documentation was updated to include a table for modified basic Create a copy of the CAFT A BEfTC/GT databases, events, new basic events, and removed basic purge these copies, then re-generate the supporting events.

tables, and include a complete listing of modified/added basic events.

FSS-A1-01 FSS-A1 MET Review of the ignition sources within the physical This F&O has been resolved .

CC 11111111 analysis units suggests that some ignition sources have been screened out from the scenario selection The documentation that was provided to the process without proper justification and Peer Review team did not explicitly disposition documentation. Question FQ-A1 was submitted each of the individual fire ignition sources that during the peer review week associated with this were identified during the fire frequency issue. The answer to the question listed specific development. A review of all identified fire reasons for screening cabinets that are not fully ignition sources was performed and a documented and justified. The answer for example disposition was provided for each item. The includes a statement -8 sources were not used given results of this effort did not result in any I

potential for fire spread. No specific fire modeling additional fire initiating events needing to be justification for no fire spread was readily available. added to the analysis. The ignition sources that were not explicitly treated were excluded Fire scenarios that may inappropriately be excluded because a postulated fire would have no from the scenario selection process and consequential impact beyond itself (loss of only quantification. the fire source) or had been subsumed by another fire initiating event. The documentation E1-95 to NL-12-0932 Basis for Proposed Change

~

Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number I Account for each fixed ignition source identified so has been enhanced to provide this additional that a proper disposition is available and the risk technical detail.

contribution from each is accounted for .. Including the risk contribution that is properly screened out.

FSS-A2-01 FSS-A2 MET This F&O has been resolved.

CC 1111/111 Instrument tubing has not been considered as a potential target in the Fire PRA. Indications to The issue of instrumentation tubing is a potential support human failure events are credited in the concern for those instances where long lengths analysis, which brings into the scope of the fire PRA of tubing are used. In such cases, the failure to the instrument tubing. Instrument tubing should be consider fire impact to the tubing could result in included as targets and evaluated accordingly. non-conservative results. Supplemental walkdowns were performed to specifically Fire impacts to indications may not be fully examine instrumentation installations. In all accounted and reflected in the quantification instances, the length of tubing was minimal and process. the treatment of the instrument itself and related cabling was sufficient to bound the tubing Identify locations and damage thresholds for the exposure. The documentation has been instrument tubing and include them in the fire enhanced to provide this additional detail.

scenario development and quantification process ..

FSS-A3-01 FSS-A3, NOT MET Review of the FRANC database identified scenarios This F&O has been resolved.

FSS-E4 where events dispositioned as Y3 in the Equipment Selection Report are excluded events. The Y3 The use of assumed routing was applied on an disposition is used for components or groups of as-needed basis. As applied, the treatment is components for which the cable routing is not more appropriately described as crediting by known. By excluding these events, credit is being exclusion. That is to say, that while the routing taken for these components, that is, it is assumed of the circuits was not explicitly determined, the cables for the component(s) are not impacted by selected locations were identified where there the scenario. Some examples are scenarios 1530 was high confidence that no related equipment E1-96 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number A 532 A, 1048-DC 8 and 2068-MQ 81 where or circuits were present. This approach and events 1FW, 11A, HPCCW and othersY3 events are treatment is consistent with Note 8 of SR CS-excluded. This is not meant to be a complete listing. A 10. The documentation that was Peer Reviewed provided a general discussion of the approach and methodology but did not provide No basis is provided for crediting components where specific details for each instance where this cable routing is not known. approach was used. The documentation was Review the excluded events table for components enhanced to explicitly address and justify those dispositioned as Y3 in the Equipment Selection instances where this approach was used. No database and provide basis for exclusion. Document technical change to the FPRA was required.

the basis or remove the credit for the component. If The treatment remains conservative and assumed routing is used as a basis for exclusion, consistent with the Standard in the context that ensure supporting requirement CS-A 11 is met. equipment and cables that are not explicitly traced are treated as failed in all plant locations unless there is reasonable confidence that they]

are not present.

FSS-A5-01 FSS-A5 NOT MET Transient fires are not consistently postulated as This F&O has been resolved.

described in the FSS notebook throughout the unscreened physical analysis units. As examples, In concert with the additional plant walkdowns I no transients have been postulated in the that were performed to support the resolution of containment, cable spreading room or switchgear other F&Os, the overall treatment of transient rooms. The walkdowns conducted during the peer fires was also re-verified for both units.

review in the cable spreading room and switchgear Additional postulated transient fire initiating rooms indicate that there are transient fire scenarios events were defined and added to the analysis.

that should be postulated as there are trays or The incorporation of these additional fire conduits within the zone of influence of a transient initiating events has a minimal effect on the CDF fire. This a systematic issue not limited to the areas and LERF results. In all cases, the walked down. consequence of the postulated transient fire is E1-97

Enclosure 1 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element I (CC)

Number I bounded by another existing fire initiating event.

The risk contribution of the different physical analysis However, the transient initiating event frequency units can be underestimated. is several orders of magnitude lower.

ConSistently postulate transient fires throughout the unscreened physical analysis units.

FSS-A5-02 FSS-A5 NOT MET This SR is associated with assigning the correct This F&O has been resolved.

ignition frequency(s) to evaluate the risk of the PAUs. A review of the PAU-Ievel ignition The observed differences in the ignition frequencies documented in Task 6 does not always frequency has been addressed and corrected as match the sum of the PAU scenario level ignition necessary in conjunction with the resolution of frequencies. F&Os FSS-A 1-01 and FSS-A5-01 .

Some of the summations of ignition frequencies for PAUs overcount the total ignition frequency for the PAU (e.g. 1044-02 has a total ignition frequency in FRANC of 7.21 E-4/yr but the summation for the PAU from Task 6 is only 6.81 E-04/yr), while others undercount the total ignition frequencies (e.g. 1140A S1 has a total ignition frequency of 4.7BE-04/yr in FRANC but the summation for the PAU from Task 6 is 2.18E-03 [1 .59E-03/yr from fixed sources]). No bases for these discrepancies were provided .

Ensure that the total ignition frequency for each PAU is partitioned appropriately. Justify excluding the ignition source contribution for any ignition sources that are "screened" from having an impact.

E1-98

Enclosure 1 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings I

Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number FSS-B1-01 FSS-B1 NOT MET The FSS notebook does not describe abandonment This F&O has been resolved.

criteria for leaving the control room due to operability. Abandonment is only based on fire The treatment of control room abandonment in generated conditions (i.e. habitability). Consider as the FPRA and the related operator interview the context of this scenario a relatively small fire in a results were re-assessed. The operator control board affecting enough controls (e.g. interviews indicated that they would generally be resulting in a high CCDP) forcing operators to leave reluctant to abandon the main control room.

the control room before fire generated conditions The existing FPRA treated this reluctance by not force abandonment. crediting possible recoveries using designed plant features supporting shutdown from outside Abandonment based on plant operability is not the main control room.

considered or evaluated.

The FPRA quantifies the fire consequence for Include in the Fire PRA modeling of control room equipment inside and outside the control room.

abandonment due to operability. The cable end Each fire scenario was reviewed to determine pOints and their mapping to basic events can provide the impact on operation from the control room.

insights on determining abandonment. Fire scenarios resulting in high consequence are control room abandonment scenarios which were treated with a 1.0 CCDP given the

.. potential of hot shorts resulting in failure of equipment relied upon when shutting down outside the control room. Other scenarios with a significant CCDP include fires at the electric power control board which may result in a station blackout condition.

Other fire scenarios inside and outside the control room result in at most the loss of a single E1-99 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number division and oHsite power. Given these failures, the control room still has full operation of the opposite division with a diesel generator.

Therefore, the FPRA addressed the concern of loss of control resulting in abandonment. The documentation was updated to further describe the review of potential fire scenarios resulting in a loss of suHicient controls resulting in abandonment.

FSS-B2-01 FSS-B2 NOT MET This F&O has been resolved.

The FRANC database distributed during the peer review week only list one transient fire for the control The resolution for this F&O is subsumed by the room. No documentation for the screening process actions associated with F&O FSS-A5-01.

of transient fires has been provided.

The risk contribution for transient fires in the control room should be considered. The control room for example should have relatively high transient influence factors for occupancy and maybe in other factors.

Add transient fire scenarios to the control room so that the contributions from such ignition sources are captured.

FSS-B2-02 FSS-B2 NOT MET This SR is associated with Main Control Room This F&O has been resolved.

(MCR) abandonment scenarios. The documentation states that all MCR abandonment scenarios have a The FPRA treatment for all abandonment cases E1-100 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number I CCDP of 1.0 (0.91 when accounting for capacity was reviewed to confirm that a correct CCDP factor), but 7 scenarios were identified that had a value was applied. The instances identified in  :

CCDP of 1.2809E-03. (ABN-1 through ABN-7). the F&O were determined to be the only cases I where an incorrect value was used. The Since these scenarios were not set to 1.0, (0.91 analysis and associated documentation have I when accounting for capacity factor) as per the been corrected as necessary.

methodology described, and additional ex-control room Human Error Probabilities (HEPs) were not reviewed for them, this is undercounting the CDF/LERF impact of MCR abandonment.

Revise the CCDPs to 1.0 (0.91 when accounting for capacity factor) for the impacted scenarios.

FSS-C4-01 FSS-C4 NOT MET The severity factors approach is based on an This F&O has been resolved.

unreviewed analysis method. Currently in the FPRA there appears to be no credit for suppression The application of the method to the FPRA is activities when the severity factor is applied. consistent with the results of the recent industry Therefore, the scenario quantification does not limit review completed subsequent to the Peer the severity of the fire from the perspective of Review. This method involves the application of suppression credit. This assessment is limited to the a modified fire factor d~veloped through a treatment of severity factors and non suppression review of the individual industry fire events that probabilities at the time of the peer review. form the basis for the generic fire frequency.

However, the FPRA has applied a conservative value - higher than the value proposed in the method reviewed.

FSS-C4-02 FSS-C4 NOT MET A review of control room abandonment scenarios This F&O has been resolved.

indicated that some of the abandonment scenarios (ABN3, ABN4, ABN6, and ABN7) have an additional All entries used in the FRANC NSP field were E1-101 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number NSP of 0.1 assigned. However, an NSP of 1E-3 is reviewed and confirmed to be appropriate. The already taken in the Control Room Abandonment review found that the instances identified in the calculation. F&O are the only cases where an incorrect value was used. The analysis and the Non-Suppression Probability (NSP) appears to be associated documentation have been corrected.

double counted for these events.

Remove the 0.1 factor or provide additional justification to show that double counting is not occurring.

FSS-C7-01 FSS-C7 NOT MET Suppression systems are credited in the Fire PRA. This F&O has been resolved.

Specific examples include the credit for manual suppression for hot gas layer scenarios and the The F&O identifies anum ber of situations where credit for fixed suppression systems in the multi the independence of multiple suppression paths compartment screening analysis. Assessment for did not adequately justify the applied failure multiple suppression paths is not provided. probabilities in the context of dependencies between those paths. The dependency involves The standard requires evaluation of potential potential failures of the fire water system that conflicts with the credit of multiple suppression could disable water based suppression systems paths. as well as the use of hose streams in support of manual suppression. A review of the overall Evaluate P&ID's for fire water and document FPRA confirmed that the only instance where independent paths for the features credited. this occurs is in the MCA considerations. The overall treatment of these multiple suppression credits in those treatments was re-assessed and the manual suppression credit was either reduced to account for dependencies as necessary. The net impact of this update did not result in an~ change in the overall anal~sis E1-102

Enclosure 1 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings I

Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number results.

FSS-D3-01 FSS-D3, NOT MET This standard requirement indicates that there needs This F&O has been resolved .

FSS-C3 to be reasonable assurance that the risk contribution is bounded or realistically characterized . There is no In concert with the additional plant walkdowns clear/consistent treatment of intervening that were performed to support the resolution of combustibles in the Fire PRA that would ensure that other F&Os, the definition of the fire scenarios risk contribution is bounded. Currently, the analysis was re-verified for both units. This re-verification includes the assumption of only two cable trays included specific consideration of secondary would be involved in a fire . This assumption has the combustibles that could participate in the potential for underestimating the temperature in a postulated fire scenarios. The result of these PAU resulting in a hot gas layer scenario been walkdowns did not identify any cases where the missed. existing fire scenarios were found to be insufficient with respect to bounding the Risk contributing scenarios can be missed or consequences of the postulated fire event.

excluded from the quantification .

Develop and apply a consistent treatment of secondary combustibles that reflect plant specific configurations.

FSS-D4-01 FSS-D4, NOT MET Key input parameters for the models, including room This F&O has been resolved.

FSS-D11, sizes and heat release rates associated with FSS-H4 intervening combustibles are not properly justified. The analysis documentation was enhanced to I Specifically, PAU sizes are referenced to the provide a more complete summary of the combustible calc (FHA) and are based on floor various fire modeling parameters. As noted in areas. No verification is included for these values. the resolution discussion of F&O FSS-D3-01 ,

The heat release rates associated with intervening this included confirmation of the appropriateness combustibles is not justified for specific scenarios. of the treatment of secondary combustibles Finally, the heat release rate for transient fires in a (intervening combustibles).

number of PAU's is assumed to be 69 kW, which E1-103 to NL-12-0932 Basis for Proposed Change

~ ~~

Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number appears to be developed from an unreviewed With respect to the heat release rate used for analysis method (no specific reference to reviewed transient fires, many of the plant locations are industry documents for this value is provided). The spatially smaller which necessarily restricts the parameters appear to be applied on an as needed floor area for the placement of any transient basis using different approaches. A table listing all ignition source and combustible. This limited the scenarios with the corresponding parameters for floor space translates to a lower credible heat each scenario is not readily available in the release rate. For the large plant locations, a documentation to show/support the parameters used larger heat release rate was used. The overall for each scenario. treatment was consistent with the latest industry guidance as developed by an EPRI sponsored Input parameters for fire modeling calculations review effort and distributed to industry. The should be justified, verified, and documented in a consensus was that NUREG/CR-6850 allowed manner that allows a reviewer to see what was for modified heat release rates for different plant applied for each scenario, and the basis for selecting locations.

the scenario-specific parameters.

Justify, verify, and document input parameters for the fire modeling analysis. At the time of the peer review, critical input parameters including PAU sizes and heat release rate intensities were not adequately justified or verified. A consolidated table listing the I fire scenarios and their specific parameters with I clear concise description of each parameter and the resulting frequency used in the scenario quantification process would be an efficient way to ensure clear, concise validation, documentation and completeness of the input parameters used in the analysis.

E1-104 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC) I Number I FSS-D7-01 FSS-D7, NOT MET Detection and suppression features are credited in This F&O has been resolved.

FSS-D8 the analysis. In addition, the multi compartment analysis credits fixed suppression systems. Since The F&O raises a concern with respect to NFPA suppression is credited in the FPRA, a review and code compliance and system availability. An documentation of governing procedures should be effort was undertaken to confirm that credited included in the FSS report. Assumptions associated fire protection features conform to the with detection and suppression activities are not requirements of the applicable fire protection properly documented. This particularly applies to codes, and that there has not been any outlier specific features at Vogtle, as the FSS report experience with respect to availability. In all essentially references generic/reviewed industry cases, no conditions were found that would methods. No documentation/justification that the require any technical (numerical) change in the generic failure probabilities are consistent with the FPRA.

installation and maintenance of the system, the system is fully operable, and meets applicable codes and standards is provided. Justification for the applicability of the generic failure probability values used in the analysis should be provided. In addition an assessment of the effectiveness of the system for the credited scenarios is required. References to specific plant procedures, pre fire plans, etc are necessary. As an example of the impact of the SR in the analysis, during the peer review walkdowns it was observed that fire protection valves are under replacement requiring fire watch in some physical analysis units.

The standard requires justification for the values selected and an assessment of the system's E1-105 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number effectiveness, including verification that plant-specific unavailability history is not an outlier with respect to industry experience.

Document Vogtle specific detection, suppression and manual suppression characteristics. Justify the detection/suppression probability values used in the analysis. Validate and document that the Vogtle plant-specific unavailability is not an outlier with respect to industry experience.

FSS-E3-01 FSS-E3 MET Southern Nuclear PRA Calculation No. This F&O has been resolved .

CCI 0293100011.11, Rev.O, Section 1.3 (Fire Scenario Selection Report) and Vogtle Fire PRA Quantification The specific issues that were identified for this Report No. 0293100011.14, Rev. 0, Section 7.0 F&O involves uncertainty associated with the provides a characterization of the uncertainty in original NUREG/CR-6850 fire frequency values various aspects of the analysis. The Monte Carlo and uncertainty associated with the input analysis in Section 9.0 of Quantification Report No. parameters to the fire modeling analyses. The 0293100011.14 does not reflect all uncertainties FPRA treatment of uncertainty included a because distributions are not identified for all treatment of all parameters that can be parameters. mathematically propagated within the capabilities of available analysis tools and Upgrade to achieve Category II. software.

Establish uncertainty distributions for all uncertainties In the case of fire frequency, uncertainty and propagate the distributions through the Fire PRA analysis was performed based on the updated model using a Monte Carlo simulation . fire frequency values (EPRI TR-1016735). The specific requirement for sensitivity using NUREG/CR-6850 values is specific to NFPA E1-106 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number 805 applications as noted in FAQ 08-0048.

In the case of fire modeling input parameters, that task was performed in a fashion that tended to skew the results towards a conservative upper bound . As such, the results are already conservative and any uncertainty analysis would tend to result in lower CDF/LERF results.

However, there is no consensus basis for establishing the uncertainty characterizations.

The associated SR was dispositioned as CC I which is judged to be sufficient given the two concerns noted.

FSS-F1-01 FSS-F1 NOT MET Southern Nuclear PRA Calculation No. This F&O has been resolved.

0293100011.11, Rev.O (Fire Scenario Selection Report), Section 4.1.7.4, indicates that plant A review and walkdown of the plant was buildings have requirements for fire ratings as performed to identify locations where exposed identified in Design Criteria DC-1000A (General structural steel was present together with a high Design Criteria - Architectural) and that this should hazard fire source. The result of this effort did constitute compliance with this supporting not identify any instances where both co-existed.

requirement. However, this supporting requirement Therefore, there was no need to modify the necessitates a detailed inspection and listing of analysis for treatment of possible structural buildings with exposed structural steel and whether collapse. The analysis documentation was any high hazard fire sources exist in those buildings. enhanced to provide this information.

This has not been provided in Calculation No.

0293100011.11. Walkdowns conducted during the peer review week suggests that there are exposed E1-107 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number ~---

structural steel elements throughout the plant. A detail inspection following the requirements in the FPRA standard may result in fire scenarios requiring quantification.

Completeness of analysis for all buildings in the global analysis boundary, and documentation of this analysis for future use.

Compliance would typically involve providing a table of all buildings in the global analysis boundary, identification of whether exposed structural steel exists in each building, and identification of whether there is a postulated fire source/fire scenario that could damage the building structure. If an assumption of building collapse is made in FSS-F2 instead of further evaluating the structural capability of a particular building, then associated scenarios must be evaluated and included in the FSS-F3 risk Quantification.

FSS-G3-01 FSS-G3 MET This F&O has been resolved.

CC 1111/111 There is no systematic process for identifying multi compartment scenarios. Specifically, a multi A comprehensive multi-compartment matrix was compartment matrix is not developed. The lack of a developed and added to the analysis rigorous process for defining multi compartment documentation. This process was also used to combinations, multi compartment scenarios can be support the resolution of other F&Os related to missed and not included in the quantification. For barrier adequacy and active fire barrier features .

example, vertical combinations of compartments The results of this enhancement of E1-108

Enclosure 1 to NL-12-0932 Basis for Proposed Change

~

I Table 8

• Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number (compartments above or below the compartment of documentation did not identify any new fire fire origin) are not considered in the analysis. scenarios that required quantification or further treatment in the FPRA.

Multi compartment scenarios may be missed and not quantified.

Generate a multi compartment matrix and apply the screening criteria to all combinations. Quantify the unscreened combinations.

FSS-G4-01 FSS-G4, NOT MET No justification is provided for the credit allowed for This F&O has been resolved.

FSS-G5 non rated barrier or active fire barriers. The documentation is not clear on how the barrier failure As noted in the resolution discussions for F&Os probabilities are assigned to each scenario. PP-B2-01, PP-B5-01, PP-B7-01, and FSS-G3 Walkdowns conducted during the peer review week 001, additional assessment, walkdowns, and where the spatial separation in the aux building documentation enhancements were performed (equipment hatch) is next to relatively large pumps. to address the associated concerns. The net result of the efforts did not change the resultant analysis in the context of CDF and LERF results.

Barrier failure probabilities are important screening Various documentation enhancements were factor and/or parameter in the risk equation. Proper also incorporated to include the related justification for its use should be documented and inform ation.

applied. The peer review finds that the application of barrier failures probability in the FPRA is inconsistent with the definition of physical analysis units in covered by the PP element. The spatial separation or active fire barrier probabilities are not consistent.

For example, spatial separation is credited in plant partitioning and the multi compartment does not E1-109 to NL-12-0932 Basis for Proposed Change Table 8

• Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number I

assign barrier failures of 1.0 to those combinations.

Document and apply a consistent method for determ ining using barrier failure probabilities.

FSS-H1-01 FSS-H1, NOT MET The fire modeling documentation appears to not be This F&O has been resolved.

FSS-H5 complete or reviewed at the time of the peer review.

For example, information important to the Additional reviews of the report were performed development of scenarios such as the handling of and corrections have been incorporated. The fire wraps which are present in the plant is not process of completing this effort did not identify included in the documentation so it is not clear any substantive change to the technical portions whether or not they are credited in the analysis. of the analysis or methodology discussions.

Additionally, there are references to the Duane Arnold Plant in the documentation, which would have I

been flagged if the documentation had been reviewed. I Peer reviewed should receive complete and reviewed Fire PRA notebooks. With the notebook in the state it is in - it is not clear what is applicable to Vogtle, and what is "leftover" from the Duane Arnold Fire PRA. .

Review and resolve comments for the FSS notebooks and ensure that all Vogtle specific information is included in the report.

FSS-H7-01 FSS-H7 NOT MET Suppression is credited in the FPRA. Review and This F&O has been resolved.

documentation of governing procedures should be E1-110 to NL-12-0932 Basis for Proposed Change Table 8

• Resolution of the VEGP Fire PRA Peer Review Findings I

Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number

,I included in the FSS report. Assumptions associated As discussed in the resolution for FSS-D7-01, with detection and suppression activities are not supplemental reviews were performed to confirm properly documented. This particularly apply to the appropriateness of the credit assigned for specific features at Vogtle, as the FSS report various fire protection system elements essentially references generic/reviewed industry (suppression, detection, barriers, etc). This methods. References to specific plant procedures, process and the associated governing plant pre fire plans, etc are necessary. This is a procedures were added to the analysis documentation issue directly related to the SR. documentation.

Document Vogtle specific detection, suppression and manual suppression characteristics .

IGN-A7-01 IGN-A7 MET This F&O has been resolved.

CC 1/111111 Walkdown documentation does not allow confirmation of all ignition sources being included in During the course of the development of the the Fire PRA process. Instances were found where FPRA, instances were identified where fire items are in the model and not in Appendix C or D of ignition sources were not explicitly included in Southern Nuclear PRA Calculation No. PRA-8C-V- the counting for fire ignition frequency. As the 12-004 (Plant Partitioning and Fire Ignition analysis was developing, it was not viewed as a Frequency, Version 0, NUREG/CR-6850 Task 1 & significant issue as it would have tended to 6). under-estimate the denominator of the fraction used in the analysis resulting in very slightly Documentation and understanding of the Fire PRA conservative fire frequency values. Given the model for future use. minor changes in the population of counted components, the exclusion of these items from Correlate the Fire PRA model and the equipment the fire ignition frequency development has items listed in the ignition frequency report. Avoid almost no impact on the fire frequency for any I having non-existent equipment items listed in the individual scenario. Instead, a per component model with a zero CCDP that are not in the ignition ~frequeDcy was used to generate a value for I E1-111 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number frequency report. any 'missing' item during the fire scenario development process. The very slight conservative biasing of the results does not have a significant impact on the FPRA results.

IGN-B1-01 IGN-B1 MET This F&O has been resolved.

CC 1111/111 Ref: Southern Nuclear PRA Calculation No. PRA BC-V-12-004 (Plant Partitioning and Fire Ignition The relationship between fire areas and zones Frequency, Version 0, NUREG/CR-6850 Task 1 & that are used in the analysis was reviewed. It 6). was confirmed that the error identified in the F&O is isolated. The analysis documentation Calculation of the ignition frequency for Fire Areas 1 has been corrected.

AB-LD-A and 1-AB-LB-A appears to have included fire zone 1035-C6 in the wrong fire area."

Incorrect PAU ignition frequency may be used in future Fire PRA calculations or risk-informed decisions.

Revise the documentation in Calculation No. PRA BC-V-12-004 to reflect the correct ignition frequency.

IGN-B4-01 IGN-B4 MET Section 3.2.1 of Southern Nuclear PRA Calculation This F&O has been resolved.

CC 11111111 No. PRA-BC-V-12-004 (Plant Partitioning and Fire Ignition Frequency, Version 0, NUREG/CR-6850 A Bayesian update of the generic fire frequency Task 1 & 6) states that Bayesian update was not values for Bins 5, 6, 11, 24, and 31 has been performed for Bins 5, 6, 11, 24, and 31. No performed. The posterior mean reflects a very justification was documented for this in the report. If slight reduction. The FPRA and related reports I there is justification for not doing a Bayesian update, have been updated to reflect these updated due to reclassifying generic events within the results . I E1-112 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number industry, this should be documented. Otherwise these Bins should be Bayesian updated.

Technical accuracy and consistency with treatment of all other BIN ignition frequencies.

Document basis for not Bayesian updating the ignition frequencies for Bins 5, 6, 11, 24, and 31 in Section 3.2.1 of Southern Nuclear PRA Calculation No. PRA-BC-V-12-004 (Plant Partitioning and Fire Ignition Frequency, Version 0, NUREG/CR-6850 Task 1 & 6).

HRA-B3-01 HRA-B3 MET For new operator actions, particularly OACONTROL- This F&O has been resolved.

CCI -AFW and OAISOLSTMTDAFW, which are listed as risk significant, the HRA evaluation does not include At the time of the Peer Review, the FPRA used a scenario description, a procedure reference, the applied both detailed and screening HRAs. The operator cues, or operator interview insights. In F&O identified a number of instances where particular, actions to correct spurious actuation screening HEPs were used that were found to would require the operators to determine that the be risk significant. The FPRA has been updated system is not needed before tripping the system. to include detailed HRA for all modeled actions.

The analysis and associated documentation Screening value for the HRAs has been justified by have been revised to reflect this update.

the cause-based decision trees. However, the assessment does not have sufficient analysis to verify that the screening value is appropriate.

For risk significant fire actions, perform a detailed HRA evaluation with operator talk-through ,

E1-113

1\ to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings I

Finding &

Capability ,

Observation Review (F&O) Element Category

,, Finding Description Resolution Number (CC) _. I procedure references and cues to perform the action and cues that may impact the diagnosis. For non-risk significant actions, provide the procedural guidance and the justification for the screening values used.

FQ-A1-01 FQ-A1 MET This F&O has been resolved.

CC 1/11/111 This SR states: for each fire scenario selected per the FSS requirements that will be quantified as a Due to the potential for spurious actuation given contributor to fire-induced plant CDF and/or LERF, a fire resulting in abandonment conditions, a TRANSLATE the equipment and cable failures, CCDP of 1.0 was assumed for abandonment including specification of the failure modes, defined scenarios. Due to the low probability of MCR per the FSS element into basic events in the Fire abandonment, the use of CCDP of 1.0 does not PRA plant response model including consideration of result in these scenarios as significant insights from the circuit failure analysis. Contrary to contributors to plant risk.

this requirement, the MCR abandonment scenarios did not translate the equipment and cable failures, Given the methodology that is used, the explicit but assumed a CCDP for the scenarios." mapping of equipment and cable failures to basic events to support CCDP quantifications is Most scenarios appeared to translate the equipment not relevant given the treatment method that is and cable failures appropriately. The exception to used.

this was the MCR abandonment scenarios which did not translate the equipment and cable failures, but used an assumed CCDP for the scenarios. A review of the information currently available for the MCR scenarios showed that the information required to do this for the MCR abandonment scenarios was readily available and could be generated .

E1-114 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP -Fire PRA Peer Review Findings I

Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number Translate the equipment and cable failures for the MCR abandonment scenarios.

FQ-B1-01 FQ-B1 NOT MET This F&O has been resolved.

This SR has back-references to the Internal Events I QU SRs. In particular, QU-B2 and QU-B3 are The F&O details refer to the lack of concerned with truncation limits and showing documentation for convergence. The FPRA was convergen ce. quantified using FRANC with a CCDP truncation of 1E-9. A formal test for convergence has been Although the quantification document does provide a performed and shows convergence occurs with discussion on truncation and convergence, the a CCDP truncation of 1E-7. Therefore, the use method used to obtain/provide evidence of of a CCDP truncation of 1E-9 is acceptable.

convergence does not technically meet the The analysis documentation has been updated requirements of these SRs. Although the document to include the FRANC CCDP truncation study.

does not specifically discuss the process used to obtain the convergence results, discussions with SNOC personnel indicate that the model was only "quantified" one time at a truncation value of 1E-9 to obtain CCDP cutsets. These cutsets were merged together and then the ignition frequencies/Severity factors/non-suppression probability values were applied to the cutsets. Once the factors were added, cutsets were removed below the specified new "truncation" values to show convergence. Although this method would show where convergence is obtained within the single quantified cutset file, it does not show convergence of the accident sequences and associated system models since only a sinqle quantification was performed. Quantification E1-115 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number - - _.

at higher/lower CCDP truncation values will provide different cutsets, that when the IF/SF/NSP values are applied and the cutsets below the new "truncation" values are determined, different results are obtained.

Quantify the FRANC model at various CCDP truncation values to verify convergence.

FQ-B1-02 FQ-B1 NOT MET This SR has back-references to the Internal Events This F&O has been resolved.

QU SRs. QU-B1 provides requirements for use of software for quantifying. The supplemental software tools that are used in the development of the FPRA include an Access This F&O does not address standard quantifying database and a cutset merging utility.

software (CAFTA, FTREX, etc.), but rather it focuses on non-standard software which is used in the The VEGP Scenario Database is an Access quantification process. Per discussions with ERIN database with simple queries that involves personnel the following software is used in the somewhat standard data relationships largely quantification process: identical to that in FRANX. The data tables were verified to ensure the queries were VEGP Scenario Database - This database is used to providing the correct data. The verification has generate scenarios. The queries in this database been included in the updated quantification could be construed to be software. report.

CTT Appender - This ERIN in-house software is A benchmarking of CTT Appender was used for appending cutset files from the various performed by not setting a truncation in CTT scenarios into a single file. In addition to appending Appender. The resulting answer matched the files together, the code adds scenario IDs to each FRANC answer. The verification has been I

cutset and creates and replaces BEs for altered included in the updated quantification report.

E1-116 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number events with a unique basic event so as to avoid loss of the altered event probability. In both cases , the favorable comparison of results is the basis to conclude that the Both of these items are subject to the requirements requirements of QU-B1 are MET.

of QU-B1.

There is no documentation that the VEGP Scenario Database and the CTT Appender meet the requirements of QU-B1 and they have been used in the quantification process.

Perform software qualification as needed and document how the requirements of QU-B1 are met for the VEGP Scenario Database and CIT Appender.

FQ-B1-03 FQ-B1 NOT MET This SR has look backs to the Internal Events QU This F&O has been resolved.

SRs. In particular, QU-B1 is concerned with ensuring that method-specific limitations and The details for this F&O refer to documentation features of software used in the quantification of software limitations. This F&O is related to, process that could impact the results are identified. and to some degree duplicates, the concerns noted in FQ-B1-02. The discussion of the A review of the documentation did not identify any resolution of FQ-B1-02 resolves the F&O related discussion associated with understanding the to software.

method-specific limitations and features of the software selected to support the Fire PRA. A review The other issue that was noted in this F&O is the I

of the limitations of software being used for this use of 0 as a value in the quantification instead analysis by the review team identified several of setting the event to FALSE. This latter issue concerns that the limitations of the software being has no known quantification concerns. The I E1-117

Enclosure 1 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

I Number used may not be fully understood by the SNOC Fire updated quantification report includes identified PRA Team . limitations of FRANC.

Identify all of the software packages, including "in house" software packages (excel spreadsheets, MSAccess queries, etc.) that have been used to directly support the development and/or quantification of the Fire PRA. Review each of the software packages and ensure that they have been demonstrated to generate appropriate results when compared to those from accepter algorithms, and that the limitations of the software packages are know and documented.

FO-81-04 FO-81 NOT MET This SR has back looks to the Internal Events OU This F&O has been resolved.

SRs. OU-89 is associated with the requirement to set logic flags to TRUE or FALSE versus 1.0/0.0 A sensitivity study was performed to address the prior to the generation of cutsets. use of sequence tracer events (FLAGS) set to 1.0 in the quantification versus setting the Since the Fire PRA does not set the logic flags to events to TRUE. The sensitivity study used the ,

FALSE, but uses a value of 0.0 instead, this SR is CAFTA cutset editor to set events with a value not met. of 1.0 to 'TRUE', to compress those events, and to minimize the updated results. The net impact I Develop a defined basis to support the no of this was a minor reduction of the total CDF applicability of the OU-89 requirement, or set the result. I flags to FALSE in the PRA model and compress the fault tree prior to using it via linking it with FRANC. With respect to the second issue related to the I use of 0 versus 'FALSE', there is no known issue i with this a~~roach.

E1-118 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number FQ-C1-02 FQ-C1 MET A review of HRA dependencies in top scenario found This F&O has been resolved.

CC 11111111 that the first HRA combination includes events OA ALT AFW ----H and OAB_TR-------H associated with The process of identifying jOint human action aligning an alternate water source for AFW and Feed dependencies was repeated following the and Bleed Injection. Based on discussions with the incorporation of changes associated with the Internal Event HRA lead, the combination is in a resolution of the F&Os. The results of this sequence that is well beyond the PSA mission time process did not identify any other instances (35 hrs vs. 30 hr time to CSD) and was identified as where dependence between action beyond the a sequence that should be removed from the model mission time occurred. This single instance in a Suggestion Level F&O in the FPIE Peer Review. noted in the F&O has been resolved.

Setting OA-ALT AFW ----H to false reduces the CDF by 5% on the top fire scenario. Therefore, it could have an impact on overall CDF. It also adds questions on the depth of HRA dependence review since this is the top HRA combination in the top CDF sequence.

Remove the accident sequences longer than 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> from the Fire PRA Quantification Model as suggested in the FPIE Peer Review.

FQ-F1-01 FQ-F1 NOT MET ThLs SR is associated with doculllenting ~ This F&O has been resolved.

E1-119 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number quantification process and its attributes. Although the process followed appears to be fairly robust, The documentation changes and updates documentation of the process is not as robust, and associated with the resolution of the other F&Os all attributes of the SR are not fully documented in a for this review element addresses and resolves manner to ensure that they are satisfied. this F&O.

Add information into the documentation to provide the required level of detail to ensure that all of the attributes of the FO SRs are met. For example, add in a discussion that explains that every Unit 2 PAU is assumed to result in a Unit 1 reactor trip, and provide a method of identifying how to determine which Unit 2 PAUs actually impact Unit 1risk (e.g. CCDP for normal Unit 1 Rx trip is xxx - so any CCDP above this value impacts Unit 1 risk). Provide a similar discussion for Unit 2. Then provide a table in the report that identifies those Unit 1 PAUs that impact Unit 2, the Unit 2 PAUs that impact Unit 1, and a list I of the PAUs that impact both Units. I I

The following SRs appear to be met technically after discussions with SNOC personnel, and walking through an actual quantification, but it is not readily apparent they are met without a lot of extra leg work and discussions .

FO-A2 - there is no discussion of Unit 2 PAUs that impact Unit 1 (or vice-versa), and there is no E1-120 to NL-12-0932 Basis for Proposed Change Table 8

• Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number

, discussion on dual unit initiators and how they are handled.

FQ-B1 - not all programs actually used in the process were identified in the report, and there is no discussion that the limitations of the software used in the process were reviewed and understood.

Additionally, the approved software list did not always have the same version of the software package as being approved for use, and some of the software packages were not include on the list at all.

FQ-B1 - Some Microsoft Access queries were developed to support the fire PRA, but there is no identification of this in the discussion of software used, and there is no documented "software quality" associated with these queries.

I FQ-B1 - The discussion on convergence does only I provides the results, but does not provide the I

process followed to actually obtain the results. Since it is obvious that the results being discussed are not a direct output of the FRANC quantification, a discussion of how the actual CDF/LERF values are obtained to show convergence needs to be provided.

If the intent of these added failures was to function as 'Tag/flag" events for use in the fire PRA only, then E1-121 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number they should use the Vogtle Tag/Flag naming convention instead of a basic event naming scheme that links to pre-defined type codes. If the intent of these added failures is to function as basic events, they need to be revisited, and data assigned to them in accordance with the requirements of this SA. In either case, the documentation needs to be updated to identify all the added failures, and how they are intended to be used in the model.

FQ-F1-02 FQ-F1 NOT MET This SR is associated with the documentation of the This F&O has been resolved.

quantification process.

This F&O addresses weaknesses in the Scenario specific quantification factors were applied documentation of the quantification process.

to the selected fire scenarios. The only record that The documentation has been revised to provide was provided that identified where these factors additional details.

were applied was the "VEGP-1-Scenario.mdb" MSAccess database. No documentation was provided that described the process followed to determine where the various quantification factors could be applied or how the applicable quantification factors were calculated for each scenario.

The establishment of the parameters used in the FRANC quantification code is heavily dependent upon the "VEGP-1-Scenario.mdb" MSAccess database. This database contains a lot of information, but there is no direct discussion of how the database works (uses inclusion versus exclusion E1-122 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution I (F&O) Element (CC)

Number logic), calculates the ignition frequencies for scenarios, identifies/applies quantification factors for scenarios, etc.

There is no discussion of how the "excluded events" were determ ined/validated for each scenario, and how this information was added into the database to verify that all of the events in the "excluded events list" should have been excluded from the associated scenario evaluation.

There is no roadmap showing how the queries are linked and run in the "VEGP-1-Scenario.mdb" MSAccess database, nor is there a roadmap explaining how the outputs of this software is imported into the FRANC database.

The CTT Appender software is used to append information to the cutsets (e.g. ignition frequencies, non-suppression probabilities, severity factors), but this software and the process followed to convert the cutsets from CCDP/CLERP to CDF/LERF is not discussed in the documentation.

Provide the documentation necessary to "verify/replicate" the information and processes followed during the quantification process. For example, document the process followed for _ _ _ -- --- - - -

E1-123 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review I Category Finding Description Resolution (F&O) Element (CC)

Number determining where the various quantification factors could be applied, and document the calculation of the applicable quantification factors for each ,

scenario.

SF-A2-01 SF-A2 MET As part of their Seismic Fire Interactions, Vogtle This F&O has been resolved.

CC 1111/111 looked at the Gaseous Systems and the dry, pre-Action Sprinkler Systems in the Auxiliary Building. In The analysis was updated to include a review of looking at the evaluation, it was determined that the same suppression system types in the I there were also dry, pre-action sprinkler systems as turbine building. The results of the review did ,

well as several wet-pipe sprinkler systems in the not identify any suppression system seismiclfire I turbine building that were not evaluated. interactions of concern. ,

I I

Vogtle needs to include an evaluation of the turbine ,

building in Section 3 of Report 0293100011.13 to I I

ensure that the potential impacts of seismic failures of these systems are considered in the seismic fire interaction evaluation.

This F&O was discussed with Vogtle staff and they plan to add an additional table to Section 3 of Report 0293100011.13 to address the turbine building sprinkler systems. A draft of the table was provided and it appears to be sufficient to fully resolve this F&O.

UNC-A1-01 UNC-A1 MET Per requirements of HLR-QU-E in section 2 of the This F&O has been resolved.

CC 11111111 Standard, OAB_TR-------H has a probability of 7.3E 2 and an EF of 5 in the HRA Calculator worksheet. A review of the entire scope of HRA results was However, in the cutset file, the event is split into two performed and all instances where an HEP of E1-124 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number events to address HRA dependencies. The first 0.10 or higher occurs was identified. For these event is OAB_TR-------H with a probability of 0.119 cases, the HRAC assigned EF was replaced and an EF of 1. The second event is recovery with a value of 3. This EF is consistent with 10VDM-OABTR2 with a probability of 0.61 and an current industry consensus practices. The EF of 1. Therefore, the EF does not match with the range of HEPs above 0.10 was also reviewed EF assessed in the HRA Calculator sheet. Based on and it was confirmed that the highest HEP was discussion with the plant team , HRA Calculator 0.25.

assigns a value of 1 for HEPs above 0.1 . (Note that this is not consistent with NUREG-1278 and should be corrected.) By breaking the HRA event to perform the HRA dependency analysis causes the EF for this HRA to be lost.

The EF for this HRA event is less than the HRA Calculator assigned EF. Other HRAs could have similar issues based on the high HEPs used for the Fire PRA.

Review the HRA events in the plant database and assign an EF consistent with the HRA Calculator (and NUREG-1278).

UNC-A2-01 UNC-A2 MET Quantification Report No. 0293100011.14, Rev. 0, This F&O has been resolved.

CC 1/11/111 Section 8.1 identifies diHerences between mean I values of CDF and LERF when using the new EPRI As noted in the resolution discussion of FSS-E3 I ignition frequencies and NUREG/CR-6850 ignition 01 , the FPRA is based on the updated EPRI fire I frequencies . However, a Monte Carlo simulation frequency values as published in NURG/CR-should be performed and uncertainty intervals 6850, Supplement 1. Uncertainty analysis for I E1-125 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number identified for CDF and LERF (along with the mean the actual fire frequency data used in the values) using the new EPRI ignition frequencies and analysis has been performed. There is no NUREG/CR-6850 ignition frequencies. This requirement to perform Monte Carlo uncertainty requirement is derived from IGN-A 10 and IGN-B5 analysis on prior data no longer used in the requirements. FPRAmodel.

Technical adequacy and completeness.

Perform Monte Carlo simulation using old and new ignition frequency data.

UNC-A2-02 UNC-A2 MET Supporting requirement FQ-F1 and FSS-E4 are not This F&O has been resolved.

CC 1/11/111 met. This results in supporting requirement UNC-A2, which references them, to be met but with this F&O. The resolution of the F&Os linked to FQ-F1 and Documentation related to the propagation of FSS-E4 effectively also closes this F&O. It is uncertainty from these supporting requirements to noted that the F&O associated with FSS-E4 is CDF and LERF uncertainty has not been fully FSS-A3-01. With respect to FSS-A3-01, the achieved. treatment remains conservative and the CDF and LERF results retain an upper bound bias Documentation to support future Fire PRA because of this bias. As a result, a treatment of application. uncertainty would reflect a risk reduction and is therefore not explicitly treated.

Upgrade the documentation related to uncertainty.

MU-C1-01 MU-C1 NOT MET The PRA configuration control process does not This F&O has been resolved .

include evaluation of the cumulative impact of pending changes on risk applications. Although this F&O is linked to FQ, the issue is more closely related to the SNC procedures for i E1-126 to NL-12-0932 Basis for Proposed Change Table 8 - Resolution of the VEGP Fire PRA Peer Review Findings Finding &

Capability Observation Review Category Finding Description Resolution (F&O) Element (CC)

Number ......

model maintenance and update. The associated procedure is in the process of being updated. Since one of the SNC plants is transitioning to NFPA 805, that plant will have the lead in generating the update to the fleet procedure for this process. This F&O has an impact on the future update to the FPRA, but at this point, this F&O has no technical impact on the FPRA.

E1-127 to NL-12-0932 Basis for Proposed Change 3.3.2.4 General Conclusions Regarding FPRA Capability The information provided in the Section 3.3.2 of this LAR demonstrates that the VEGP at-power FPRA models conforms to the standard at CC-II, which satisfies the guidance of RG 1.200, Revision 2.

The VEGP FPRA model technical capability evaluations described above provide a robust basis for concluding that the FPRA models are suitable for use in supporting the implementation of 10 CFR 50.69.

3.4 Risk Evaluations (10 CFR 50.69{b){2){iv>>

3.4.1 Rule Requirements Per 10 CFR 50.69(b)(2)(iv) an LAR must include the following: "A description of, and basis for acceptability of, the evaluations to be conducted to satisfy

§ 50.69(c)(1 )(iv). The evaluations must include the effects of common cause interaction susceptibility, and the potential impacts from known degradation mechanisms for both active and passive functions, and address internally and externally initiated events and plant operating modes (e.g., full power and shutdown conditions)."

Per 10 CFR 50.69(c)(1 )(iv) the following is required: "Include evaluations that provide reasonable confidence that for SSCs categorized as RISC-3, sufficient safety margins are maintained and that any potential increases in core damage frequency (CDF) and large early release frequency (LERF) resulting from changes in treatment permitted by implementation of

§ 50.69(b)(1) and (d)(2) are small."

Per 10 CFR 50.69(b)(1) licensees are allowed to use alternate ways for complying with the requirements outlined in 10 CFR 50.69(b)(1) for RISC-3 and RISC-4 SSCs.

For RISC-3 SSCs, 10 CFR 50.69(d)(2) states that, "The licensee or applicant shall ensure, with reasonable confidence, that RISG-3 SSCs remain capable of performing their safety-related functions under design basis conditions, including seismic conditions and environmental conditions and effects throughout their service life. The treatment of RISC-3 SSCs must be consistent with the categorization process.

Inspection and testing, and corrective action shall be provided for RISC-3 SSCs.

(i) Inspection and testing. Periodic inspection and testing activities must be conducted to determine that RISC-3 SSGs will remain capable of performing their safety-related functions under design basis conditions; and (ii) Corrective action. Conditions that would prevent a RISC-3 SSG from performing its safety-related functions under design basis E1-128 to I\lL-12-0932 Basis for Proposed Change conditions must be corrected in a timely manner. For significant conditions adverse to quality, measures must be taken to provide reasonable confidence that the cause of the condition is determined and corrective action taken to preclude repetition."

The SNC categorization process for 10 CFR 50.69, its proposed implementation, and this submittal, meet the above requirements, as described in the following sections.

3.4.2 Explanation of Process The process that SNC will use at VEGP for categorization of SSCs into the four risk-informed safety categories (RISC-1, RISC-2, RISC-3, and RISC-4) defined in 10 CFR 50.69 is specified in SNC procedures. The guidance in SNC procedures is consistent with and is intended to implement the categorization process guidance provided in NEI 00-04 as endorsed by RG 1.201, with clarification. The process is risk-informed, in that it combines risk insights with traditional engineering prinCiples of maintaining defense-in-depth and adequate safety margins, and employing an lOP, as described in section 3.1 of this LAR, to ensure that the system functions and operating experience have been appropriately considered in the categorization process. The following discussion explains how the SNC process meets 10 CFR 50.69(b)(2)(iv) requirements.

3.4.2.1 Evaluations for Safety Margin, Change in CDF/LERF, and the Potential Impacts from Known Degradation Mechanisms In accordance with I\lEI-00-04 section 8, SNC's 10 CFR 50.69 process includes performance of a risk sensitivity study to confirm that the categorization process results in acceptably small increases to CDF and LERF. The set of SSCs whose unreliability is adjusted in the final sensitivity study to assess the acceptability of the categorization results is determined through the overall process defined in the SNC 10 CFR 50.69 procedures and subordinate instructions, which are consistent with NEI 00-04 sections 2 through 7. The integrated risk sensitivity study conservatively increases the failure rate by a factor of 3 for all candidate RISC-3 PRA modeled SSCs simultaneously to ensure that potential increases in CDF and LERF due to changes in treatment are small.

The SNC 10 CFR 50.69 process defense in depth evaluation is as defined in NEI00-04. For SSCs that are identified as candidate RISC-3, and also for any redundant identical RISC-3 SSCs within the system, the evaluation assures that key safety functions are still maintained assuming that the candidate RISC-3 SSCs do not perform their function.

As part of the 10 CFR 50.69 implementation, a performance monitoring process will be defined and implemented at VEGP to ensure that potential increases in failure rates of categorized components will be detected and addressed before reaching the rate assumed in the sensitivity study. Performance monitoring of categorized SSCs, and PRA updates that SNC will implement during program E1-129 to NL-12-0932 Basis for Proposed Change implementation, in accordance with the rule requirements, will continue to capture failure data for RISC-3 SSCs, and will allow for the timely identification of any important new degradation mechanisms that may have a bearing on the categorization.

Paragraph 10 CFR 50.69(d)(2)(ii}, Corrective Action, states that conditions that would prevent a RISC-3 SSC from performing its safety-related functions under design basis conditions must be corrected in a timely manner. For significant conditions adverse to quality, measures must be taken to provide reasonable confidence that the cause of the condition is determined and corrective action taken to preclude repetition. The primary intent of this provision is to address the possible effects of potential common cause failures and degradation mechanisms following implementation, as discussed in paragraph 10 CFR 50.69(b)(2)(iv}. The VEGP corrective action process will be followed regarding potential conditions adverse to quality (including common cause failures). Per this process, the cause of the condition must be determined and corrective action taken in a timely manner to preclude repetition.

3.4.2.2 Evaluation for Common Cause Interaction Susceptibility Common cause interactions are addressed in the SNC 10 CFR 50.69 process as follows.

The SNC 10 CFR 50.69 categorization procedure specifies the process as defined in NEI 00-04. This process requires consideration of common cause risk importance measures, both risk achievement worth (RAW) and Fussell-Vesely (F-V). This requires that groups of components with potentially high common cause impacts based on quantitative PRA models are maintained in the RISC-1 or RISC-2 categories (i.e., high safety significant).

SNC's 10 CFR 50.69 categorization process addresses both known degradation mechanisms and common cause interactions for both active and passive functions, and meets the requirements of 10 CFR 50.69(b)(2)(iv}. The failure rates for equipment and initiating event frequencies used in the VEGP PRA include the quantifiable impacts from known degradation mechanisms, as well as other mechanisms (e.g., design errors, manufacturing defiCiencies, human errors, etc.).

Common cause treatment in the VEGP base (internal events at power) PRA used to support the 10 CFR 50.69 process meets the requirements in the ASMEIANS PRA Standard. For example, the criteria for treatment of common cause are delineated in the supporting requirements in the ASMEIANS PRA Standard associated with high level requirement HLR-SY-B. The VEGP internal events at power PRA includes a robust evaluation of common cause, equivalent to that specified for Capability Category II of the ASME PRA Standard for the relevant supporting requirements. This has been confirmed via industry peer review of the VEGP PRA, as described in section 3.1 of this Enclosure 1.

E1-130 to I\lL-12-0932 Basis for Proposed Change The specific NRC regulatory positions on NEI 00-04 as identified in section C of RG1.201 are addressed relative to the SNC 50.69 process proposed for implementation at VEGP via this LAR in Appendix A. As indicated in Appendix A, no exceptions to those regulatory positions are taken.

3.4.3 Conclusion The SNC 10 CFR 50.69 process implements the guidance in each section of NEI 00-04. It includes: the SSC categorization process defined in sections 2 through 7 and 10; the overall risk sensitivity defined in section 8 that is used to confirm that the categorization process results in acceptably small increases to COF and LERF; the lOP function, defined in section 9, of reviewing and ensuring that the system functions and operating experience have been appropriately considered in the process; and the processes, defined in sections 11 and 12, to provide reasonable confidence that the validity of the categorization process (including the risk sensitivity study) is maintained. The process includes consideration of the effects of common cause interaction susceptibility, and the potential impacts from known degradation mechanisms for both active and passive functions, considering internally and externally initiated events at full power and shutdown conditions. For passive components, the SI\lC 10 CFR 50.69 process utilizes the guidance in EPRI TR-112657, Revision B-A, as detailed in section 3.1.2.

Given the above, the SNC 10 CFR 50.69 process to be implemented at VEGP addresses all aspects of the guidance in NEI 00-04. The process provides reasonable confidence that, for SSCs categorized as RISC-3, sufficient safety margins are maintained and that any potential increases in COF and LERF resulting from changes in treatment permitted by implementation of 10 CFR 50.69(b)(1) and 10 CFR 50.69(d)(2) are small, and that that the requirements of 10 CFR 50.69(c)(1 )(iv) are met.

4.0 REGULATORY EVALUATION

4.1 Significant Hazards Consideration The proposed VEGP Units 1 and 2 OL LCs will allow for the voluntary implementation of 10 CFR 50.69. Implementation of 10 CFR 50.69 allows for application of a risk-informed categorization process per 10 CFR 50.69(c) to modify the scope of SSCs subject to special treatment requirements. Alternative treatments per 10 CFR 50.69(b)(1) and 10 CFR 50.69(d)(2) can then be applied consistent with the categorization of the SSCs. Implementation of 10 CFR 50.69 will allow the licensee and the NRC to better focus attention and resources on SSCs that have safety significance, resulting in improved plant safety.

SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

E1-131 to NL-12-0932 Basis for Proposed Change

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed VEGP Units 1 and 2 OL LCs will allow for the voluntary implementation of 10 CFR 50.69. The SNC risk-informed categorization process has been documented per the requirements of 10 CFR 50.69(b)(2) and meets the requirements of 10 CFR 50.69(c). The SNC risk-informed categorization process will be used to modify the scope of SSCs subject to special treatment requirements. Alternative treatments permitted per 10 CFR 50.69(b)(1) and 10 CFR 50.69(d)(2) can then be applied consistent with the categorization of the SSCs. The process provides reasonable confidence that, for SSCs categorized as RISC-3, sufficient safety margins are maintained and that any potential increases in CDF and LERF resulting from changes in treatment are small per 10 CFR 50.69(c)(1 )(iv). The proposed OL LCs do not result in or require any physical or operational changes to VEGP SSCs, including SSCs intended for the prevention or mitigation of accidents. Implementation of 10 CFR 50.69 in compliance with 10 CFR 50.69 requirements ensures that RISC-1 and RISC-3 SSCs remain capable of performing their design basis functions, including safety-related functions, under design basis conditions. In addition, the process ensures that RISC-2 SSCs are capable of performing their safety significant functions.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed VEGP Units 1 and 2 OL LCs will allow for the voluntary implementation of 10 CFR 50.69. The SNC risk-informed categorization process has been documented per the requirements of 10 CFR 50.69(b)(2) and meets the requirements of 10 CFR 50.69(c). The SNC risk-informed categorization process will be used to modify the scope of SSCs subject to special treatment requirements. Alternative treatments permitted per 10 CFR 50.69(b)(1) and 10 CFR 50.69(d)(2) can then be applied consistent with the categorization of the SSCs. The process provides reasonable confidence that, for SSCs categorized as RISC-3, sufficient safety margins are maintained and that any potential increases in CDF and LERF resulting from changes in treatment are small per 10 CFR 50.69(c)(1 )(iv). The proposed OL LCs do not result in or require any phYSical or operational changes to VEGP SSCs, including SSCs intended for the prevention or mitigation of accidents. Implementation of 10 CFR 50.69 in compliance with 10 CFR 50.69 requirements ensures that RISC-1 and RISC-3 SSCs remain capable of performing their design basis functions, including safety-related functions, under design basis conditions. In addition, the process ensures that RISC-2 SSCs are capable of performing their safety significant functions.

E1-132

1 to NL-12-0932 for Proposed Change Therefore, the proposed does not create of a new or different kind of accident from previously evaluated.

proposed manl'1.'nQ,nT involve a significant in a margin of Response: No.

proposed VEGP Units 1 and 2 OL LCs will allow for voluntary implementation of 10 50.69. The SNC risk-informed categorization process has been documented the requirements of 10 50.69{b){2) and meets the requirements of 10 CFR 50.69(c). The risk-informed categorization process will to modify the subject to soeclal treatment Alternative per 10 50.69(b){1) 50.69(d)(2) can consistent with categorization of The process reasonable confidence that, for as RISC-3, sufficient safety margins are maintained and that potential increases in and resulting from changes in treatment are small per 10 CFR 50.69(c)(1)(iv). The only requirements that are relaxed for SSCs, consistent with their categorization, are those related to The safety margins with SSCs basis functions technical requirements unchanged.

Additionally, it is required there be reasonable that any potential increases in be small from changes in reliability resulting from changes by 10 CFR 50.69.

a result individual to be of their design functions. It is that sufficient safety are preserved.

Therefore, the proposed does not involve a significant reduction in a margin of safety.

Ca~:ieu on the above, SNC that the proposed involve a significant hazards consideration under the 10 50.92(c), and, accordingly, a finding of "no significant "Ar'',rilor<::,1"'A,n is justified.

Applicable Regulatory Requirements/Criteria proposed VEGP Units 1 and 2 OL LCs will allow for the voluntary implementation of 10 CFR The SNC risk-informed categorization process documented in this per the requirements of 10 CFR 50.69{b)(2) meets the requirements of 10 CFR 50.69(c). The SNC risk-informed process will be to modify the of subject to treatment requirements. Alternative treatments per 50.69{b)(1) and 10 50.69{d}(2) can then applied consistent with categorization of the Implementation of 10 will be in with 10 to ensure and RISC-3 capable of their design including safety

-133 to NL-1;<;;..-v,CJuc;..

Basis for Proposed Change related functions, under design conditions. In addition, the process ensures that RISC-2 SSCs are capable of performing their significant functions.

The categorization described in the LAR conforms to the guidance in NRC RG 1.201, "Guidelines for Categorizing Structures, and Components in Nuclear Power According to their Safety Significance,"

Revision 1 dated May 2006 (Reference 3). The categorization also conforms to the guidance in NEI 00-04, "10 CFR 50.69 Categorization Guideline," Revision 0 dated July 2005 (Reference 4), as endorsed by 1.201.

Conformance of the SNC categorization process to RG 1 is documented in Appendix A of this enclosure. indicated in Appendix A, no exceptions are taken to RG 1.201.

4.3 Precedent NRC, by dated June 1 2011 (Reference 2), in toSNC December 6, 2010, granted pilot status 50.69 LAR which is the initial requested amendment to allow for the voluntary implementation of 10 50.69. are no directly applicable precedents.

4.4 Conclusions In conclusion, on the considerations above, (1) reasonable assurance the health and safety of public will not endangered by operation in the proposed manner, (2) such activities will conducted in compliance with the Commission's regulations, and (3) of amendment will not be inimical to the common defense security or the health and of the public.

5.0 ENVIRONMENTAL CONSIDERATION

proposed Units 1 and 2 OL LCs will allow for the voluntary implementation 10 CFR 50.69. The SNC risk-informed categorization process has documented in this LAR requirements of 10 CFR 50.69(b)(2) and meets the requirements of 10 50.69(c). The risk-informed categorization process will used to modify the scope subject to speCial treatment requirements. Alternative treatments permitted per 10 50.69(b)(1) and 10 CFR 50.69(d)(2) can then applied with the categorization of SSCs. Implementation of 10 CFR 50.69 will in compliance with 10 50.69 requirements to ensure that RISC-1 and RISC-3 SSCs capable of performing their design basis functions, including safety related functions, under basis conditions. In addition, the process ensures that RISC-2 are capable of performing their safety significant functions.

proposed amendments do not involve (i) a significant hazards consideration, (ii) a significant in the types or a significant in the amounts of any effluents that may be released offsite, or (iii) a significant increase in E1-134 to NL-12-0932 Basis for Proposed Change individual or cumulative occupational radiation exposure. Accordingly, the proposed amendments meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors", November 22, 2004.

2. NRC letter to Mark J. Ajluni, Director, Nuclear Licensing, SNC, June 17, 2011

[ML11171A084].

3. NRC Regulatory Guide 1.201 (for Trial Use), "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," Revision 1, May 2006.

4. NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0, Nuclear Energy Institute, July 2005.

5. ASME, "Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities", Code Case N-660, Revision 0,Section XI, Division 1, American Society of Mechanical Engineers, 2004.

6. EPRI TR-112657, Revision B-A, Revised EPRI Risk-Informed In-service Inspection Evaluation Procedure, Electric Power Research Institute, 1999.

7. WCAP-16308-NP-A, Revision 0, "Pressurized Water Reactor Owners Group 10 CFR 50.69 Pilot Program - Categorization Process - Wolf Creek Generating Station," August 2009 (PA-SEE-0027)" [ML092430185 and IIJ1L092430186].

8. "Final Safety Evaluation by the Office of Nuclear Reactor Regulation, Topical Report WCAP-16308-NP, Revision 0, "Pressurized Water Reactor Owners Group 10 CFR 50.69 Pilot Program - Categorization Process - Wolf Creek Generating Station", Nuclear Energy Institute Project No. 689," March 26, 2009 [ML090260674].

9. Generic Letter 88-20, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), Supplement 4," USNRC, June 1991

10. NUREG-1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities",

US Nuclear Regulatory Commission, June 1991.

11 . ASMEIANS RA-Sa-2009, "Standard for Level1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant E1-135 to I\JL-12-0932 Basis for Proposed Change Applications", Addendum A to RA-S-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.

12. NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management, December," 1991.

13. NUREG-1855, "Guidelines on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making", US Nuclear Regulatory Commission, March 2009.

14. VEGP PRA-BC-V-07-003, VEGP Level 1 and Level 2 PRA Model Revision 4

- at power, internal events, December 15, 2010).

15. EPRI TR-1 016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments," Electric Power Research Institute, Final Report, December 2008.

16. NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities", US Nuclear Regulatory Commission, September 2005

17. EPRI, NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," Electric Power Research Institute, Revision 1, August 1991.

18. Regulatory Guide 1.60, "Design Response Spectra for Seismic Design of Nuclear Power Plants," US Nuclear Regulatory Commission, 1973.

19. Letter from Dr. John Reed, Jack R. Benjamin & Associates, Inc., to Mr. Keith D. Wooten, Southern Company Services, Inc., Vogtle Units 1 and 2 Seismic Evaluation Peer Review, JBA Project No. 191-020, Plant Walkdown Status Report, January 10, 1994.

20. NUREG/CR-0098, "Development of Criteria for Seismic Review of Selected Nuclear Power Plants", US Nuclear Regulatory Commission, May 1978.

21. NUREG-1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities,"

US Nuclear Regulatory Commission, June 1991

22. NUREG 75-087, "Standard Review Plan for the Review of Safety Analysis Report for Nuclear Power Plants," LWR edition, US Nuclear Regulatory Commission, December 1975.

23. INPO 06-08, Guidelines for the Conduct of Outages an Nuclear Power Plants", Institute of Nuclear Power Operators, December 2006.

24. SOER 09-01, Shutdown Safety, August 31,2009.

25. NRC Inspection Manual Chapter 0609, Appendix G, "Shutdown Operations Significance Determination Process," 5/25/04.

E1-136 to NL-12-0932 Basis for Proposed Change

26. ASME RA-Sc-2007, "ASME RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications", Addenda to ASME RA-S 2002, ASME, New York, NY, August 31,2007.

27. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities", Revision 1, US Nuclear Regulatory Commission, January 2007.

28. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities", Revision 2, US Nuclear Regulatory Commission, March 2009.

29. NEI 00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," Nuclear Energy Institute, 2000.

30. ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, April 2002 and Addenda to Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," ASME RA-Sa-2003, American Society of Mechanical Engineers, 2003.

31. NEI 07-012, "Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, Revision 1, Nuclear Energy Institute, 2010.

E1-137 to NL-12-0932 Basis for Proposed Change APPENDIX A Regulatory Guide (RG) 1.201 Revision 1 Trial Version Conformance Matrix RG Section RG Position Vogtle Position Comments Revision 0 of NEI 00-04 references numerous other documents, but the NRC's endorsement of Revision 0 of NEI 00-04 does not constitute an endorsement of C.1 those other referenced documents Conforms See C.3 Revision 0 of NEI 00-04 includes examples to supplement the guidance. However, the NRC's endorsement of Revision 0 of NEI 00-04 does not constitute a determination that the examples are applicable for all licensees. A licensee or applicant SNC understands that the NRC's endorsement must ensure that a given example is applicable to its of Revision 0 of NEI 00-04 does not constitute a particular circumstances before implementing the determination that the examples are applicable C.2 guidance as described in that example. Conforms for all licensees.

SNC has used methodology explained in NEI 00-04 (Revision 0) for categorization of active components. SNC has used a method other than code case N-660 (Revision 0) for categorization of passive components. The SNC selected method is based on the EPRI risk informed lSI methodology (EPRI TR 112657 Revision B-A). Tables 1A and 1B in this enclosure provide a comparison of the SNC To meet the requirements of §50.69 for categorization methodology and WCAP-16308-NP-A (including of SSCs, licensees may use methods other than those the NRC SER). The selected method is set forth in Revision 0 of NEI 00-04. The NRC staff will conservative and yields realistic results . This determine the acceptability of such other methods by method was developed after NEI 00-04 was C.3 evaluating them against the requirements of §50.69. Conforms approved.

When PRAs have not been performed, NEI 00-04 (Rev At the present time, SNC has a peer reviewed

0) allows the use of non-PRA-type evaluations (e.g., internal events (including internal flooding) PRA fire-induced vulnerability evaluation (FIVE), seismic model and Fire PRA model. SNC will use I margins analysis (SMA), and NEI guidance in qualitative approaches for other risks - Seismic CA -

clil-JMARC 91 -06, "Guidelines for Industr~ Actions to Conforms risk, Other External Events, and Shutdown - I E1-138 to NL-12-0932 Basis for Proposed Change APPENDIX A R~gulatory Guide (RG) 1.201 Revision 1 Trial Version Conformance Matrix RG Section RG Position VC!9lle Position Comments Assess Shutdown Management," to address shutdown until PRA models are developed. SNC has operations) . Such non-PRA-type evaluations will result dedicated resources to develop Vogtle seismic in more conservative categorization. PRA model.

Technical Adequacy Attributes of Analyses Implementing Revision 0 of NEI 00-04 The licensee or applicant is expected to document the technical adequacy of their internal events PRA for 50.69 application per RG 1.200, which endorses NEI 00-02 and ASME Standard RA-S-2002.

The RG 1.201 Rev 1.0 further states that "However, the documents mentioned above currently cover only internal events at full power. There is not currently a similarly endorsed standard for the external events, internal fires, and low-power and shutdown PRAs, or for non-PRA-type analyses (e.g. , FIVE, SMA, NUMARC 91-06) , and Section 3.3 of Revision 0 of NEI 00-04 provides only limited guidance for determining the technical adequacy attributes required for these types of analyses for this specific application .

Therefore, for §50.69 submittals that are received before the NRC endorses standards for external events, internal fires, and low-power and shutdown PRAs, as well as non-PRA-type analyses, the NRC staff expects the licensee or applicant to document the bases for why the method employed is technically adequate for this application . Toward that end, as part of the plant-specific application requesting to implement §50.69, the licensee or applicant will provide Refer to section 3.2 of Enclosure 1 to NL-12 C.5 the bases supJ)ortingJlle technical adequacy of its I Conforms 0932 .

E1-139 to NL-12-0932 Basis for Proposed Change APPENDIX A Regulatory Guide (RG) 1.201 Revision 1 Trial Version Conformance Matrix RG Section RG Position --

Vogtle Position Comments external events, internal fires, and low-power and shutdown PRAs, and non-PRA-type analyses for this application."

Uncertainty Considerations in Revision 0 of NEI 00-04 The staff notes that the purpose of the sensitivity studies performed as part of the risk categorization process is to address the impact of parameter and model uncertainties on the categorization. The staff understands the phrase "applicable sensitivity studies identified in the characterization of PRA adequacy" (in Tables 5.2 through 5.5 of Revision 0 of NEI 00-04), as meaning those uncertainties not addressed by the other sensitivity studies in Tables 5.2 through 5.5.

These uncertainties are typically identified via PRA peer reviews or self-assessments that are associated SNC agrees that the phrase "applicable with the licensee's choice of specific models and sensitivity studies identified in the assumptions, as discussed in Section 2.2.5.5 of characterization of PRA adequacy" (in Tables Regulatory Guide 1.174, "An Approach for Using 5.2 through 5.5 of Revision 0 of NEI 00-04), as Probabilistic Risk Assessment in Risk-Informed meaning those uncertainties not addressed by Decisions on Plant-Specific Changes to the licensing the other sensitivity studies in Tables 5.2 C.6 Basis." Conforms through 5.5.

"Common-Cause Failure and Degradation Mechanism The noted issues will be primarily addressed via Considerations in NEI 00-04 (Rev 0) a process consistent with that described in Section 12.4 of NEI-00-04. As noted in that The NRC staff notes that mechanisms that could lead section, various elements of the risk to large increases in core damage frequency (CDF) categorization process, which are defined in and large early release frequency (LERF), which could SNC 10 CFR 50.69 procedures, ensure that the potentially invalidate the assumptions underlying the potential for common cause failures of RISC-3 categorization process, including the risk sensitivity Conforms (via SSCs is appropriately considered. The study, are the emergence of extensive common-cause Implementation categorization process elements that C.7 failures (CCFs) impacting multiple systems and Process) accomplish this include base PRA model E1 -140 to NL-12-0932 Basis for Proposed Change

~

APPENDIX A R~ulatory Guide (RG) 1.201 Revision 1 Trial Version Conformance Matrix RG Section RG Position Vogtle Position Comments significant unmitigated degradation. However, for requirements, consideration of common cause these types of impacts to occur, the mechanisms that risk importance measures (RAW and FV),

lead to failure, in the absence or relaxation of defense-in-depth evaluation, and the integrated treatment, would have to be sufficiently rapidly risk sensitivity study. In addition to the developing or not self-revealing such that there would categorization process itself, the requirements be few opportunities for early detection and corrective of the rule for RISC-3 treatment, including test action. Section 12.4 of NEI 00-04 describes an and inspection (§50.69(d)(2)(i)), periodic acceptable performance-based approach to address evaluation (§50.69(e)) and corrective action these concerns. (§50.69(d)(2)(ii)), provide important defenses against the potential for common cause failures Alternatively, those aspects of treatment that are going undetected. In accordance with the necessary to prevent significant SSC degradation or process described in NEI-00-04 Section 12.4, failure from known mechanisms, to the extent that the performance monitoring of RISC-3 SSCs, as results of the risk sensitivity study would be invalidated, required by 10 CFR 50.69(e)(3), will be could be identified by the licensee or applicant, and established to detect and address potential such aspects of treatment would be retained. This increases in failure rates before the rate alternative approach would require an understanding of assumed in the categorization integrated the degradation and common-cause failure sensitivity study is reached. Failures of RISC-3 mechanisms and the elements of treatment that are SSCs will be identified and tracked in the VEGP sufficient to prevent them. As an example of how this corrective action program. As part of the alternative approach might be implemented, the known corrective action program, failures of RISC-3 existence of certain degradation mechanisms affecting SSCs will be reviewed periodically to determine pressure boundary SSC integrity could be used to the extent of condition (i.e. , whether this failure support retain ing the current requirements regarding is indicative of a potential common cause inspections or examinations or use of the risk-informed failure). As part of assessing data from the ASME Code Cases, as accepted by the NRC's corrective action program for impact associated I regulatory process. As another example, changing with alternate treatment, failures will be I

levels of treatment on several similar SSCs that might assessed for groups of like component types be sensitive to potential CCF would require (e.g. , motor operated valves, air operated consideration of whether the planned monitoring and valves, motor-driven pumps, etc.), regardless of i corrective action program, or other aspects of system. The intent of the Reriodic review is E1-141 to NL-12-0932 Basis for Proposed Change APPENDIX A Regulatory Guide (RG) 1.201 Revision 1 Trial Version Conformance Matrix RG Section RG Position Vogtle Position Comments treatment, would be effective to sufficiently minimize twofold:

the potential for CCFs impacting multiple systems, First, to ensure that the failure rate of RISC-such that the categorization process (including the risk 3 SSCs in a given time period has not sensitivity study) remains valid." unacceptably increased due to the changes in treatment. The periodic review will validate that the rate of RISC-3 SSC equipment failures has not increased by a factor greater than that used in the integrated risk sensitivity study to confirm acceptability of categorization results.

Second, the review of component group failure data will be performed to detect the potential occurrence of inter-system common cause failures, and to allow timely corrective action if necessary, as required by

§50.69(d)(2)(ii). Since most RISC-3 components have low failure rates, noted increases to these rates will be most readily detected through grouping of components.

If failure rate increases are noted, attention will be focused on common treatment changes to groups of components to ensure that the potential for inter-system common cause failure remains low. This corrective action review will also consider previous component performance history.

Criteria will be established for detecting adverse failure trends prior to exceeding the factor used in the categorization sensitivity study. If the number of failures for a group of SSCs exceeds E1-142 to NL-12-0932 Basis for Proposed Change APPENDIX A Regulatory Guide (RG) 1.201 Revision 1 Trial Version Conformance Matrix RG Section RG Position Vogtle Position Comments the criterion, e.g., a factor of three increase over the expected (historical) number of failures, a potential adverse trend will be identified requiring further assessment. The failure criterion will be selected to assure an assessment is initiated prior to exceeding the factor used in the risk sensitivity study.

Appropriate actions (which could include changes in treatment or categorization) will then be taken to preclude reaching unacceptable performance.

If deemed to be appropriate to address the potential for significant component degradation due to particular known degradation mechanisms, or to address possible concerns regarding ability to adequately and promptly detect such mechanisms , SNC may determine that certain aspects of existing treatment that might otherwise be relaxed under this program should be voluntarily retained. This would be done on a case-by-case basis, with consideration of issues such as those noted by NRC in C.7.

The NRC staff notes that the implementation of all SNC procedures follow all aspects of NEI 00-04 processes described in NEI 00-04 (i.e., Sections 2 Conforms (sections 2 - to achieve reasonable confidence in the through 12) is integral to providing reasonable 7,9, and 10) evaluations required by §50.69(c)(1 )(iv). As confidence in the evaluations required by noted in C.3, SNC has elected to use a method

§50.69(c)(1 )(iv). All aspects of the guidance are Conforms (via other than the Code Case N660 (Rev 0) for important and interrelated. Implementation categorization of passive components. This Process) for sections method is based on the EPRI RI-ISI C.B Sections 2 through 7 and Section 10 of NEI 00-04 11 and 12 methodology (EPRI TR 112657 Rev B-A). It is E1-143 to NL-12-0932 Basis for Proposed Change APPENDIX A Regulatory Guide (RG) 1.201 Revision 1 Trial Version Conformance Matrix RG Section . RG Position Vogtle Position Comments describe the processes used to determ ine the set of SNC's position that the selected method is SSCs, for which unreliability is adjusted in the risk superior than the Code Case N660 (Rev 0) that sensitivity study described in Section 8, which is used is mentioned in NEI 00-04. The selected to confirm that the categorization process results in method is conservative and yields realistic acceptably small increases to CDF and LERF. results. This method was developed after NEI 00-04 was approved. This selection has no Section 9 describes the integrated decision-making impact on achieving reasonable confidence.

panel (lOP) function of reviewing and ensuring that the system functions and operating experience have been appropriately considered in the process.

Finally, Sections 11 and 12 describe the processes that provide reasonable confidence that the validity of the categorization process (including the risk sensitivity study) is maintained. Thus, all aspects of NEI 00-04 must be followed to achieve reasonable confidence in the evaluations required by §50.69(c)(1 )(iv).

The NRC staff understands that term "important-to- SNC procedures require the lOP to consider the safety" refers to non safety-related SSCs that have SSC function(s) that caused it to be originally been determined to be important. These non safety- classified as important-to-safety in order for an related SSCs will be categorized as either RISC-2 or LSS categorization to be justified.

RISC-4, as determined by their safety significance, in Notwithstanding this additional consideration, accordance with the §50.69 categorization process. these nonsafety-related SSCs will be categorized as either RISC-2 if they are The NRC staff understands that the use of phrase determined to be High Safety Significant (HSS)

" ... blends risk insights, new technical information and or RISC-4 if they are determined to be Low operational feedback .... " and similar phrases (e.g., the Safety Significant (LSS), as shown in Section third guiding prinCiple in Section 1.3), as meaning that 6.14 of the NMP procedure.

the integrated decision-making process must C.9; systematically consider the quantitative and qualitative SNC's understanding of phrase" ... blends risk Section 1.2 information available regarding the various modes of Conforms insights, new technical information and E1-144 to NL-12-0932 Basis for Proposed Change APPENDIX A Regulatory Guide (RG) 1.201 Revision 1 Trial Version Conformance Matrix RG Section RG Position Vogtle Position Comments plant operation and initiating events, including: operational feedback .... " and similar phrases are in line with NRC understanding.

PRA quantitative risk results and insights (e.g., CDF, LERF, and importance measures);

deterministic, traditional engineering factors and insights (e.g., defense-in-depth, safety margins, and containment integrity); and any other pertinent information (e.g., industry and plant-specific operational and performance experience, feedback, and corrective actions program) in the categorization of SSCs."

The second guiding principle in Section 1.3 states that deterministic or qualitative information should be used if no PRA information exists related to a particular hazard or operating mode. This principle is not to be understood to mean that deterministic or qualitative information should be used only when no PRA information exists. The NRC staff believes that the integrated decision-making process must systematically consider the quantitative and qualitative information available regarding the various modes of operation and initiating events, including PRA, quantitative risk results and insights; deterministic, traditional engineering factors and insights, and any Qualitative information is reviewed for all C9; other pertinent information in the categorization of components in the categorized system, not just I Section 1.3 SSCs. Conforms those that are not modeled.

In Section 4.0 and Section 5.1, NEI 00-04 references As stated in C.3, SNC has elected to use a ASME Code Case N-660, "Risk-Informed Safety method other than code case N660 (Rev 0) for Classification for Use in Risk-Informed categorization of passive components. This Repair/Replacement Activities," as an approach for method is based on the EPRI RI-ISI C9; addressing the pressure-retaining function or passive methodology (EPRI TR 112657 Rev 8-A). It is Section 4.0 function of active components. The version of ASME Conforms SNC's position that the selected method is E1-145

Enclosure 1 to NL-12-0932 Basis for Proposed Change I APPENDIX A Regulatory Guide (RG) 1.201 Revision 1 Trial Version Conformance Matrix RG Section RG Position Vogtle Position Comments Code Case N-660 that is acceptable to the NRC staff superior to code case N660 (Rev 0) that is for use in this application is the version identified in RG mentioned in the NEI 00-04. The selected 1.147, "Inservice Inspection Code Case Acceptability, method is conservative but still provides ASME Section XI, Division 1," subject to any conditions sufficiently realistic insights with regard to or limitations specified therein. Alternatives to this Code categorization. This method was developed Case may be submitted for NRC review and approval after NEI 00-04 was approved.

as part of a specific §50.69 application.

In Section 6.2, the NEI 00-04 guidance contains criteria for confirming that an SSC is LSS (or recategorizing it as safety-significant) based on defense-in-depth considerations, which include criteria related to containment bypass, containment isolation , early hydrogen burns, and long-term containment integrity.

The containment isolation criteria listed in this section of NEI 00-04 are applicable to containment penetrations. The NRC staff understands the use of the phrase "containment penetration" as including electrical penetrations, air locks, equipment hatches, and piping penetrations (including containment isolation valves). Further, the staff notes SNC believes that this comment refers to that the containment isolation criteria in this section of alternative treatment. When developing NEI 00-04 are separate and distinct from those set alternative treatments, SNC will ensure that the forth in §50.69(b)(1 )(x). The criteria in §50.69(b)(1 )(x) criteria mentioned in §50.69(b)(1 )(x) are used in are to be used in determining which containment determining which containment penetrations penetrations and valves may be exempted from the and valves may be exempted from the Type B Type B and Type C leakage testing requirements in and Type C leakage testing requirements in both Options A and B of Appendix J to 10 CFR Part 50, both Options A and B of Appendix J to 10 CFR but the §50.69(b)(1 )(x) criteria are not used to Conforms (via Part 50, but the §50.69(b)(1 )(x) criteria are not C9; determine the proper RISC category for containment 1m plem entation used to determine the proper RISC category for Section 6.2 isolation valves or penetrations. Process) containment isolation valves orRenetrations.

C9: ~ risk sensitivity study addresses the impact of Conforms~ The VEGP PRA model includes common cause E1-146 to NL-12-0932 Basis for Proposed Change APPENDIX A Regulatory Guide (RG) 1.201 Revision 1 Trial Version Conformance Matrix RG Section RG Position Vogtle Position Comments Section 8 potential increases in the failure rates of all RISC-3 Implementation failures, i"n accordance with the criteria stated in SSCs resulting from the change in treatment. Section 8 Process) the ASMEIANS PRA Standard. As noted in the of NEI 00*04 includes commentary on the NRC comments on NEI-00-04, modeling of consideration of known degradation mechanisms and inter-system common cause failure is not an common-cause interactions and failures in PRAs, expectation in the PRA Standard, and is not which includes the observation that intersystem addressed in the VEG P PRA. The SNC common-cause failures are not typically modeled alternative treatment process, which has not yet because factors such as design diversity and different been defined, will include performance service environments ensure they are negligible monitoring, categorization feedback, and contributors to risk . The NRC staff notes that because corrective action steps. These will ensure that if intersystem common-cause failures are typically not a common cause failure mechanism (inter- or included in PRA models and, therefore, are not intra-system) were to develop, it would be addressed by the risk sensitivity study. Therefore, the detected (through the performance monitoring licensee or applicant relies upon the alternative steps to be included in the implementation treatment and feedback requirements, including process), evaluated for categorization impact (in corrective action provisions, established in §50.69 and accordance with the process defined in SNC discussed in Section 12 of NEI 00-04, to ensure that procedures), and promptly addressed through any significant intersystem common-cause failure the VEG P corrective action process. Depending mechanisms would be identified and corrected so that on the significance of the failure mechanism, the the assumptions underlying the categorization are not resulting action might include recategorization, invalidated. modification to the alternative treatment for the SSCs in question, or additional (e.g., more frequent, or re-focused) performance monitoring.

Section 9.2 of NEI 00-04 limits the lOP review of risk information to active functions and SSCs. The NRC staff believes that this limitation in review scope is attributable to the reliance of NEI 00-04 on ASME Code Case N-660 to address passive functions, which is SNC procedures are developed such that 50.69 C9; performed by an expert panel. The expert panel used lOP is responsible for reviewing results of active Section 9.2 in performing ASME Code Case N-660 may be the Conforms and passive component categorization.

E1-147

Enclosure 1 to NL-12-0932 Basis for Proposed Change APPENDIX A I Regulatory Guide (RG) 1.201 Revision 1 Trial Version Conformance Matrix

! RG Section RG Position Vogtle Position Comments same panel as the lOP used in the §50.69 categorization process; however, it is not required to be the same panel. As such, the IDP review of risk information should address both active and passive functions and SSCs.

The NEI 00-04 guidance allows licensees to implement different approaches, depending on the scope of their PRA (e.g., the approach if a seismic margins analyses is relied upon is different and more limiting than the approach if a seismic PRA is used). As part of the NRC's review and approval of a licensee's or applicant's application requesting to implement §50.69, the NRC staff intends to impose a license condition that will explicitly address the scope of the PRA and non-PRA methods used in the licensee's categorization approach . If a licensee or applicant wishes to change its categorization approach and the change is outside the bounds of the NRC's license condition (e.g., switch from a seismic margins analysis to a seismic PRA), the licensee or applicant will need to seek NRC approval, via a license amendment, of the implementation of the SNC understands that SNC will have to submit a new approach in their categorization process. The license amendment to obtain NRC approval focus of the NRC staff's review and approval will be on prior to implementing a new approach (e.g.,

C9; the technical adequacy of the methodology and switch from a seismic margins analysis to a Section 11 .2 analyses relied upon for this application. Conforms seismic PRA) in the categorization process.

The guidance in Section 12 of NEI 00-04 refers to the need to update the risk information and categorization I process if the categorization results are " ... more than minimally affected ." The NRC staff understands that C9; being "more than minimally affected" would include a SNC's understanding is in line with NRC staff's Section 12.1 situation in wh ich there is indication that an SSC that is Conforms understanding.

E1-148 to NL-12-0932 Basis for Proposed Change APPENDIX A Regulatory Guide {RG).1.201 Revision 1 Trial Version Conformance Matrix RG Section RG Position V~gtle Position Comments categorized as low safety significant would be changed to safety-significant. The NRC staff also recognizes that the licensee or applicant may change the categorization and/or treatment aspects of SSCs so that there is reasonable confidence that the cumulative risk increase from implementing §50.69 is maintained acce~small.

The guidance in Section 12.4 of NEI 00-04 defines CCF as " ... the simultaneous failure of more than one SSC to perform its function, due to the same cause ... ,"

and Appendix B to NEI 00-04 provides a similar definition, but with "simultaneous" replaced by "during a short period of time." These definitions are derived from their use in a PRA context, where the emphasis is on failure of more than one SSC during a specified mission time. The staff notes that the licensee's or applicant's corrective action program associated with the implementation of §50.69 should address the potential for SSC failures at different times resulting from a common cause, even if they are revealed at different times. In addition, the staff notes that the SNC has not developed treatment guidance at guidance in Section 12.4 that potential adverse trends the present time as the LAR subm ittal per 10 need not be evaluated until the number of expected CFR 50.69 (b)(2)(i) - (iv) does not require failures for a group of SSCs doubles may not be Conforms (via licensee to address treatment guidance in the C9; practical for SSCs with low failure rates assumed in the Implementation LAR. When SNC develops treatment Section 12.4 PRA. Process) guidelines, SNC will address this comment.

IMPLEMENTATION To evaluate licensee compliance with the requirements of §50.69 for the categorization of SSCs, NRC will use D either the methods described and/or endorsed in this N/A N/A E1-149 to NL-12-0932 Basis for Proposed Change APPENDIX A Regulatory Guide (RG) 1.201 Revision 1 Trial Version Conformance Matrix RG Section RG Position Vogtle Position Comments guide OR NRC approved alternative method(s) (that were submitted as part of the licensee's LAR).

E1-150

Vogtle Electric Generating Plant Pilot 10 CFR 50.69 License Amendment Request Enclosure 2 Operating Licenses Marked-up Pages

Amendment Additional Condition Implementation Number 102 . The licensee will implement all applicable crane, load and Before and rigging and load guidelines NUREG-0612 and during ANSI Standard B30.2, as described in the letters operations, as September 4, 1997, May 19 and June 12,1998, and evaluated in appropriate.

the Safety 29, 1 154 Upon implementation of the Amendment adopting TSTF-448, As stated in the 3, the determination of CRE unfiltered air nle,akcloe Additional required by SR 3.7.10.5, in accordance with TS Condition the measurement of CFE pressure as required by 5.5.20,d, shall be met. Following implementation:

of SR 3.7.10.5, in accordance with

,","',:ll'lT''*r1"n 5,5.20.c.(i), shall be within the frequency of 6 years, plus the i8-month allowance of SR as measured from March 2004, the date of the most recent successful tracer test, as stated in the June 16, 2004 letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent successful tracer gas test is than 6 years, (b) The first performance of the periodic assessment of CRE habitability, specification 5.5.20.c.(ii), shall be within 3 years, plus the 9-month allowance of SR as measured from March 2004, the date of the most recent successful tracer gas as stated in the June 16, 2004 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer test is greater than 3 years.

The first performance of the periodic measurement of CRE pressure, specification shall be within 18 months, plus the 138 allowance of SR as measured from March the date of the most recent successful pressure measurement test, or within 138 days if not rTf"I/*m",.n previously.

ogtle Unit 1 Amendment No.

'- Insert 1 here

Insert 1 Amendment Additional Condition Implementation Number Date Southern Nuclear Operating Company (SNC) is approved to As stated in the implement 10 CFR 50 .69 using the processes for categorization of Additional Condition Risk-Informed Safety Class (RISC)-1 , RISC-2 , RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the licensee amendment request dated (and supplements dated ) and as approved in the safety evaluation report dated (and supplements dated ).

NRC prior approval is required for a change to a categorization process that is outside the bounds specified above (eg ., change from a seismic margins approach to a seismic probabilistic risk assessment).

Amendment Additional Condition

• Implementation Number Date 78 The shall implement a procedure will prohibit Prior to into an extended Emergency Diesel Generator Allowed Outage implementation of time (14 days), for scheduled maintenance purposes, if severe Amendment No. 78.

weather are as in the application dated January 22, 1 as supplemented by letter dated March 18, 1998, and evaluated in the staffs Safety Evaluation Ma 1 80 The UFSAR will be updated to include the heat load that will To be included in ensure the limit of 1 will not be as the next appropriate well as the requirement to perform a heat load evaluation before UFSAR update

. transferring irradiated fuel to either pool, as described in the following the 4,1 May 19 and June 12, installation of the 1998, and evaluated in the staffs Safety Evaluation dated June Unit 1 spent fuel 29, 1998. racks.

135 Upon implementation of the Amendment adopting TSTF-448, As stated in the Revision 3, the determination of CRE unfiltered air inleakage as Additional Condition required by SR 3.7.10.5, in accordance with TS 5.5.20.c.(i), and the measurement of CFE pressure as required by SpeCification 5.5.20.d, shall be considered met. Following implementation:

The first performance of SR 3.7.10.5, in accordance with Specification 5.5.20.c.(i), shall be within the specified frequency of 6 years, plus the 18-month allowance of SR as measured from March 2004, the date of the most recent successful tracer gas test, as stated in the June 16, 2004 letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 The first of the periodiC assessment of CRE habitability, specification shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured from March 23, 2004, the date of the most recent successful tracer gas as stated in the June 16, 2004 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is than 3 years.

The first of the periodic measurement of CRE pressure, specification shall be within 18 months, plus the 138 allowance of SR 3.0.2, as measured March 23, 2004, the date of the most recent successful pressure measurement test, or within 138 days if not erformed reviousl.

Amendment No.

Insert 2 Amendment Additional Condition Implementation Number Date Southern Nuclear Operating Company (SNC) is approved to As stated in the implement 10 CFR 50.69 using the processes for categorization of Additional Condition Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the licensee amendment request dated (and supplements dated ) and as approved in the safety evaluation report dated (and supplements dated ).

NRC prior approval is required for a change to a categorization process that is outside the bounds specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment) .

Vogtle Electric Generating Plant Pilot 10 CFR 50.69 license Amendment Request Enclosure 3 Operating licenses Clean Typed Pages

Amendment Additional Condition Implementation Number Date 102 The licensee will implement all applicable crane, load path and Before and height, rigging and load testing guidelines of NUREG-0612 and during reracking ANSI Standard B30 .2, as described in the licensee's letters dated operations, as September 4, 1997, May 19 and June 12,1998, and evaluated in appropriate.

the staffs Safety Evaluation dated June 29, 1998 154 Upon implementation of the Amendment adopting TSTF-448, As stated in the Revision 3, the determination of CRE unfiltered air inleakage as Additional required by SR 3.7.10.5, in accordance with TS 5.5.20.c.(i), and Condition the measurement of CFE pressure as required by Specification 5.5.20.d, shall be considered met. Following implementation:

(a) The first performance of SR 3.7.10.5, in accordance with Specification 5.5.20.c.(i), shall be within the specified frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from March 23, 2004, the date of the most recent successful tracer gas test, as stated in the June 16, 2004 letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, specification 5.5.20.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured from March 23, 2004 , the date of the most recent successful tracer gas test, as stated in the June 16, 2004 letter response to Generic Letter 2003-01 , or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

The first performance of the periodic measurement of CRE pressure, specification 5.5.20.d, shall be within 18 months, plus the 138 days allowance of SR 3.0.2, as measured from March 23, 2004 , the date of the most recent successful pressure measurement test, or within 138 days if not performed previously.

Southern Nuclear Operating Company (SNC) is approved to As stated in the implement 10 CFR 50.69 using the processes for categorization of Additional Condition Risk-Informed Safety Class (RISC)-1 , RISC-2, RISC-3, and RISC-4 structures , systems, and components (SSCs) specified in the licensee amendment request dated (and supplements dated ) and as approved in the safety evaluation report dated (and supplements dated ).

NRC prior approval is required for a change to a categorization process that is outside the bounds specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment).

Vogtle Unit 1 Amendment No.

Amendment Additional Condition Implementation Number Date Southern Nuclear Operating Company (SNC) is approved to As stated in the implement 10 CFR 50.69 using the processes for categorization of Additional Condition Risk-Informed Safety Class (RISCH, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the licensee amendment request dated (and supplements dated ) and as approved in the safety evaluation report dated (and supplements dated NRC prior approval is required for a change to a categorization process that is outside the bounds specified above (e.g., change

• from a seismic margins approach to a seismic probabilistic risk I assessment),

Vogtle Unit 2 Amendment No.