NL-14-1016, Response to Request for Additional Information on Plant Vogtle License Amendment Request to Revise Tech Specs to Implement NEI 06-09, Rev. 0, Risk Informed Tech Specs Initiative 4b, Risk-Managed Tech Specs Guidelines.

Response to Request for Additional Information on Plant Vogtle License Amendment Request to Revise Tech Specs to Implement NEI 06-09, Rev. 0, Risk Informed Tech Specs Initiative 4b, Risk-Managed Tech Specs Guidelines.

ML14198A574
Person / Time
Site:	Vogtle
Issue date:	07/17/2014
From:	Pierce C Southern Co, Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NEI 06-09, NL-14-1016
Download: ML14198A574 (68)

Text

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

40 Inverness Center Parkway Post Office Box 1295 A

Birmingham, AL 35242 Tel 205.992.7872 Fax 205.992.7601 SOUTHERN COMPANY July 17, 2014 Docket Nos.: 50-424 NL-14-1016 50-425 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant, Units 1 and 2 Response to Request for Additional Information on Plant Vogtle License Amendment Request to Revise Technical Specifications to Implement NEI 06-09, Revision 0, "Risk Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines"

References:

1. Southern Nuclear Operating Company letter, NL-12-1344, dated September 13, 2012, License Amendment Request to Revise Technical Specifications to Implement NEI 06-09, Revision 0, "Risk Informed Technical Specifications Initiative 4b, Risk Managed Technical Specifications (RMTS)

Guidelines'

2. NEI 06-09-A, Risk Informed Technical Specifications Initiative 4b, Risk Managed Technical Specifications (RMTS)

Guidelines, November, 2006

3. NRC Letter dated May 16, 2013, Vogtle Electric Generating Plant, Units 1 and 2 (VEGP) Request for Additional Information, (TAC Nos. ME9555 and ME9556)

4. Southern Nuclear Operating Company letter, NL-13-1540, dated August 2, 2013, Vogtle Electric Generating Plant Response to Request for Additional Information on Plant Vogtle License Amendment Request to Revise Technical Specifications to Implement NEI 06-09, Revision 0, "Risk Informed Technical Specifications Initiative 4b, Risk Managed Technical Specifications (RMTS) Guidelines"

5. NRC Letter dated June 9, 2014, Vogtle Electric Generating Plant, Units 1 and 2, Request for Additional Information (TAC Nos. ME9555 and ME9556)

6. NRC Letter dated June 25, 2014, Vogtle Electric Generating Plant, Units 1 and 2 - Corrected - Request for Additional Information, (TAC Nos. ME9555 and ME 9556)

U.S. Nuclear Regulatory Commission NL-14-1016 Page 2 of 3 Ladies and Gentlemen, By Reference 1, Southern Nuclear Operating Company (SNC) submitted a license amendment request for the Vogtle Electric Generating Plant (VEGP)

Technical Specifications (TS) to permit the use of the Risk Managed Technical Specifications per Reference 2.

Per Reference 3, the NRC requested additional information (RAI) to facilitate their review. SNC provided a response to that request by Reference 4.

On April 15 and 16, 2014, the NRC visited the SNC offices in Birmingham to perform an audit of the SNC VEGP license amendment request as presented in References 1 and 4. From that audit, additional questions were generated, some of which required follow-up beyond the audit meeting. Accordingly, the NRC addressed those questions in a request for additional information per Reference 5, as amended by Reference 6. Enclosure 1 to this letter contains SNC's responses to the NRC questions, which are transcribed prior to each SNC response. Enclosure 2 and Enclosure 3 provide additional information for specific responses, as documented in Enclosure 1.

Furthermore, SNC proposes to revise the TS mark-up submitted as part of the Reference 1 letter. The changes to the Reference 1 letter are described in of this letter as well as in the response to the first NRC RAI provided by Reference 4. Upon staff agreement with the proposed changes, SNC will promptly provide the updated package to the NRC.

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

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

Respectfully submitted, C.f(~ -.

C. R. Pierce Regulatory Affairs Director CRP/OCV

... ~ .,... .. .. ***

Sworn to me~n s scribed before me this 17~ay of ::T vi; ,2014 CJL~

Notary Public My commission expires: 1/z./~o/~

~*

U.S. Nuclear Regulatory Commission NL-14-1016 Page 3 of 3

Enclosures:

1. Response to Nuclear Regulatory Commission Questions

2. Comparison of Regulatory Guide (RG) 1.200 Revisions 1 and 2

3. Mark-Up of Technical Specifications (TS) Limiting Condition of Operation (LCO) 3. 7.9 cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. T. E. Tynan, Vice President- Vogtle 1 & 2 Mr. B. L. lvey, Vice President- Regulatory Affairs Mr. D. R. Madison, Vice President- Fleet Operations Mr. B. J. Adams, Vice President - Engineering Mr. S. C. Waldrup, Regulatory Affairs Manager- Vogtle RType: CVC7000 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. R. E. Martin, NRR Senior Project Manager- Vogtle 1 & 2 Mr. L. M. Cain, Senior Resident Inspector- Vogtle 1 & 2

Vogtle Electric Generating Plant Response to Request for Additional Information on Plant Vogtle License Amendment Request to Revise Technical Specifications to Implement NEI 06-09, Revision 0, "Risk Informed Technical Specifications Initiative 4b, Risk Managed Technical Specifications (RMTS) Guidelines" Enclosure 1 Response to Nuclear Regulatory Commission Questions to NL-14-1016 Response to NRC Questions Technical Specification Review RAis In the LAR, SNC has requested changes that are considered a deviation from what was approved by the NRC staff in Technical Specification Task Force (TSTF) Traveler TSTF-505.

Specifically, the deviations are:

• TS Limiting Condition for Operation (LCO) 3.5.5, "Seal Injection Flow," Condition A,

• TS LCO 3.7.3, "Main Feedwater Isolation Valves (MFIVs) and Main Feedwater Regulation valves (MFRVs) and Associated Bypass Valves," Conditions A, B, C, and D, and,

• TS LCO 3.8.3, "Diesel Fuel Oil, Lube Oil, Starting Oil, and Ventilation," Condition A.

NRC Technical Specification Review Question #1 Provide a technical basis for NRC staff review for the requested change to TS LCO 3.5.5, "Seal Injection Flow," Condition A or remove the requested change from the LAR.

NRC Technical Specification Review Question #2 Provide a technical basis for NRC staff review for the requested change toTS LCO 3.7.3, "Main Feedwater Isolation Valves (MFIVs) and Main Feedwater Regulation Valves (MFRVs) and Associated Bypass Valves," Conditions A, B, C, and D or remove the requested changes from the LAR.

NRC Technical Specification Reyiew Question #3 Provide a technical basis for NRC Staff review for the requested change to TS LCO 3.8.3, "Diesel Fuel Oil, Lube Oil, Starting Air, and Ventilation," Condition A or remove the requested changes from the LAR.

SNC Response to Technical Specifications Review Questions #1, #2, and #3 LCOs 3.5.5, 3.7.3 and 3.8.3 will be removed from the SNC VEGP Risk Informed Completion Time Program. The LAR will be revised to reflect these changes.

NRC Technical Specification Review Question #4 An oversight occurred during the NRC review of TSTF-505, Revision 1 and the

"-A" was omitted from the reference to NEI 06-09, Revision 0, in the Risk Informed Completion Time Program in proposed TS Section 5.5.22. Please provided a revised TS 5.5.22, with the reference as follows: Nuclear Energy Institute, NEI 06-09, "Risk Informed Technical Specifications Initiative 4b: Risk Managed Technical Specification (RMTS)," Revision 0-A, October 2012.

SNC Response to Technical Specification Review Question #4 A revision to the LAR will be submitted to add the "-A" to the NEI 06-09 reference in the proposed addition to TS Section 5.5.22, "Risk Informed Completion Time Program."

E1-1 to NL-14-1016 Response to NRC Questions NRC Technical Specification Review Question #5 An oversight occurred during the NRC review of TSTF-505, Revision 1, and a specific scenario was not satisfactorily addressed. SNC is requested to address the following scenario.

For this scenario, the TS system is comprised of train A and train B and performs two associated Probabilistic Risk Assessment (PRA) success criteria, called PRA function 1 and PRA function 2.

In an emergent condition, with both TS system train A and train B TS inoperable and the associated PRA success criteria considered PRA functional with train A able to perform PRA function 1 and train B able to perform function 2 (i.e., neither train by itself can perform PRA functions 1 and 2 but both trains together maintain PRA functionality), the NEI 06-09 guidelines will allow a risk informed completion time to be entered in this scenario, however there is no way to repair either train A or train B without losing PRA functionality. In this scenario, NEI 06-09 allows delaying a plant shutdown up to 30 days, depending on the system's risk significance, to repair the necessary system. This scenario was overlooked during the NRC staff's review of TSTF-505, and although the NRC staff approved NEI 06-09 and TSTF-505 it was not our intention to allow delaying plant shutdown in this type of scenario.

Please provide changes to the proposed "Risk Informed Completion Time Program," in VEGP TS 5.5.22, which prevents entry into a risk informed completion time for this specific scenario.

SNC Response to Technical Specification Review Question #5 Section 5.5.22.f will be added to prevent a RICT entry, or to exit a RICT entry already made, for the condition stated in the above question. It will state the following:

• ~ RICT entry is not permitted, or a RICT entry made shall be exited, for any condition involving a TS Loss of Function if a PRA Functionality determination that reflects the plant configuration concludes that the LCO cannot be restored without placing the TS inoperable trains in an alignment which results in a loss of functional/eve/ PRA success criteria."

E1-2 to NL-14-1016 Response to NRC Questions Probabilistic Risk Assessment Review RAis NRC Probabilistic Risk Assessment Review Question #1, Internal Events PRA Peer Review In Enclosure 2 of the LAR (page E2-4), it is stated that, "In May 2009, the VEGP PRA internal events model Revision 4 (including internal flooding) was reviewed against the requirements of the 2007 version of the PRA standard ... as amended by RG [Regulatory Guide] 1.200, Revision 1 ...."

Please summarize the peer review conducted in May 2009 and clarify if it was a full peer review where the team met the guidelines outlined in NEI 00-02 (e.g., 5 or 6 members that included the full range of experience required to perform an internal events PRA), followed the process outlined in NEI 00-02 (e.g., offsite preparation, one week onsite review, and post review documentation), and reviewed the PRA against all the elements in the ASME 2009 standard. If the review was not a full peer review, please describe the review in detail and provide all earlier Findings and Observations (F&Os) from any previous reviews.

SNC Response to Probabilistic Assessment Review Question #1 The VEGP PRA peer review conducted in May 2009 was performed using the process defined in Nuclear Energy Institute (NEI) 05-04 and it was a full-scope peer review. NEI 05-04 guidance supplants NEI 00-02 guidance for conducting a peer review. Although there is no technical impact, SNC provides the following clarification.

The VEGP License Amendment Request (LAR) made a reference to the American Society of Mechanical Engineers (ASME) RA-Sc-2007, which is not technically correct because the May 2009 peer review report references RA-Sb-2005. Both RA-Sb-2005 and RA-Sc-2007 are Addenda to ASME PRA Standard RA-S-2002. Addendum c of RA-S-2002 made only relatively minor changes to Addendum b, and these changes do not have any technical impact on the capability of the PRA. The main changes of interest between these two addenda are in Section 5 (Configuration Control), particularly sections 5.5 (Pending Changes) and 5.6 (Previous PRA Applications). In the RA-Sc-2007 addenda, additional verbiage was incorporated to provide further clarifications to a user in these two sections. Hence, the changes made in these two sections were clarification type changes. The RA-Sc-2007 addenda also added non-mandatory Appendix A to provide examples of PRA Maintenance, PRA Upgrade, and the Advisability of Peer Review, and made editorial corrections to several references in Section 4.

In February 2009, the ASME and the American Nuclear Society (ANS) approved the new, combined PRA Standard (ASMEIANS RA-Sa-2009, "Addendum to ASMEIANS RA-S-2008-Standard for Levei1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," the American Society of Mechanical Engineers and the American Nuclear Society, February, 2009). The NRC endorsed this new standard in Revision 2 of RG 1.200, which was issued in March 2009. The new standard was not generally available in time to support the VEGP peer review, so the peer review was performed against the version of the ASME PRA Standard (RA-Sb-2005) that was used in the PWR Owners Group peer reviews of internal events at power PRAs up to that point. Table E2.3 in Enclosure 2 of the VEGP LAR compares the 2007 version (RA-Sc-2007) against the 2009 version (RA-Sa-2009). Table E2.3 has been revised to make it consistent with Section 3.3 of NEI 05-04 "Process for Performing Internal Events PRA Peer Reviews Using the ASMEIANS PRA Standard," Revision 3, November 2009. The revised Table E2.3 is enclosed E1-3 to NL-14-1016 Response to NRC Questions as Enclosure 2 to this letter. The peer review concludes that the VEGP model satisfies the guidance of RG 1.200, Revision 2. The most significant change in the PRA Standard between RA-S-2002 (and addenda) and RA-S-2008 (and addenda) was the addition of requirements for PRAs for other than internal events at power. That is, the requirements for internal events at power PRAs in RA-Sb-2009 are substantially the same as those in RA-Sb-2005 (and RA-Sc-2007).

Summary of May 2009 Peer Review:

The scope of the peer review conducted in May 2009 was a full scope PRA peer review of the VEGP internal events at power PRA to determine compliance with ASME PRA Standard (RA-Sb-2005, "Addenda to ASME RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," American Society of Mechanical Engineers, New York, NY, December 2005) and RG 1.200, Revision 1. This peer review was performed using the process defined in Nuclear Energy Institute (NEI) 05-04, "Process for Performing Follow-on PRA Peer Reviews Using the ASME PRA Standard." NEI 05-04, Revision 1 has been endorsed by Revision 1 and Revision 2 of RG 1.200. Note that, although the title of NEI 05-04 includes the phrase "Follow-on," this document provides a process appropriate for both full and partial peer reviews of internal events at power PRAs against the requirements in the PRA Standard and RG 1.200.

The peer review was conducted during the week of May 4 through May 8, 2009. It covered all nine technical elements from the ASME PRA Standard plus the configuration control element.

The model that was reviewed was the VEGP Level 1 and Level 2 PRA Model Revision 4 - at power, internal events."

The peer review team consisted of six reviewers having a full range of experience required to perform the peer review. Each reviewer was assigned a lead role for a review element. The lead reviewer was assisted by two reviewers acting in a support role. The documents were supplied in advance to the peer review team members. During the week of May 4 through May 8, 2009, the peer review team was onsite, performed the review, and provided preliminary results on May 8, 2009. A final report was issued on November 10, 2009. The following table summarizes the results of the peer review.

E1-4 to NL-14-1016 Response to NRC Questions Table 4-1 Summary of Capability Category Assessment by PRA Element Capability Category SR Not Met Met CC-I CC-II CC-III CC-IIII CC-11/111 NIA TOTAL Initiating Event (IE) Total 21 0 5 0 5 0 2 33 Accident Sequence Analysis (AS) Total 17 0 1 2 0 0 1 21 Success Criteria (SC) Total 10 0 1 0 0 3 0 14 Systems Analysis (SV) Total 32 0 2 0 2 3 3 42 Human Reliability (HR) Total 1 19 0 5 1 2 6 1 35 Data Analysis (DA) Total 17 0 5 2 2 4 4 34 Internal Flooding (IF) Total 39 0 2 1 3 2 3 50 Quantification (QU) Total 1 28 0 2 1 0 2 1 35 Large Early Release Frequency (LE) Total 1 17 0 15 0 0 4 5 42 Maintenance & Update/Configuration Control 10 0 0 0 0 0 0 10 (MU) Total GRAND TOTALS 3 210 0 38 7 14 24 20 316 E1-5 to NL-14-1016 Response to NRC Questions NRC Probabilistic Assessment Review Question #2. Unreviewed Analysis Methods CUAMsl Please identify and provide technical justification for any PRA methodology that has not been formally accepted by the NRC staff. The NRC staff has formally accepted methods during resolution of UAMs as well as NUREG/CR-6850 (as supplemented) or the National Fire Protection Association Standard (NFPA) 805, "Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," frequently asked question (FAQ) guidance.

If a position on a method has been established by the NRC, please confirm that the accepted version of the method is used per the NRC position and, if not, then provide an acceptable alternative.

SNC Response to Probabilistic Assessment Review Question #2 In response to VEGP 50.69 RAI 27, SNC identified the following methods used in VEGP fire PRA that may not be consistent with those endorsed by the NRC:

• Not using 0.001 as lowest value for failure of manual suppression (using values lower than 0.001, even 0)

• Not using lower failure threshold for sensitive electronics

• Electrical cabinet heat release rate or severity factors and cabinet to cabinet fires (one method rejected, one method commented, others exist)

• The VEGP fire PRA has used alignment factor values that have not been accepted by NRC staff. Although an endorsed method has been used for alignment factor for oil pump fires, the alignment factor used by SNC for oil pump fires is slightly different than the factor accepted by the NRC staff in the NRC endorsement letter dated June 21' 2012.

SNC revised the VEGP PRA model by incorporating the following changes to remove unendorsed methods:

• 0.001 was used as the lowest value for failure of manual suppression,

• Fire modeling for sensitive electronics model was revised to be consistent with the approach in FAQ 13-004, "On Clarifications regarding Treatment of Sensitive Electronics," which was accepted on December 3, 2013 by a Letter from Hossain G.

Hamzehee at NRR,

• The use of electrical cabinet heat release rate or severity factors and cabinet to cabinet fires was removed and the model was revised using the NRC approved/endorsed methods, and

• The values suggested in the NRC endorsement letter dated June 21, 2012 were used as alignment factors for oil pump fires.

E1-6 to NL-14-1016 Response to NRC Questions SNC intends to use the revised fire PRA model that reflects the above changes to support the VEGP RICT program implementation. The PRA model used to support VEGP RICT Program will be controlled by SNC procedures to reflect the as-built, as-operated plant condition. Also, the PRA model is planned to be updated in the future as necessary for incorporating new consensus methodologies/data endorsed by the NRC.

NRC Probabilistic Assessment Review Question #3, LAR Page E2-34. F&O FSS-G4-01 Please discuss the method for assigning barrier failure probabilities of different types. If it does not follow NUREG/CR-6850, please provide the basis for it. If NUREG/CR-6850, Section 11.5.4, Table 11-3 fire barrier type failure probabilities are used, is the sum of the barrier failure probabilities used in the Fire PRA? If not, please explain how the Fire PRA will be updated to account for barriers that require summing these barrier type probabilities.

SNC Response to Probabilistic Assessment Review Question #3 In the VEGP Fire PRA, an applied barrier failure probability is based on the sum of the failure probability of each barrier element.

Reference:

Appendix C of "A.W. Vogtle Electric Generating Plant Units 1 and 2 Fire Probabilistic Risk Assessment Fire Scenario Selection Report," August, 2012.

NRC Probabilistic Assessment Review Question #4. Fire Ignition Frequencies For purposes of 4b, the Fire PRA should be updated with an NRC-accepted updated fire ignition frequency database. As such please propose an implementation item to update the Fire PRA with NRC-accepted updated fire ignition frequencies and to compare the new total baseline CDF and LEAF to the RG 1.174 acceptance criteria.

Please explain the process that will be used to stay within the risk acceptance criteria in RG 1.174 when future changes or PRA updates, such as the fire ignition frequency update, may result in exceeding those risk guidelines (e.g., baseline risk acceptance criteria of 1E-4/yr (CDF) or 1E-5/yr (LEAF), etc.).

SNC Response to Probabilistic Assessment Review Question #4 SNC will update the Fire PRA with NRC-accepted updated fire ignition frequencies when available and to compare the new baseline CDF and LEAF from quantified sources to the RG 1.174 acceptance criteria.

The PRA update process will be controlled by SNC procedures that include maintaining the total CDF and LEAF mean values from all quantified sources documented in the September 13, 2012 LAR, including impact of changes to fire ignition frequency updates, within the RG 1.174 bounds of 1E-04/yr (CDF) or 1E-05/yr (LEAF).

NRC Probabilistic Assessment Review Question #5, PRA Functional: Safety Margins The LAR states in the "Significant Hazards Considerations" section: The proposed change permits the extension of Completion Times provided risk is assessed and managed within the E1-7 to NL-14-1016 Response to NRC Questions Risk Informed Completion Time Program. The proposed change implements a risk-informed configuration management program to assure that adequate margins of safety are maintained."

A TS operability determination is made before a PRA functionality determination. A structure, system, and component (SSC) which is inoperable may be found to be outside limits in the TS or applicable code performance parameter(s), according to Inspection Manual Part 9900 (now Inspection Manual Chapter 0326, Reference 16); therefore, a PRA functional SSC may have parameter(s) outside certain limits. As such the RICT program must assure that safety margins are maintained for a PRA functional condition.

Safety margins assessment includes an assessment of assumptions or inputs to a safety analysis as discussed in RG 1.177:

Safety analysis acceptance criteria in the Final Safety Analysis Report (FSAR) are met or proposed revisions provide sufficient margin to account for analysis and data uncertainties (e.g., the proposed TS CT [Completion Time] or SF [Surveillance Frequency] change does not adversely affect any assumptions or inputs to the safety analysis, or, if such inputs are affected, justification is provided to ensure sufficient safety margin will continue to exist).

For TS CT changes, an assessment should be made of the effect on the FSAR acceptance criteria assuming the plant is in the condition addressed by the proposed CT (i.e., the subject equipment is inoperable) and there are no additional failures. Such an assessment should result in the identification of all situations in which entry into the condition addressed by the proposed CT could result in failure to meet an intended safety function.

Please explain if the RICT program considers these factors related to safety margin as noted above for a PRA functionality determination.

SNC Response to Probabilistic Assessment Review Question #5 The NRC SER for NEI 06-09, Section 3.2, "Evaluation," states the following regarding impact on safety margins:

"The proposed change maintains sufficient safety margins. The design, operation, testing methods, and acceptance criteria for SSCs, specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by risk-informed changes to the CTs. Similarly, there is no impact to safety analysis acceptance criteria, as described in the plant licensing basis. Thus, safety margins are maintained by the proposed methodology, and the third key safety principle of RG 1.277 is satisfied."

RG 1.174, Section 2.1.2 Safety Margin, provides the following basis for the above NRC SER on NEI 06-09 statement:

"... With sufficient safety margins, the following are true:

• Codes and standards or their alternatives approved for use by the NRC are met.

E1-8 to NL-14-1016 Response to NRC Questions

• Safety analysis acceptance criteria in the LB (e.g., FSAR, supporting analyses) are met or proposed revisions provide sufficient margin to account for analysis and data uncertainty."

The above conclusions of the Safety Evaluation are applicable to the VEGP RICT Program as explained below:

1) The extended CT depends on the existing plant configuration at the time the TS component becomes inoperable and throughout the time that the component remains inoperable; entry into the extended CT will not change the plant configuration.

2) Entering the extended CT via the RICT program involves no physical changes to the inoperable component or to its redundant systems and subsystems.

3) Entry into the extended CT will not alter any existing operating, testing, or abnormal operating procedures.

4) Entry into the extended CT will not alter any FSAR acceptance criteria for the Operable and Functional redundant components and will, in fact, not alter the acceptance criteria for the inoperable component although that acceptance criteria may not be met due to the component's inoperable status.

5) The Configuration Risk Management (CRM) Program will ensure that prior to entering a RICT, adequate margins of safety are maintained by ensuring that systems and subsystems redundant to the inoperable component are available, and that the PRA success criteria are met. Margins to safety are also maintained by the implementation of Risk Management Actions (RMAs), although usually no quantitative credit will be taken for the RMAs for the RICT calculation.

In summary, entry into the CT will not change the plant equipment configuration, nor will it change plant procedures, and it will not involve physical changes to plant equipment redundant to the inoperable component. Consequently, entering an extended CT does nothing to affect the existing safety margin. Furthermore, any RMAs put in place will not increase the likelihood of an initiating event or prevent the mitigation of transients or accidents because an RMA, by definition, is meant to control and/or compensate any risk increase from extending the CT when certain cumulative risk thresholds are reached.

Finally, with regard to identifying all situations in which entry into the CT could result in failure to meet an intended safety function, those situations have already been identified as "loss of function" conditions in the SNC VEGP Risk Informed TS. Each of those conditions are identified in the SNC September 13, 2012 initial submittal and the SNC RAI response of August 2, 2013.

The loss of function condition, as provided for in TSTF-505, cannot be entered voluntarily.

NRC Probabilistic Assessment Review Question #6, PRA Functionality: Success Criteria From NEI 06-09, Section 11.1, "If a component is declared inoperable due to degraded performance parameters, but the affected parameter does not and will not impact the success criteria of the PRA model, then the component may be considered PRA functional for purposes of the RICT calculation ...."

E1-9 to NL-14-1016 Response to NRC Questions Since NEI 06-09 states that the methodology is consistent with RG 1.174 philosophy, success criteria used for the RICT program must be consistent with maintaining safety margin and defense in depth philosophy as described in RG 1.174.

Success criteria information could be at a detailed or a high level {e.g., parameter, component, train, system, or other). The appropriate success criteria should be considered for a PRA functionality determination when evaluating the impact on success criteria of the PRA model.

For example, a PRA functionality determination may need to consider more than train-level success criteria {e.g., potentially both system and train level success criteria, and possibly other success criteria information). Please explain how your RICT program ensures the appropriate success criteria are considered in a PRA functionality determination.

SNC Response to Probabilistic Assessment Review Question #6 ASMEIANS RA-Sb-2013 defines success criteria as follows:

"Success criteria: criteria for establishing the minimum number or combinations of systems or components required to operate, or minimum levels of performance per component during a specific period of time, to ensure that the safety functions are satisfied... "

There are two parts to this definition:

• Part 1: Criteria for establishing the minimum number or combinations of systems or components required to operate to ensure safety functions are satisfied, or

• Part 2: Minimum levels of performance per component during a specific period of time to ensure safety functions are satisfied.

The system success criteria documented in the peer reviewed PRA address the two parts, as applicable, that support the VEGP RICT Program and are summarized as part of "CRM System Guidelines" for use during PRA Functionality determinations.

CRM System Guidelines document the PRA success criteria for all the VEGP TS systems included in the scope of the VEGP RICT Program. These include the following:

• Auxiliary Feedwater {AFW) system

• Containment Cooling Units {CCU) system

• Component Cooling Water {CCW) system

• Emergency Core Cooling Systems {ECCS)

• Condensate and Feedwater system

• Containment Isolation {CI) system

• Main Steam {MS) system

• Electrical systems

• Heating, Ventilation, and Cooling {HVAC) system

• Pressurizer Pressure Relief {PPR) system For example, the Part 1 success criteria for AFW system success criteria documented in the "CRM System Guidelines" include specific information, such as the following {this information is subject to change and is only provided to illustrate the scope of detail provided in the CRM System Guidelines):

E1-10

Enclosure 1 to NL-14-1016 Response to NRC Questions Part 1 PRA Success Criteria:

• 1 of 2 main steam lines provides steam to the turbine driven AFW pump

• 1 of 3 AFW pumps provides 570 gpm flow to 2 of 4 SGs Part 2 PRA Success Criteria:

• AFW inlet temperature to the steam generator= 120°F

• CST initial water mass = 2.8E+6 lb PRA Functionality determinations are performed based on the SNC procedure~ tha~ refer to the CRM System Guidelines as a reference source. The results of the PRA Funct1onahty determinations are documented as QA records and retained for the life of the plant.

NRC Probabilistic Assessment Review Question #7, SSC Performance Levels Per 10 CFR 50.36(c)(2) Limiting Conditions for operation. (i) Limiting Conditions for Operations are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

Section 3 of the TSTF-505 Model Safety Evaluation (SE) states:

The RICT program provides the necessary administrative controls to permit extension of CTs and thereby delay reactor shutdown or remedial actions, if risk is assessed and managed within specified limits and programmatic requirements. The specified safety function or performance levels of TS required SSCs are unchanged, and the remedial actions, including the requirement to shut down the reactor, are also unchanged; only the CTs for the Required Actions are extended by the RICT Program.

Please discuss how your RICT program addresses how required SSCs performance levels are unchanged.

SNC Response to Probabilistic Assessment Review Question #7 The VEGP RICT program is consistent with the NRC-approved guidance of NEI 06-09. It does not alter the system or train Operability requirements with respect to the number of systems and trains required to be Operable. Neither does it change the stated TS performance criteria, such as flow rates, response times, stroke times, setpoints, etc. The RICT program only serves to provide a risk-informed CT based on the specific plant configuration and the CDF and LERF metrics. Furthermore, the RICT program does not change the LCO Conditions, the Required Actions, or the default Conditions should the CT be exceeded. Finally, the RICT program will include a performance monitoring program, per the requirements of RG 1.174. This program is used to ensure that use of the RICT program, for a specific sse, does not degrade the performance of that SSC over time. The VEGP Performance Monitoring Program is described in detail in the September 13, 2012 LAR, Enclosure 9 (Reference 1).

E1-11 to NL-14-1016 Response to NRC Questions NRC Probabilistic Assessment Review Question #8, PRA Functionality: Reliability Section 3 of the TSTF-505 Model SE states: The proposed RICT Program uses plant-specific operating experience for component reliability and availability data. Thus the allowances permitted by the RICT Program are directly reflective of actual component performance in conjunction with component risk significance."

In some instances a PRA functional sse would be assigned its nominal unreliability in the PRA model for a RICT calculation. Depending on the specific situation and the degradation mechanism for the PRA functional SSC, the nominal reliability model may not be applicable.

Please address the following:

a. Explain how the RICT calculation would be performed, considering PRA functionality, if an evaluation determined that the SSC's reliability may be less than nominal (e.g., an increase in the failure rate over the PRA mission time). Would it be considered as non-functional for the RICT calculation (i.e., not taking credit for PRA functionality in the RICT)?

b. If it was determined to use a nominal reliability for a PRA functional SSC, would the potential for uncertainty in the reliability be considered in the RleT program? For example, does the RICT program include performing a sensitivity analysis to identify additional Risk Management Actions (RMAs) for defense in depth?

SNC Response to Probabilistic Assessment Review Question #8 In cases where sse degradation is the cause of inoperabilities, PRA Functionality determinations are performed (consistent with the following NEI 06-09 guidance):

"The PRA function may be considered in cases that involve SSG inoperabilities which, while degraded, do not involve a potential for further degrading component performance. In most cases, degrading SSCs may not be considered to be PRA functional while inoperable. For example, a pump which fails its surveillance test for required discharge pressure is declared inoperable. It cannot be considered functional for calculation of a RICT, since the cause of the degradation may be unknown, further degradation may occur, and since the safety margin established by the pump's operability requirements may no longer be met. As a counter example, a valve with a degrading stroke time may be considered PRA functional if the stroke time is not relevant to the performance of the safety function of the valve; for example, if the valve is required to close and is secured in the closed position."

As a result, the failure probability need not be increased depending on the failure mechanism causing the degraded condition. Given an inoperable condition caused by a degraded condition, the VEGP RleT Program allows only two choices to be made in the eRM Tool:

• Either a "PRA non-functional" or "PRA functional" condition to represent the TS degraded condition.

o If the inoperability is evaluated as a "PRA non-functional" condition, eRM Tool will treat the sse as failed, or E1-12 to NL-14-1016 Response to NRC Questions o If the inoperability is evaluated as a "PRA functional" condition, CRM Tool will treat the sse with the nominal base-case failure probability.

The rationale for using the nominal reliability for a PRA functional SSC includes the determination that the base case PRA results are still applicable, the degraded condition has been demonstrated to meet the PRA success criteria, and the sse is considered fully available.

No additional uncertainty or sensitivity analysis is planned to be performed during a RICT entry, which is consistent with the expectations of NEI 06-09 and the NRC SER on NEI 06-09. A one-time RICT uncertainty analysis has been performed and discussed in detail in the September 13, 2012 LAR, Enclosure 7, "Key Assumptions and Sources of Uncertainty."

The RICT Program implements RMAs prior to exceeding the RMAT per the guidance in NEI 06-09 as means of ensuring adequate defense-in-depth. The RMAT risk thresholds embody the significance of equipment unavailability and not equipment availability. As a result, NEI 06-09 does not require additional RMAs when PRA Functionality has been demonstrated for an inoperable TS Condition.

NRC Probabilistic Assessment Review Question #9. PRA Functionality: Process A PRA Functionality determination may include a number of steps, both non-PRA and PRA related, which can take time to complete. For example, a TS Operability determination must be made before a PRA Functionality determination, among other steps. Describe the process to support a PRA Functionality determination and whether this process can be accomplished within short LCO completion time frames, such as those associated with a loss-of-function, in order to support a quality decision on the RICT. If there is not enough time to step through the process, consider removing the RICT program from the loss-of-function LCOs.

SNC Response to Probabilistic Assessment Review Question #9 The PRA Functional evaluation is supported by a SNC procedure that references use of CRM System Guidelines that document the PRA success criteria required for making an expedited PRA Functionality determination. A CRM System Guideline document exists for each system included in the scope of the VEGP RICT Program. The PRA Functionality determinations are supported by site and corporate PRA analysts. In addition, the CRM Tool is highly automated and is capable of calculating a RICT for the inoperable condition in a relatively short time, within minutes for most cases. Use of the CRM Tool is supported by SNC procedures, and users undergo periodic continued training.

Regarding Operability determinations, it is true that they are made before PRA Functionality determinations; however, they will not cut into the front stop CT frame available to determine PRA Functionality. The Operability determination is made for the purpose of determining the state of the component; is it Operable or not? Therefore, the declaration of lnoperability will mark the beginning of the CT for the particular LCO. In other words, a determination of Operability will not be made during the LCO CT period; that period will be entered once it has been determined that the particular component is inoperable.

In the case of a loss-of-function situation, the 1-hour provided prior to beginning the transition to Mode 3 is considered sufficient if the cause is known. However, if a timely CRM calculation cannot be performed, a RICT will not be entered.

E1-13 to NL-14-1016 Response to NRC Questions NRC Probabilistic Assessment Review Question #1 o. Common Cause Failure Modeling RG 1.177 Appendix A describes an acceptable method for treating CCF for equipment out of service for preventive maintenance or for corrective maintenance. The CCF treatment discussed on page E6-7 of the LAR does not appear to be consistent with this guidance.

Provide assurance that RICT calculations will follow RG 1.177 guidance for CCF for preventive and corrective maintenance, or an acceptable alternative.

SNC Response to Probabilistic Assessment Review Question #1 0 The SNC RICT calculations will be performed to meet the guidance provided in NEI 06-09, which is consistent with the RG 1.177 guidance. Specifically, the treatment of CCF in the CAM Tool will be as described below:

Planned Configurations:

For planned configurations the RICT calculations will be performed consistent with NEI 06-09, Section 3.3.6, "Common Cause Failure Consideration," guidance on the treatment of CCF:

"For all RICT assessments of planned configurations, the treatment of common cause failures in the quantitative CRM Tools may be performed by considering only the removal of the planned equipment and not adjusting common cause failure terms."

This approach will result in slightly shorter completion times than if RICTs were calculated using the RG 1.177 approach (i.e., it is conservative), and it will prevent deviation from the NRC's approved approach of NEI 06-09.

Emergent Configurations:

For emergent configurations, the RICT Program will abide by NEI 06-09, Section 3.3.6, "Common Cause Failure Consideration," guidance on the treatment of CCF:

"For RICT assessments involving unplanned or emergent conditions, the potential for common cause failure is considered during the operability determination process. This assessment is more accurately described as an 'extent of condition' assessment."

"In addition to a determination of operability on the affected component, the operator should make a judgment with regard to whether the operability of similar or redundant components might be affected."

"The components are considered functional in the PRA unless the operability evaluation determines otherwise."

An "extent of condition" evaluation together with an operability evaluation will provide an assessment of the vulnerability of the operable redundant components to any common cause failure potential. The RICT determination process for an emergent configuration will be consistent with the following guidance provided in the NRC SEA for NEI 06-09:

"Emergent Failures. During the time when an RICT is in effect and risk is being assessed and managed, it is possible that emergent failures of SSCs may occur, and E1-14 to NL-14-1016 Response to NRC Questions these must be assessed to determine the impact on the RICT. If a failed component is one of two or more redundant components in separate trains of a system, then there is potential for a common cause failure mechanism. Licensees must continue to assess the remaining redundant components to determine there is reasonable assurance of their continued operability, and this is not changed by implementation of the RMTS. If a licensee concludes that the redundant components remain operable, then these components are functional for purposes of the RICT. However, the licensee is required to consider and implement additional risk management actions (RMAs), due to the potential for increased risks from common cause failure of similar equipment. The staff interprets TR NEI 06-09, Revision 0, as requiring consideration of such RMAs whenever the redundant components are considered to remain operable, but the licensee has not completed the extent of condition evaluations ... "

In keeping with the above NRC guidance, if it is determined that redundant components remain operable, these components are considered PRA functional for purpose of RICT determinations. However, SNC will consider and implement additional RMAs, due to the potential for increased risks from common cause failure of similar equipment, whenever the redundant components are considered to remain operable but an extent of condition evaluation has not yet been completed. The consideration and implementation of additional RMAs, according to the NRC SER on NEI 06-09, is considered to be consistent with the guidance of RG 1.177 regarding the treatment of increased risks from common cause failures.

In summary, the following three outcomes are expected from an "extent of condition" evaluation, and the corresponding impacts on the RICT determination process are provided below:

• Outcome 1 - The "extent of condition" evaluation concludes that redundant equipment is unaffected by the emergent equipment failure and is considered TS Operable.

o The CRM Tool will reflect the redundant equipment as PRA functional. However, additional RMAs are not considered since the components are PRA functional.

• Outcome 2 -The "extent of condition" evaluation is incomplete prior to the expiration of the TS Front Stop and the redundant equipment is considered TS Operable.

o The CRM Tool will reflect the redundant equipment as PRA functional. However, additional RMAs are considered and implemented due to the potential for increased risks from common cause failure of similar components.

• Outcome 3 - The "extent of condition" evaluation concludes that redundant equipment is TS Inoperable because of the emergent equipment failure.

o The CRM Tool will reflect the redundant equipment either as PRA functional or PRA non-functional

• PRA functional: Additional RMAs are not considered since the components are considered PRA functional.

• PRA non-functional: This condition will result in TS LOF and if the functional level PRA success criteria are not met the RICT will be exited.

E1-15 to NL-14-1016 Response to NRC Questions NRC Probabilistic Assessment Review Question #11, SSCs Not in the PRA

a. In some instances SSCs may not be in the PRA model which may have an impact on CDF and LEAF, and are necessary to quantify a RICT. Examples are the SSCs noted in Table E6.1 of the LAR. In such instances, how does your RICT program consider quantification of a RICT for those SSCs?

b. In other instances, SSCs may not be in the PRA model as a result of analyses showing no impact on CDF or LEAF. In such instances a RICT may not be quantifiable for the SSCs.

Please address the following for TS LCOs 3.6.6, and 3.6.2.

i. TS LCO 3.6.6 Containment Spray (CS) and Cooling Systems. Table E1.1 of the LAR (page E1-8) notes that CS is not credited in the PRA for containment heat removal. It also notes that the CS system does contribute to the draw down on the RWST which impacts CDF, but inoperability of CS has no severe risk impact and is included in the scope of the RICT program. The CS system also appears to be not modeled in the PRA due to low probability. Please explain if the impact on the RWST level is incorporated into the PRA model.

ii. TS LCO 3.6.2 Containment Air Locks. The LCO required actions apply to restoring the air lock to operable status after verifying a door is closed. According to LAR Table E1.1 (page E1-7) the containment air locks are in the PRA model. However, it is not clear if there is an impact expected on CDF. LAR Table E1.2 (page E1-20) appears to indicate there is no impact on CDF. However, according toLAR Table E1.3 (page E1-32), and "Note 2," (page E1-41), the calculated RICT value is less than the front-stop CT, so the RICT program cannot be entered for this instance," which appears to indicate that the impact on LEAF would not allow the RICT to be applied to the TS LCO. Please explain how these observations support the inclusion of this TS LCO in the RICT program.

SNC Response to Probabilistic Assessment Review Question #11 (a) The SSCs listed in the September 13, 2012 LAR, Enclosure 6, Table E6.1 ,*"CAMP Model Changes for Additional RICT Program SSCs," are all reflected in the CRM model. The VEGP CRM Tool is able to quantify a RICT when they are removed from service. If a sse is not in the PRA logic model but the PRA has demonstrated that it is not needed to prevent core damage or large early release, like the ESF room cooling system, the CRM Tool will be able to quantify a 30-day back stop because a zero delta CDF/delta LEAF is returned by the CRM Tool when this system is out of service.

(b) (i) The CS system is designed to provide containment atmosphere cooling to limit post-accident pressure and temperature in addition to removing iodine to reduce fission product release. Therefore, its function is synonymous with ECCS, which in "PRA functionality space is credited for prevention and mitigation of a core damage accident. As a result, CS can be included in the scope of RICT Program.

CS is not credited in PRA for containment heat removal because it has no heat exchanger for heat sink. A realistic PRA success criterion has demonstrated that the impact of CS operation more rapidly reduces the RWST level and reduces the time available for ECCS injection, which adversely impacts CDF.

E1-16 to NL-14-1016 Response to NRC Questions Unavailability of CS is programed in the CRM Tool to cause no change in CDF when CS is out of service (OOS),which is the lowest achievable change in CDF returned by the CRM Tool when a SSC is OOS, because it has been demonstrated that an unavailable CS will result in a CDF lower than the base case CDF (i.e., delta CDF will decrease when CS is OOS unlike most other PRA modeled SSCs where the delta CDF will increase when a sse is OOS, assuming no other plant systems modeled in the PRA are OOS).

In the VEGP PRA model, the operation of CS initiates LERF scenarios through CS lines. If CS started to operate and then later CS pumps fail to run, it will create potential LERF pathways if isolation of the CS penetrations fails. If CS does not start, the CS penetrations will remain isolated. The start of CS would in-fact increase LERF in VEGP. Therefore, unavailability of CS is programed in the CRM Tool to cause no change in LERF (lowest achievable change in LERF when a SSC is OOS) when CS is OOS, assuming no other plant systems modeled in the PRA are OOS.

Impact on the RWST level modeled in PRA In the VEGP PRA, accelerated depletion of RWST inventory when CS is actuated is considered through the time available for operator action to switch to ECCS recirculation. CS will actuate if containment pressure reaches the CS actuation setpoint. The increase in containment pressure after a LOCA may be lessened by containment cooling units (CCUs). In the VEGP PRA, it is conservatively assumed that CS will actuate and run without failure if the CS actuation setpoint is reached.

CS would actuate in Large and Medium LOCA cases, even if a number of CCUs are running (note: MAAP runs to support VEGP large and medium LOCA PRA modeling assumed 3 of 8 CCUs are running). CS actuation may not occur if more than 3 CCUs are running (even in large or medium LOCA cases). Therefore, the available times for operator action to switch to ECCS recirculation for large and medium LOCA cases are based on the MAAP result where CS actuated when the CS actuation setpoint is reached.

For small LOCA cases, MAAP results for VEGP PRA modeling showed that CS will not actuate if 3 of 8 CCUs are running. CS will actuate if less than 3 CCUs are running, and RWST depletion is accelerated significantly when CS pumps drain water from the RWST. For example, for a 1.6 inch diameter small LOCA break, the time to the RWST level reaching 29% (or the lo-lo level setpoint) was estimated to be approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> when CS is actuated and approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> when CS is not actuated. The significant difference in the available time for operator action to switch to ECCS recirculation resulted in different human error probabilities for the associated operator action. Therefore, in the VEGP PRA model, for the small LOCA case, two operator error basic events are modeled to represent the two cases described above.

The condition that CS is actuated in the PRA model is represented by the failure of CCUs (3 of 8 CCUs success criteria). The human error probability for the case where CS is actuated is 2.3E-3 which is much higher than the human error probability for the case where CS is not actuated, which is 2.1 E-4.

(ii) As indicated in the September 13, 2012 LAR, Enclosure 1, Table E1.2, "Unit 1/Unit 2 TS RICT Estimate Based on CDF Limit," and Table E1.3, "Unit 1/Unit 2 TS RICT E1-17 to NL-14-1016 Response to NRC Questions Estimate Based on LERF Limit," the values shown are meant to demonstrate the effect on CDF and LERF, respectively, for each individual condition to which the RICT Program applies. The actual RICT values during program implementation will be calculated based on the actual plant configuration and the on-record version of the PRA model available. This TS Condition is included in the scope of the VEGP RICT program because it is modeled in the PRA and is included in the scope of TSTF-505.

By including this TS Condition in the scope of the VEGP RICT Program, SNC is able to calculate a longer RICT for TS inoperable conditions that are considered PRA Functional.

NRC Probabilistic Assessment Review Question #12. Variable Limits According to TSTF-505, Rev 1, Scope Item 9 the following is assumed for the Traveler:

The Traveler will not modify Required Actions in Conditions in which variables are not within limit unless a modeled system could be used as a surrogate in calculating a RICT (e.g.,

using the modeled pressurizer as a surrogate for pressurizer level).

According to the LAR, the following LCOs cite variable limits:

3.5.1 Accumulators Condition A 3.5.4 Refueling Water Storage Tank Conditions A, B, and C 3.6.3 Containment Isolation Valves Condition A 3.7.6 (Unit 1) Condensate Storage Tank (CST) Condition A 3.7.6 (Unit 2) Condensate Storage Tank (CST) Condition A The RAI response letter dated August 2, 2013, notes on page E-11 that for the Unit 1 and Unit 2 TS LCO 3. 7.6, the CST itself is the surrogate. Does this mean the CST will be unavailable in the PRA model to calculate a RICT? For the other TS LCOs above, please also discuss the surrogate and its ability to be used for a RICT calculation. Also note if the surrogate provides a conservative estimate.

SNC Response to Probabilistic Assessment Review Question #12 3.5.1 Accumulators Condition A: "Accumulator Outlet Check Valve to RCS Cold Leg" is set to be failed in the CRM Tool.

• This Surrogate provides a conservative estimate by assuming loss of Accumulator in CRM Tool resulting in a conservatively shorter RICT.

• Loss of Accumulator is the worst case surrogate for this degraded condition.

3.5.4 Refueling Water Storage Tank (RWST) Condition A: "Refueling Water Storage Tank" is set to be failed in the CRM Tool.

E1-18 to NL-14-1016 Response to NRC Questions

• This Surrogate provides a conservative estimate by assuming loss of RWST in the CRM Tool resulting in a conservatively shorter RICT.

• Loss of RWST is the worst case surrogate for this degraded condition.

3.6.3 Containment Isolation Valves Conditions A, B, and C: There are approximately 34 types of containment isolation valve groups modeled in the PRA. The degraded (valve stroke time failure) valve(s) is (are) set to fail in the CRM Tool.

• The surrogate provides a conservative estimate by assuming loss of valve(s) in the CRM Tool resulting in a conservatively shorter RICT.

• Loss of valve(s) is the worst case surrogate for this degraded condition.

3.7.6 (Unit 1) Condensate Storage Tank (CST) Condition A: "CST is set to be failed in the CRM Tool.

• The surrogate provides a conservative estimate by assuming loss of CST in the CRM Tool resulting in a conservatively shorter RICT.

• Loss of CST is the worst case surrogate for this degraded condition.

3.7.6 (Unit 2) Condensate Storage Tank (CST) Condition A: "CST is set to be failed in the CRM Tool.

• The surrogate provides a conservative estimate by assuming loss of CST in the CRM Tool resulting in a conservatively shorter RICT.

• Loss of CST is the worst case surrogate for this degraded condition.

NRC Probabilistic Assessment Review Question #13. Programmatic: Success Criteria Differs from Design Basis The LAR, Enclosure 1, was provided to address the following submittal guidance:

The comparison should justify that the scope of the PRA model, including applicable success criteria such as number of SSCs required, flow rate, etc., are consistent [with the] licensing basis assumptions (i.e., 50.46 ECCS flow rates) for each of the TS requirements, or an appropriate disposition or programmatic restriction will be provided." identified LCOs with differences between the design basis and the PRA success criteria. For such LCOs, please address the following:

a. In a number of instances, the disposition in Table E1.1 justifies such differences as PRA success criteria representing "more realistic success criteria." Since the PRA success criteria differ in some instances from design basis success criteria, please confirm that the PRA success criteria is up-to-date, clearly and fully documented for the "4b" application to the level of detail necessary for the RICT program, and appropriate review processes are being implemented for the supporting calculations.

E1-19 to NL-14-1016 Response to NRC Questions

b. Discuss any applicable programmatic restrictions.

SNC Response to Probabilistic Assessment Review Question #13 (a) Success criteria are documented as part of the PRA documentation and included in the scope of the WOG Peer Review. PRA success criteria for each system included in the scope of the RICT Program are further documented in the "CRM System Guidelines" including flow rates, where applicable, for ease of use during PRA Functionality evaluations when a RICT is entered.

The PRA success criteria are documented in a SNC calculation, which is governed by SNC Calculation procedures. The success criteria calculations are living documents and are maintained to reflect the as-built, as-operated plant condition. SNC calculations are performed by qualified individuals and include a preparer, a reviewer, and an approver.

(b) Table E1.1 of the September 13, 2012 LAR documents the TS LCO Conditions included in the scope of the VEGP RICT Program for a comparison between the design basis and PRA success criteria. It also documents in the "Disposition Column" of Table E1.1 a satisfactory disposition where a difference was identified. Since all differences, as documented in Table E1.1, were satisfactorily resolved, no programmatic restrictions were necessary.

NRC Probabilistic Assessment Review Question #14. RMAs for Infrequently Tested SSCs Some SSCs or SSC function(s) in the PRA model may have relatively long times between surveillances or tests. Some SSCs or sse function(s) may be under the TSTF-425 program for surveillance test intervals which have a monitoring program requirement. For such SSCs or sse function(s) would a check of the results of the standby SSC's monitoring program be made prior to establishing a RICT, or would defense in depth RMAs related to the sse be a consideration, depending on the sse and standby time between surveillances or tests, for a RICT?

SNC Response to Probabilistic Assessment Review Question #14 VEGP has implemented the Surveillance Frequency Control Program (SFCP) under TSTF-425.

However, with or without the SFCP, the time between surveillances for a particular SSC, or group of SSCs, would not be a consideration when entering a RICT. This is not a requirement of the Standard Technical Specifications (STS) or of NEI 06-09.

Under the current STS, a check of surveillances, or performance of surveillances on SSCs, which may be, for example, redundant to the inoperable component, is not normally performed upon entering a TS Action statement. Neither is a check of the monitoring program for SSCs under the SFCP.

However, there are times when a LCO Action explicitly requires performance of a surveillance on another component. For example, LCO 3.8.1 Action 8.1 of the VEGP STS requires performing breaker alignments for the offsite circuits per SR 3.8.1.1 when a diesel generator is discovered inoperable. In such a case, the SR would be performed regardless of whether the intent was to enter a RICT or to remain within the front stop.

E1-20 to NL-14-1016 Response to NRC Questions Per SR 3.0.1, SRs must be met during the Modes or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the LCO. The SRs must be performed at the required TS frequency in accordance with the controlling programs. Requiring that an SR be met does not necessarily suggest that it must be performed; this holds true when entering Action statements (i.e., there is no requirement to perform SRs on SSCs that are redundant to the inoperable component unless explicitly directed by the TS RAS).

The above, notwithstanding the Bases to SR 3.0.1, states:

"Systems and Components are assumed to be OPERABLE when the associated SRs have been met. Nothing in this Specification, however, is to be construed as implying that systems or components are OPERABLE when:

a. The systems or components are known to be inoperable, although still meeting the SRs; or,

b. The requirements of the Surveillance(s) are known not to be met between required Surveillance performances."

Accordingly, when entering an Action within its front stop, or if the intent is to use an extended CT, if either a. or b. above is the case for a redundant component (or for a support or supported system), that component should be declared inoperable and a determination for a possible loss of safety function should commence. With respect to RMAs, they may be implemented during RICTs to perform surveillances or other activities to reduce risk. However, putting RMAs in place to perform SRs only because a particular SR has a long frequency (or has not been performed for a long period of time) will not necessarily be done.

The following examples are surveillance-type activities that may be implemented as RMAs to decrease risk:

1) During a diesel generator outage, the condition of the offsite power supply, the switchyard, and the grid may be evaluated prior to entering a RICT.

2) During a safety-related battery RICT, the remaining battery capacity may be evaluated and its ability to perform its safety function protected. Additionally, the battery float voltage on the remaining batteries may be verified on a periodic basis to ensure it is over the minimum required float voltage.

NRC Probabilistic Assessment Review Question #15. RMAs The EPRI Equipment Out of Service (EOOS) Tool provides insights such as the important equipment available during the RICT to help in identifying RMAs. However, other insights (e.g.,

important fire areas) may also be checked for potential RMAs. Please describe how your RICT program guidance considers insights, other than the EOOS tool-generated list of equipment to identify RMAs.

SNC Response to Probabilistic Assessment Review Question #15 Some maintenance activities could involve increased fire risk (performing hot work, for example). These activities are coordinated with other activities that may remove equipment from service considered important for mitigation of core damage risk. These activities require E1-21 to NL-14-1016 Response to NRC Questions coordination between the 10 CFR 50.65(a)(4) risk assessors and individuals scheduling activities on fire protection equipment to determine if RMAs are appropriate, or if any of the work should be re-scheduled. Hot work in the 'A' ECCS room while simultaneously removing a 'B' ECCS pump from service, is an example of the type of situation requiring communication and coordination between the risk assessors in Work Controls and fire protection personnel.

Other types of RMAs include the following:

• Shift briefs

• Protecting equipment

• Minimizing activities that could cause a plant transient

• Establishing alternate plant alignments

• Staging temporary equipment General guidance for RMA protected equipment development includes the following considerations:

• Protect the remaining train of a two train system including its power supplies and initiation/control logic.

• Protect equipment identified through a review of CAM Tooi"Remain in Service" list as important to the specific plant configuration.

• Consider protecting other systems that perform the same function.

• Consider placing standby equipment in service.

• If available, stage temporary equipment, such as B.5.b equipment, and perform a shift brief on its use.

Examples of measures that are established for SSCs in TSs are provided below:

• Measures that may be established during a diesel generator (DG) RICT, so that the increased risk is reduced and to ensure adequate defense in depth, are:

o The condition of the offsite power supply, the switchyard, and the grid is evaluated prior to entering a RICT, and RMAs are implemented; particularly during times of high grid stress conditions, such as during high demand conditions; o Deferral of switchyard maintenance, such as deferral of discretionary maintenance on the main, auxiliary, or startup transformers associated with the unit; o Deferral of maintenance that affects the reliability of the trains associated with the operable DGs; E1-22 to NL-14-1016 Response to NRC Questions o Deferral of planned maintenance activities on station blackout mitigating systems, and treating those systems as protected equipment; o Contacting the dispatcher on a periodic basis to provide information on the DG status and the power needs of the facility.

• Measures that could be considered during the extended period that a safety related battery is inoperable for elective maintenance, so that the increased risk is reduced and to ensure adequate defense in depth, are:

o Consider limiting the immediate discharge of the affected battery, if possible; o Consider recharging the affected battery to float voltage conditions using a spare battery charger; o Evaluate the remaining battery capacity and protect its ability to perform its safety function; o Periodically verify battery float voltage is equal to or greater than the minimum required float voltage for remaining batteries.

NRC Probabilistic Assessment Review Question #16. Risk Management Action Times CRMATsl NEI 06-09 guidance notes that if an emergent condition occurs such that a RMAT is exceeded that RMAs shall be identified and implemented. Discuss how your RICT program addresses the unlikely scenario of exceeding 1E-3/yr (CDF) and 1E-4/yr (LERF) as a result of an emergent condition, how the reduction in risk will be known given that NEI 06-09 does not require RMAs to be quantified, and the immediate steps your RICT program would perform if such an event were to occur.

SNC Response to Probabilistic Assessment Review Question #16 The 1E-3/yr (CDF) and 1E-4/yr (LERF) are instantaneous limits that are monitored when a RICT is in place, and an emergent condition causes the instantaneous CDF/LERF limits to be breached. The total RICT calculated prior to a RICT entry is significantly reduced when this limit is breached. When an emergent condition exceeds1 E-3/yr (CDF) and 1E-4/yr (LERF), the ICDP/ILERP will begin to accumulate at a significantly faster rate (steeper slope), the originally calculated RICT will be reduced considerably, and exit from the RICT will be realized sooner than originally anticipated if some of the OOS equipment is not returned to service (RTS). In other words, the remnant of the originally calculated RICT could be significantly shorter when instantaneous CDF and LERF limits exceed 1E-3 and 1E-4, respectively, because of an emergent condition. The point of the above discussion is to explain that there is no need to know the quantitative implication of implementing RMAs because the ICDP/ILERP limits of 1E-5/1 E-6 are continuously monitored and will be reached quickly once the 1E-3/yr (CDF) and 1E-4/yr (LERF) instantaneous limits are breached.

NEI 06-09, Section 3.4.1, "Risk Management Action Incorporation in a RMTS Program," state the following regarding the need for RMA quantification:

E1-23 to NL-14-1016 Response to NRC Questions "RMAs may be quantified to determine revised RICT values, but this quantification of RMAs is neither expected nor required, as omission of this RMA quantification results in conservative RICT values."

"For evolutions where compensatory RMAs are planned in support of maintenance (e.g.,

temporary diesels), it may be beneficial to quantify RMAs, to determine realistic RICT values."

SNC intends to abide by the above guidance regarding the need for quantification of RMAs.

NRC Probabilistic Assessment Review Question #17. RG 1.1n Tier 2 A Tier 2 assessment would be performed for each RICT prior to entry into the TS LCO. Discuss your RICT program Tier 2 requirements and implementation. Also, provide assurance that the Tier 2 results are reviewed for reasonableness.

SNC Response to Probabilistic Assessment review Question #17 RG 1.177 describes Tier 2 requirement as:

'7he licensee should provide reasonable assurance that risk-significant plant equipment outage configurations will not occur."

The NRC SER on NEI 06-09 states the following regarding meeting the requirements of Tier 2:

'7R NEI 06-09, Revision 0, does not permit high risk configurations which would exceed instantaneous CDF and LERF limits. It further requires implementation of RMAs when the actual or anticipated risk accumulation during a RICT will exceed 10 percent of the ICDP or ILERP limit. Such RMAs may include rescheduling planned activities to lower risk periods or implementing risk reduction measures. The limits established for entry into a RICT and for RMA implementation are consistent with the guidance of NUMARC 93-01 endorsed by RG

1. 182 as applicable to plant maintenance activities. These TR requirements are consistent with the principle of Tier 2 to avoid risk-significant configurations."

The VEGP RICT Program is consistent with the NRC SER on TR NEI 06-09, Revision 0 as stated below:

The VEGP RICT Program uses CRM functionality to limit voluntary RICT entry when instantaneous limits of 1E-3/yr (CDF) and 1E-4 (LERF) are exceeded. Additionally, as required, the VEGP RICT Program uses CRM functionality to identify RMA candidates requiring RMAs and tracks the time (RMAT) from the start of the RICT (time zero) when the ICDP/LERP exceeds 1E-6/1 E-7, the trigger for RMA implementation. The CRM Tool development has undergone testing and validations to ensure the calculated results have a high degree of confidence. In addition, the users will perform a qualitative review to ensure reasonableness of results.

NRC Probabilistic Assessment Review Question #18. Scope of Equipment It is understood that some components or equipment may not be considered for a risk-informed completion time (RICT) calculation, i.e., not within the scope of the RICT program. Please E1-24 to NL-14-1016 Response to NRC Questions explain if there are certain types of equipment which the licensee considers to be outside the scope of the RICT program (e.g., fire protection equipment, seismic-related equipment such as snubbers, etc.).

SNC Response to Probabilistic Assessment Review Question #18 Those SSCs modeled in the PRA are those SSCs being monitored in the Maintenance Rule program, as per 10 CFR 50.65. Equipment for which a RICT can be calculated is equipment within the TS that has aCT that includes the statement, "OR In accordance with the Risk Informed Completion Time Program." Systems that do not have this statement cannot use a RICT in lieu of the fixed TS CT. However, the plant configuration is taken into consideration for a RICT calculation, so systems that are not within the RICT Program that are modeled in the PRA will be part of the overall RICT evaluation. For example, fire detection and fire suppression equipment is not explicitly modeled in the PRA as a basic event, so they cannot be made OOS within the CRM Tool based on their actual operational status. However, their failure probabilities may be implicitly reflected in the fire CDF/LERF quantification that forms the basis of the RICT evaluation.

With respect to seismic contributions to the RICT calculation, VEGP will use a bounding approach, as described in the September 13, 2012 LAR.

NRC Probabilistic Assessment Review Question #19. Scope of the Implementation and Monitoring Program The implementation and monitoring program described in Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," Revision 1, includes maintenance rule control forTS equipment affected by the TS change. Please provide assurance that the proposed TS limiting condition for operation (LCO) equipment for the RICT program is covered under the licensee's maintenance rule program.

SNC Response to Probabilistic Assessment Review Question #19 SSCs in the scope of the RICT Program that have their CT extended by entry into the RICT Program are monitored to ensure their safety performance is not degraded. The RICT Program is covered under the existing Maintenance Rule (MR) monitoring program. If MR performance criteria are exceeded for any sse, a CR is generated with a cause determination conducted. If it is determined that the use of a RICT was the cause of a sse performance deficiency, the Performance Monitoring procedure directs that corrective action(s) be taken. One such action may include a moratorium on future entry into pre-planned RICTs for some period of time. In addition, it should be noted that the VEGP operating history will reflect the impact of the RICT Program, which will be accounted in future VEGP PRA model updates. The post-RICT Program implementation impact on the average annual CDF/LERF will be monitored as required per NEI 06-09 to ensure the RG 1.174 CDF/LERF limits are maintained.

Also see SNC Response to Probabilistic Assessment Review Question #7.

NRC Probabilistic Assessment Review Question #20. RICT for Multiple SSCs Inoperable In some of the proposed TS LCOs, multiple SSCs may be inoperable per TS; however, the applicable TS required actions require restoring a subset of the number of SSCs. An example E1-25 to NL-14-1016 Response to NRC Questions is TS LCO 3.4.11 Required Actions F.2 and F.3. Condition F is a condition for more than one block valve inoperable. The RICT is being proposed for F.2, restore one block valve to operable status, and F.3, restore remaining block valve to operable status. Another example is TS LCO 3.7.4 Condition B, two or more required Atmospheric Relief Valve (ARV) lines inoperable.

Required Action B.1 is to restore at least two ARV lines to operable status. In such cases a RICT is being requested. If a TS LCO Condition is entered due to multiple SSCs inoperable (e.g., greater than one block valve forTS LCO 3.4.11, or greater than two ARVs forTS 3.7.4),

the RICT for restoring multiple SSCs is expected to be shorter than the RICT for restoring a lesser number of SSCs (e.g., at least 1 block valve, or at least two ARVs). Please clarify how a RICT will be calculated for restoration of a lesser number of SSCs than for which the LCO was entered.

SNC Response for Probabilistic Assessment Review Question #20 For the case where a RICT is being proposed initially for F.2, restore one block valve to operable status, and subsequently for F.3, restore remaining block valve to operable status, the following provides the RICT calculation steps in a chronological order.

If an initial RICT is entered forTS LCO 3.4.11 Condition F - more than one block valve inoperable, the following is applicable:

Condition F.2- Restore one block valve to OPERABLE status in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or in accordance with the Risk Informed Completion Time Program.

The CAM Tool will initially calculate a RICT assuming two block valves inoperable if they are PRA non-functional. This will yield a worst-case RICT (assuming for this illustration that no other equipment is OOS), i.e., the cumulative ICDP/ILERP is being accumulated at a faster rate (steep slope). If subsequently one block valve is declared operable prior to exceeding the ICP/ILERP limit of 1E-5/1 E-6, the RICT will be recalculated assuming one block valve is returned to service. This will yield a longer overall RICT than the original RICT calculated for both block valves being inoperable and non-functional (assuming for this illustration that no other equipment is OOS), i.e., the cumulative ICDP/ILERP is being accumulated at a lesser rate (less steep slope).

The TS LCO Condition applicable for this case is as follows:

Condition F.3- Restore remaining block valve to OPERABLE status in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or in accordance with the Risk Informed Completion Time Program.

For these types of configurations, the RICT time zero does not reset, and a RICT is not exited and reentered in the CAM Tool until all overlapping OOS equipment (TS and non-TS) has been returned to service.

NRC Probabilistic Assessment Review Question #21. TS LCO 3.7.2 Required Actions D. 1 and E.1 TS LCO 3.7.2 Required Actions 0.1 and E.1 require verification" for Main Steam Isolation Valves (MSIVs) versus "restoration." In such a case, please explain how a RICT would be performed if a MSIV was considered PRA functional and there is no required action to "restore" to operability. Consider, for example, if PRA functionality was related to the MSIV closure E1-26 to NL-14-1016 Response to NRC Questions status (e.g., not fully closed). Without a "restore" required action how is a time length determined for a RICT calculation?

SNC Response to Probabilistic Assessment Review Question #21 TS LCO 3.7.2 Required Actions 0.1 and E.1 will be removed from the VEGP RICT Program.

NRC Probabilistic Assessment Review Question #22. TS LCO 3.7.9 Condition B Success Criteria License Amendments No. 170 and 152 for Vogtle Units 1 and 2, respectively, revise TS 3.7.9 by changing the criteria for nuclear service cooling water (NSCW) tower three-and four-fan operation and provides a 7-day CT for one fan/spray cell being inoperable under certain conditions.

Please describe whether this will have any impact on the PRA success criteria and explain how you plan to meet the minimum performance level of the equipment. If it is determined that the PRA success criteria needs to be updated forTS LCO 3.7.9 Condition B, please explain how you plan to update the PRA success criteria.

Note that, per LAR Table E1.1 (page E1-12), the current PRA success criteria forTS LCO 3. 7.9 Condition B requires: "4 of 4 fans when Sl signal has actuated." (This is discussed further in Table E6.2 on page E6-5.) This seems to indicate that 4 fans are currently required to ensure the minimum performance level of the equipment in the event of a LOCA, per the PRA success criteria.

SNC Response to Probabilistic Assessment Review Question #22 In the September 13, 2012 LAR, Enclosure 1, Table E1.1, "Revised TS LCO Conditions to Corresponding PRA Functions," discusses the PRA success criteria for the TS LCO 3.7.9, "Ultimate Heat Sink." The "Disposition" column of Table E1.1 states the following:

"The PRA success criteria do not credit low wet bulb temperatures to reduce the number of required fans, but are based on scenario specific heat removal requirements ... "

As a result, the addition of the 7-day CT for one fan/spray cell being inoperable under certain conditions will not impact the currently documented PRA success criteria in Table E1.1. The calculated RICT for either Condition B or Condition C will be identical assuming the plant configurations are also identical. to this letter, "Markup of TS LCO 3.7.9, 'Ultimate Heat Sink,"' reflects License Amendments No. 170 and 152 for VEGP Units 1 and 2, respectively, and shows where "In accordance with the Risk Informed Completion Time Program" applies. A change to LAR Table E1.1 will be required to agree with the current TS LCO 3.7.9, "Ultimate Heat Sink," and reflect the revision documented in Enclosure 3.

E1-27

Vogtle Electric Generating Plant Response to Request for Additional information on Plant Vogtle License Amendment Request to Revise Technical Specifications to Implement NEI 06-09, Revision 0, "Risk Informed Technical Specifications Initiative 4b, Risk Managed Technicai_Specifications (RMTS) Guidelines" Enclosure2 Comparison of Regulatory Guide (RG) 1.200 Revisions 1 and 2 Internal Events PRA Requirements to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 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, as amended by RG 1.200, Revision 1. RG 1.200, Revision 2 was issued in March 2009. So it would be prudent to review the VEGP PRAto the guidance of RG 1.200, Revision 2.

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

NEI 05-04, Revision 3, "Process for Performing Internal Events PRA Peer Reviews Using the ASMEIANS PRA Standard," provides guidance on performing a gap assessment between Addendum B of the ASME!ANS PRA Standard and RG 1.200, Revision 1, and RG 1.200, Revision 2. Section 3.3 of NEI 05-04 identifies those changes to SRs that would require are-evaluation of the PRA against the PRA Standard requirements. These SRs are included in Table E2.3 below. Note that differences considered typographical, editorial, or providing additional descriptions of the SRs were not considered technically significant and were excluded.

E2-1 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The revised wording in The VEGP internal the 2009 PRA Standard events PRA provides CC-1/11/111: CC-1/11/111: provides additional mean values of HEPs PROVIDE an assessment of the CHARACTERIZE the uncertainty in description for the SR. with associated uncertainty in the HEPs consistent the estimates of the HEPs consistent uncertainty parameters.

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

E2-2 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The revised wording in The VEGP internal the 2009 PRA Standard events PRA uses the CC-I: CC-I: provides additional HRA Calculator which USE an approach that takes the USE an approach that takes the description for the SR. takes into account plant-following into account: following into account: specific shaping factors related to the cues for (a) the complexity of the response (a) the complexity of detection, event detection, and diagnosis, decision-making and diagnosis, decision-The ASEP Approach is an executing the required response making.

acceptable approach. ,,

The ASEP Approach [2-6] is an The revised SR does not acceptable approach. require a revised or additional scope of work CC-11/111: CC-11/111: or change the conclusion When estimating HEPs EVALUATE When estimating HEPs EVALUATE of the peer review for this the impact of the following plant- the impact of the following plant- SR.

specific and scenario-specific specific and scenario-specific performance shaping factors: performance shaping factors:

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

E2-3 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The additional SR was added as part of the RG Revision 1 was CC-1/11/111: ee-1/11/111: 1.200 Revision 1 considered in the VEGP For each sse for which repair is For each SSe for which repair is to clarifications to the 2007 peer review. The VEGP to be modeled, ESTIMATE, based be modeled, ESTIMATE, based on standard and was internal events PRA does on the data collected in DA-e14, the data collected in DA-e15, the subsequently not include any sses for the probability of failure to repair probability of failure to repair the renumbered by RG 1.200 which repair is modeled.

the sse in time to prevent core sse in time to prevent core damage Revision 2.

damage as a function of the as a function of the accident accident sequence in which the sequence in which the sse failure sse failure QU-A2a: The LERF requirement Section 10.3.2 of the was added by RG 1.200 internal events PRA CC-1/11/111: ee-1/11/111: Revision 2. calculation presents PROVIDE estimates of the individual PROVIDE estimates of the individual estimates for individual sequences in a manner consistent sequences in a manner consistent The updated SR explicitly LERF sequence cutsets.

with the estimation of total eDF ... with the estimation of total eDF (and requires consideration of LERF.

E2-4 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The phrase, "from internal The peer review based events," was deleted from on the 2007 version of CC-I: CC-I: the 2009 version of the the PRA Standard ESTIMATE the point estimate CDF ESTIMATE the point estimate CDF PRA Standard. The addressed these LERF from internal events. (and LEAF). LERF requirement was requirements. Section added by RG 1.200 10.3.2 of the internal CC-11: Revision 2. events PRA calculation ESTIMATE the mean CDF from CC-11: presents the mean CDF internal events, accounting for the ESTIMATE the mean CDF (and The SR explicitly requires and LERF results.

"state-of-knowledge" correlation LEAF), accounting for the state-of- consideration of LERF.

between event probabilities [Note knowledge correlation between However, per the note in (1 )]. event probabilities [Note (1 )]. 2007 SR LE-E4 and LE-F3, LERF was addressed CC-III: in applicable CALCULATE the mean CDF from CC-III: requirements of Table internal events by propagating the CALCULATE the mean CDF (and 4.5.8, which includes all uncertainty distributions, ensuring LEAF) by propagating the QU SRs. Thus, the peer that the "state-of-knowledge" uncertainty distributions, ensuring review using the 2007 correlation between event that the state-of-knowledge version of the PRA probabilities is taken into account. correlation between event Standard addressed probabilities is taken into account. these LERF irements.

E2-5 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 QU-B5: QU-B5: The revised wording in The VEGP internal the 2009 PRA Standard events PRA logic loops CC-1/11/111: CC-1/11/111: provides less ambiguous were broken according to Fault tree linking and some other Fault tree linking and some other wording for the SR. the guidance provided modeling approaches may result in modeling approaches may result in without the introduction of circular logic that must be broken circular logic that must be broken unnecessary before the model is solved. BREAK before the model is solved. BREAK conservatisms or non-the circular logic appropriately. the circular logic appropriately. conservatisms.

Guidance for breaking logic loops is Guidance for breaking logic loops is provided in NUREG/CR-2728 [Note provided in NUREG/CR-2728 [2-13]. The revised SR does not (1 )]. When resolving circular logic, When resolving circular logic, DO require a revised or AVOID introducing unnecessary NOT introduce unnecessary additional scope of work conservatisms or non- conservatisms or non- or change the conclusion conservatisms. conservatisms. of the peer review for this SR.

E2-6 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The LEAF requirement The peer review based was added by RG 1.200 on the 2007 version of CC-1/11/111: CC-1/11/111: Revision 2. the PRA Standard ACCOUNT for system successes in ACCOUNT for system successes in addressed these LEAF addition to system failures in the addition to system failures in the The SA explicitly requires requirements. The Level evaluation of accident sequences to evaluation of accident sequences to consideration of LEAF. 2 PRA event trees the extent needed for realistic the extent needed for realistic However, per the note in presented in Section 9.2 estimation of CDF. This accounting estimation of CDF or LERF. This 2007 SR LE-E4 and LE- of the internal events may be accomplished by using accounting may be accomplished by F3, LEAF was addressed PRA calculation explicitly numerical quantification of success using numerical quantification of in applicable account for system probability, complementary logic, or success probability, complementary requirements of Table successes.

a delete term approximation and logic, or a delete term approximation 4.5.8, which includes all includes the treatment of transfers and includes the treatment of QU SAs. Thus, the peer among event trees where the transfers among event trees where review using the 2007 "successes" may not be transferred the "successes" may not be version of the PRA between event trees. transferred between event trees. Standard addressed these LEAF uirements.

E2-7 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 QU-E3: QU-E3: The LERF requirement The peer review based was added by RG 1.200 on the 2007 version of CC-I: CC-I: Revision 2. the PRA Standard ESTIMATE the uncertainty interval ESTIMATE the uncertainty interval addressed these LERF of the COF results. Provide a basis of the COF (and LERF) results. The SR explicitly requires requirements. Section for the estimate consistent with the Provide a basis for the estimate consideration of LERF. 10.4 of the internal events characterization parameter consistent with the characterization However, per the Note in PRA calculation presents uncertainties (OA-03, HR-06, HR- parameter uncertainties (OA-03, 2007 SR LE-E4 and LE- the uncertainty intervals G8, IE-C15). HR-06, HR-G8, IE-C15). F3, LERF was addressed for both COF and LERF, in applicable with consideration of the CC-11: CC-11: requirements of Table state-of-knowledge ESTIMATE the uncertainty interval ESTIMATE the uncertainty interval 4.5.8, which includes all correlation.

of the COF results. ESTIMATE the of the COF (and LERF) results. QU SRs. Thus, the peer uncertainty intervals associated with ESTIMATE the uncertainty intervals review using the 2007 parameter uncertainties (OA-03, associated with parameter version of the PRA HR-06, HR-G8, IE-C15), taking into uncertainties (OA-03, HR-06, HR- Standard addressed account the state-of-knowledge G8, IE-C15), taking into account the these LERF correlation. state-of-knowledge correlation. requirements.

CC-III: CC-III:

PROPAGATE parameter PROPAGATE parameter uncertainties (OA-03, HR-06, HR- uncertainties (OA-03, HR-06, HR-IE-C1 .... no IE-C1 ch E2-8 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 QU-E4: QU-E4: Separate requirements No action, CC-II met for for CC-I, II, and Ill were 2007 version of the PRA CC-I: CC-11111111: collapsed into a single Standard.

PROVIDE an assessment of the For each source of model requirement for CC-1/11/111 impact of the model uncertainties uncertainty and related assumption in the 2009 version of the and assumptions on the results of identified in QU-E1 and QU-E2, PRA Standard. The the PRA. respectively, IDENTIFY how the reference to Note 1 was PRA model is affected {e.g., deleted by RG 1.200 CC-11: introduction of a new basic event, Revision 2.

EVALUATE the sensitivity of the changes to basic event probabilities, results to model uncertainties and change in success criterion, The updated SR assigns key assumptions using sensitivity introduction of a new initiating the same requirement to analyses [Note {1 )]. event). all three CCs. Meeting CC-II in the 2007 version CC-III: of the PRA Standard EVALUATE the sensitivity of the assures that the new SR results to uncertain model boundary is met.

conditions and other assumptions using sensitivity analyses except where such sources of uncertainty have been adequately treated in the quantitative uncertainty analysis E2-9 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IF-F1: IFPP-81: The 2009 PRA Standard The VEG P internal divides the SR for flooding report CC-1/11/111: CC-1/11/111: documentation that adequately documents DOCUMENT the internal flooding DOCUMENT the internal flood plant facilitates applications, each of the elements of analysis in a manner that facilitates partitioning in a manner that upgrades, and review into the internal flooding PRA applications, upgrades, and facilitates PRA applications, separate SRs for each of analysis.

peer review. upgrades, and peer review. the internal flooding elements (plant The revised SR does not IFS0-81: partitioning, flood require a revised or sources, flood scenarios, additional scope of work CC-1/11/111: initiating events, and or change the conclusion DOCUMENT the internal flood quantification). of the peer review for this sources in a manner that facilitates SR.

PRA applications, upgrades, and peer review.

IFSN-81:

CC-1/11/111:

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

IFEV-81:

CC-1/11/111:

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

E2-10 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 CC-1/11/111: CC-1/11/111:

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

IF-F2: The requirement to Section 5 and Appendix A document walkdowns the internal flooding CC-1/11/111: CC-1/11/111: performed in support of PRA document the DOCUMENT the process used to DOCUMENT the process used to plant partitioning was

• walkdowns performed to identify ... flood areas, . . . For identify flood areas. For example, added to the 2009

• validate information example, this documentation this documentation typically includes version of the PRA related to flood areas, typically includes Standard. flood sources, SSCs, (a) flood areas used in the analysis mitigation and other flood (b) flood areas used in the analysis and the reason for eliminating The updated SR cites related features in the and the reason for eliminating areas from further analysis examples of acceptable flood areas.

areas from further analysis documentation of the (b) any walkdowns performed in process to identify flood support of the plant partitioning sources.

Since documentation of walkdowns was not in the 2007 version of the PRA Standard, it was not reviewed as part of the peer review conducted using that version of the PRA Standard.

E2-11 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IF-F2: IFS0-82: The requirement to The internal flooding PRA document walkdowns documents the CC-1/11/11 I: CC-1/11/11 I: performed in support of walkdowns performed to DOCUMENT the process used to DOCUMENT the process used to the identification or validate information identify applicable flood sources. For identify applicable flood sources. For screening of flood related to flood areas, example, this documentation example, this documentation sources was added to flood sources, SSCs, typically includes typically includes 2009 version of the PRA mitigation and other flood Standard. related features in the (a) flood sources identified in the (a) flood sources identified in the flood areas.

analysis, rules used to screen out analysis, rules used to screen out The updated SR cites these sources, and the resulting these sources, and the resulting examples of acceptable list of sources to be further list of sources to be further documentation of the examined examined process to identify flood sources.

(f) screening criteria used in the (b)screening criteria used in the analysis analysis Since documentation of walkdowns was not in the (j) calculations or other analyses (c) calculations or other analyses 2007 version of the PRA used to support or refine the used to support or refine the Standard, it was not flooding evaluation flooding evaluation reviewed as part of the peer review conducted (d) any walkdowns performed in using that version of the support of the identification or PRA Standard.

sc of flood sources E2-12 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IF-F2: IFSN-82: The requirement to The internal flooding PRA document walkdowns documents the CC-1/11/111: CC-1/11/111: performed in support of walkdowns performed to DOCUMENT the process used to DOCUMENT the process used to the identification or validate information identify applicable flood scenarios. identify applicable flood scenarios. screening of flood related to flood areas, For example, this documentation For example, this documentation scenarios was added to flood sources, SSCs, typically includes typically includes 2009 version of the PRA mitigation and other flood (a) propagation pathways ... Standard. related features in the (c) propagation pathways ... flood areas.

(b) accident mitigating features and The updated SR cites (d) accident mitigating features and barriers credited ... examples of acceptable barriers credited ... documentation of the (c) assumptions or calculations used process to identify flood (e) assumptions or calculations used in the determination of ... flood- scenarios.

in the determination of ... flood- induced effects on equipment induced effects on equipment operability Since documentation of operability walkdowns was not in the (d) screening criteria used in the 2007 version of the PRA (f) screening criteria used in the analysis Standard, it was not analysis reviewed as part of the (e) flooding scenarios considered, peer review conducted (g) flooding scenarios considered, screened, and retained using that version of the screened, and retained PRA Standard.

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

modified ...

(g) calculations or other analyses (j) calculations or other analyses used to support or refine the used to support or refine the flooding evaluation flooding evaluation (h) any walkdowns performed in support of the identification or Ct~*~-nll"'ln of flood ct~l:lns:trll'tct to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IF-F2: IFEV-82: The 2009 PRA Standard The VEGP internal divides the SR for typical flooding report CC-1/11/111: CC-1/11/111: items included in the adequately documents DOCUMENT the process used ... DOCUMENT the process used to internal flooding analysis each of the elements of internal flood model development. identify applicable flood-induced process documentation the internal flooding For example, this documentation initiating events. For example, this into separate SRs for analysis.

typically includes documentation typically includes each of the internal flooding elements (plant The revised SR does not (f) screening criteria used in the (a) flood frequencies, component partitioning, flood require a revised or analysis unreliabilities/unavailabilities, and sources, flood scenarios, additional scope of work HEPs used in the analysis (i.e., initiating events, and or change the conclusion (i) flood frequencies, component the data values unique to the quantification). While of the peer review for this unreliabilities/unavailabilities, and flooding analysis) most elements (PP, SO, SR.

HEPs used in the analysis (i.e., SN, QU) included the data values unique to the (b) calculations or other analyses additional requirements to flooding analysis) used to support or refine the document walkdowns, the flooding evaluation EV element included no (j) calculations or other analyses change in requirements.

used to support or refine the (c) screening criteria used in the flooding evaluation analysis E2-14 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The requirement to The internal flooding PRA document walkdowns documents the CC-1/11/111: CC-1/11/111: performed in support of walkdowns performed to DOCUMENT the process used to DOCUMENT the process used to internal flood accident validate information define the applicable internal flood define the applicable internal flood sequence quantification related to flood areas, accident sequences and their accident sequences and their was added in 2009 flood sources, SSCs, associated quantification. For associated quantification. For version of the PRA mitigation and other flood example, this documentation example, this documentation Standard. related features in the typically includes: typically includes: flood areas that are The updated SR cites considered in flood (j) calculations or other analyses (a) calculations or other analyses examples of acceptable sequence definition.

used to support or refine the used to support or refine the documentation of the flooding evaluation flooding evaluation process to identify flood related features (f) screening criteria used in the (b) screening criteria used in the considered in flood analysis analysis sequence quantification.

(i) flooding scenarios considered, (c) flooding scenarios considered, Since documentation of screened, and retained screened, and retained walkdowns was not in the 2007 version of the PRA (k) results of the internal flood (d) results of the internal flood Standard, it was not analysis, consistent with the analysis, consistent with the reviewed as part of the quantification requirements quantification requirements peer review conducted provided in HLR-QU-D provided in HLR-QU-D using that version of the PRA Standard.

(e) any walkdowns performed in support of internal flood accident sequence ntification E2-15 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IF-F3: IFPP-83: The 2009 PRA Standard The VEGP internal divides the SR for flooding report did not CC-1/11/111: CC-1/11/111: documentation of include documentation of DOCUMENT the assumptions and DOCUMENT sources of model assumptions and sources sources of model sources of uncertainty associated uncertainty and related assumptions of uncertainty into uncertainty for the with the internal flooding analysis. (as identified in QU-E1 and QU-E2) separate SRs for each of internal flooding analysis.

associated with the internal flood the internal flooding plant partitioning. elements (plant However, as part of this partitioning, flood License Amendment IFS0-83: sources, flood scenarios, Request, identification initiating events, and and documentation of the CC-1/11/111: quantification). sources of model DOCUMENT sources of model uncertainty in all uncertainty and related assumptions elements of the VEGP (as identified in QU-E1 and QU-E2) internal events Level 1 associated with the internal flood and Level2 PRA sources. (including internal flooding) was performed.

IFSN-83:

CC-1/11/111:

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

E2-16 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IF-F3: IFEV-83: The 2009 PRA Standard The VEGP internal divides the SR for flooding report did not CC-1/11/111: CC-1/11/111: documentation of include documentation of DOCUMENT the assumptions and Document sources of model assumptions and sources sources of model sources of uncertainty associated uncertainty and related of uncertainty into uncertainty for the with the internal flooding analysis. assumptions (as identified in QU- separate SRs for each of internal flooding analysis.

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

and E2-17 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The clarifications to the The VEGP internal 2007 PRA Standard flooding report met the CC-I: CC-I: provided in RG 1.200, combined CC-I/II be For the SSCs identified in IF-C2c, For the SSCs identified in IFSN-A5, Revision 1 included the describing the

... (no change) ... (no change) separation of CC-I and C- mechanisms considered,

11. The revised wording in including submergence CC-II: CC-II: the 2009 PRA Standard and spray; thereby INCLUDE failure by submergence For the SSCs identified in IFSN- provides additional meeting CC-11. However, and spray in the identification A5, IDENTIFY the susceptibility of description for the SR. there was not sufficient process. each sse in a flood area to flood- consideration of humidity, ASSESS qualitatively the impact of induced failure mechanisms. condensation, flood-induced mechanisms that are INCLUDE failure by submergence temperature, etc. to merit not formally addressed (e.g., using and spray in the identification CC-III.

the mechanisms listed under process.

Capability Category Ill of this ASSESS qualitatively the impact of requirement), by using conservative flood-induced mechanisms that are assumptions. not formally addressed (e.g., using the mechanisms listed under CC-III: Capability Category Ill of this For the SSCs identified in IF-C2c, requirement), by using conservative

... (no change) assumptions.

CC-III:

For the SSCs identified in IFSN-A5,

... (no change)

E2-18 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IE-D3 (also AS-C3. SC-C3. SY-C3. IE-D3 (also AS-C3. SC-C3, SY-C3. The NRC recommended The Recovery Analysis HR-13. DA-E3. LE-G4l: HR-13. DA-E3. LE-G4): wording changes on "key and Uncertainty Analysis assumptions and sources for the VEGP Level 1 and CC-1/11/11 I: CC-1/111111: of uncertainty were Level2 Model (Chapter DOCUMENT the key assumptions DOCUMENT the sources of model reflected in the 2009 PRA 10) included and key sources uncertainty uncertainty and related Standard. documentation on the associated with the initiating event assumptions (as identified in QU- sources of modeling analysis (also accident sequence E1 and QU-E2 or LE-F3) Since the wording uncertainty.

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

E2-19 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The NRC recommended The Recovery Analysis wording changes on "key and Uncertainty Analysis cc 1/11/11: cc 1/11/11: assumptions and sources for the VEGP Level 1 and IDENTIFY key sources of model IDENTIFY sources of model of uncertainty were Level2 Model (Chapter uncertainty. uncertainty. reflected in the 2009 PRA 10) included Standard. documentation on the sources of modeling Since the wording uncertainty.

changes were not in the 2007 version of the PRA In addition, as part of this standard, the wording License Amendment changes were not directly Request, an assessment considered during the and characterization of peer review conducted the sources of model using the 2007 PRA uncertainty in the VEGP Standard. internal events Level1 and Level 2 PRA was performed according to the guidance in NUREG-1855 and EPRI TR-1016737.

E2-20 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 QU-E2: QU-E2: The NRC recommended The Recovery Analysis wording changes on "key and Uncertainty Analysis cc 1/11/11: cc 1/11/11: assumptions and sources for the VEGP Level1 and IDENTIFY key assumptions made in IDENTIFY assumptions made in the of uncertainty were Level 2 Model (Chapter the development of the PRA model. development of the PRA model. reflected in the 2009 PRA 10) included Standard. documentation on the sources of modeling Since the wording uncertainty.

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

E2-21 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 QU-F4: QU-F4: The NRC recommended The Recovery Analysis wording changes on "key and Uncertainty Analysis cc 1/11/11: cc 1/11/11: assumptions and sources for the VEGP Level1 and DOCUMENT key assumptions and DOCUMENT the characterization of uncertainty were Level 2 Model (Chapter key sources of uncertainty, such as: of the sources of model reflected in the 2009 PRA 10) included possible optimistic or uncertainty and related assumptions Standard. documentation on the conservative success criteria, (as identified in QU-E4). sources of modeling suitability of the reliability data, Since the wording uncertainty.

possible modeling uncertainties changes were not in the (modeling limitations due to the 2007 version of the PRA In addition, as part of this method selected), degree of standard, the wording License Amendment completeness in the selection of changes were not directly Request, an assessment initiating events, possible spatial considered during the and characterization of dependencies, etc. peer review conducted the sources of model using the 2007 PRA uncertainty in the VEGP Standard. internal events Level 1 and Level 2 PRA was.

performed according to the guidance in NUREG-1855 and EPRI TR-1016737.

E2-22 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 LE-06: LE-07: The SR wording and The peer review based numbering changes were on the 2007 version of Cl: Cl: incorporated in the 2009 the PRA Standard PERFORM containment isolation PERFORM containment isolation version of the PRA addressed these LERF analysis in a conservative analysis in a conservative Standard. requirements and is not manner ... (unchanged) manner ... (unchanged) affected by the wording The difference is change.

C II: C II: considered editorial only.

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

CC Ill: CC Ill:

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

E2-23 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IE-A4a IE-A6 The wording change is a This SR was met.. ....

reduction in CCII and CC Ill: CCII and CC Ill: requirements, and does When performing the systematic When performing the systematic not need to be re-evaluation required in IE-A4, evaluation required in IE-A5, assessed unless the SR INCLUDE initiating events resulting INCLUDE initiating events resulting was not met in the peer from multiple failures, if the from multiple failures, if the review.

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

E2-24 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IE-B3: IE-B3: The revised wording in The VEGP internal the 2009 PRA Standard events PRA grouping of Cl: Cl: provides less ambiguous initiating events was GROUP initiating events only when GROUP initiating events only when wording for the SR. found to be appropriate.

the following is true: the following can be ensured:

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

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

GROUP initiating events only when GROUP initiating events only when the following is true: the following can be ensured:

(items (a) and (b) unchanged) (items (a) and (b) unchanged)

E2-25 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 AS-A10 AS-A10 The clarifications to the The VEGP internal 2007 PRA Standard were events PRA accident CCII: CCII: provided in RG 1.200, sequence construction In constructing the accident In constructing the accident Revision 1 and were was assigned CC Ill by sequence models, INCLUDE, for sequence models, INCLUDE, for subsequently also the peer review team for each modeled initiating event, each modeled initiating event, included in the 2009 PRA this SR.

sufficient detail that signifiGant sufficient detail that differences in Standard.

differences in requirements on requirements on systems and The revised SR does not systems and required operator required operator interactions (e.g., require a revised or responses interactions (e.g., systems initiations or valve additional scope of work systems initiations or valve alignment) are captured. Where or change the conclusion alignments) are captured. Where diverse systems and/ or operator of the peer review for this diverse systems and/ or operator actions provide a similar function, if SR.

actions provide a similar function, if choosing one over another changes choosing one over another changes the requirements for operator the requirements for operator intervention or the need for other intervention or the need for other systems, MODEL each separately.

MODEL each HR-E2: The clarifications to the The VEGP human 2007 PRA Standard were reliability analysis cc 1/11/111: cc 1/11/111: provided in RG 1.200, includes both the IDENTIFY IDENTIFY those actions Revision 1 and were cognitive (diagnosis) and (a) those actions required to initiate (a) required to initiate ... (not subsequently also execution portions of

... (not changed) changed) included in the 2009 PRA control room actions.

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

E2-26 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 DA-C14 DA-C15 The clarifications to the The VEGP internal 2007 PRA Standard were events PRA does not c 1/11/111: c 1/11/111: provided in RG 1.200, include any SSCs for For each SSC for which repair is to For each SSC for which repair is to Revision *1 and were which repair is modeled.

be modeled (see SY-A22), be modeled (see SY-A22), subsequently also IDENTIFY instances of plant- IDENTIFY instances of plant-specific included in the The SR is judged to be specific experience and, when experience and, when that is renumbered SR in RG not applicable that is insufficient to estimate insufficient to estimate failure to 1.200 Revision 2.

failure to repair consistent with repair consistent with DA-09 ... (not OA-08 ... chan DA-D1: DA-D1: The clarifications to the Parameter estimates for 2007 PRA Standard were most failure data in the CCII and Ill: CC II and Ill: provided in RG 1.200, VEGP internal events

... USE a Bayes update process 9F ... USE a Bayes update process . Revision 1 and were PRA were based on a equivalent statistical process that CHOOSE. .. subsequently also Bayesian update process.

assigns that assigns appropriate included in the 2009 PRA A few estimates were weight to tho statistical significance Standard to remove based on published of tho generic and plant specific ambiguity. generic data only.

evidence and provides an appropriate characterization of tho The revised SR does not uncertainty. change the conclusion of CHOOSE. .. the peer review for this SR.

E2-27 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IF-81: IFSO-A1: The requirement to Potential flood sources include the fire protection identified in Section 5 of CC-1/11/111: CC-1/11/111: system in Item (a) as a the internal flooding PRA For each flood area, IDENTIFY the For each flood area, IDENTIFY the potential flooding source reviewed as part of 2009 potential sources of flooding [Note potential sources of flooding [Note was added by RG 1.200 peer review against 2007 (1 )]. INCLUDE: (1 )]. INCLUDE Revision 1. This version of the PRA requirement was standard amended by RG (a) equipment (e.g., piping, valves, (a) equipment (e.g., piping, valves, addressed in the peer 1.200, Revision 1 include pumps) located in the area that pumps) located in the area that review, which used the RCS-connected systems are connected to fluid systems are connected to fluid systems 2007 version of the PRA - chemical and volume (e.g., circulating water system, (e.g., circulating water system, Standard amended by control system (CVCS),

service water system, fire service water system, fire RG 1.200 Revision 1. containment spray (CS),

protection system, component protection system, component residual heat removal cooling water system, feedwater cooling water system, feedwater The requirement to (RHR), reactor coolant system, condensate and steam system, condensate and steam include the reactor system drain tank systems) systems, and reactor coolant coolant system in Item (a) (RCSDT), safety injection system) as a potential flooding (SI), and reactor water source was added to the makeup system (RMWS).

2009 version of the PRA As outlined in the Plant Standard. Thus, it was Vogtle Internal Flooding not reviewed as part of notebook, the the peer review Containment Building conducted using that (and RCS components version of the PRA therein) is not included in Standard. the scope of the internal flooding analysis.

E2-28 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The clarifications to the Flood calculations for the 2007 PRA Standard were VEGP internal flooding cc 1/11/111: cc 1/11/111: provided in RG 1.200, analysis consider a range For each source and its identified For each source and its identified Revision 1 and were of flow rates.

failure mechanism, IDENTIFY the failure mechanism, IDENTIFY the subsequently also characteristic of release and the characteristic of release and the included in the re- The revised SR does not capacity of the source. INCLUDE: capacity of the source. INCLUDE: numbered SR in the 2009 change the conclusion of PRA Standard. the peer review for this (b) range of flow rates of ¥later (b) range of flow rates SR.

IF-C3b: IFSN-AS The clarifications to the Flood calculations for the 2007 PRA Standard were VEGP internal flooding C II: C II: provided in RG 1.200, analysis consider drain IDENTIFY inter-area propagation IDENTIFY inter-area propagation Revision 1 and were lines and other pathways through the normal flow path from through the normal flow path from subsequently removed not directly related to one area to another via drain lines; one area to another via drain lines; from CC II for the re- room opening.

and areas connected via back flow and areas connected via backflow numbered SR in the 2009 through drain lines involving failed through drain lines involving failed PRA Standard. The revised SR requires check valves, pipe and cable check valves, pipe and cable a reduced scope of word penetrations (including cable trays), penetrations (including cable trays), and does not change the doors, stairwells, hatchways, and doors, stairwells, hatchways, and conclusion of the peer HVAC ducts. INCLUDE potential for HVAC ducts. INCLUDE potential for review for this SR.

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

potential for barrier unavailability, maintenance activities.

E2-29

Enclosure 2 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IF-D3: IFEV-A2 The revised wording in The VEGP internal the 2009 PRA Standard flooding analysis provides less ambiguous grouping (and some wording for the SR. consequential AVOID subsuming scenarios into a DO NOT SUBSUME scenarios into subsuming) was found to group unless a group unless be appropriate.

The revised SR does not require a revised or additional scope of work or change the conclusion of the peer review for this SR.

IF-E6a: IFQU-A8 RG 1.200 Revision 2 The VEGP internal removed the phrase "due flooding quantification cc 1/11/111: cc 1/11/111: to causes independent of was performed using the INCLUDE, in the quantification, the INCLUDE, in the quantification, the the flooding" which did same basic events for combined effects of failures caused combined effects of failures caused not change the intent of other non-flooding by flooding and those coincident with by flooding and those coincident with the requirement. failures as used in the the flooding due to causes the flooding due to independent internal events model.

independent causes of the flooding causes including equipment including equipment failures, failures, unavailability due to The revised SR does not unavailability due to maintenance, maintenance, common-cause require a revised or common-cause failures, and other failures, and other credible causes. additional scope of work credible causes. or change the conclusion of the peer review for this SR.

E2-30 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The sentences were NUREG/CR-6928 is used clarifications provided in as the source for generic CC-1/11/111: CC-1/11/111: RG 1.200 Revision 1 and data priors in Revision 4 Revision 2, respectively. of the VEGP internal An example of an acceptable An example of an acceptable events PRA.

generic data sources is NUREG/CR- generic data sources is NUREG/CR- The updated SR cites a 5750 [Note 1]. 6928 [Note 1]. more recent example of Per NEI 05-0, Revision 3, an acceptable generic this SR does not require data source. re-evaluation.

DA-C1: DA-C1: Reference NUR 715 NU R-6928 is used was added by RG 1.200 as the source for generic CC-1/11/111: CC-1/11/111: Revision 1; References data priors in Revision 4 NUREG-1715 and of the VEGP internal Examples of parameter estimates Examples of parameter estimates NUREG/CR-6928 were events PRA.

and associated sources include: and associated sources include included in the 2009 (a) component failure rates and (a) component failure rates and version of the PRA Per NEI 05-0, Revision 3, probabilities: NUREG/CR-4639 probabilities: NUREG/CR-4639 Standard. this SR does not require

[Note (1 )], NUREG/CR-4550 [2-7], NUREG/CR-4550 [2-3], re-evaluation.

[Note (2)], NUREG-1715 [Note 7] NUREG-1715 [2-21], The updated SR cites NUREG/CR-6928 [2-20] more recent examples of acceptable generic data sources.

E2-31 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 Separate requirements No action, CC-II met for for CC-I, II, and Ill were 2007 version of the PRA CC-I: CC-I/lVIII: collapsed into a single Standard.

PROVIDE a qualitative assessment IDENTIFY and CHARACTERIZE the requirement for CC-1/11/111 of the key sources of uncertainty. LERF sources of model uncertainty in the 2009 version of the Examples: and related assumptions, in a PRA Standard. The (a) Identify bounding assumptions. manner consistent with the difference was not (b) Identify conservative treatment of applicable requirements of Tables 2- included in NEI 05-04, phenomena. 2.7-2(d) and 2-2.7-2(e). Revision 3.

CC-11:

PROVIDE uncertainty analysis that The updated SR assigns identifies the key sources of the same requirement to uncertainty and includes sensitivity all three CCs. Meeting studies for the significant CC-II in the 2007 version contributors to LERF. of the PRA Standard assures that the new SR CC-III: is met.

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

E2-32 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 RG 1.200 Revision 2 No action, This SR was removed the phrase met for 2007 version of C-1/11/111 : C-1/11/111 : "including uncertainty and the PRA Standard.

sensitivity analyses, as DOCUMENT the process used to DOCUMENT the process used to appropriate for the level identify plant damage states and identify plant damage states and of detail of the analysis,"

accident progression contributors, accident progression contributors, which is a reduction in the define accident progression define accident progression requirement.

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

uncertainty and sensitivity analyses, as appropriate for the level of detail of the ~n:::uu~:t*R E2-33 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 SY-815: The sentences were As noted in Table 9.2-1 of provided in the 2007 and the internal events PRA CC-1/11/111: CC-IIII/III: 2009 versions of the PRA calculation, failure of the Standard, respectively. containment boundary (h) harsh environments induced by (h) harsh environments induced by The difference was not due to venting is not containment venting, or failure containment venting, failure of included in NEI 05-04, applicable to the VEGP that may occur prior to the onset the containment venting ducts, Revision 3. large, dry, of core damage. or failure of the containment subatmospheric boundary that may occur prior to The updated SR explicitly containment.

the onset of core damage requires consideration of containment venting ducts and failure of the containment boundary prior to core damage.

E2-34

Vogtle Electric Generating Plant Response to Request for Additional Information on Plant Vogtle License Amendment Request to Revise Technical Specifications to Implement NEI 06-09, Revision 0, "Risk Informed Technical Specifications Initiative 4b, Risk Managed Technical Specifications (RMTS) Guidelines" Enclosure3 Mark-Up of Technical Specifications (TS)

Limiting Condition of Operation (LCO) 3.7.9

UHS 3.7.9

3. 7 PLANT SYSTEMS 3.7.9 Ultimate Heat Sink (UHS)

LCO 3.7.9 The UHS shall be OPERABLE. The fans/spray cells shall be as specified in Figure 3.7.9-1.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Nuclear A.1 Restore water 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Service Cooling Water temperature(s) and water (NSCW) basins with level(s) to within limits.

water temperature and/or water level not within limits.

B. One NSCW cooling B.1 Restore fan to 7days tower with one required OPERABLE status.

fan/spray cell inoperable I

when operating in four fan/spray cell required region of Figure 3.7.9-1.

c. One NSCW cooling C.1 Restore fan(s)/spray 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> tower with one or more cell(s) to OPERABLE required fans/spray cells inoperable for reasons other than Condition B.

status.

/

v v (continued)

OR In accordance with the Risk Informed Completion Time Program Vogtle Units 1 and 2 3.7.9-1 Amendment No.,:~~~ (Unit 1)

Amendment No. +e2 (Unit 2)

Text

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

40 Inverness Center Parkway Post Office Box 1295 A

Birmingham, AL 35242 Tel 205.992.7872 Fax 205.992.7601 SOUTHERN COMPANY July 17, 2014 Docket Nos.: 50-424 NL-14-1016 50-425 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant, Units 1 and 2 Response to Request for Additional Information on Plant Vogtle License Amendment Request to Revise Technical Specifications to Implement NEI 06-09, Revision 0, "Risk Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines"

References:

1. Southern Nuclear Operating Company letter, NL-12-1344, dated September 13, 2012, License Amendment Request to Revise Technical Specifications to Implement NEI 06-09, Revision 0, "Risk Informed Technical Specifications Initiative 4b, Risk Managed Technical Specifications (RMTS)

Guidelines'

2. NEI 06-09-A, Risk Informed Technical Specifications Initiative 4b, Risk Managed Technical Specifications (RMTS)

Guidelines, November, 2006

3. NRC Letter dated May 16, 2013, Vogtle Electric Generating Plant, Units 1 and 2 (VEGP) Request for Additional Information, (TAC Nos. ME9555 and ME9556)

4. Southern Nuclear Operating Company letter, NL-13-1540, dated August 2, 2013, Vogtle Electric Generating Plant Response to Request for Additional Information on Plant Vogtle License Amendment Request to Revise Technical Specifications to Implement NEI 06-09, Revision 0, "Risk Informed Technical Specifications Initiative 4b, Risk Managed Technical Specifications (RMTS) Guidelines"

5. NRC Letter dated June 9, 2014, Vogtle Electric Generating Plant, Units 1 and 2, Request for Additional Information (TAC Nos. ME9555 and ME9556)

6. NRC Letter dated June 25, 2014, Vogtle Electric Generating Plant, Units 1 and 2 - Corrected - Request for Additional Information, (TAC Nos. ME9555 and ME 9556)

U.S. Nuclear Regulatory Commission NL-14-1016 Page 2 of 3 Ladies and Gentlemen, By Reference 1, Southern Nuclear Operating Company (SNC) submitted a license amendment request for the Vogtle Electric Generating Plant (VEGP)

Technical Specifications (TS) to permit the use of the Risk Managed Technical Specifications per Reference 2.

Per Reference 3, the NRC requested additional information (RAI) to facilitate their review. SNC provided a response to that request by Reference 4.

On April 15 and 16, 2014, the NRC visited the SNC offices in Birmingham to perform an audit of the SNC VEGP license amendment request as presented in References 1 and 4. From that audit, additional questions were generated, some of which required follow-up beyond the audit meeting. Accordingly, the NRC addressed those questions in a request for additional information per Reference 5, as amended by Reference 6. Enclosure 1 to this letter contains SNC's responses to the NRC questions, which are transcribed prior to each SNC response. Enclosure 2 and Enclosure 3 provide additional information for specific responses, as documented in Enclosure 1.

Furthermore, SNC proposes to revise the TS mark-up submitted as part of the Reference 1 letter. The changes to the Reference 1 letter are described in of this letter as well as in the response to the first NRC RAI provided by Reference 4. Upon staff agreement with the proposed changes, SNC will promptly provide the updated package to the NRC.

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

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

Respectfully submitted, C.f(~ -.

C. R. Pierce Regulatory Affairs Director CRP/OCV

... ~ .,... .. .. ***

Sworn to me~n s scribed before me this 17~ay of ::T vi; ,2014 CJL~

Notary Public My commission expires: 1/z./~o/~

~*

U.S. Nuclear Regulatory Commission NL-14-1016 Page 3 of 3

Enclosures:

1. Response to Nuclear Regulatory Commission Questions

2. Comparison of Regulatory Guide (RG) 1.200 Revisions 1 and 2

3. Mark-Up of Technical Specifications (TS) Limiting Condition of Operation (LCO) 3. 7.9 cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. T. E. Tynan, Vice President- Vogtle 1 & 2 Mr. B. L. lvey, Vice President- Regulatory Affairs Mr. D. R. Madison, Vice President- Fleet Operations Mr. B. J. Adams, Vice President - Engineering Mr. S. C. Waldrup, Regulatory Affairs Manager- Vogtle RType: CVC7000 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. R. E. Martin, NRR Senior Project Manager- Vogtle 1 & 2 Mr. L. M. Cain, Senior Resident Inspector- Vogtle 1 & 2

Vogtle Electric Generating Plant Response to Request for Additional Information on Plant Vogtle License Amendment Request to Revise Technical Specifications to Implement NEI 06-09, Revision 0, "Risk Informed Technical Specifications Initiative 4b, Risk Managed Technical Specifications (RMTS) Guidelines" Enclosure 1 Response to Nuclear Regulatory Commission Questions to NL-14-1016 Response to NRC Questions Technical Specification Review RAis In the LAR, SNC has requested changes that are considered a deviation from what was approved by the NRC staff in Technical Specification Task Force (TSTF) Traveler TSTF-505.

Specifically, the deviations are:

• TS Limiting Condition for Operation (LCO) 3.5.5, "Seal Injection Flow," Condition A,

• TS LCO 3.7.3, "Main Feedwater Isolation Valves (MFIVs) and Main Feedwater Regulation valves (MFRVs) and Associated Bypass Valves," Conditions A, B, C, and D, and,

• TS LCO 3.8.3, "Diesel Fuel Oil, Lube Oil, Starting Oil, and Ventilation," Condition A.

NRC Technical Specification Review Question #1 Provide a technical basis for NRC staff review for the requested change to TS LCO 3.5.5, "Seal Injection Flow," Condition A or remove the requested change from the LAR.

NRC Technical Specification Review Question #2 Provide a technical basis for NRC staff review for the requested change toTS LCO 3.7.3, "Main Feedwater Isolation Valves (MFIVs) and Main Feedwater Regulation Valves (MFRVs) and Associated Bypass Valves," Conditions A, B, C, and D or remove the requested changes from the LAR.

NRC Technical Specification Reyiew Question #3 Provide a technical basis for NRC Staff review for the requested change to TS LCO 3.8.3, "Diesel Fuel Oil, Lube Oil, Starting Air, and Ventilation," Condition A or remove the requested changes from the LAR.

SNC Response to Technical Specifications Review Questions #1, #2, and #3 LCOs 3.5.5, 3.7.3 and 3.8.3 will be removed from the SNC VEGP Risk Informed Completion Time Program. The LAR will be revised to reflect these changes.

NRC Technical Specification Review Question #4 An oversight occurred during the NRC review of TSTF-505, Revision 1 and the

"-A" was omitted from the reference to NEI 06-09, Revision 0, in the Risk Informed Completion Time Program in proposed TS Section 5.5.22. Please provided a revised TS 5.5.22, with the reference as follows: Nuclear Energy Institute, NEI 06-09, "Risk Informed Technical Specifications Initiative 4b: Risk Managed Technical Specification (RMTS)," Revision 0-A, October 2012.

SNC Response to Technical Specification Review Question #4 A revision to the LAR will be submitted to add the "-A" to the NEI 06-09 reference in the proposed addition to TS Section 5.5.22, "Risk Informed Completion Time Program."

E1-1 to NL-14-1016 Response to NRC Questions NRC Technical Specification Review Question #5 An oversight occurred during the NRC review of TSTF-505, Revision 1, and a specific scenario was not satisfactorily addressed. SNC is requested to address the following scenario.

For this scenario, the TS system is comprised of train A and train B and performs two associated Probabilistic Risk Assessment (PRA) success criteria, called PRA function 1 and PRA function 2.

In an emergent condition, with both TS system train A and train B TS inoperable and the associated PRA success criteria considered PRA functional with train A able to perform PRA function 1 and train B able to perform function 2 (i.e., neither train by itself can perform PRA functions 1 and 2 but both trains together maintain PRA functionality), the NEI 06-09 guidelines will allow a risk informed completion time to be entered in this scenario, however there is no way to repair either train A or train B without losing PRA functionality. In this scenario, NEI 06-09 allows delaying a plant shutdown up to 30 days, depending on the system's risk significance, to repair the necessary system. This scenario was overlooked during the NRC staff's review of TSTF-505, and although the NRC staff approved NEI 06-09 and TSTF-505 it was not our intention to allow delaying plant shutdown in this type of scenario.

Please provide changes to the proposed "Risk Informed Completion Time Program," in VEGP TS 5.5.22, which prevents entry into a risk informed completion time for this specific scenario.

SNC Response to Technical Specification Review Question #5 Section 5.5.22.f will be added to prevent a RICT entry, or to exit a RICT entry already made, for the condition stated in the above question. It will state the following:

• ~ RICT entry is not permitted, or a RICT entry made shall be exited, for any condition involving a TS Loss of Function if a PRA Functionality determination that reflects the plant configuration concludes that the LCO cannot be restored without placing the TS inoperable trains in an alignment which results in a loss of functional/eve/ PRA success criteria."

E1-2 to NL-14-1016 Response to NRC Questions Probabilistic Risk Assessment Review RAis NRC Probabilistic Risk Assessment Review Question #1, Internal Events PRA Peer Review In Enclosure 2 of the LAR (page E2-4), it is stated that, "In May 2009, the VEGP PRA internal events model Revision 4 (including internal flooding) was reviewed against the requirements of the 2007 version of the PRA standard ... as amended by RG [Regulatory Guide] 1.200, Revision 1 ...."

Please summarize the peer review conducted in May 2009 and clarify if it was a full peer review where the team met the guidelines outlined in NEI 00-02 (e.g., 5 or 6 members that included the full range of experience required to perform an internal events PRA), followed the process outlined in NEI 00-02 (e.g., offsite preparation, one week onsite review, and post review documentation), and reviewed the PRA against all the elements in the ASME 2009 standard. If the review was not a full peer review, please describe the review in detail and provide all earlier Findings and Observations (F&Os) from any previous reviews.

SNC Response to Probabilistic Assessment Review Question #1 The VEGP PRA peer review conducted in May 2009 was performed using the process defined in Nuclear Energy Institute (NEI) 05-04 and it was a full-scope peer review. NEI 05-04 guidance supplants NEI 00-02 guidance for conducting a peer review. Although there is no technical impact, SNC provides the following clarification.

The VEGP License Amendment Request (LAR) made a reference to the American Society of Mechanical Engineers (ASME) RA-Sc-2007, which is not technically correct because the May 2009 peer review report references RA-Sb-2005. Both RA-Sb-2005 and RA-Sc-2007 are Addenda to ASME PRA Standard RA-S-2002. Addendum c of RA-S-2002 made only relatively minor changes to Addendum b, and these changes do not have any technical impact on the capability of the PRA. The main changes of interest between these two addenda are in Section 5 (Configuration Control), particularly sections 5.5 (Pending Changes) and 5.6 (Previous PRA Applications). In the RA-Sc-2007 addenda, additional verbiage was incorporated to provide further clarifications to a user in these two sections. Hence, the changes made in these two sections were clarification type changes. The RA-Sc-2007 addenda also added non-mandatory Appendix A to provide examples of PRA Maintenance, PRA Upgrade, and the Advisability of Peer Review, and made editorial corrections to several references in Section 4.

In February 2009, the ASME and the American Nuclear Society (ANS) approved the new, combined PRA Standard (ASMEIANS RA-Sa-2009, "Addendum to ASMEIANS RA-S-2008-Standard for Levei1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," the American Society of Mechanical Engineers and the American Nuclear Society, February, 2009). The NRC endorsed this new standard in Revision 2 of RG 1.200, which was issued in March 2009. The new standard was not generally available in time to support the VEGP peer review, so the peer review was performed against the version of the ASME PRA Standard (RA-Sb-2005) that was used in the PWR Owners Group peer reviews of internal events at power PRAs up to that point. Table E2.3 in Enclosure 2 of the VEGP LAR compares the 2007 version (RA-Sc-2007) against the 2009 version (RA-Sa-2009). Table E2.3 has been revised to make it consistent with Section 3.3 of NEI 05-04 "Process for Performing Internal Events PRA Peer Reviews Using the ASMEIANS PRA Standard," Revision 3, November 2009. The revised Table E2.3 is enclosed E1-3 to NL-14-1016 Response to NRC Questions as Enclosure 2 to this letter. The peer review concludes that the VEGP model satisfies the guidance of RG 1.200, Revision 2. The most significant change in the PRA Standard between RA-S-2002 (and addenda) and RA-S-2008 (and addenda) was the addition of requirements for PRAs for other than internal events at power. That is, the requirements for internal events at power PRAs in RA-Sb-2009 are substantially the same as those in RA-Sb-2005 (and RA-Sc-2007).

Summary of May 2009 Peer Review:

The scope of the peer review conducted in May 2009 was a full scope PRA peer review of the VEGP internal events at power PRA to determine compliance with ASME PRA Standard (RA-Sb-2005, "Addenda to ASME RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," American Society of Mechanical Engineers, New York, NY, December 2005) and RG 1.200, Revision 1. This peer review was performed using the process defined in Nuclear Energy Institute (NEI) 05-04, "Process for Performing Follow-on PRA Peer Reviews Using the ASME PRA Standard." NEI 05-04, Revision 1 has been endorsed by Revision 1 and Revision 2 of RG 1.200. Note that, although the title of NEI 05-04 includes the phrase "Follow-on," this document provides a process appropriate for both full and partial peer reviews of internal events at power PRAs against the requirements in the PRA Standard and RG 1.200.

The peer review was conducted during the week of May 4 through May 8, 2009. It covered all nine technical elements from the ASME PRA Standard plus the configuration control element.

The model that was reviewed was the VEGP Level 1 and Level 2 PRA Model Revision 4 - at power, internal events."

The peer review team consisted of six reviewers having a full range of experience required to perform the peer review. Each reviewer was assigned a lead role for a review element. The lead reviewer was assisted by two reviewers acting in a support role. The documents were supplied in advance to the peer review team members. During the week of May 4 through May 8, 2009, the peer review team was onsite, performed the review, and provided preliminary results on May 8, 2009. A final report was issued on November 10, 2009. The following table summarizes the results of the peer review.

E1-4 to NL-14-1016 Response to NRC Questions Table 4-1 Summary of Capability Category Assessment by PRA Element Capability Category SR Not Met Met CC-I CC-II CC-III CC-IIII CC-11/111 NIA TOTAL Initiating Event (IE) Total 21 0 5 0 5 0 2 33 Accident Sequence Analysis (AS) Total 17 0 1 2 0 0 1 21 Success Criteria (SC) Total 10 0 1 0 0 3 0 14 Systems Analysis (SV) Total 32 0 2 0 2 3 3 42 Human Reliability (HR) Total 1 19 0 5 1 2 6 1 35 Data Analysis (DA) Total 17 0 5 2 2 4 4 34 Internal Flooding (IF) Total 39 0 2 1 3 2 3 50 Quantification (QU) Total 1 28 0 2 1 0 2 1 35 Large Early Release Frequency (LE) Total 1 17 0 15 0 0 4 5 42 Maintenance & Update/Configuration Control 10 0 0 0 0 0 0 10 (MU) Total GRAND TOTALS 3 210 0 38 7 14 24 20 316 E1-5 to NL-14-1016 Response to NRC Questions NRC Probabilistic Assessment Review Question #2. Unreviewed Analysis Methods CUAMsl Please identify and provide technical justification for any PRA methodology that has not been formally accepted by the NRC staff. The NRC staff has formally accepted methods during resolution of UAMs as well as NUREG/CR-6850 (as supplemented) or the National Fire Protection Association Standard (NFPA) 805, "Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," frequently asked question (FAQ) guidance.

If a position on a method has been established by the NRC, please confirm that the accepted version of the method is used per the NRC position and, if not, then provide an acceptable alternative.

SNC Response to Probabilistic Assessment Review Question #2 In response to VEGP 50.69 RAI 27, SNC identified the following methods used in VEGP fire PRA that may not be consistent with those endorsed by the NRC:

• Not using 0.001 as lowest value for failure of manual suppression (using values lower than 0.001, even 0)

• Not using lower failure threshold for sensitive electronics

• Electrical cabinet heat release rate or severity factors and cabinet to cabinet fires (one method rejected, one method commented, others exist)

• The VEGP fire PRA has used alignment factor values that have not been accepted by NRC staff. Although an endorsed method has been used for alignment factor for oil pump fires, the alignment factor used by SNC for oil pump fires is slightly different than the factor accepted by the NRC staff in the NRC endorsement letter dated June 21' 2012.

SNC revised the VEGP PRA model by incorporating the following changes to remove unendorsed methods:

• 0.001 was used as the lowest value for failure of manual suppression,

• Fire modeling for sensitive electronics model was revised to be consistent with the approach in FAQ 13-004, "On Clarifications regarding Treatment of Sensitive Electronics," which was accepted on December 3, 2013 by a Letter from Hossain G.

Hamzehee at NRR,

• The use of electrical cabinet heat release rate or severity factors and cabinet to cabinet fires was removed and the model was revised using the NRC approved/endorsed methods, and

• The values suggested in the NRC endorsement letter dated June 21, 2012 were used as alignment factors for oil pump fires.

E1-6 to NL-14-1016 Response to NRC Questions SNC intends to use the revised fire PRA model that reflects the above changes to support the VEGP RICT program implementation. The PRA model used to support VEGP RICT Program will be controlled by SNC procedures to reflect the as-built, as-operated plant condition. Also, the PRA model is planned to be updated in the future as necessary for incorporating new consensus methodologies/data endorsed by the NRC.

NRC Probabilistic Assessment Review Question #3, LAR Page E2-34. F&O FSS-G4-01 Please discuss the method for assigning barrier failure probabilities of different types. If it does not follow NUREG/CR-6850, please provide the basis for it. If NUREG/CR-6850, Section 11.5.4, Table 11-3 fire barrier type failure probabilities are used, is the sum of the barrier failure probabilities used in the Fire PRA? If not, please explain how the Fire PRA will be updated to account for barriers that require summing these barrier type probabilities.

SNC Response to Probabilistic Assessment Review Question #3 In the VEGP Fire PRA, an applied barrier failure probability is based on the sum of the failure probability of each barrier element.

Reference:

Appendix C of "A.W. Vogtle Electric Generating Plant Units 1 and 2 Fire Probabilistic Risk Assessment Fire Scenario Selection Report," August, 2012.

NRC Probabilistic Assessment Review Question #4. Fire Ignition Frequencies For purposes of 4b, the Fire PRA should be updated with an NRC-accepted updated fire ignition frequency database. As such please propose an implementation item to update the Fire PRA with NRC-accepted updated fire ignition frequencies and to compare the new total baseline CDF and LEAF to the RG 1.174 acceptance criteria.

Please explain the process that will be used to stay within the risk acceptance criteria in RG 1.174 when future changes or PRA updates, such as the fire ignition frequency update, may result in exceeding those risk guidelines (e.g., baseline risk acceptance criteria of 1E-4/yr (CDF) or 1E-5/yr (LEAF), etc.).

SNC Response to Probabilistic Assessment Review Question #4 SNC will update the Fire PRA with NRC-accepted updated fire ignition frequencies when available and to compare the new baseline CDF and LEAF from quantified sources to the RG 1.174 acceptance criteria.

The PRA update process will be controlled by SNC procedures that include maintaining the total CDF and LEAF mean values from all quantified sources documented in the September 13, 2012 LAR, including impact of changes to fire ignition frequency updates, within the RG 1.174 bounds of 1E-04/yr (CDF) or 1E-05/yr (LEAF).

NRC Probabilistic Assessment Review Question #5, PRA Functional: Safety Margins The LAR states in the "Significant Hazards Considerations" section: The proposed change permits the extension of Completion Times provided risk is assessed and managed within the E1-7 to NL-14-1016 Response to NRC Questions Risk Informed Completion Time Program. The proposed change implements a risk-informed configuration management program to assure that adequate margins of safety are maintained."

A TS operability determination is made before a PRA functionality determination. A structure, system, and component (SSC) which is inoperable may be found to be outside limits in the TS or applicable code performance parameter(s), according to Inspection Manual Part 9900 (now Inspection Manual Chapter 0326, Reference 16); therefore, a PRA functional SSC may have parameter(s) outside certain limits. As such the RICT program must assure that safety margins are maintained for a PRA functional condition.

Safety margins assessment includes an assessment of assumptions or inputs to a safety analysis as discussed in RG 1.177:

Safety analysis acceptance criteria in the Final Safety Analysis Report (FSAR) are met or proposed revisions provide sufficient margin to account for analysis and data uncertainties (e.g., the proposed TS CT [Completion Time] or SF [Surveillance Frequency] change does not adversely affect any assumptions or inputs to the safety analysis, or, if such inputs are affected, justification is provided to ensure sufficient safety margin will continue to exist).

For TS CT changes, an assessment should be made of the effect on the FSAR acceptance criteria assuming the plant is in the condition addressed by the proposed CT (i.e., the subject equipment is inoperable) and there are no additional failures. Such an assessment should result in the identification of all situations in which entry into the condition addressed by the proposed CT could result in failure to meet an intended safety function.

Please explain if the RICT program considers these factors related to safety margin as noted above for a PRA functionality determination.

SNC Response to Probabilistic Assessment Review Question #5 The NRC SER for NEI 06-09, Section 3.2, "Evaluation," states the following regarding impact on safety margins:

"The proposed change maintains sufficient safety margins. The design, operation, testing methods, and acceptance criteria for SSCs, specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by risk-informed changes to the CTs. Similarly, there is no impact to safety analysis acceptance criteria, as described in the plant licensing basis. Thus, safety margins are maintained by the proposed methodology, and the third key safety principle of RG 1.277 is satisfied."

RG 1.174, Section 2.1.2 Safety Margin, provides the following basis for the above NRC SER on NEI 06-09 statement:

"... With sufficient safety margins, the following are true:

• Codes and standards or their alternatives approved for use by the NRC are met.

E1-8 to NL-14-1016 Response to NRC Questions

• Safety analysis acceptance criteria in the LB (e.g., FSAR, supporting analyses) are met or proposed revisions provide sufficient margin to account for analysis and data uncertainty."

The above conclusions of the Safety Evaluation are applicable to the VEGP RICT Program as explained below:

1) The extended CT depends on the existing plant configuration at the time the TS component becomes inoperable and throughout the time that the component remains inoperable; entry into the extended CT will not change the plant configuration.

2) Entering the extended CT via the RICT program involves no physical changes to the inoperable component or to its redundant systems and subsystems.

3) Entry into the extended CT will not alter any existing operating, testing, or abnormal operating procedures.

4) Entry into the extended CT will not alter any FSAR acceptance criteria for the Operable and Functional redundant components and will, in fact, not alter the acceptance criteria for the inoperable component although that acceptance criteria may not be met due to the component's inoperable status.

5) The Configuration Risk Management (CRM) Program will ensure that prior to entering a RICT, adequate margins of safety are maintained by ensuring that systems and subsystems redundant to the inoperable component are available, and that the PRA success criteria are met. Margins to safety are also maintained by the implementation of Risk Management Actions (RMAs), although usually no quantitative credit will be taken for the RMAs for the RICT calculation.

In summary, entry into the CT will not change the plant equipment configuration, nor will it change plant procedures, and it will not involve physical changes to plant equipment redundant to the inoperable component. Consequently, entering an extended CT does nothing to affect the existing safety margin. Furthermore, any RMAs put in place will not increase the likelihood of an initiating event or prevent the mitigation of transients or accidents because an RMA, by definition, is meant to control and/or compensate any risk increase from extending the CT when certain cumulative risk thresholds are reached.

Finally, with regard to identifying all situations in which entry into the CT could result in failure to meet an intended safety function, those situations have already been identified as "loss of function" conditions in the SNC VEGP Risk Informed TS. Each of those conditions are identified in the SNC September 13, 2012 initial submittal and the SNC RAI response of August 2, 2013.

The loss of function condition, as provided for in TSTF-505, cannot be entered voluntarily.

NRC Probabilistic Assessment Review Question #6, PRA Functionality: Success Criteria From NEI 06-09, Section 11.1, "If a component is declared inoperable due to degraded performance parameters, but the affected parameter does not and will not impact the success criteria of the PRA model, then the component may be considered PRA functional for purposes of the RICT calculation ...."

E1-9 to NL-14-1016 Response to NRC Questions Since NEI 06-09 states that the methodology is consistent with RG 1.174 philosophy, success criteria used for the RICT program must be consistent with maintaining safety margin and defense in depth philosophy as described in RG 1.174.

Success criteria information could be at a detailed or a high level {e.g., parameter, component, train, system, or other). The appropriate success criteria should be considered for a PRA functionality determination when evaluating the impact on success criteria of the PRA model.

For example, a PRA functionality determination may need to consider more than train-level success criteria {e.g., potentially both system and train level success criteria, and possibly other success criteria information). Please explain how your RICT program ensures the appropriate success criteria are considered in a PRA functionality determination.

SNC Response to Probabilistic Assessment Review Question #6 ASMEIANS RA-Sb-2013 defines success criteria as follows:

"Success criteria: criteria for establishing the minimum number or combinations of systems or components required to operate, or minimum levels of performance per component during a specific period of time, to ensure that the safety functions are satisfied... "

There are two parts to this definition:

• Part 1: Criteria for establishing the minimum number or combinations of systems or components required to operate to ensure safety functions are satisfied, or

• Part 2: Minimum levels of performance per component during a specific period of time to ensure safety functions are satisfied.

The system success criteria documented in the peer reviewed PRA address the two parts, as applicable, that support the VEGP RICT Program and are summarized as part of "CRM System Guidelines" for use during PRA Functionality determinations.

CRM System Guidelines document the PRA success criteria for all the VEGP TS systems included in the scope of the VEGP RICT Program. These include the following:

• Auxiliary Feedwater {AFW) system

• Containment Cooling Units {CCU) system

• Component Cooling Water {CCW) system

• Emergency Core Cooling Systems {ECCS)

• Condensate and Feedwater system

• Containment Isolation {CI) system

• Main Steam {MS) system

• Electrical systems

• Heating, Ventilation, and Cooling {HVAC) system

• Pressurizer Pressure Relief {PPR) system For example, the Part 1 success criteria for AFW system success criteria documented in the "CRM System Guidelines" include specific information, such as the following {this information is subject to change and is only provided to illustrate the scope of detail provided in the CRM System Guidelines):

E1-10

Enclosure 1 to NL-14-1016 Response to NRC Questions Part 1 PRA Success Criteria:

• 1 of 2 main steam lines provides steam to the turbine driven AFW pump

• 1 of 3 AFW pumps provides 570 gpm flow to 2 of 4 SGs Part 2 PRA Success Criteria:

• AFW inlet temperature to the steam generator= 120°F

• CST initial water mass = 2.8E+6 lb PRA Functionality determinations are performed based on the SNC procedure~ tha~ refer to the CRM System Guidelines as a reference source. The results of the PRA Funct1onahty determinations are documented as QA records and retained for the life of the plant.

NRC Probabilistic Assessment Review Question #7, SSC Performance Levels Per 10 CFR 50.36(c)(2) Limiting Conditions for operation. (i) Limiting Conditions for Operations are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

Section 3 of the TSTF-505 Model Safety Evaluation (SE) states:

The RICT program provides the necessary administrative controls to permit extension of CTs and thereby delay reactor shutdown or remedial actions, if risk is assessed and managed within specified limits and programmatic requirements. The specified safety function or performance levels of TS required SSCs are unchanged, and the remedial actions, including the requirement to shut down the reactor, are also unchanged; only the CTs for the Required Actions are extended by the RICT Program.

Please discuss how your RICT program addresses how required SSCs performance levels are unchanged.

SNC Response to Probabilistic Assessment Review Question #7 The VEGP RICT program is consistent with the NRC-approved guidance of NEI 06-09. It does not alter the system or train Operability requirements with respect to the number of systems and trains required to be Operable. Neither does it change the stated TS performance criteria, such as flow rates, response times, stroke times, setpoints, etc. The RICT program only serves to provide a risk-informed CT based on the specific plant configuration and the CDF and LERF metrics. Furthermore, the RICT program does not change the LCO Conditions, the Required Actions, or the default Conditions should the CT be exceeded. Finally, the RICT program will include a performance monitoring program, per the requirements of RG 1.174. This program is used to ensure that use of the RICT program, for a specific sse, does not degrade the performance of that SSC over time. The VEGP Performance Monitoring Program is described in detail in the September 13, 2012 LAR, Enclosure 9 (Reference 1).

E1-11 to NL-14-1016 Response to NRC Questions NRC Probabilistic Assessment Review Question #8, PRA Functionality: Reliability Section 3 of the TSTF-505 Model SE states: The proposed RICT Program uses plant-specific operating experience for component reliability and availability data. Thus the allowances permitted by the RICT Program are directly reflective of actual component performance in conjunction with component risk significance."

In some instances a PRA functional sse would be assigned its nominal unreliability in the PRA model for a RICT calculation. Depending on the specific situation and the degradation mechanism for the PRA functional SSC, the nominal reliability model may not be applicable.

Please address the following:

a. Explain how the RICT calculation would be performed, considering PRA functionality, if an evaluation determined that the SSC's reliability may be less than nominal (e.g., an increase in the failure rate over the PRA mission time). Would it be considered as non-functional for the RICT calculation (i.e., not taking credit for PRA functionality in the RICT)?

b. If it was determined to use a nominal reliability for a PRA functional SSC, would the potential for uncertainty in the reliability be considered in the RleT program? For example, does the RICT program include performing a sensitivity analysis to identify additional Risk Management Actions (RMAs) for defense in depth?

SNC Response to Probabilistic Assessment Review Question #8 In cases where sse degradation is the cause of inoperabilities, PRA Functionality determinations are performed (consistent with the following NEI 06-09 guidance):

"The PRA function may be considered in cases that involve SSG inoperabilities which, while degraded, do not involve a potential for further degrading component performance. In most cases, degrading SSCs may not be considered to be PRA functional while inoperable. For example, a pump which fails its surveillance test for required discharge pressure is declared inoperable. It cannot be considered functional for calculation of a RICT, since the cause of the degradation may be unknown, further degradation may occur, and since the safety margin established by the pump's operability requirements may no longer be met. As a counter example, a valve with a degrading stroke time may be considered PRA functional if the stroke time is not relevant to the performance of the safety function of the valve; for example, if the valve is required to close and is secured in the closed position."

As a result, the failure probability need not be increased depending on the failure mechanism causing the degraded condition. Given an inoperable condition caused by a degraded condition, the VEGP RleT Program allows only two choices to be made in the eRM Tool:

• Either a "PRA non-functional" or "PRA functional" condition to represent the TS degraded condition.

o If the inoperability is evaluated as a "PRA non-functional" condition, eRM Tool will treat the sse as failed, or E1-12 to NL-14-1016 Response to NRC Questions o If the inoperability is evaluated as a "PRA functional" condition, CRM Tool will treat the sse with the nominal base-case failure probability.

The rationale for using the nominal reliability for a PRA functional SSC includes the determination that the base case PRA results are still applicable, the degraded condition has been demonstrated to meet the PRA success criteria, and the sse is considered fully available.

No additional uncertainty or sensitivity analysis is planned to be performed during a RICT entry, which is consistent with the expectations of NEI 06-09 and the NRC SER on NEI 06-09. A one-time RICT uncertainty analysis has been performed and discussed in detail in the September 13, 2012 LAR, Enclosure 7, "Key Assumptions and Sources of Uncertainty."

The RICT Program implements RMAs prior to exceeding the RMAT per the guidance in NEI 06-09 as means of ensuring adequate defense-in-depth. The RMAT risk thresholds embody the significance of equipment unavailability and not equipment availability. As a result, NEI 06-09 does not require additional RMAs when PRA Functionality has been demonstrated for an inoperable TS Condition.

NRC Probabilistic Assessment Review Question #9. PRA Functionality: Process A PRA Functionality determination may include a number of steps, both non-PRA and PRA related, which can take time to complete. For example, a TS Operability determination must be made before a PRA Functionality determination, among other steps. Describe the process to support a PRA Functionality determination and whether this process can be accomplished within short LCO completion time frames, such as those associated with a loss-of-function, in order to support a quality decision on the RICT. If there is not enough time to step through the process, consider removing the RICT program from the loss-of-function LCOs.

SNC Response to Probabilistic Assessment Review Question #9 The PRA Functional evaluation is supported by a SNC procedure that references use of CRM System Guidelines that document the PRA success criteria required for making an expedited PRA Functionality determination. A CRM System Guideline document exists for each system included in the scope of the VEGP RICT Program. The PRA Functionality determinations are supported by site and corporate PRA analysts. In addition, the CRM Tool is highly automated and is capable of calculating a RICT for the inoperable condition in a relatively short time, within minutes for most cases. Use of the CRM Tool is supported by SNC procedures, and users undergo periodic continued training.

Regarding Operability determinations, it is true that they are made before PRA Functionality determinations; however, they will not cut into the front stop CT frame available to determine PRA Functionality. The Operability determination is made for the purpose of determining the state of the component; is it Operable or not? Therefore, the declaration of lnoperability will mark the beginning of the CT for the particular LCO. In other words, a determination of Operability will not be made during the LCO CT period; that period will be entered once it has been determined that the particular component is inoperable.

In the case of a loss-of-function situation, the 1-hour provided prior to beginning the transition to Mode 3 is considered sufficient if the cause is known. However, if a timely CRM calculation cannot be performed, a RICT will not be entered.

E1-13 to NL-14-1016 Response to NRC Questions NRC Probabilistic Assessment Review Question #1 o. Common Cause Failure Modeling RG 1.177 Appendix A describes an acceptable method for treating CCF for equipment out of service for preventive maintenance or for corrective maintenance. The CCF treatment discussed on page E6-7 of the LAR does not appear to be consistent with this guidance.

Provide assurance that RICT calculations will follow RG 1.177 guidance for CCF for preventive and corrective maintenance, or an acceptable alternative.

SNC Response to Probabilistic Assessment Review Question #1 0 The SNC RICT calculations will be performed to meet the guidance provided in NEI 06-09, which is consistent with the RG 1.177 guidance. Specifically, the treatment of CCF in the CAM Tool will be as described below:

Planned Configurations:

For planned configurations the RICT calculations will be performed consistent with NEI 06-09, Section 3.3.6, "Common Cause Failure Consideration," guidance on the treatment of CCF:

"For all RICT assessments of planned configurations, the treatment of common cause failures in the quantitative CRM Tools may be performed by considering only the removal of the planned equipment and not adjusting common cause failure terms."

This approach will result in slightly shorter completion times than if RICTs were calculated using the RG 1.177 approach (i.e., it is conservative), and it will prevent deviation from the NRC's approved approach of NEI 06-09.

Emergent Configurations:

For emergent configurations, the RICT Program will abide by NEI 06-09, Section 3.3.6, "Common Cause Failure Consideration," guidance on the treatment of CCF:

"For RICT assessments involving unplanned or emergent conditions, the potential for common cause failure is considered during the operability determination process. This assessment is more accurately described as an 'extent of condition' assessment."

"In addition to a determination of operability on the affected component, the operator should make a judgment with regard to whether the operability of similar or redundant components might be affected."

"The components are considered functional in the PRA unless the operability evaluation determines otherwise."

An "extent of condition" evaluation together with an operability evaluation will provide an assessment of the vulnerability of the operable redundant components to any common cause failure potential. The RICT determination process for an emergent configuration will be consistent with the following guidance provided in the NRC SEA for NEI 06-09:

"Emergent Failures. During the time when an RICT is in effect and risk is being assessed and managed, it is possible that emergent failures of SSCs may occur, and E1-14 to NL-14-1016 Response to NRC Questions these must be assessed to determine the impact on the RICT. If a failed component is one of two or more redundant components in separate trains of a system, then there is potential for a common cause failure mechanism. Licensees must continue to assess the remaining redundant components to determine there is reasonable assurance of their continued operability, and this is not changed by implementation of the RMTS. If a licensee concludes that the redundant components remain operable, then these components are functional for purposes of the RICT. However, the licensee is required to consider and implement additional risk management actions (RMAs), due to the potential for increased risks from common cause failure of similar equipment. The staff interprets TR NEI 06-09, Revision 0, as requiring consideration of such RMAs whenever the redundant components are considered to remain operable, but the licensee has not completed the extent of condition evaluations ... "

In keeping with the above NRC guidance, if it is determined that redundant components remain operable, these components are considered PRA functional for purpose of RICT determinations. However, SNC will consider and implement additional RMAs, due to the potential for increased risks from common cause failure of similar equipment, whenever the redundant components are considered to remain operable but an extent of condition evaluation has not yet been completed. The consideration and implementation of additional RMAs, according to the NRC SER on NEI 06-09, is considered to be consistent with the guidance of RG 1.177 regarding the treatment of increased risks from common cause failures.

In summary, the following three outcomes are expected from an "extent of condition" evaluation, and the corresponding impacts on the RICT determination process are provided below:

• Outcome 1 - The "extent of condition" evaluation concludes that redundant equipment is unaffected by the emergent equipment failure and is considered TS Operable.

o The CRM Tool will reflect the redundant equipment as PRA functional. However, additional RMAs are not considered since the components are PRA functional.

• Outcome 2 -The "extent of condition" evaluation is incomplete prior to the expiration of the TS Front Stop and the redundant equipment is considered TS Operable.

o The CRM Tool will reflect the redundant equipment as PRA functional. However, additional RMAs are considered and implemented due to the potential for increased risks from common cause failure of similar components.

• Outcome 3 - The "extent of condition" evaluation concludes that redundant equipment is TS Inoperable because of the emergent equipment failure.

o The CRM Tool will reflect the redundant equipment either as PRA functional or PRA non-functional

• PRA functional: Additional RMAs are not considered since the components are considered PRA functional.

• PRA non-functional: This condition will result in TS LOF and if the functional level PRA success criteria are not met the RICT will be exited.

E1-15 to NL-14-1016 Response to NRC Questions NRC Probabilistic Assessment Review Question #11, SSCs Not in the PRA

a. In some instances SSCs may not be in the PRA model which may have an impact on CDF and LEAF, and are necessary to quantify a RICT. Examples are the SSCs noted in Table E6.1 of the LAR. In such instances, how does your RICT program consider quantification of a RICT for those SSCs?

b. In other instances, SSCs may not be in the PRA model as a result of analyses showing no impact on CDF or LEAF. In such instances a RICT may not be quantifiable for the SSCs.

Please address the following for TS LCOs 3.6.6, and 3.6.2.

i. TS LCO 3.6.6 Containment Spray (CS) and Cooling Systems. Table E1.1 of the LAR (page E1-8) notes that CS is not credited in the PRA for containment heat removal. It also notes that the CS system does contribute to the draw down on the RWST which impacts CDF, but inoperability of CS has no severe risk impact and is included in the scope of the RICT program. The CS system also appears to be not modeled in the PRA due to low probability. Please explain if the impact on the RWST level is incorporated into the PRA model.

ii. TS LCO 3.6.2 Containment Air Locks. The LCO required actions apply to restoring the air lock to operable status after verifying a door is closed. According to LAR Table E1.1 (page E1-7) the containment air locks are in the PRA model. However, it is not clear if there is an impact expected on CDF. LAR Table E1.2 (page E1-20) appears to indicate there is no impact on CDF. However, according toLAR Table E1.3 (page E1-32), and "Note 2," (page E1-41), the calculated RICT value is less than the front-stop CT, so the RICT program cannot be entered for this instance," which appears to indicate that the impact on LEAF would not allow the RICT to be applied to the TS LCO. Please explain how these observations support the inclusion of this TS LCO in the RICT program.

SNC Response to Probabilistic Assessment Review Question #11 (a) The SSCs listed in the September 13, 2012 LAR, Enclosure 6, Table E6.1 ,*"CAMP Model Changes for Additional RICT Program SSCs," are all reflected in the CRM model. The VEGP CRM Tool is able to quantify a RICT when they are removed from service. If a sse is not in the PRA logic model but the PRA has demonstrated that it is not needed to prevent core damage or large early release, like the ESF room cooling system, the CRM Tool will be able to quantify a 30-day back stop because a zero delta CDF/delta LEAF is returned by the CRM Tool when this system is out of service.

(b) (i) The CS system is designed to provide containment atmosphere cooling to limit post-accident pressure and temperature in addition to removing iodine to reduce fission product release. Therefore, its function is synonymous with ECCS, which in "PRA functionality space is credited for prevention and mitigation of a core damage accident. As a result, CS can be included in the scope of RICT Program.

CS is not credited in PRA for containment heat removal because it has no heat exchanger for heat sink. A realistic PRA success criterion has demonstrated that the impact of CS operation more rapidly reduces the RWST level and reduces the time available for ECCS injection, which adversely impacts CDF.

E1-16 to NL-14-1016 Response to NRC Questions Unavailability of CS is programed in the CRM Tool to cause no change in CDF when CS is out of service (OOS),which is the lowest achievable change in CDF returned by the CRM Tool when a SSC is OOS, because it has been demonstrated that an unavailable CS will result in a CDF lower than the base case CDF (i.e., delta CDF will decrease when CS is OOS unlike most other PRA modeled SSCs where the delta CDF will increase when a sse is OOS, assuming no other plant systems modeled in the PRA are OOS).

In the VEGP PRA model, the operation of CS initiates LERF scenarios through CS lines. If CS started to operate and then later CS pumps fail to run, it will create potential LERF pathways if isolation of the CS penetrations fails. If CS does not start, the CS penetrations will remain isolated. The start of CS would in-fact increase LERF in VEGP. Therefore, unavailability of CS is programed in the CRM Tool to cause no change in LERF (lowest achievable change in LERF when a SSC is OOS) when CS is OOS, assuming no other plant systems modeled in the PRA are OOS.

Impact on the RWST level modeled in PRA In the VEGP PRA, accelerated depletion of RWST inventory when CS is actuated is considered through the time available for operator action to switch to ECCS recirculation. CS will actuate if containment pressure reaches the CS actuation setpoint. The increase in containment pressure after a LOCA may be lessened by containment cooling units (CCUs). In the VEGP PRA, it is conservatively assumed that CS will actuate and run without failure if the CS actuation setpoint is reached.

CS would actuate in Large and Medium LOCA cases, even if a number of CCUs are running (note: MAAP runs to support VEGP large and medium LOCA PRA modeling assumed 3 of 8 CCUs are running). CS actuation may not occur if more than 3 CCUs are running (even in large or medium LOCA cases). Therefore, the available times for operator action to switch to ECCS recirculation for large and medium LOCA cases are based on the MAAP result where CS actuated when the CS actuation setpoint is reached.

For small LOCA cases, MAAP results for VEGP PRA modeling showed that CS will not actuate if 3 of 8 CCUs are running. CS will actuate if less than 3 CCUs are running, and RWST depletion is accelerated significantly when CS pumps drain water from the RWST. For example, for a 1.6 inch diameter small LOCA break, the time to the RWST level reaching 29% (or the lo-lo level setpoint) was estimated to be approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> when CS is actuated and approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> when CS is not actuated. The significant difference in the available time for operator action to switch to ECCS recirculation resulted in different human error probabilities for the associated operator action. Therefore, in the VEGP PRA model, for the small LOCA case, two operator error basic events are modeled to represent the two cases described above.

The condition that CS is actuated in the PRA model is represented by the failure of CCUs (3 of 8 CCUs success criteria). The human error probability for the case where CS is actuated is 2.3E-3 which is much higher than the human error probability for the case where CS is not actuated, which is 2.1 E-4.

(ii) As indicated in the September 13, 2012 LAR, Enclosure 1, Table E1.2, "Unit 1/Unit 2 TS RICT Estimate Based on CDF Limit," and Table E1.3, "Unit 1/Unit 2 TS RICT E1-17 to NL-14-1016 Response to NRC Questions Estimate Based on LERF Limit," the values shown are meant to demonstrate the effect on CDF and LERF, respectively, for each individual condition to which the RICT Program applies. The actual RICT values during program implementation will be calculated based on the actual plant configuration and the on-record version of the PRA model available. This TS Condition is included in the scope of the VEGP RICT program because it is modeled in the PRA and is included in the scope of TSTF-505.

By including this TS Condition in the scope of the VEGP RICT Program, SNC is able to calculate a longer RICT for TS inoperable conditions that are considered PRA Functional.

NRC Probabilistic Assessment Review Question #12. Variable Limits According to TSTF-505, Rev 1, Scope Item 9 the following is assumed for the Traveler:

The Traveler will not modify Required Actions in Conditions in which variables are not within limit unless a modeled system could be used as a surrogate in calculating a RICT (e.g.,

using the modeled pressurizer as a surrogate for pressurizer level).

According to the LAR, the following LCOs cite variable limits:

3.5.1 Accumulators Condition A 3.5.4 Refueling Water Storage Tank Conditions A, B, and C 3.6.3 Containment Isolation Valves Condition A 3.7.6 (Unit 1) Condensate Storage Tank (CST) Condition A 3.7.6 (Unit 2) Condensate Storage Tank (CST) Condition A The RAI response letter dated August 2, 2013, notes on page E-11 that for the Unit 1 and Unit 2 TS LCO 3. 7.6, the CST itself is the surrogate. Does this mean the CST will be unavailable in the PRA model to calculate a RICT? For the other TS LCOs above, please also discuss the surrogate and its ability to be used for a RICT calculation. Also note if the surrogate provides a conservative estimate.

SNC Response to Probabilistic Assessment Review Question #12 3.5.1 Accumulators Condition A: "Accumulator Outlet Check Valve to RCS Cold Leg" is set to be failed in the CRM Tool.

• This Surrogate provides a conservative estimate by assuming loss of Accumulator in CRM Tool resulting in a conservatively shorter RICT.

• Loss of Accumulator is the worst case surrogate for this degraded condition.

3.5.4 Refueling Water Storage Tank (RWST) Condition A: "Refueling Water Storage Tank" is set to be failed in the CRM Tool.

E1-18 to NL-14-1016 Response to NRC Questions

• This Surrogate provides a conservative estimate by assuming loss of RWST in the CRM Tool resulting in a conservatively shorter RICT.

• Loss of RWST is the worst case surrogate for this degraded condition.

3.6.3 Containment Isolation Valves Conditions A, B, and C: There are approximately 34 types of containment isolation valve groups modeled in the PRA. The degraded (valve stroke time failure) valve(s) is (are) set to fail in the CRM Tool.

• The surrogate provides a conservative estimate by assuming loss of valve(s) in the CRM Tool resulting in a conservatively shorter RICT.

• Loss of valve(s) is the worst case surrogate for this degraded condition.

3.7.6 (Unit 1) Condensate Storage Tank (CST) Condition A: "CST is set to be failed in the CRM Tool.

• The surrogate provides a conservative estimate by assuming loss of CST in the CRM Tool resulting in a conservatively shorter RICT.

• Loss of CST is the worst case surrogate for this degraded condition.

3.7.6 (Unit 2) Condensate Storage Tank (CST) Condition A: "CST is set to be failed in the CRM Tool.

• The surrogate provides a conservative estimate by assuming loss of CST in the CRM Tool resulting in a conservatively shorter RICT.

• Loss of CST is the worst case surrogate for this degraded condition.

NRC Probabilistic Assessment Review Question #13. Programmatic: Success Criteria Differs from Design Basis The LAR, Enclosure 1, was provided to address the following submittal guidance:

The comparison should justify that the scope of the PRA model, including applicable success criteria such as number of SSCs required, flow rate, etc., are consistent [with the] licensing basis assumptions (i.e., 50.46 ECCS flow rates) for each of the TS requirements, or an appropriate disposition or programmatic restriction will be provided." identified LCOs with differences between the design basis and the PRA success criteria. For such LCOs, please address the following:

a. In a number of instances, the disposition in Table E1.1 justifies such differences as PRA success criteria representing "more realistic success criteria." Since the PRA success criteria differ in some instances from design basis success criteria, please confirm that the PRA success criteria is up-to-date, clearly and fully documented for the "4b" application to the level of detail necessary for the RICT program, and appropriate review processes are being implemented for the supporting calculations.

E1-19 to NL-14-1016 Response to NRC Questions

b. Discuss any applicable programmatic restrictions.

SNC Response to Probabilistic Assessment Review Question #13 (a) Success criteria are documented as part of the PRA documentation and included in the scope of the WOG Peer Review. PRA success criteria for each system included in the scope of the RICT Program are further documented in the "CRM System Guidelines" including flow rates, where applicable, for ease of use during PRA Functionality evaluations when a RICT is entered.

The PRA success criteria are documented in a SNC calculation, which is governed by SNC Calculation procedures. The success criteria calculations are living documents and are maintained to reflect the as-built, as-operated plant condition. SNC calculations are performed by qualified individuals and include a preparer, a reviewer, and an approver.

(b) Table E1.1 of the September 13, 2012 LAR documents the TS LCO Conditions included in the scope of the VEGP RICT Program for a comparison between the design basis and PRA success criteria. It also documents in the "Disposition Column" of Table E1.1 a satisfactory disposition where a difference was identified. Since all differences, as documented in Table E1.1, were satisfactorily resolved, no programmatic restrictions were necessary.

NRC Probabilistic Assessment Review Question #14. RMAs for Infrequently Tested SSCs Some SSCs or SSC function(s) in the PRA model may have relatively long times between surveillances or tests. Some SSCs or sse function(s) may be under the TSTF-425 program for surveillance test intervals which have a monitoring program requirement. For such SSCs or sse function(s) would a check of the results of the standby SSC's monitoring program be made prior to establishing a RICT, or would defense in depth RMAs related to the sse be a consideration, depending on the sse and standby time between surveillances or tests, for a RICT?

SNC Response to Probabilistic Assessment Review Question #14 VEGP has implemented the Surveillance Frequency Control Program (SFCP) under TSTF-425.

However, with or without the SFCP, the time between surveillances for a particular SSC, or group of SSCs, would not be a consideration when entering a RICT. This is not a requirement of the Standard Technical Specifications (STS) or of NEI 06-09.

Under the current STS, a check of surveillances, or performance of surveillances on SSCs, which may be, for example, redundant to the inoperable component, is not normally performed upon entering a TS Action statement. Neither is a check of the monitoring program for SSCs under the SFCP.

However, there are times when a LCO Action explicitly requires performance of a surveillance on another component. For example, LCO 3.8.1 Action 8.1 of the VEGP STS requires performing breaker alignments for the offsite circuits per SR 3.8.1.1 when a diesel generator is discovered inoperable. In such a case, the SR would be performed regardless of whether the intent was to enter a RICT or to remain within the front stop.

E1-20 to NL-14-1016 Response to NRC Questions Per SR 3.0.1, SRs must be met during the Modes or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the LCO. The SRs must be performed at the required TS frequency in accordance with the controlling programs. Requiring that an SR be met does not necessarily suggest that it must be performed; this holds true when entering Action statements (i.e., there is no requirement to perform SRs on SSCs that are redundant to the inoperable component unless explicitly directed by the TS RAS).

The above, notwithstanding the Bases to SR 3.0.1, states:

"Systems and Components are assumed to be OPERABLE when the associated SRs have been met. Nothing in this Specification, however, is to be construed as implying that systems or components are OPERABLE when:

a. The systems or components are known to be inoperable, although still meeting the SRs; or,

b. The requirements of the Surveillance(s) are known not to be met between required Surveillance performances."

Accordingly, when entering an Action within its front stop, or if the intent is to use an extended CT, if either a. or b. above is the case for a redundant component (or for a support or supported system), that component should be declared inoperable and a determination for a possible loss of safety function should commence. With respect to RMAs, they may be implemented during RICTs to perform surveillances or other activities to reduce risk. However, putting RMAs in place to perform SRs only because a particular SR has a long frequency (or has not been performed for a long period of time) will not necessarily be done.

The following examples are surveillance-type activities that may be implemented as RMAs to decrease risk:

1) During a diesel generator outage, the condition of the offsite power supply, the switchyard, and the grid may be evaluated prior to entering a RICT.

2) During a safety-related battery RICT, the remaining battery capacity may be evaluated and its ability to perform its safety function protected. Additionally, the battery float voltage on the remaining batteries may be verified on a periodic basis to ensure it is over the minimum required float voltage.

NRC Probabilistic Assessment Review Question #15. RMAs The EPRI Equipment Out of Service (EOOS) Tool provides insights such as the important equipment available during the RICT to help in identifying RMAs. However, other insights (e.g.,

important fire areas) may also be checked for potential RMAs. Please describe how your RICT program guidance considers insights, other than the EOOS tool-generated list of equipment to identify RMAs.

SNC Response to Probabilistic Assessment Review Question #15 Some maintenance activities could involve increased fire risk (performing hot work, for example). These activities are coordinated with other activities that may remove equipment from service considered important for mitigation of core damage risk. These activities require E1-21 to NL-14-1016 Response to NRC Questions coordination between the 10 CFR 50.65(a)(4) risk assessors and individuals scheduling activities on fire protection equipment to determine if RMAs are appropriate, or if any of the work should be re-scheduled. Hot work in the 'A' ECCS room while simultaneously removing a 'B' ECCS pump from service, is an example of the type of situation requiring communication and coordination between the risk assessors in Work Controls and fire protection personnel.

Other types of RMAs include the following:

• Shift briefs

• Protecting equipment

• Minimizing activities that could cause a plant transient

• Establishing alternate plant alignments

• Staging temporary equipment General guidance for RMA protected equipment development includes the following considerations:

• Protect the remaining train of a two train system including its power supplies and initiation/control logic.

• Protect equipment identified through a review of CAM Tooi"Remain in Service" list as important to the specific plant configuration.

• Consider protecting other systems that perform the same function.

• Consider placing standby equipment in service.

• If available, stage temporary equipment, such as B.5.b equipment, and perform a shift brief on its use.

Examples of measures that are established for SSCs in TSs are provided below:

• Measures that may be established during a diesel generator (DG) RICT, so that the increased risk is reduced and to ensure adequate defense in depth, are:

o The condition of the offsite power supply, the switchyard, and the grid is evaluated prior to entering a RICT, and RMAs are implemented; particularly during times of high grid stress conditions, such as during high demand conditions; o Deferral of switchyard maintenance, such as deferral of discretionary maintenance on the main, auxiliary, or startup transformers associated with the unit; o Deferral of maintenance that affects the reliability of the trains associated with the operable DGs; E1-22 to NL-14-1016 Response to NRC Questions o Deferral of planned maintenance activities on station blackout mitigating systems, and treating those systems as protected equipment; o Contacting the dispatcher on a periodic basis to provide information on the DG status and the power needs of the facility.

• Measures that could be considered during the extended period that a safety related battery is inoperable for elective maintenance, so that the increased risk is reduced and to ensure adequate defense in depth, are:

o Consider limiting the immediate discharge of the affected battery, if possible; o Consider recharging the affected battery to float voltage conditions using a spare battery charger; o Evaluate the remaining battery capacity and protect its ability to perform its safety function; o Periodically verify battery float voltage is equal to or greater than the minimum required float voltage for remaining batteries.

NRC Probabilistic Assessment Review Question #16. Risk Management Action Times CRMATsl NEI 06-09 guidance notes that if an emergent condition occurs such that a RMAT is exceeded that RMAs shall be identified and implemented. Discuss how your RICT program addresses the unlikely scenario of exceeding 1E-3/yr (CDF) and 1E-4/yr (LERF) as a result of an emergent condition, how the reduction in risk will be known given that NEI 06-09 does not require RMAs to be quantified, and the immediate steps your RICT program would perform if such an event were to occur.

SNC Response to Probabilistic Assessment Review Question #16 The 1E-3/yr (CDF) and 1E-4/yr (LERF) are instantaneous limits that are monitored when a RICT is in place, and an emergent condition causes the instantaneous CDF/LERF limits to be breached. The total RICT calculated prior to a RICT entry is significantly reduced when this limit is breached. When an emergent condition exceeds1 E-3/yr (CDF) and 1E-4/yr (LERF), the ICDP/ILERP will begin to accumulate at a significantly faster rate (steeper slope), the originally calculated RICT will be reduced considerably, and exit from the RICT will be realized sooner than originally anticipated if some of the OOS equipment is not returned to service (RTS). In other words, the remnant of the originally calculated RICT could be significantly shorter when instantaneous CDF and LERF limits exceed 1E-3 and 1E-4, respectively, because of an emergent condition. The point of the above discussion is to explain that there is no need to know the quantitative implication of implementing RMAs because the ICDP/ILERP limits of 1E-5/1 E-6 are continuously monitored and will be reached quickly once the 1E-3/yr (CDF) and 1E-4/yr (LERF) instantaneous limits are breached.

NEI 06-09, Section 3.4.1, "Risk Management Action Incorporation in a RMTS Program," state the following regarding the need for RMA quantification:

E1-23 to NL-14-1016 Response to NRC Questions "RMAs may be quantified to determine revised RICT values, but this quantification of RMAs is neither expected nor required, as omission of this RMA quantification results in conservative RICT values."

"For evolutions where compensatory RMAs are planned in support of maintenance (e.g.,

temporary diesels), it may be beneficial to quantify RMAs, to determine realistic RICT values."

SNC intends to abide by the above guidance regarding the need for quantification of RMAs.

NRC Probabilistic Assessment Review Question #17. RG 1.1n Tier 2 A Tier 2 assessment would be performed for each RICT prior to entry into the TS LCO. Discuss your RICT program Tier 2 requirements and implementation. Also, provide assurance that the Tier 2 results are reviewed for reasonableness.

SNC Response to Probabilistic Assessment review Question #17 RG 1.177 describes Tier 2 requirement as:

'7he licensee should provide reasonable assurance that risk-significant plant equipment outage configurations will not occur."

The NRC SER on NEI 06-09 states the following regarding meeting the requirements of Tier 2:

'7R NEI 06-09, Revision 0, does not permit high risk configurations which would exceed instantaneous CDF and LERF limits. It further requires implementation of RMAs when the actual or anticipated risk accumulation during a RICT will exceed 10 percent of the ICDP or ILERP limit. Such RMAs may include rescheduling planned activities to lower risk periods or implementing risk reduction measures. The limits established for entry into a RICT and for RMA implementation are consistent with the guidance of NUMARC 93-01 endorsed by RG

1. 182 as applicable to plant maintenance activities. These TR requirements are consistent with the principle of Tier 2 to avoid risk-significant configurations."

The VEGP RICT Program is consistent with the NRC SER on TR NEI 06-09, Revision 0 as stated below:

The VEGP RICT Program uses CRM functionality to limit voluntary RICT entry when instantaneous limits of 1E-3/yr (CDF) and 1E-4 (LERF) are exceeded. Additionally, as required, the VEGP RICT Program uses CRM functionality to identify RMA candidates requiring RMAs and tracks the time (RMAT) from the start of the RICT (time zero) when the ICDP/LERP exceeds 1E-6/1 E-7, the trigger for RMA implementation. The CRM Tool development has undergone testing and validations to ensure the calculated results have a high degree of confidence. In addition, the users will perform a qualitative review to ensure reasonableness of results.

NRC Probabilistic Assessment Review Question #18. Scope of Equipment It is understood that some components or equipment may not be considered for a risk-informed completion time (RICT) calculation, i.e., not within the scope of the RICT program. Please E1-24 to NL-14-1016 Response to NRC Questions explain if there are certain types of equipment which the licensee considers to be outside the scope of the RICT program (e.g., fire protection equipment, seismic-related equipment such as snubbers, etc.).

SNC Response to Probabilistic Assessment Review Question #18 Those SSCs modeled in the PRA are those SSCs being monitored in the Maintenance Rule program, as per 10 CFR 50.65. Equipment for which a RICT can be calculated is equipment within the TS that has aCT that includes the statement, "OR In accordance with the Risk Informed Completion Time Program." Systems that do not have this statement cannot use a RICT in lieu of the fixed TS CT. However, the plant configuration is taken into consideration for a RICT calculation, so systems that are not within the RICT Program that are modeled in the PRA will be part of the overall RICT evaluation. For example, fire detection and fire suppression equipment is not explicitly modeled in the PRA as a basic event, so they cannot be made OOS within the CRM Tool based on their actual operational status. However, their failure probabilities may be implicitly reflected in the fire CDF/LERF quantification that forms the basis of the RICT evaluation.

With respect to seismic contributions to the RICT calculation, VEGP will use a bounding approach, as described in the September 13, 2012 LAR.

NRC Probabilistic Assessment Review Question #19. Scope of the Implementation and Monitoring Program The implementation and monitoring program described in Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," Revision 1, includes maintenance rule control forTS equipment affected by the TS change. Please provide assurance that the proposed TS limiting condition for operation (LCO) equipment for the RICT program is covered under the licensee's maintenance rule program.

SNC Response to Probabilistic Assessment Review Question #19 SSCs in the scope of the RICT Program that have their CT extended by entry into the RICT Program are monitored to ensure their safety performance is not degraded. The RICT Program is covered under the existing Maintenance Rule (MR) monitoring program. If MR performance criteria are exceeded for any sse, a CR is generated with a cause determination conducted. If it is determined that the use of a RICT was the cause of a sse performance deficiency, the Performance Monitoring procedure directs that corrective action(s) be taken. One such action may include a moratorium on future entry into pre-planned RICTs for some period of time. In addition, it should be noted that the VEGP operating history will reflect the impact of the RICT Program, which will be accounted in future VEGP PRA model updates. The post-RICT Program implementation impact on the average annual CDF/LERF will be monitored as required per NEI 06-09 to ensure the RG 1.174 CDF/LERF limits are maintained.

Also see SNC Response to Probabilistic Assessment Review Question #7.

NRC Probabilistic Assessment Review Question #20. RICT for Multiple SSCs Inoperable In some of the proposed TS LCOs, multiple SSCs may be inoperable per TS; however, the applicable TS required actions require restoring a subset of the number of SSCs. An example E1-25 to NL-14-1016 Response to NRC Questions is TS LCO 3.4.11 Required Actions F.2 and F.3. Condition F is a condition for more than one block valve inoperable. The RICT is being proposed for F.2, restore one block valve to operable status, and F.3, restore remaining block valve to operable status. Another example is TS LCO 3.7.4 Condition B, two or more required Atmospheric Relief Valve (ARV) lines inoperable.

Required Action B.1 is to restore at least two ARV lines to operable status. In such cases a RICT is being requested. If a TS LCO Condition is entered due to multiple SSCs inoperable (e.g., greater than one block valve forTS LCO 3.4.11, or greater than two ARVs forTS 3.7.4),

the RICT for restoring multiple SSCs is expected to be shorter than the RICT for restoring a lesser number of SSCs (e.g., at least 1 block valve, or at least two ARVs). Please clarify how a RICT will be calculated for restoration of a lesser number of SSCs than for which the LCO was entered.

SNC Response for Probabilistic Assessment Review Question #20 For the case where a RICT is being proposed initially for F.2, restore one block valve to operable status, and subsequently for F.3, restore remaining block valve to operable status, the following provides the RICT calculation steps in a chronological order.

If an initial RICT is entered forTS LCO 3.4.11 Condition F - more than one block valve inoperable, the following is applicable:

Condition F.2- Restore one block valve to OPERABLE status in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or in accordance with the Risk Informed Completion Time Program.

The CAM Tool will initially calculate a RICT assuming two block valves inoperable if they are PRA non-functional. This will yield a worst-case RICT (assuming for this illustration that no other equipment is OOS), i.e., the cumulative ICDP/ILERP is being accumulated at a faster rate (steep slope). If subsequently one block valve is declared operable prior to exceeding the ICP/ILERP limit of 1E-5/1 E-6, the RICT will be recalculated assuming one block valve is returned to service. This will yield a longer overall RICT than the original RICT calculated for both block valves being inoperable and non-functional (assuming for this illustration that no other equipment is OOS), i.e., the cumulative ICDP/ILERP is being accumulated at a lesser rate (less steep slope).

The TS LCO Condition applicable for this case is as follows:

Condition F.3- Restore remaining block valve to OPERABLE status in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or in accordance with the Risk Informed Completion Time Program.

For these types of configurations, the RICT time zero does not reset, and a RICT is not exited and reentered in the CAM Tool until all overlapping OOS equipment (TS and non-TS) has been returned to service.

NRC Probabilistic Assessment Review Question #21. TS LCO 3.7.2 Required Actions D. 1 and E.1 TS LCO 3.7.2 Required Actions 0.1 and E.1 require verification" for Main Steam Isolation Valves (MSIVs) versus "restoration." In such a case, please explain how a RICT would be performed if a MSIV was considered PRA functional and there is no required action to "restore" to operability. Consider, for example, if PRA functionality was related to the MSIV closure E1-26 to NL-14-1016 Response to NRC Questions status (e.g., not fully closed). Without a "restore" required action how is a time length determined for a RICT calculation?

SNC Response to Probabilistic Assessment Review Question #21 TS LCO 3.7.2 Required Actions 0.1 and E.1 will be removed from the VEGP RICT Program.

NRC Probabilistic Assessment Review Question #22. TS LCO 3.7.9 Condition B Success Criteria License Amendments No. 170 and 152 for Vogtle Units 1 and 2, respectively, revise TS 3.7.9 by changing the criteria for nuclear service cooling water (NSCW) tower three-and four-fan operation and provides a 7-day CT for one fan/spray cell being inoperable under certain conditions.

Please describe whether this will have any impact on the PRA success criteria and explain how you plan to meet the minimum performance level of the equipment. If it is determined that the PRA success criteria needs to be updated forTS LCO 3.7.9 Condition B, please explain how you plan to update the PRA success criteria.

Note that, per LAR Table E1.1 (page E1-12), the current PRA success criteria forTS LCO 3. 7.9 Condition B requires: "4 of 4 fans when Sl signal has actuated." (This is discussed further in Table E6.2 on page E6-5.) This seems to indicate that 4 fans are currently required to ensure the minimum performance level of the equipment in the event of a LOCA, per the PRA success criteria.

SNC Response to Probabilistic Assessment Review Question #22 In the September 13, 2012 LAR, Enclosure 1, Table E1.1, "Revised TS LCO Conditions to Corresponding PRA Functions," discusses the PRA success criteria for the TS LCO 3.7.9, "Ultimate Heat Sink." The "Disposition" column of Table E1.1 states the following:

"The PRA success criteria do not credit low wet bulb temperatures to reduce the number of required fans, but are based on scenario specific heat removal requirements ... "

As a result, the addition of the 7-day CT for one fan/spray cell being inoperable under certain conditions will not impact the currently documented PRA success criteria in Table E1.1. The calculated RICT for either Condition B or Condition C will be identical assuming the plant configurations are also identical. to this letter, "Markup of TS LCO 3.7.9, 'Ultimate Heat Sink,"' reflects License Amendments No. 170 and 152 for VEGP Units 1 and 2, respectively, and shows where "In accordance with the Risk Informed Completion Time Program" applies. A change to LAR Table E1.1 will be required to agree with the current TS LCO 3.7.9, "Ultimate Heat Sink," and reflect the revision documented in Enclosure 3.

E1-27

Vogtle Electric Generating Plant Response to Request for Additional information on Plant Vogtle License Amendment Request to Revise Technical Specifications to Implement NEI 06-09, Revision 0, "Risk Informed Technical Specifications Initiative 4b, Risk Managed Technicai_Specifications (RMTS) Guidelines" Enclosure2 Comparison of Regulatory Guide (RG) 1.200 Revisions 1 and 2 Internal Events PRA Requirements to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 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, as amended by RG 1.200, Revision 1. RG 1.200, Revision 2 was issued in March 2009. So it would be prudent to review the VEGP PRAto the guidance of RG 1.200, Revision 2.

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

NEI 05-04, Revision 3, "Process for Performing Internal Events PRA Peer Reviews Using the ASMEIANS PRA Standard," provides guidance on performing a gap assessment between Addendum B of the ASME!ANS PRA Standard and RG 1.200, Revision 1, and RG 1.200, Revision 2. Section 3.3 of NEI 05-04 identifies those changes to SRs that would require are-evaluation of the PRA against the PRA Standard requirements. These SRs are included in Table E2.3 below. Note that differences considered typographical, editorial, or providing additional descriptions of the SRs were not considered technically significant and were excluded.

E2-1 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The revised wording in The VEGP internal the 2009 PRA Standard events PRA provides CC-1/11/111: CC-1/11/111: provides additional mean values of HEPs PROVIDE an assessment of the CHARACTERIZE the uncertainty in description for the SR. with associated uncertainty in the HEPs consistent the estimates of the HEPs consistent uncertainty parameters.

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

E2-2 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The revised wording in The VEGP internal the 2009 PRA Standard events PRA uses the CC-I: CC-I: provides additional HRA Calculator which USE an approach that takes the USE an approach that takes the description for the SR. takes into account plant-following into account: following into account: specific shaping factors related to the cues for (a) the complexity of the response (a) the complexity of detection, event detection, and diagnosis, decision-making and diagnosis, decision-The ASEP Approach is an executing the required response making.

acceptable approach. ,,

The ASEP Approach [2-6] is an The revised SR does not acceptable approach. require a revised or additional scope of work CC-11/111: CC-11/111: or change the conclusion When estimating HEPs EVALUATE When estimating HEPs EVALUATE of the peer review for this the impact of the following plant- the impact of the following plant- SR.

specific and scenario-specific specific and scenario-specific performance shaping factors: performance shaping factors:

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

E2-3 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The additional SR was added as part of the RG Revision 1 was CC-1/11/111: ee-1/11/111: 1.200 Revision 1 considered in the VEGP For each sse for which repair is For each SSe for which repair is to clarifications to the 2007 peer review. The VEGP to be modeled, ESTIMATE, based be modeled, ESTIMATE, based on standard and was internal events PRA does on the data collected in DA-e14, the data collected in DA-e15, the subsequently not include any sses for the probability of failure to repair probability of failure to repair the renumbered by RG 1.200 which repair is modeled.

the sse in time to prevent core sse in time to prevent core damage Revision 2.

damage as a function of the as a function of the accident accident sequence in which the sequence in which the sse failure sse failure QU-A2a: The LERF requirement Section 10.3.2 of the was added by RG 1.200 internal events PRA CC-1/11/111: ee-1/11/111: Revision 2. calculation presents PROVIDE estimates of the individual PROVIDE estimates of the individual estimates for individual sequences in a manner consistent sequences in a manner consistent The updated SR explicitly LERF sequence cutsets.

with the estimation of total eDF ... with the estimation of total eDF (and requires consideration of LERF.

E2-4 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The phrase, "from internal The peer review based events," was deleted from on the 2007 version of CC-I: CC-I: the 2009 version of the the PRA Standard ESTIMATE the point estimate CDF ESTIMATE the point estimate CDF PRA Standard. The addressed these LERF from internal events. (and LEAF). LERF requirement was requirements. Section added by RG 1.200 10.3.2 of the internal CC-11: Revision 2. events PRA calculation ESTIMATE the mean CDF from CC-11: presents the mean CDF internal events, accounting for the ESTIMATE the mean CDF (and The SR explicitly requires and LERF results.

"state-of-knowledge" correlation LEAF), accounting for the state-of- consideration of LERF.

between event probabilities [Note knowledge correlation between However, per the note in (1 )]. event probabilities [Note (1 )]. 2007 SR LE-E4 and LE-F3, LERF was addressed CC-III: in applicable CALCULATE the mean CDF from CC-III: requirements of Table internal events by propagating the CALCULATE the mean CDF (and 4.5.8, which includes all uncertainty distributions, ensuring LEAF) by propagating the QU SRs. Thus, the peer that the "state-of-knowledge" uncertainty distributions, ensuring review using the 2007 correlation between event that the state-of-knowledge version of the PRA probabilities is taken into account. correlation between event Standard addressed probabilities is taken into account. these LERF irements.

E2-5 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 QU-B5: QU-B5: The revised wording in The VEGP internal the 2009 PRA Standard events PRA logic loops CC-1/11/111: CC-1/11/111: provides less ambiguous were broken according to Fault tree linking and some other Fault tree linking and some other wording for the SR. the guidance provided modeling approaches may result in modeling approaches may result in without the introduction of circular logic that must be broken circular logic that must be broken unnecessary before the model is solved. BREAK before the model is solved. BREAK conservatisms or non-the circular logic appropriately. the circular logic appropriately. conservatisms.

Guidance for breaking logic loops is Guidance for breaking logic loops is provided in NUREG/CR-2728 [Note provided in NUREG/CR-2728 [2-13]. The revised SR does not (1 )]. When resolving circular logic, When resolving circular logic, DO require a revised or AVOID introducing unnecessary NOT introduce unnecessary additional scope of work conservatisms or non- conservatisms or non- or change the conclusion conservatisms. conservatisms. of the peer review for this SR.

E2-6 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The LEAF requirement The peer review based was added by RG 1.200 on the 2007 version of CC-1/11/111: CC-1/11/111: Revision 2. the PRA Standard ACCOUNT for system successes in ACCOUNT for system successes in addressed these LEAF addition to system failures in the addition to system failures in the The SA explicitly requires requirements. The Level evaluation of accident sequences to evaluation of accident sequences to consideration of LEAF. 2 PRA event trees the extent needed for realistic the extent needed for realistic However, per the note in presented in Section 9.2 estimation of CDF. This accounting estimation of CDF or LERF. This 2007 SR LE-E4 and LE- of the internal events may be accomplished by using accounting may be accomplished by F3, LEAF was addressed PRA calculation explicitly numerical quantification of success using numerical quantification of in applicable account for system probability, complementary logic, or success probability, complementary requirements of Table successes.

a delete term approximation and logic, or a delete term approximation 4.5.8, which includes all includes the treatment of transfers and includes the treatment of QU SAs. Thus, the peer among event trees where the transfers among event trees where review using the 2007 "successes" may not be transferred the "successes" may not be version of the PRA between event trees. transferred between event trees. Standard addressed these LEAF uirements.

E2-7 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 QU-E3: QU-E3: The LERF requirement The peer review based was added by RG 1.200 on the 2007 version of CC-I: CC-I: Revision 2. the PRA Standard ESTIMATE the uncertainty interval ESTIMATE the uncertainty interval addressed these LERF of the COF results. Provide a basis of the COF (and LERF) results. The SR explicitly requires requirements. Section for the estimate consistent with the Provide a basis for the estimate consideration of LERF. 10.4 of the internal events characterization parameter consistent with the characterization However, per the Note in PRA calculation presents uncertainties (OA-03, HR-06, HR- parameter uncertainties (OA-03, 2007 SR LE-E4 and LE- the uncertainty intervals G8, IE-C15). HR-06, HR-G8, IE-C15). F3, LERF was addressed for both COF and LERF, in applicable with consideration of the CC-11: CC-11: requirements of Table state-of-knowledge ESTIMATE the uncertainty interval ESTIMATE the uncertainty interval 4.5.8, which includes all correlation.

of the COF results. ESTIMATE the of the COF (and LERF) results. QU SRs. Thus, the peer uncertainty intervals associated with ESTIMATE the uncertainty intervals review using the 2007 parameter uncertainties (OA-03, associated with parameter version of the PRA HR-06, HR-G8, IE-C15), taking into uncertainties (OA-03, HR-06, HR- Standard addressed account the state-of-knowledge G8, IE-C15), taking into account the these LERF correlation. state-of-knowledge correlation. requirements.

CC-III: CC-III:

PROPAGATE parameter PROPAGATE parameter uncertainties (OA-03, HR-06, HR- uncertainties (OA-03, HR-06, HR-IE-C1 .... no IE-C1 ch E2-8 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 QU-E4: QU-E4: Separate requirements No action, CC-II met for for CC-I, II, and Ill were 2007 version of the PRA CC-I: CC-11111111: collapsed into a single Standard.

PROVIDE an assessment of the For each source of model requirement for CC-1/11/111 impact of the model uncertainties uncertainty and related assumption in the 2009 version of the and assumptions on the results of identified in QU-E1 and QU-E2, PRA Standard. The the PRA. respectively, IDENTIFY how the reference to Note 1 was PRA model is affected {e.g., deleted by RG 1.200 CC-11: introduction of a new basic event, Revision 2.

EVALUATE the sensitivity of the changes to basic event probabilities, results to model uncertainties and change in success criterion, The updated SR assigns key assumptions using sensitivity introduction of a new initiating the same requirement to analyses [Note {1 )]. event). all three CCs. Meeting CC-II in the 2007 version CC-III: of the PRA Standard EVALUATE the sensitivity of the assures that the new SR results to uncertain model boundary is met.

conditions and other assumptions using sensitivity analyses except where such sources of uncertainty have been adequately treated in the quantitative uncertainty analysis E2-9 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IF-F1: IFPP-81: The 2009 PRA Standard The VEG P internal divides the SR for flooding report CC-1/11/111: CC-1/11/111: documentation that adequately documents DOCUMENT the internal flooding DOCUMENT the internal flood plant facilitates applications, each of the elements of analysis in a manner that facilitates partitioning in a manner that upgrades, and review into the internal flooding PRA applications, upgrades, and facilitates PRA applications, separate SRs for each of analysis.

peer review. upgrades, and peer review. the internal flooding elements (plant The revised SR does not IFS0-81: partitioning, flood require a revised or sources, flood scenarios, additional scope of work CC-1/11/111: initiating events, and or change the conclusion DOCUMENT the internal flood quantification). of the peer review for this sources in a manner that facilitates SR.

PRA applications, upgrades, and peer review.

IFSN-81:

CC-1/11/111:

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

IFEV-81:

CC-1/11/111:

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

E2-10 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 CC-1/11/111: CC-1/11/111:

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

IF-F2: The requirement to Section 5 and Appendix A document walkdowns the internal flooding CC-1/11/111: CC-1/11/111: performed in support of PRA document the DOCUMENT the process used to DOCUMENT the process used to plant partitioning was

• walkdowns performed to identify ... flood areas, . . . For identify flood areas. For example, added to the 2009

• validate information example, this documentation this documentation typically includes version of the PRA related to flood areas, typically includes Standard. flood sources, SSCs, (a) flood areas used in the analysis mitigation and other flood (b) flood areas used in the analysis and the reason for eliminating The updated SR cites related features in the and the reason for eliminating areas from further analysis examples of acceptable flood areas.

areas from further analysis documentation of the (b) any walkdowns performed in process to identify flood support of the plant partitioning sources.

Since documentation of walkdowns was not in the 2007 version of the PRA Standard, it was not reviewed as part of the peer review conducted using that version of the PRA Standard.

E2-11 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IF-F2: IFS0-82: The requirement to The internal flooding PRA document walkdowns documents the CC-1/11/11 I: CC-1/11/11 I: performed in support of walkdowns performed to DOCUMENT the process used to DOCUMENT the process used to the identification or validate information identify applicable flood sources. For identify applicable flood sources. For screening of flood related to flood areas, example, this documentation example, this documentation sources was added to flood sources, SSCs, typically includes typically includes 2009 version of the PRA mitigation and other flood Standard. related features in the (a) flood sources identified in the (a) flood sources identified in the flood areas.

analysis, rules used to screen out analysis, rules used to screen out The updated SR cites these sources, and the resulting these sources, and the resulting examples of acceptable list of sources to be further list of sources to be further documentation of the examined examined process to identify flood sources.

(f) screening criteria used in the (b)screening criteria used in the analysis analysis Since documentation of walkdowns was not in the (j) calculations or other analyses (c) calculations or other analyses 2007 version of the PRA used to support or refine the used to support or refine the Standard, it was not flooding evaluation flooding evaluation reviewed as part of the peer review conducted (d) any walkdowns performed in using that version of the support of the identification or PRA Standard.

sc of flood sources E2-12 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IF-F2: IFSN-82: The requirement to The internal flooding PRA document walkdowns documents the CC-1/11/111: CC-1/11/111: performed in support of walkdowns performed to DOCUMENT the process used to DOCUMENT the process used to the identification or validate information identify applicable flood scenarios. identify applicable flood scenarios. screening of flood related to flood areas, For example, this documentation For example, this documentation scenarios was added to flood sources, SSCs, typically includes typically includes 2009 version of the PRA mitigation and other flood (a) propagation pathways ... Standard. related features in the (c) propagation pathways ... flood areas.

(b) accident mitigating features and The updated SR cites (d) accident mitigating features and barriers credited ... examples of acceptable barriers credited ... documentation of the (c) assumptions or calculations used process to identify flood (e) assumptions or calculations used in the determination of ... flood- scenarios.

in the determination of ... flood- induced effects on equipment induced effects on equipment operability Since documentation of operability walkdowns was not in the (d) screening criteria used in the 2007 version of the PRA (f) screening criteria used in the analysis Standard, it was not analysis reviewed as part of the (e) flooding scenarios considered, peer review conducted (g) flooding scenarios considered, screened, and retained using that version of the screened, and retained PRA Standard.

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

modified ...

(g) calculations or other analyses (j) calculations or other analyses used to support or refine the used to support or refine the flooding evaluation flooding evaluation (h) any walkdowns performed in support of the identification or Ct~*~-nll"'ln of flood ct~l:lns:trll'tct to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IF-F2: IFEV-82: The 2009 PRA Standard The VEGP internal divides the SR for typical flooding report CC-1/11/111: CC-1/11/111: items included in the adequately documents DOCUMENT the process used ... DOCUMENT the process used to internal flooding analysis each of the elements of internal flood model development. identify applicable flood-induced process documentation the internal flooding For example, this documentation initiating events. For example, this into separate SRs for analysis.

typically includes documentation typically includes each of the internal flooding elements (plant The revised SR does not (f) screening criteria used in the (a) flood frequencies, component partitioning, flood require a revised or analysis unreliabilities/unavailabilities, and sources, flood scenarios, additional scope of work HEPs used in the analysis (i.e., initiating events, and or change the conclusion (i) flood frequencies, component the data values unique to the quantification). While of the peer review for this unreliabilities/unavailabilities, and flooding analysis) most elements (PP, SO, SR.

HEPs used in the analysis (i.e., SN, QU) included the data values unique to the (b) calculations or other analyses additional requirements to flooding analysis) used to support or refine the document walkdowns, the flooding evaluation EV element included no (j) calculations or other analyses change in requirements.

used to support or refine the (c) screening criteria used in the flooding evaluation analysis E2-14 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The requirement to The internal flooding PRA document walkdowns documents the CC-1/11/111: CC-1/11/111: performed in support of walkdowns performed to DOCUMENT the process used to DOCUMENT the process used to internal flood accident validate information define the applicable internal flood define the applicable internal flood sequence quantification related to flood areas, accident sequences and their accident sequences and their was added in 2009 flood sources, SSCs, associated quantification. For associated quantification. For version of the PRA mitigation and other flood example, this documentation example, this documentation Standard. related features in the typically includes: typically includes: flood areas that are The updated SR cites considered in flood (j) calculations or other analyses (a) calculations or other analyses examples of acceptable sequence definition.

used to support or refine the used to support or refine the documentation of the flooding evaluation flooding evaluation process to identify flood related features (f) screening criteria used in the (b) screening criteria used in the considered in flood analysis analysis sequence quantification.

(i) flooding scenarios considered, (c) flooding scenarios considered, Since documentation of screened, and retained screened, and retained walkdowns was not in the 2007 version of the PRA (k) results of the internal flood (d) results of the internal flood Standard, it was not analysis, consistent with the analysis, consistent with the reviewed as part of the quantification requirements quantification requirements peer review conducted provided in HLR-QU-D provided in HLR-QU-D using that version of the PRA Standard.

(e) any walkdowns performed in support of internal flood accident sequence ntification E2-15 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IF-F3: IFPP-83: The 2009 PRA Standard The VEGP internal divides the SR for flooding report did not CC-1/11/111: CC-1/11/111: documentation of include documentation of DOCUMENT the assumptions and DOCUMENT sources of model assumptions and sources sources of model sources of uncertainty associated uncertainty and related assumptions of uncertainty into uncertainty for the with the internal flooding analysis. (as identified in QU-E1 and QU-E2) separate SRs for each of internal flooding analysis.

associated with the internal flood the internal flooding plant partitioning. elements (plant However, as part of this partitioning, flood License Amendment IFS0-83: sources, flood scenarios, Request, identification initiating events, and and documentation of the CC-1/11/111: quantification). sources of model DOCUMENT sources of model uncertainty in all uncertainty and related assumptions elements of the VEGP (as identified in QU-E1 and QU-E2) internal events Level 1 associated with the internal flood and Level2 PRA sources. (including internal flooding) was performed.

IFSN-83:

CC-1/11/111:

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

E2-16 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IF-F3: IFEV-83: The 2009 PRA Standard The VEGP internal divides the SR for flooding report did not CC-1/11/111: CC-1/11/111: documentation of include documentation of DOCUMENT the assumptions and Document sources of model assumptions and sources sources of model sources of uncertainty associated uncertainty and related of uncertainty into uncertainty for the with the internal flooding analysis. assumptions (as identified in QU- separate SRs for each of internal flooding analysis.

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

and E2-17 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The clarifications to the The VEGP internal 2007 PRA Standard flooding report met the CC-I: CC-I: provided in RG 1.200, combined CC-I/II be For the SSCs identified in IF-C2c, For the SSCs identified in IFSN-A5, Revision 1 included the describing the

... (no change) ... (no change) separation of CC-I and C- mechanisms considered,

11. The revised wording in including submergence CC-II: CC-II: the 2009 PRA Standard and spray; thereby INCLUDE failure by submergence For the SSCs identified in IFSN- provides additional meeting CC-11. However, and spray in the identification A5, IDENTIFY the susceptibility of description for the SR. there was not sufficient process. each sse in a flood area to flood- consideration of humidity, ASSESS qualitatively the impact of induced failure mechanisms. condensation, flood-induced mechanisms that are INCLUDE failure by submergence temperature, etc. to merit not formally addressed (e.g., using and spray in the identification CC-III.

the mechanisms listed under process.

Capability Category Ill of this ASSESS qualitatively the impact of requirement), by using conservative flood-induced mechanisms that are assumptions. not formally addressed (e.g., using the mechanisms listed under CC-III: Capability Category Ill of this For the SSCs identified in IF-C2c, requirement), by using conservative

... (no change) assumptions.

CC-III:

For the SSCs identified in IFSN-A5,

... (no change)

E2-18 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IE-D3 (also AS-C3. SC-C3. SY-C3. IE-D3 (also AS-C3. SC-C3, SY-C3. The NRC recommended The Recovery Analysis HR-13. DA-E3. LE-G4l: HR-13. DA-E3. LE-G4): wording changes on "key and Uncertainty Analysis assumptions and sources for the VEGP Level 1 and CC-1/11/11 I: CC-1/111111: of uncertainty were Level2 Model (Chapter DOCUMENT the key assumptions DOCUMENT the sources of model reflected in the 2009 PRA 10) included and key sources uncertainty uncertainty and related Standard. documentation on the associated with the initiating event assumptions (as identified in QU- sources of modeling analysis (also accident sequence E1 and QU-E2 or LE-F3) Since the wording uncertainty.

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

E2-19 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The NRC recommended The Recovery Analysis wording changes on "key and Uncertainty Analysis cc 1/11/11: cc 1/11/11: assumptions and sources for the VEGP Level 1 and IDENTIFY key sources of model IDENTIFY sources of model of uncertainty were Level2 Model (Chapter uncertainty. uncertainty. reflected in the 2009 PRA 10) included Standard. documentation on the sources of modeling Since the wording uncertainty.

changes were not in the 2007 version of the PRA In addition, as part of this standard, the wording License Amendment changes were not directly Request, an assessment considered during the and characterization of peer review conducted the sources of model using the 2007 PRA uncertainty in the VEGP Standard. internal events Level1 and Level 2 PRA was performed according to the guidance in NUREG-1855 and EPRI TR-1016737.

E2-20 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 QU-E2: QU-E2: The NRC recommended The Recovery Analysis wording changes on "key and Uncertainty Analysis cc 1/11/11: cc 1/11/11: assumptions and sources for the VEGP Level1 and IDENTIFY key assumptions made in IDENTIFY assumptions made in the of uncertainty were Level 2 Model (Chapter the development of the PRA model. development of the PRA model. reflected in the 2009 PRA 10) included Standard. documentation on the sources of modeling Since the wording uncertainty.

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

E2-21 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 QU-F4: QU-F4: The NRC recommended The Recovery Analysis wording changes on "key and Uncertainty Analysis cc 1/11/11: cc 1/11/11: assumptions and sources for the VEGP Level1 and DOCUMENT key assumptions and DOCUMENT the characterization of uncertainty were Level 2 Model (Chapter key sources of uncertainty, such as: of the sources of model reflected in the 2009 PRA 10) included possible optimistic or uncertainty and related assumptions Standard. documentation on the conservative success criteria, (as identified in QU-E4). sources of modeling suitability of the reliability data, Since the wording uncertainty.

possible modeling uncertainties changes were not in the (modeling limitations due to the 2007 version of the PRA In addition, as part of this method selected), degree of standard, the wording License Amendment completeness in the selection of changes were not directly Request, an assessment initiating events, possible spatial considered during the and characterization of dependencies, etc. peer review conducted the sources of model using the 2007 PRA uncertainty in the VEGP Standard. internal events Level 1 and Level 2 PRA was.

performed according to the guidance in NUREG-1855 and EPRI TR-1016737.

E2-22 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 LE-06: LE-07: The SR wording and The peer review based numbering changes were on the 2007 version of Cl: Cl: incorporated in the 2009 the PRA Standard PERFORM containment isolation PERFORM containment isolation version of the PRA addressed these LERF analysis in a conservative analysis in a conservative Standard. requirements and is not manner ... (unchanged) manner ... (unchanged) affected by the wording The difference is change.

C II: C II: considered editorial only.

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

CC Ill: CC Ill:

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

E2-23 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IE-A4a IE-A6 The wording change is a This SR was met.. ....

reduction in CCII and CC Ill: CCII and CC Ill: requirements, and does When performing the systematic When performing the systematic not need to be re-evaluation required in IE-A4, evaluation required in IE-A5, assessed unless the SR INCLUDE initiating events resulting INCLUDE initiating events resulting was not met in the peer from multiple failures, if the from multiple failures, if the review.

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

E2-24 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IE-B3: IE-B3: The revised wording in The VEGP internal the 2009 PRA Standard events PRA grouping of Cl: Cl: provides less ambiguous initiating events was GROUP initiating events only when GROUP initiating events only when wording for the SR. found to be appropriate.

the following is true: the following can be ensured:

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

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

GROUP initiating events only when GROUP initiating events only when the following is true: the following can be ensured:

(items (a) and (b) unchanged) (items (a) and (b) unchanged)

E2-25 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 AS-A10 AS-A10 The clarifications to the The VEGP internal 2007 PRA Standard were events PRA accident CCII: CCII: provided in RG 1.200, sequence construction In constructing the accident In constructing the accident Revision 1 and were was assigned CC Ill by sequence models, INCLUDE, for sequence models, INCLUDE, for subsequently also the peer review team for each modeled initiating event, each modeled initiating event, included in the 2009 PRA this SR.

sufficient detail that signifiGant sufficient detail that differences in Standard.

differences in requirements on requirements on systems and The revised SR does not systems and required operator required operator interactions (e.g., require a revised or responses interactions (e.g., systems initiations or valve additional scope of work systems initiations or valve alignment) are captured. Where or change the conclusion alignments) are captured. Where diverse systems and/ or operator of the peer review for this diverse systems and/ or operator actions provide a similar function, if SR.

actions provide a similar function, if choosing one over another changes choosing one over another changes the requirements for operator the requirements for operator intervention or the need for other intervention or the need for other systems, MODEL each separately.

MODEL each HR-E2: The clarifications to the The VEGP human 2007 PRA Standard were reliability analysis cc 1/11/111: cc 1/11/111: provided in RG 1.200, includes both the IDENTIFY IDENTIFY those actions Revision 1 and were cognitive (diagnosis) and (a) those actions required to initiate (a) required to initiate ... (not subsequently also execution portions of

... (not changed) changed) included in the 2009 PRA control room actions.

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

E2-26 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 DA-C14 DA-C15 The clarifications to the The VEGP internal 2007 PRA Standard were events PRA does not c 1/11/111: c 1/11/111: provided in RG 1.200, include any SSCs for For each SSC for which repair is to For each SSC for which repair is to Revision *1 and were which repair is modeled.

be modeled (see SY-A22), be modeled (see SY-A22), subsequently also IDENTIFY instances of plant- IDENTIFY instances of plant-specific included in the The SR is judged to be specific experience and, when experience and, when that is renumbered SR in RG not applicable that is insufficient to estimate insufficient to estimate failure to 1.200 Revision 2.

failure to repair consistent with repair consistent with DA-09 ... (not OA-08 ... chan DA-D1: DA-D1: The clarifications to the Parameter estimates for 2007 PRA Standard were most failure data in the CCII and Ill: CC II and Ill: provided in RG 1.200, VEGP internal events

... USE a Bayes update process 9F ... USE a Bayes update process . Revision 1 and were PRA were based on a equivalent statistical process that CHOOSE. .. subsequently also Bayesian update process.

assigns that assigns appropriate included in the 2009 PRA A few estimates were weight to tho statistical significance Standard to remove based on published of tho generic and plant specific ambiguity. generic data only.

evidence and provides an appropriate characterization of tho The revised SR does not uncertainty. change the conclusion of CHOOSE. .. the peer review for this SR.

E2-27 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IF-81: IFSO-A1: The requirement to Potential flood sources include the fire protection identified in Section 5 of CC-1/11/111: CC-1/11/111: system in Item (a) as a the internal flooding PRA For each flood area, IDENTIFY the For each flood area, IDENTIFY the potential flooding source reviewed as part of 2009 potential sources of flooding [Note potential sources of flooding [Note was added by RG 1.200 peer review against 2007 (1 )]. INCLUDE: (1 )]. INCLUDE Revision 1. This version of the PRA requirement was standard amended by RG (a) equipment (e.g., piping, valves, (a) equipment (e.g., piping, valves, addressed in the peer 1.200, Revision 1 include pumps) located in the area that pumps) located in the area that review, which used the RCS-connected systems are connected to fluid systems are connected to fluid systems 2007 version of the PRA - chemical and volume (e.g., circulating water system, (e.g., circulating water system, Standard amended by control system (CVCS),

service water system, fire service water system, fire RG 1.200 Revision 1. containment spray (CS),

protection system, component protection system, component residual heat removal cooling water system, feedwater cooling water system, feedwater The requirement to (RHR), reactor coolant system, condensate and steam system, condensate and steam include the reactor system drain tank systems) systems, and reactor coolant coolant system in Item (a) (RCSDT), safety injection system) as a potential flooding (SI), and reactor water source was added to the makeup system (RMWS).

2009 version of the PRA As outlined in the Plant Standard. Thus, it was Vogtle Internal Flooding not reviewed as part of notebook, the the peer review Containment Building conducted using that (and RCS components version of the PRA therein) is not included in Standard. the scope of the internal flooding analysis.

E2-28 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The clarifications to the Flood calculations for the 2007 PRA Standard were VEGP internal flooding cc 1/11/111: cc 1/11/111: provided in RG 1.200, analysis consider a range For each source and its identified For each source and its identified Revision 1 and were of flow rates.

failure mechanism, IDENTIFY the failure mechanism, IDENTIFY the subsequently also characteristic of release and the characteristic of release and the included in the re- The revised SR does not capacity of the source. INCLUDE: capacity of the source. INCLUDE: numbered SR in the 2009 change the conclusion of PRA Standard. the peer review for this (b) range of flow rates of ¥later (b) range of flow rates SR.

IF-C3b: IFSN-AS The clarifications to the Flood calculations for the 2007 PRA Standard were VEGP internal flooding C II: C II: provided in RG 1.200, analysis consider drain IDENTIFY inter-area propagation IDENTIFY inter-area propagation Revision 1 and were lines and other pathways through the normal flow path from through the normal flow path from subsequently removed not directly related to one area to another via drain lines; one area to another via drain lines; from CC II for the re- room opening.

and areas connected via back flow and areas connected via backflow numbered SR in the 2009 through drain lines involving failed through drain lines involving failed PRA Standard. The revised SR requires check valves, pipe and cable check valves, pipe and cable a reduced scope of word penetrations (including cable trays), penetrations (including cable trays), and does not change the doors, stairwells, hatchways, and doors, stairwells, hatchways, and conclusion of the peer HVAC ducts. INCLUDE potential for HVAC ducts. INCLUDE potential for review for this SR.

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

potential for barrier unavailability, maintenance activities.

E2-29

Enclosure 2 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 IF-D3: IFEV-A2 The revised wording in The VEGP internal the 2009 PRA Standard flooding analysis provides less ambiguous grouping (and some wording for the SR. consequential AVOID subsuming scenarios into a DO NOT SUBSUME scenarios into subsuming) was found to group unless a group unless be appropriate.

The revised SR does not require a revised or additional scope of work or change the conclusion of the peer review for this SR.

IF-E6a: IFQU-A8 RG 1.200 Revision 2 The VEGP internal removed the phrase "due flooding quantification cc 1/11/111: cc 1/11/111: to causes independent of was performed using the INCLUDE, in the quantification, the INCLUDE, in the quantification, the the flooding" which did same basic events for combined effects of failures caused combined effects of failures caused not change the intent of other non-flooding by flooding and those coincident with by flooding and those coincident with the requirement. failures as used in the the flooding due to causes the flooding due to independent internal events model.

independent causes of the flooding causes including equipment including equipment failures, failures, unavailability due to The revised SR does not unavailability due to maintenance, maintenance, common-cause require a revised or common-cause failures, and other failures, and other credible causes. additional scope of work credible causes. or change the conclusion of the peer review for this SR.

E2-30 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 The sentences were NUREG/CR-6928 is used clarifications provided in as the source for generic CC-1/11/111: CC-1/11/111: RG 1.200 Revision 1 and data priors in Revision 4 Revision 2, respectively. of the VEGP internal An example of an acceptable An example of an acceptable events PRA.

generic data sources is NUREG/CR- generic data sources is NUREG/CR- The updated SR cites a 5750 [Note 1]. 6928 [Note 1]. more recent example of Per NEI 05-0, Revision 3, an acceptable generic this SR does not require data source. re-evaluation.

DA-C1: DA-C1: Reference NUR 715 NU R-6928 is used was added by RG 1.200 as the source for generic CC-1/11/111: CC-1/11/111: Revision 1; References data priors in Revision 4 NUREG-1715 and of the VEGP internal Examples of parameter estimates Examples of parameter estimates NUREG/CR-6928 were events PRA.

and associated sources include: and associated sources include included in the 2009 (a) component failure rates and (a) component failure rates and version of the PRA Per NEI 05-0, Revision 3, probabilities: NUREG/CR-4639 probabilities: NUREG/CR-4639 Standard. this SR does not require

[Note (1 )], NUREG/CR-4550 [2-7], NUREG/CR-4550 [2-3], re-evaluation.

[Note (2)], NUREG-1715 [Note 7] NUREG-1715 [2-21], The updated SR cites NUREG/CR-6928 [2-20] more recent examples of acceptable generic data sources.

E2-31 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 Separate requirements No action, CC-II met for for CC-I, II, and Ill were 2007 version of the PRA CC-I: CC-I/lVIII: collapsed into a single Standard.

PROVIDE a qualitative assessment IDENTIFY and CHARACTERIZE the requirement for CC-1/11/111 of the key sources of uncertainty. LERF sources of model uncertainty in the 2009 version of the Examples: and related assumptions, in a PRA Standard. The (a) Identify bounding assumptions. manner consistent with the difference was not (b) Identify conservative treatment of applicable requirements of Tables 2- included in NEI 05-04, phenomena. 2.7-2(d) and 2-2.7-2(e). Revision 3.

CC-11:

PROVIDE uncertainty analysis that The updated SR assigns identifies the key sources of the same requirement to uncertainty and includes sensitivity all three CCs. Meeting studies for the significant CC-II in the 2007 version contributors to LERF. of the PRA Standard assures that the new SR CC-III: is met.

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

E2-32 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 RG 1.200 Revision 2 No action, This SR was removed the phrase met for 2007 version of C-1/11/111 : C-1/11/111 : "including uncertainty and the PRA Standard.

sensitivity analyses, as DOCUMENT the process used to DOCUMENT the process used to appropriate for the level identify plant damage states and identify plant damage states and of detail of the analysis,"

accident progression contributors, accident progression contributors, which is a reduction in the define accident progression define accident progression requirement.

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

uncertainty and sensitivity analyses, as appropriate for the level of detail of the ~n:::uu~:t*R E2-33 to NL-14-1016 Comparison RG 1.200, Revisions 1 and 2 SY-815: The sentences were As noted in Table 9.2-1 of provided in the 2007 and the internal events PRA CC-1/11/111: CC-IIII/III: 2009 versions of the PRA calculation, failure of the Standard, respectively. containment boundary (h) harsh environments induced by (h) harsh environments induced by The difference was not due to venting is not containment venting, or failure containment venting, failure of included in NEI 05-04, applicable to the VEGP that may occur prior to the onset the containment venting ducts, Revision 3. large, dry, of core damage. or failure of the containment subatmospheric boundary that may occur prior to The updated SR explicitly containment.

the onset of core damage requires consideration of containment venting ducts and failure of the containment boundary prior to core damage.

E2-34

Vogtle Electric Generating Plant Response to Request for Additional Information on Plant Vogtle License Amendment Request to Revise Technical Specifications to Implement NEI 06-09, Revision 0, "Risk Informed Technical Specifications Initiative 4b, Risk Managed Technical Specifications (RMTS) Guidelines" Enclosure3 Mark-Up of Technical Specifications (TS)

Limiting Condition of Operation (LCO) 3.7.9

UHS 3.7.9

3. 7 PLANT SYSTEMS 3.7.9 Ultimate Heat Sink (UHS)

LCO 3.7.9 The UHS shall be OPERABLE. The fans/spray cells shall be as specified in Figure 3.7.9-1.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Nuclear A.1 Restore water 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Service Cooling Water temperature(s) and water (NSCW) basins with level(s) to within limits.

water temperature and/or water level not within limits.

B. One NSCW cooling B.1 Restore fan to 7days tower with one required OPERABLE status.

fan/spray cell inoperable I

when operating in four fan/spray cell required region of Figure 3.7.9-1.

c. One NSCW cooling C.1 Restore fan(s)/spray 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> tower with one or more cell(s) to OPERABLE required fans/spray cells inoperable for reasons other than Condition B.

status.

/

v v (continued)

OR In accordance with the Risk Informed Completion Time Program Vogtle Units 1 and 2 3.7.9-1 Amendment No.,:~~~ (Unit 1)

Amendment No. +e2 (Unit 2)