ML12061A245
| ML12061A245 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 03/05/2012 |
| From: | Patrick Boyle Plant Licensing Branch II |
| To: | Ajluni M Southern Nuclear Operating Co |
| Boyle P | |
| References | |
| TAC ME7763 | |
| Download: ML12061A245 (43) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 5, 2012 Mr. M. J. Ajluni Nuclear licensing Director Southern Nuclear Operating Company, Inc.
P.O. Box 1295 Bin - 038 Birmingham, Alabama 35201-1295 SUB..IECT: VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2, AUDIT REPORT FOR THE PROCESS BEING DEVELOPED TO SUPPORT A LICENSE AMENDMENT REQUEST TO IMPLEMENT RISK INFORMED CATEGORIZATION OF SYSTEMS, STRUCTURES, AND COMPONENTS (TAC NO. ME7763)
Dear Mr. Ajluni:
By letter to the U.S. Nuclear Regulatory Commission (NRC) dated December 6, 2010, Southern Nuclear Operating Company, Inc.(SNC), stated that they plan to request a license amendment to implement Title 10 Code of the Code of Federal Regulations (10 CFR) 50.69, "Risk-informed categorization and treatment of systems, structures, and components [SSCs] for nuclear power reactors," for the Vogtle Electric Generating Plant, Units 1 and 2 (VEGP) during 2012.
A principle step in implementing 10 CFR 50.69 is a meeting of the licensee's integrated decision-making panel (lOP) where the final safety-significant categories for components are chosen. During a March 29, 2011, meeting between SNC and the NRC, SNC informed the NRC staff that the lOP would meet in November 2011, and January 2012, and invited the NRC staff to observe one of the meetings. The NRC staff observed the November 29, 2011, lOP meeting. VEGP and NRC staff also met the afternoon of November 28, 2011, to discuss the draft categorization procedures and any other issues related to the categorization process.
Enclosed is an audit report summarizing the NRC staff observations of the categorization process.
M. Ajluni -2 Please feel free to contact me with questions regarding this audit report.
Sincerely, Patrick G. Boyle, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-424 and 50-425
Enclosures:
- 1. Audit Report
- 2. Meeting Attendees
- 3. Meeting Slides cc: Distribution via Listserv
AUDIT BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO CATEGORIZATION PROCESS TO IMPLEMENET 10 CFR 50.69 SOUTHERN NUCLEAR OPERATING COMPANY, INC.
VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-424 AND 50-425
1.0 BACKGROUND
By letter dated December 6, 2010 (Agencywide Documents Access and Management System, (ADAMS) Accession No. ML103420287), Southern Nuclear Operating Company, Inc. (SNC) informed the NRC that it plans a 2012 request to implement Title 10 of the Code of Federal Regulations (10 CFR), Section 50.69 at its Vogtle Electric Generating Plant, Units 1 and 2 (VEGP) . SNC also requested pilot status and a waiver of review fees. By letter dated June 17, 2011 (ADAIVIS Accession No. ML11171A084), the NRC granted the request for pilot status and fee waiver.
On March 29, 2011 , the NRC staff and SNC met to review SNC's planned approach for implementation of 10 CFR 50.69, risk-informed categorization and treatment of structures, systems, and components (SSCs) for nuclear power reactors. SNC informed the NRC staff about its ongoing trial categorization of three VEGP systems to test the efficacy of the categorization process prior to documenting the process in the VEGP 10 CFR 50.69 license amendment request. By letter dated August 17, 2011 (ADAMS Accession No. ML112300122),
SNC provided the draft categorization procedures to the NRC.
At the March 29, 2011, meeting, SNC invited the NRC staff to observe one of the integrated decision-making panel (lOP) meetings. The NRC staff observed the November 29,2011, lOP meeting . VEGP and NRC staff also met the afternoon of November 28, 2011, to discuss the draft categorization procedures and any other issues related to the categorization process.
The NRC has issued or endorsed the following categorization guidelines for use in implementation of 10 CFR 50.69:
- Regulatory Guide (RG) 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance," (For Trial Use) dated May 2006, endorses, with comments and clarifications, Revision 0 of ENCLOSURE 1
-2 NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," dated July 2005 (ADAMS Accession No. ML061090627).
- RG 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1,"
dated October 2010, conditionally accepts ASME Code Case N-660, "Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities,Section XI, Division 1," dated July 23, 2002 (ADAMS Accession No . ML101800536).
- Topical Report WCAP-16308-NP-A, "Pressurized Water Reactor Owners Group 10 CFR 50.69 Pilot Program - Categorization Process - Wolf Creek Generating Station,"
Revision 0, August 2009 (ADAMS Accession No. ML092430186) .
2.0 PURPOSE A principle step in the categorization is a meeting of the licensee's lOP where the final safety significant categories for SSCs are chosen . SNC informed the NRC that the lOP would meet in November 2011 , and January 2012, and invited the NRC staff to observe one of the meetings.
NRC staff traveled to Augusta, GA, and the Vogtle plant, to meet with SNC on November 28, 2011, and observe the November 29, 2011 , lOP panel.
3.0 AUDIT DETAILS The audit was conducted on November 28, 2011, in Augusta , GA, and November 29, 2011, at the Vogtle plant site. The discussions identified a number of issues but resolution of the issues was not attempted during the audit. The NRC staff attending the audit included:
Donnie Harrison, Chief, Probabilistic Risk Assessment Licensing Branch (APLA)
Stephen Dinsmore, Senior Reliability and Risk Analyst (APLA)
Dan O'Neal, Reliability and Risk Analyst (APLA)
November 28, 2011 The audit began in the morning of November 28, 2011, with the NRC staff going through the draft procedures provided and consolidating comments. In the afternoon, the Licensee provided an overview and a description of the process. Attendees at this presentation are included in and the presentation is included in Enclosure 3.
The licensee's categorization process is implemented through multiple procedures. Applying the procedures result in assigning high or low safety significance (HSS and LSS respectively) to each structure, system, and component (SSCs) in a system . Additional procedures provide for post implementation monitoring required by 10 CFR 50.69. The licensee's procedures rely, often word for word, on NEI 00-04 "10 CFR 50.69 SSC Categorization Guideline," July 2005 (ADAMS Accession No. ML052910035) guidance, which is endorsed with comments and clarifications in RG 1.201.
NEI 00-04 provides alternative methods to evaluate the safety-significance of SSCs with respect to fire, seismic and other types of hazards. VEGP's procedures included the alternative methods from NEI 00-04. It was unclear if VEGP would use different options on, for example ,
-3 different systems. The licensee clarified that they were completing their fire probabilistic risk assessment (PRA) and had scheduled a Peer Review of the fire PRA for February 2012. Until their fire PRA was complete, they would use the fire safe shutdown list (SSL) option but after the PRA is completed they intended to use the fire PRA. This plan led to the observation that the categorization process may substantively change over time. The NRC staff opined that the categorization process needed to be approved by the NRC and therefore changing the process (e.g., switching from SSL to fire PRA) might require an additional LAR.
The licensee indicated that its licensing action request (LAR) to implement 10 CFR 50.69 would not necessarily include a request to approve the procedures, but to approve the methods incorporated in the procedures. The precise change that will be requested in the LAR had not yet been determined.
Several of the methodologies in NEI 00-04 that are reflected in the procedures have not been exercised. These included:
- Weighted average of reliability achievement worth (RAW) and Fussell-Vesely (FV) across initiating events
- Selecting an unavailability increase factor and developing monitoring action guidelines based on that factor The following general observations were collected by the NRC staff and communicated to the licensee during the presentation. The NRC staff and the licensee did not engage in discussions about how the observations may, or may not, impact the licensee's LAR.
- The NRC staff indicated that the LAR should specify as precisely as possible what change is being requested, what methods are applied, and what methods may be applied
- Given the numerous options that are available in the guidance documents, the development and review of an LAR that retained all the options could be very complex November 29, 2011 The NRC audit team and most of the other participants from the November 28 meeting traveled to the Vogtle plant to observe the lOP meeting where the NRC audit team was able to view the culmination of the categorization effort and gain good insights regarding both the robustness of the categorization process in general and the lOP decision-making process. The observers did not participate in the lOP deliberations. There were two opportunities for interactions between the lOP and the observers, once after each of the two systems completed the lOP decision making process.
The lOP meeting addressed two systems: the Containment Spray (CS) system and the Radiation Monitoring (RM) system. A system is addressed by walking through each system's "notebook" which is prepared by the licensee staff and contractors for the lOP to review. The notebooks ("10 CFR 50.69 Categorization: Containment Spray System" and "10 CFR 50.69
- 4 Categorization: Radiation Monitoring System") identify system functions, system components, and the results of the various evaluations that are described in the categorization procedures.
They provide a traceable path that ends with a final table of all system components with assigned safety significance.
The lOP had previously reviewed the CS system notebook and this second review was completed by the lOP approving the final notebook (with comments) . The lOP had not previously reviewed the RM system notebook. The lOP identified some actions for the members and/or licensee support personnel to investigate to prepare for the second meeting on the RM system.
The flowing general observations were collected by the NRC staff and communicated to the licensee. The NRC staff and the licensee did not engage in discussions about how the observations may, or may not, impact the licensee's categorization process.
- The categorizations procedure as described in NEI 00-04 are fairly general and do not address all the detailed evaluations and corresponding documentation presented to the lOP. It was not obvious to the NRC audit team that following the procedures would always result in an equivalent level of detailed information developed and provided in the notebooks. Therefore, either more detailed procedures or both the procedures and examples will be needed to unambiguously identify the process as implemented at the facility.
- Not all the methods in the procedures were exercised to develop the notebooks. Some of the different optional methods to categorize the SSCs and all of the post categorization methods such as changing importance measure guidelines and confirming the factors in the risk increase evaluation are accomplished were not exercised. Implementation of these non-exercised methods is expected to be as f.lexible as the methods exercised during the audit. Therefore, details about these items may be necessary in the LAR to unambiguously indentify the process as implemented at the facility.
- System functions can be complicated to define but provide a vital structure to the process. The example system notebooks defined functions in two different sections, an initial section that appeared to identify high level safety-related functions, and, some dozens of pages later, a second section that sub-divided these functions and added additional functions. This separation was confusing and does not simplify the already complex process. Additionally, not all functions seemed to be identified . For example, some CS SSCs were identified on the fire SSL as possible diversions of fire fighting water but there was no corresponding CS function such as prevention of diversion of fire fighting water.
-5
- The evaluations presented to the lOP need to clearly address external hazards risk if a PRA exists, or to make it clear to the lOP that there is not an external hazards PRA and as a result the SSL SSCs are, for example, all automatically HSS.
- The lOP discussed whether the likelihood of pressure boundary rupture (or use of the consequences for a more likely leakage instead of rupture) could be included in the categorization process. The guidance in the passive categorization process clearly identifies the limited circumstances where likelihood of rupture or the consequences of a leak can be included.
Radiation monitoring
- No quantitative PRA results were possible for the RM system and there appeared to be some difficulty in applying the qualitative guidelines. It was often noted that a set of SSCs (e.g., stack monitors) do not provide the "sole means" to support a function (e.g .,
offsite emergency activity) and are therefore LSS. There was, however, no systematic identification of the alternative means. For example, loss of some RM SSCs disabled information relied upon to identify an "adverse containment condition ." Depending on the accident scenario other SSCs, such as temperature and/or pressure indicators, could provide an alternative means but these indicators are not always activated and therefore perhaps not equivalent to the primary means.
- Similar to the lack of information about specific alternative means, the methodology had not explicitly evaluated whether operators could still reliably perform actions given the loss of information that might occur after failures of RM SSCs.
- As discussed explicitly during the lOP deliberation, the categorization evaluation was developed assuming that the defense-in-depth philosophy is applied solely to protecting physical barriers (Le., clad, reactor coolant system, and containment). However, as discussed in RG 1.174 and other documents, the NRC staff considers that the defense in-depth philosophy also applies to the successive layers to prevent accidents from occurring, mitigation of accidents should they occur, and emergency preparedness to minimize public health consequences if releases occur. Although the importance of the RM system to support emergency preparedness and minimize the consequences of release was repeatedly identified in the qualitative analysis, this importance did not seem to be fully reflected in the results.
-6
- The defense-in-depth containment isolation questions were applied to containment penetrations by RM SSCs. However, the impact of postulated RM SSCs failures that would, in turn, lead to the failure of RM signals to isolate non-RM SSC penetrations should also have been considered. The lOP directed that this consideration be added to the results.
4.0 AUDIT
SUMMARY
The audit demonstrated that a comprehensive categorization structure can be developed and implemented. Once the structure is in place and the preparatory work for each system completed it appears that an lOP should be able to disposition a system in one or two meetings.
The collection of SSCs into a system is, to some extent, arbitrary, and therefore individual SSCs may perform functions not generally associated with the system when all design bases functions and functions credited for mitigation and prevention of severe accidents are included.
Identification of all functions is key to a structured categorization process.
The qualitative categorization process, generally implemented by responding to questions related to SSC functions, would benefit from improvements in the available guidance.
This audit did not review VEGP's categorization process against the categorization guidelines issued or endorsed by NRC, but the NRC audit team is familiar with the guidelines and believes that, in general, the VEGP process meets the intent of the guidance. However, the complexity of the process does not appear to be fully captured in the general methodology issued or endorsed by the NRC and therefore simply repeating the general method description from NEI 00-04 in a plant-specific LAR will not be sufficient for the staff to conclude the culmination of the categorization process meets the requirements in 50.69(c). This will be an area of further interaction between the NRC staff and the licensee during the pilot activities.
MEETING PARTICIPANTS Ralph Chackal, Associated Engineering Resources Patrick O'Regan, EPRI Barry Sloane, Erin Engineering Biff Bradley, NEI Stephen Dinsmore, NRC Donnie Harrison, NRC T. Mark Hickox, SNC Ken McElroy, SNC Dan O'Neal, NRC Vish Patel, SNC Erick Sweet, SNC Jesse Thomas, SNC ENCLOSURE 2
Risk-Informed Categorization of SSCs for Southern Nuclear NRC/SNC Meeting November 28,2011 SNC Risk-Informed Engineering ENCLOSURE 3
Purpose o Purpose Communicate SNC's Approach for Categorization of SSCs using NEI 00-04
- Active components
- Passive components Communicate purpose of the lOP meeting on 11/29/11 and deliberation during the lOP meeting o Desired Outcome Develop Common Understanding of:
SNC's Approach for Categorization of SSCs using NEI 00-04
- Purpose of and deliberation during the IDP meeting
Agenda
- o Welcome/Introduction o INPO Nuclear Safety Principle/Target Zero message o Project status o Overview of SNC 50.69 procedures (NMP-ES-065 & 066) o Passive component categorization method (NMP-ES-065-002) o Break o Categorization approach (NMP-ES-065-003) o Pilot Feedback on Application of NEI Document o Discussion of NRC review comments / Pilot Feedback on Application of NEI Document o lOP meeting on 11/29/11
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Vogtle 10 CFR 50.69 Timeline D Overall Project Duration Submit LAR Obtain NRC Approval Begin Implementation Dec 2013 Begin Implementation Jan 2012 Jan 2013 Jan 2011 Jan 2011 Jun 2012 Nov 2013 Project starts Submit LAR NRC Approval
Project Statu s o Procedures developed NE100-04 o lOP established D Trial categorization of 3 systems Containment Spray
- Teleconferences and trial lOP meeting Radiation Monitoring eves D More information in the hand out
Objectives
-D Two meetings - 11/28 and 11/29 11/28: Communication of our methodology 11/29: Observation of lOP
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Overview of 50 .69 Proced ures D Overview of 50.69 procedures at SNC (NMP-ES-065)
NMP-ES-065
'10 CFR 50 ,69 Program (Categorization and Treatment of SSCs) 1.-
Provides ovelview of the 50,69 process and contains ALL definitions 7
NMP-ES-065-003 Integrated Risk of Components Develops Deterministic Risk and combines results with those of NMP-E 5-065-001 and NMP-ES-065-002, Bins each component IIlto RISC-1 through 4 ':4 categories, T11ese results are sent to lOP (NMP-ES-066-002) for review and approval NMP-ES-066-002 50.69 IDP Reviav",
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Review and approve preliminary LSSIHSS designation of AL!..
components For LSS, review the risk information, defense-in-depth, and Active Component Risk NM P-ES-065-002 safety margins Analyze 5 risks vi;) PRA model AND Passive Component Risk qu;)/itative approach Assigns LSS or HSS based on Components not modeled Jre neither risk evaluation of pressure LSS or HSS. boundary risk.
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Passive Component Categorization o Founded on the Consequence Evaluation portion of the EPRI traditional RI-ISI methodology o EPRI traditional RI-ISI methodology extended to address break exclusion region (BER, EPRI-1006936 Rev O-A)
If not in Tech Spec, RI-BER can be implemented via 50.59 o EPRI RI-ISI / RI-BER used extensively BWRs PWRs-B&W, PWRs-CE & PWRs-W o NMP-ES-065-002 incorporates lessons learned from N660 rO, ra & rb, draft N752, WCAP-16308 and ANO-2 SER
Passive Component Categorization Consequence Evaluation D Component failure assumed with a probability of 1.0 spectrum of break sizes evaluated, unless precluded by design small to large highest consequence ranking is used o Impacts Initiati ng event Mitigative trains/systems Containment performance Combination of impacts
Passive Component Categorization Classification D Impacts are ranked as either c High, medium, low or none D Final classification High consequence rank = HSS Medium, low or none consequence rank
- Additional considerations (e.g. defense in depth)
- If false, HSS is assigned
- If true, LSS is assigned
Passive Com ponent Categorization Comparison to SER on WCAP-16308 o there exist conflicts in the wording and requirements in the SER, Chapter 4 and Appendix A of the WCAP and is therefore at times hard to follow o consistent with EPRI TR-112657 Rev B-A, the SNC procedure requires that all safety functions supported by a system be completely evaluated as part of that system's categorization, while the WCAP would allow an "interim' categorization o operator actions, when credited, need to meet the requirements of NRC approved EPRI TR-112657, Rev 8-A (e.g. GS, GF & lOP not allowed) o unless precluded by design, a spectrum of break sizes needs to be evaluated and the one with the highest consequence rank used o SNC procedure currently limits the application to Class 2 and 3 components (i.e. Class 1 is always HSS) o SNC procedure requires that all relevant configurations be assessed as pa rt of the categorization process
Passive Component Categorization
- RCIC - LPCI "A" f-- roo- RHR "A" f-
- HPCI - LPCI "B" r- RHR "B" f-Initiating
--+ Success Event Main
- Feedwater - Core Spray "A" r-Condenser f-
- 2 of 11 SRV - - Core Spray "B" ----J Venting
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High Pressure Makeup
& Pressure Control Low Pressure Makeup Heat Removal
Passive Component Categorization Table 6 Definition of Consequence Impact Groups and Configurations CONSEQUENCES Impact Group Configuration Description Initiating Operating A PBF* occurs in an operating (pressurized)
Event system resulting in an initiating event Loss of Mitigating Standby A PBF occurs in a standby system and does not result in an Ability initiating event, but degrades the mitigating capabilities of a system or train. After failure is discovered, the plant enters the applicable Allowed Outage Time defined in the Technical Specification Demand A PBF occurs when system/train operation is required by an independent demand Combination Operating A PBF causes an initiating event with an additional loss of mitigating ability (in addition to the expected mitigating degradation due to the initiator)
Containment Any A PBF, in addition to the above impacts, also affects containment performance
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Categorization Approach D Qualitative Assessment Develop a list of functions performed by the system Identify the components within the system For each component, identify the system function(s) that it supports Qualitative Risk Assessment of System Functions Qualitative Risk Assessment of Components D Overall Risk Assessment of. Components - highest of:
PRA integrated risk for modeled components Risk for non-modeled hazards (e.g., seismic)
[J Passive risk Qualitative risk
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Pilot Feedback on Application of NEI Document o Note that this is a feedback only.
o SNC has NOT deviated from the NEI document.
Pilot Feedback on Application of NE I Docu ment o Does not allow consideration of component redundancy as a possible risk decreasing factor (e.g., two 100 % capacity redundant pumps).
In PRA model, we use AND gate for such pumps.
o For non-modeled hazards, uses qualitative screening with no allowance for lOP justification CS System Manual boundary valve, 11206U6117, is on Fire SSEL Initial Position: Closed Desired Position during fire: Closed What is the probability of manual valve spuriously opening during fire?
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DP Meeting o lOP meeting on 11/29/11 8:30 to 3:30 ET o Containment Spray system first May not take long as this would be the second meeting Review past meeting minutes o Radiation Monitoring System second Conducted a teleconference to review and agree risk ranking of functions
- We will still cover this in the lOP meeting First meeting face-to-face on RM System
lOP Meeting o Request during NRC Observation Provide comments/questions at designed time End of the CS System discussion
- End of the Radiation Monitoring System discussion Withhold making comments during other times o Discuss next step
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Project Statu s o Risk Characterization Internal Events with Internal Flooding
- Updating in 2011. Safe Shutdown Equipment List (SSEL) review complete.
Other External Events
- Updated in 2011 Shutdown Qualitative DID model developed in 2010.
Project Status o Vogtle Fire PRA Working model available in mid-December 2011
- It will be applied to es and eves categorization As an interim measure, Fire SSEL has been used for es and eves o Peer Review scheduled in Feb 2012 F&O Resolution
- LAR will show how each F&O was/will be resolved
Project Status D 5 Risk Contributors Internal Fire Seismic Other External Shutdown Events Risk Risk Risk Events Risk Risk Internal Events Fire Seism ic Other Shutdown (includi ng
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PRA PRA External PRA Internal Events PRA
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Updated Updated SID DID FIVE Seismic IPEEE Model via Margin Screening EOOS Analysis
M. Ajluni -2 Please feel free to contact me with questions regarding this audit report.
Sincerely, IRA!
Patrick G. Boyle, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-424 and 50-425
Enclosures:
- 1. Audit Report
- 2. Meeting Attendees
- 3. Meeting Slides cc: Distribution via Listserv DISTRISUTION:
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