L-MT-14-049, Response to NRC Request for Additional Information Regarding Request RR-003, Implementation of Risk-Informed/Safety-Based Lnservice Inspection Program Based on ASME Boiler and Pressure Code Case N-716 for ASME Class 1 & 2 Piping Welds...
| ML14139A233 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 05/19/2014 |
| From: | Fili K D Northern States Power Co, Xcel Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-MT-14-049, TAC MF3050 | |
| Download: ML14139A233 (8) | |
Text
Xcel Energy@ May 19, 2014 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket 50-263 Renewed Facility Operating License No. DPR-22 Monticello Nuclear Generating Plant 2807 W County Road 75 Monticello, MN 55362 L-MT-14-049 10 CFR 50.55a(g)
Subject:
Response to NRC Request for Additional Information Regarding Request RR-003, Implementation of a Risk-Informed/Safety-Based lnservice Inspection Program Based on ASME Boiler and Pressure Code Case N-716 for ASME Class 1 and 2 Piping Welds (TAC No. MF3050)
References:
- 1) Letter from Northern States Power Company, a Minnesota corporation (NSPM) d/b/a Xcel Energy to Document Control Desk, "Request to Utilize an Alternative to the Requirements of 10 CFR 50.55a(g) for Implementation of a Risk-Informed, Safety-Based lnservice Inspection Program Based on ASME Code Case N-716", dated October 31, 2013 (ADAMS Accession Number ML 13308A390).
- 2) E-mail from Terry Beltz to Randy Rippy, "Monticello Nuclear Generating Plant-Draft Requests for Additional Information (APLA) re: Request RR-003 (TAC No. MF3050)", dated April11, 2014 (ADAMS Accession Number ML 14119A012).
By letter dated October 31, 2013, Northern States Power Company, a Minnesota corporation (NSPM), d/b/a Xcel Energy requested authorization to implement a risk-informed, safety-based inservice inspection (lSI) program based on American Society of Mechanical Engineers (ASME) Code Case N-716, "Alternative Piping Classification and Examination Requirements,Section XI, Division 1 ,"for the Monticello Nuclear Generating Plant (Reference 1 ). Subsequently, by e-mail dated April11, 2014 (Reference 2), the NRC provided draft Requests for Additional Information (RAI) for additional information required to complete its review. In an e-mail dated April 11, 2014, NSPM accepted the draft RAI. The formal response is provided in Enclosure
- 1. In addition, typographical errors identified in Reference 1 are corrected in the Errata in Enclosure
- 2. If you have any questions or require additional information,.
please contact Mr. Randy Rippy at 612-330-6911.
Document Control Desk Page 2 Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.
Karen D. Fili Site Vice President, Monticello Nuclear Generating Plant Northern States Power Company-Minnesota Enclosures (2) cc: Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC Minnesota Department of Commerce ENCLOSURE 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING RR-003 RAI APLA-1 The Safety Evaluation Report (SER) related to EPRI Risk-Informed lnservice Inspection Evaluation Procedure (EPRI TR-112657, Rev. B) states that EPRI TR-112657, Rev. B, does not include a detailed discussion of the specific assumptions to be used to guide the assessment of the direct and indirect effects of segment failures.
The SER states that specific assumptions regarding the direct and indirect effects of pipe segment failure should be developed by the individual licensees and should form part of the onsite documentation.
NSPM states in the license amendment request (LAR) submittal dated October 31, 2013, that under the solution for F&O [Facts & Observations]
originating from Supporting Requirement IFSN-A6 that a qualitative assessment was performed for the following flood failure mechanisms:
submergence; spray; jet impingement; pipe whip; humidity; condensation and temperature.
a) Please clarify if specific guidelines and assumptions used for determining direct and indirect effects of flooding, including assumptions on the failure of components affected by the pipe break, have been developed for this application.
b) Please clarify if the Joss of mitigating ability (as discussed in Section 3.2.4 of the SER), where segment failures occur that only cause failure of mitigating functions but do not cause a plant trip, has been considered in consequence evaluations for this application in accordance with EPRI TR-112657.
Response to RAI APLA-1 a) Per the resolution requested from the April 2013 Peer Review, a qualitative assessment was performed for the following flood failure mechanisms to meet Capability Category II criteria:
-Submergence -Spray -Jet Impingement -Pipe Whip -Humidity
-Condensation -Temperature Spray and submergence effects are specifically discussed and accounted for throughout the Internal Flooding Accident Sequence Notebook for those internal flooding scenarios that are impacted.
All internal flood initiators account for submergence and spray. Page 1 of 3 Jet Impingement and Pipe Whip are assumed to have no effect on the internal flooding analysis unless specifically stated in the internal flooding scenario.
This assumption is based on the research provided in NUREG/CR-3231 which found that these potential impacts of pipe breaks require very specific pipe spacing/orientation, rarely result in a complete (guillotine) break, and even if pipe break occurs, significant reduction in cross sectional area of the target pipe follows. The instances which all these conditions are met are few in Monticello Nuclear Generating Plant (MNGP) and the probability of these types of events coincident with a flood initiator is sufficiently remote. Assumption
- 5 was added to the "General Assumption" Section 2.2 of the Internal Flooding Accident Sequence Notebook, PRA-MT-IF-AS Rev. 3.1. Humidity, condensation, and temperature effects are assumed to be encompassed in the bounding, conservative assumption that all spray floods in the MNGP flood model fail the entire room where the flood exists irrespective of room size or pipe orientation.
b) The Internal Flooding analysis performed for the MNGP PRA model update included all flooding sources which could lead to an automatic plant trip or a Technical Specification required manual shutdown of the plant (mitigating impact, but no automatic plant trip). This methodology is consistent with the ASME/ANS RA-Sa-2009 standard as endorsed by Regulatory Guide (RG) 1.200, Revision 2, and applied in the proposed Risk-Informed/Safety-Based (RIS_B) lnservice Inspection program based on ASME Code Case N-716. RAI APLA-2 According to Regulatory Issue Summary 2007-06, "Regulatory Guide 1.200 Implementation," the NRC staff expects licensees to fully address all scope elements with Revision 2 of Regulatory Guide (RG) 1.200 by the end of its implementation period (i.e., one year after the issuance of Revision 2 of RG 1.200). Revision 2 of RG 1.200 endorses, with exceptions and clarifications, the combined American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) probabilistic risk assessment (PRA) standard (ASMEIANS RA-Sa-2009).
Given that the implementation date of Revision 2 to RG 1.200 was in April 2010, and your LAR was submitted in October 2013, please describe any gaps between the 2013 focused scope peer review of the PRA model used in this application and RG 1.200, Revision 2, that are relevant to this submittal and the impact on the application.
Response to RAI APLA-2 When reviewing RAI APLA-2, NSPM identified two typographical errors in the "Request to Utilize an Alternative to the Requirements of 10 CFR 50.55a(g) for Implementation of a Risk-Informed, Safety-Based lnservice Inspection Program Page 2 of 3 Based on ASM E Code Case N-716", dated October 31, 2013 (ADAMS Accession Number ML 13308A390).
The original submittal made references to RG 1.200, Revision 1, which should have been RG 1.200, Revision 2, as specified in the References section.
Please see Enclosure 2 for the corrected pages. This issue has been entered into the Monticello corrective action program. The MNGP Probabilistic Risk Assessment Peer Review was performed in April 2013 using the NEI 05-04 process, the ASME PRA Standard (ASME/ANS RA-Sa-2009), and Regulatory Guide 1.200, Revision 2. Therefore, any gaps from RG 1.200, Revision 2 have been identified in the form of Findings and Observations resulting from the Peer Review. The April 2013 Peer Review was an internal events, excluding fire, Peer Review facilitated by the Boiling Water Reactor Owner's Group (BWROG). Prior to the LAR submittal in October 2013, all Peer Review Findings have been resolved and incorporated into the PRA model used to support the LAR submittal.
Page 3 of 3 ENCLOSURE 2 ERRATA REGARDING 10 CFR 50.55a REQUEST RR-003 NSPM identified two typographical errors in the "Request to Utilize an Alternative to the Requirements of 10 CFR 50.55a(g) for Implementation of a Risk-Informed, Safety-Based lnservice Inspection Program Based on ASME Code Case N-716", dated October 31, 2013 (ADAMS Accession Number ML 13308A390).
The original submittal made references to RG 1.200, Revision 1, which should have been RG 1.200, Revision 2. The corrected pages follow. This issue has been entered into the Monticello corrective action program. 2 pages follow MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)
- 1. INTRODUCTION Monticello Nuclear Generating Plant (Monticello) is currently in Period 1 of the Fifth In service Inspection (lSI) Interval as defined by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Section XI Code. Monticello plans to implement a informed/safety-based inservice inspection (RIS_B) program for the Fifth lSI Interval.
The Fifth lSI Interval began September 1, 2012. The ASME Section XI Code of record for the Fifth lSI Interval is the 2007 Edition through the 2008 Addenda for Examination Category B-F, B-J, C-F-1, and C-F-2 Class 1 and 2 piping components.
The RIS_B process used in this submittal is based upon ASME Code Case N-716, Alternative Piping Classification and Examination Requirements,Section XI Division 1, which is founded in large part on the RI-ISI process as described in Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A, Revised Risk-Informed lnservice Inspection Evaluation Procedure.
1.1 Relation
to NRC Regulatory Guides 1.174 and 1.178 As a risk-informed application, this submittal meets the intent and principles of Regulatory Guide 1.17 4, An Approach for Using Probabilistic Risk Assessment in Informed Decisions On Plant-Specific Changes to the Licensing Basis, and Regulatory Guide 1.178, An Approach for Plant-Specific Risk-Informed Decision making lnservice Inspection of Piping. Additional information is provided in Section 3.4.2 relative to defense-in-depth.
1.2 Probabilistic
Safety Assessment (PSA) Quality The methodology in Code Case N-716 provides for examination of a generic population of high safety significant (HSS) segments, supplemented with a rigorous flooding analysis to identify if any plant-specific HSS segments need to be added. Satisfying the requirement for the plant-specific analysis requires confidence that the flooding PRA is capable of successfully identifying any significant flooding contributors that are not identified in the generic population.
The Monticello PRA is based on a detailed model of the plant that was originally developed from the Individual Plant Examination (IPE) and Individual Plant Examination for External Events (IPEEE) projects.
The original model was reviewed by the NRC and underwent Boiling Water Reactor Owner's Group (BWROG) certification in 1997. NRC reviews of the IPE and IPEEE are documented in the NRC Staff Evaluations on IPE in May 1994 (TAC No M74435) and IPEEE dated April2000 (TAC No M83644). The NRC concluded that the Monticello process is capable of identifying the most likely severe accidents and no significant impacts on the PRA were identified.
The Monticello PRA has since been upgraded.
It is a Level 2, at-power model. A major upgrade of the internal events model to meet the guidance of RG 1.200, Revision -4 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," as well as the American 8 of 31 MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)
Society of Mechanical Engineers and American National Standard (ASME/ANS)
PRA Standard RA-Sa-2009 was completed in 2013. A formal, BWROG-sponsored industry peer review of the upgraded internal events model was completed in April 2013. The peer review utilized the process described in Nuclear Energy Institute document NEI 05-04, "Process for Performing Follow-on PRA Peer Reviews Using the ASME PRA Standard," January 2005, and the ASME/ANS PRA Standard.
This review to ASME Capability Category II requirements confirmed that the PRA model met the requirements of RG 1.200, Revision -4 2, and ASME/ANS RA-Sa-2009.
There were twenty two Finding Level F&Os identified by the peer review team. Attachment A contains a summary of these findings and their resolution.
To date, all of the Peer Review Findings have been resolved and inserted into the PRA model or dispositioned for this submittal.
The original RI-ISI evaluation concluded external events are not likely to impact the consequence ranking. This position is further supported by Section 2 of EPRI Report 1021467, "Nondestructive Evaluation:
Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs" which concludes that quantification of these events will not change the conclusions derived from the RI-ISI process. As a result, there is no need to further consider these events. The PRA model used for development of the RIS_B evaluation accounts for conditions applicable to the Extended Power Uprate which Monticello intends to implement following NRC approval.
- 2. PROPOSED ALTERNATIVE TO CURRENT lSI PROGRAMS 2.1 ASME Section XI ASME Section XI Examination Categories B-F, B-J, C-F-1, and C-F-2 currently contain requirements for the nondestructive examination (NDE) of Class 1 and 2 piping components.
The .alternative RIS_B Program for piping is described in Code Case N-716. The RIS_B Program will be substituted for the current program for Class 1 and 2 piping (Examination Categories B-F, B-J, C-F-1 and C-F-2) in accordance with 10 CFR 50.55a(a)(3)(i) by alternatively providing an acceptable level of quality and safety. Other non-related portions of the ASME Section XI Code will be unaffected.
2.2 Augmented
Programs The impact of the RIS_B application on the various plant augmented inspection programs listed below were considered.
This section documents only those plant augmented inspection programs that address common piping with the RIS_B application scope (e.g., Class 1 and 2 piping).
- The plant augmented inspection program for high-energy line breaks outside containment has not been revised by this application.
A separate evaluation and program in accordance with the risk-informed break exclusion region methodology (RI-BER) described in EPRI Report 1006937, Extension of EPRI Risk Informed IS/ Methodology to Break Exclusion Region Programs has not yet been established.
9 of31