L-11-293, CFR 50.55a Request for Alternative Examination Requirements for ASME Class 1 Piping Welds

CFR 50.55a Request for Alternative Examination Requirements for ASME Class 1 Piping Welds

ML113180450
Person / Time
Site:	Perry
Issue date:	11/14/2011
From:	Bezilla M FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-11-293
Download: ML113180450 (8)

Text

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^^^^ 10 Center Road FirstEnergy Nuclear Operating Company Perry, Ohio 44081 MarkB.Bazllla 440-280-5382 Vice President Fax: 440-280-8029 November 14, 2011 L-11-293 10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Perry Nuclear Power Plant Docket No. 50-440, License No. NPF-58 10 CFR 50.55a Request for Alternative Examination Requirements for ASME Class 1 Piping Welds Pursuant to 10 CFR 50.55a(a)(3), FirstEnergy Nuclear Operating Company (FENOC) requests Nuclear Regulatory Commission (NRC) approval for continued use of the existing Perry Nuclear Power Plant (PNPP) risk-informed inservice inspection program, with updates, relative to certain non-destructive examination requirements associated with American Society of Mechanical Engineers (ASME) Class 1 piping welds.

The enclosed proposed alternative would be implemented during the PNPP third inservice inspection interval. FENOC requests approval of the alternative by October 31, 2012 to support planning and scheduling of work for the PNPP spring 2013 refueling outage.

There are no regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Phil H. Lashley, Supervisor-Fleet Licensing, at (330) 315-6808.

Sincen Mark B. Bezilla

Enclosure:

Perry Nuclear Power Plant, 10 CFR 50.55a Request IR-049, Revision 1 cc: NRC Region III Administrator NRC Resident Inspector Nuclear Reactor Regulation Project Manager

Perry Nuclear Power Plant 10 CFR 50.55a Request IR-049, Revision 1 Page 1 of 7 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

-Alternative Provides Acceptable Level of Quality and Safety-

1. American Society of Mechanical Engineers (ASME) Code Components Affected ASME Code Class 1 piping welds as listed in Table 1

2. Applicable Code Edition and Addenda

ASME Code Section XI, 2001 Edition through the 2003 Addenda

3. Applicable Code Requirement

IWB-2500, Examination and Pressure Test Requirements Table IWB-2500-1, Examination Categories Class 1 Piping Welds Category B-F, Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles Item Nos. B5.10, B5.20, B5.30 Category B-J, Pressure Retaining Welds in Piping Item Nos. B9.11, B9.21, B9.31, B9.32, B9.40

4. Reason for Request

On October 17, 2001, Nuclear Regulatory Commission (NRC) staff approved the Perry Nuclear Power Plant (PNPP) RI-ISI Program for use during the second and third periods of the second 10-year ISI interval. In its approval, NRC staff concluded the RI-ISI program is consistent with Electric Power Research Institute, Inc. (EPRI)

TR-112657, "Revised Risk-Informed Inservice Inspection Evaluation Procedure,"

Revision B-A, and is an acceptable alternative to the requirements of ASME Code,Section XI, for inservice inspection of ASME Class 1 piping, examination categories B-F and B-J. The proposed RI-ISI program, with updates, provides a 63 percent reduction in required examinations compared to ASME Section XI. As such, FirstEnergy Nuclear Operating Company (FENOC) requests that the PNPP RI-ISI program, with updates, be approved for continued use during the third 10-year ISI interval.

Perry Nuclear Power Plant 10 CFR 50.55a Request IR-049, Revision 1 Page 2 of 7

5. Proposed Alternative and Basis for Use Pursuant to 10 CFR 50.55a, FENOC requests NRC approval of the PNPP RI-ISI program as an alternative to the ASME inspection requirements for Class 1 Examination Category B-F and B-J piping welds. The PNPP RI-ISI program was developed in accordance with the methodology contained in EPRITR-112657 and was approved for use at PNPP during the second and third periods of the second 10-year inspection interval. The PNPP-specific RI-ISI program is summarized in Table 1. The RI-ISI program has been updated, consistent with the intent of Nuclear Energy Institute (NEI) 04-05, "Living Program Guidance To Maintain Risk-Informed Inservice Inspection Programs For Nuclear Plant Piping Systems," and continues to meet EPRI TR-112657 and Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Sepecific Changes to the Licensing Basis," Revision 2, risk acceptance criteria. The RI-ISI program update incorporated new input information such as plant modifications and changes to the probabilistic risk assessment (PRA) model.

Based on the PRA model, the consequence ranking changed. The RI-ISI program update increased the consequence ranking for certain pipe segments in the following systems from a medium consequence rank to a high consequence rank: reactor pressure vessel (RPV) nozzles and connections, nuclear boiler - main steam reactor recirculation, residual heat removal, reactor core isolation cooling, reactor water cleanup, and feedwater. The impact of this change is reflected in the additional number of inspection locations for the new RI-ISI interval compared to the previous interval as identified in Table 1.

The PNPP PRA model was formally reviewed in 1997 and has since undergone multiple assessments. The most recent assessment, using the ASME RA-Sb-2005 PRA Standard, was performed in 2008. The PRA model meets Capability Category II for all supporting requirements regarding Level 1 internal events only with exception of internal flooding, and is compliant with Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 1. Additionally, the latest model update was performed to meet the supporting requirements for Level 1 internal events with exception of internal flooding for the latest revision of the ASME and American Nuclear Society (ANS) PRA Standard, ASME/ANS RA-Sa-2009. The current PRA model of record is PRA-PY1-FP-R0 and has a core damage frequency (CDF) value of 4.29E-6.

The model does not include large early release frequency (LERF), internal flood, fire, seismic, or external events. A review of the RI-ISI risk impact results identifies that even if a conditional large early release probability (CLERP) existed, given core damage was set to 1.0, which would bound any impact, the risk acceptance criteria of EPRI TR-112657 would still be met. Therefore, not having an updated LERF calculation as a part of the updated PRA model, does not affect the RI-ISI program update. Internal flooding is not required as the PRA is not used to evaluate the spatial impacts of pressure boundary failures.

Perry Nuclear Power Plant 10 CFR 50.55a Request IR-049, Revision 1 Page 3 of 7 With respect to Revision 2 of Regulatory Guide 1.200, EPR11021467, "Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs," Section 2.2 provides justification for RI-ISI supporting analyses based only on internal events PRAs while excluding other hazard groups. The basis for excluding these hazard groups is discussed below using the applicable information from EPR11021467.

• Internal fire events - The potential contribution of piping failure to internal fire risk is insignificant as the failure probability of piping is insignificant compared to the failure probability of other systems, structures and components (SSCs), such as pumps, valves and power supplies. Fire events are also not likely to present significantly different challenges to the piping in the scope of this application.

Meeting defense-in-depth and safety margin principles provides additional assurance that this conclusion will remain valid. ISI is an integral part of defense-in-depth, and the RI-ISI process will maintain the basic intent of ISI (that is, identifying and repairing flaws), and thus provide reasonable assurance of an ongoing substantive assessment of piping condition. In addition, there are no changes to design basis events, and thus safety margins are maintained.

• Seismic events - Well engineered systems and structures (for example, piping systems) are seismically rugged. Individual plant examinations of external events (IPEEE) and other industry and NRC studies (for example, EPRI TR-1000895, NUREG/CR-5646) have shown piping systems to have seismic fragility capacities greater than the screening values typically used in seismic assessment and are not considered likely to fail during a seismic event. ISI is not considered in establishing fragilities of such SSCs. Meeting defense-in-depth and safety margin principles provides assurance that this conclusion will remain valid. ISI is an integral part of defense in depth, and the RI-ISI process will maintain the basic intent of ISI (that is, identifying and repairing flaws) and thus provide reasonable assurance of an ongoing substantive assessment of piping condition. In addition, there are no changes to design basis events, and thus safety margins are maintained.

• High winds, external floods, and other external hazards - The purpose of developing a RI-ISI program is to define an alternative inservice inspection strategy for piping systems. Other hazards (for example, high wind, external floods) are not considered in the development of an inservice inspection program for piping. The reasons include: the structural ruggedness of the piping systems, location, as relevant systems are typically inside well engineered structures, and the consequence assessment for internal events already includes the consideration of spatial impacts. In addition, the substantial industry experience with plants implementing RI-ISI programs has not identified changes based upon insight from the evaluation of these other external hazards. The very small potential impact on the potential for piping failure of a RI-ISI process, and the approaches to maintaining defense-in-depth and safety margins summarized above, provide confidence in this conclusion.

Perry Nuclear Power Plant 10 CFR 50.55a Request IR-049, Revision 1 Page 4 of 7 In summary, excluding hazard groups such as internal fires, seismic events, high winds, external floods, and other external hazards as addressed in the ASME/ANS PRA Standard (RA-Sa-2009) has an insignificant impact on the RI-ISI analysis and will not change the conclusions derived from the RI-ISI process.

Augmented Programs:

In accordance with EPRI TR-112657, piping welds identified as Category A per NUREG-0313, "Technical Report on Material Selection and Processing Guidelines for BWR [Boiling Water Reactor] Coolant Pressure Boundary Piping," are considered resistant to inter-granular stress corrosion cracking (IGSCC), and as such, are assigned a low failure potential provided no other damage mechanisms are present.

The augmented inspection program for the other NUREG-0313 (Category C and E) piping welds at PNPP are performed in accordance with BWR Vessel and Internals Project (BWRyiP) BWRVIP-75-A, "BWRVIP-75-A: BWR Vessel and Internals Project Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules." The augmented inspection program for flow accelerated corrosion (FAC) in accordance with Generic Letter 89-08, "Erosion/Corrosion-Induced Pipe Wall Thinning," is relied upon to manage this damage mechanism but is not otherwise affected or changed by the proposed RI-ISI program. The augmented program for piping welds in the high energy break exclusion region, which is performed in accordance with the PNPP Updated Safety Analysis Report, including the use of EPRI TR-1006937, "Extension of the EPRI Risk-Informed Inservice Inspection (RI-ISI) Methodology to Break Exclusion Region (BER) Programs," Revision 0-A, is also not affected or changed by the proposed RI-ISI program.

With the increased number of inspection locations resulting from the change in consequence ranking, no examinations for defense-in-depth considerations are necessary.

Pursuant to 10 CFR 50.55a(a)(3)(i), the proposed alternative to the ASME Code Section XI examination requirements will continue to provide an acceptable level of quality and safety.

6. Duration of Proposed Alternative This proposed alternative shall be utilized during the third 10-year inservice inspection interval currently scheduled to expire May 17, 2019.

7. Precedent Nuclear Regulatory Commission letter to FirstEnergy Nuclear Operating Company, October 17, 2001,

Subject:

Safety Evaluation of Relief Request IR-049 Associated with the Second 10-Year Interval Inservice Testing Program (TAC No. MB1174).

Perry Nuclear Power Plant 10 CFR 50.55a Request IR-049, Revision 1 Page 5 of 7

8. References EPRI TR-112657, Revised Risk-Informed Inservice Inspection Evaluation Procedure, Revision B-A NEI 04-05, Living Program Guidance To Maintain Risk-Informed Inservice Inspection Programs For Nuclear Plant Piping Systems Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis, Revision 2 Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 1 ASME Ra-Sb-2005, ASME PRA Standard ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications EPR11021467, Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs TR-1000895, Individual Plant Examination for External Events (IPEEE) Seismic Insights: Revision to EPRI Report TR-112932 NUREG/CR-5646, Piping System Response During High-Level Simulated Seismic Tests at the Heissdampfreaktor Facility (SHAM Test Facility)

NUREG-0313, Technical Report on Material Selection and Processing Guidelines for BWR [Boiling Water Reactor] Coolant Pressure Boundary Piping BWRVIP-75-A, BWR Vessel and Internals Project Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules Generic Letter 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning EPRI TR-1006937, Extension of the EPRI Risk-Informed Inservice Inspection (RI-ISI)

Methodology to Break Exclusion Region (BER) Programs, Revision 0-A

Perry Nuclear Power Plant 10 CFR 50.55a Request IR-049, Revision 1 Page 6 of 7 Table 1: Inspection Location Selection Comparison

^Approved RI-ISI Risk<2) Consequence Failure Potential121 Code Weld New RI-ISI Interval System Interval Rank Category Count Category Rank DMs Rank RI-ISI Other11* RI-ISI Other" TASCS, TT, 1B13 2(1) High (High) High Medium (High) B-F 6 1 5 1 2 (IGSCC, FAC) 1B13 2 High High TT Medium B-J 1 1 - 1 -

1B13 4(2) Medium (High) High None (IGSCC) Low (Medium) B-F 17 3 14 3 4 B-F 9 1 - 0 -

1B13 4 Medium High None Low B-J 2 0 - 0 1 B-F 2 0 2 0 1 1B13 6a (5a) Low (Medium) Medium None (IGSCC) Low (Medium)

B-J 2 0 2 0 1 B-F 1 0 - 0 -

1B13 6a Low Medium None Low B-J 2 0 - 0 -

1B21 4 Medium High None Low B-J 77 1 - 8 -

1B21 6a Low Medium None Low B-J 44 0 - 0 - .

1B21 7a Low Low None Low B-J 2 0 - 0 -

1B33 4 Medium High None Low B-J 112 11 - 12 -

1C41 4 Medium High None Low B-J 7 1 - 1 -

B-F 1 0 - 0 -

1C41 6a Low Medium None Low B-J 26 0 - 0 -

1E12 2 High High TT Medium B-J 4 1 - 1 -

B-F 1 0 - 0 -

1E12 4 Medium High None Low B-J 81 8 - 9 -

1E12 6a Low Medium None Low B-J 18 0 - 0 -

1E12 6b Low Low TT Medium B-J 10 0 - 0 -

1E21 4 Medium High None Low B-J 29 3 - 3 -

1E21 6a Low Medium None Low B-J 4 0 - 0 -

1E22 4 Medium High None Low B-J 26 3 - 3 -

1E22 6a Low Medium None Low B-J 4 0 - 0 -

Perry Nuclear Power Plant 10 CFR 50.55a Request IR-049, Revision 1 Page 7 of 7 Table 1: Inspection Location Selection Comparison (Conf d) 1st Approved RI-ISI Risk'21 Consequence Failure Potential'2' Code Weld New RI-ISI Interval System Interval Rank Category Count Category Rank DMs Rank RI-ISI Other' RNSI Other" 1E32 4 Medium High None Low B-J 12 2 - 2 -

1E51 2 High High TT Medium B-J 8 2 - 2 -

1E51 4 Medium High None Low B-J 9 1 - 1 -

1E51 5a Medium Medium TT Medium B-J 24 3 - 3 -

1E51 6a Low Medium None Low B-J 2 0 - 0 -

1G33 4 Medium High None Low B-J 16 1 - 2 -

B-F 8 1 - 0 -

1G33 6a Low Medium None Low B-J 97 1 - 0 -

1G33 7a Low Low None Low B-J 6 0 - 0 -

1N22 2 High High TT Medium B-J 2 1 - 1 -

1N22 5a Medium Medium TT Medium B-J 63 7 - 7 -

1N22 6b Low Low TT Medium B-J 1 0 - 0 -

TASCS, TT, 1N27 2(1) High (High) High Medium (High) B-J 20 5 - 5 -

(FAC) 1N27 2(1) High (High) High TT, (FAC) Medium (High) B-J 39 1 - 10 -

1N27 4(1) Medium (High) High None (FAC) Low (High) B-J 3 0 - 1 -

1N27 5a (3) Medium (High) Medium TT, (FAC) Medium (High) B-J 2 1 - 1 -

Totals 800 60 23 77 9 Notes

1. The column labeled "Other* is generally used to identify augmented inspection program locations credited per Section 3.6.5 of EPRITR-112657. The EPRI methodology allows augmented inspection program locations to be credited if the inspection locations selected strictly for RI-ISI purposes produce less than a 10 percent sampling of the overall Class 1 weld population. Due to the additional required examinations with the proposed RI-ISI program, no defense-in-depth locations are necessary, and the locations identified in this column reflect those performed in accordance with BWRVIP-75-A. The "Other" column has been retained in this table solely for uniformity purposes with the other RI-ISI application template submittals.

2. Risk categorization is presented with and without consideration of FAC and IGSCC damage mechanisms; parentheses enclose those with consideration of FAC and IGSCC.

Systems Degradation Mechanisms (DMs) 1B13 - RPV Nozzles and Connections 1E22 - High Pressure Core Spray FAC - flow accelerated corrosion 1B21 - Nuclear Boiler- Main Steam 1E32 - Main Steam Isolation Valve Leakage Control System IGSCC - intergranular stress corrosion cracking 1B33 - Reactor Recirculation 1E51 - Reactor Core Isolation Cooling TASCS - thermal stratification, cycling and striping 1C41 - Standby Liquid Control 1G33 - Reactor Water Cleanup TT - thermal transient 1E12 - Residual Heat Removal 1N22 - Miscellaneous Drains - Main Steam Drains 1E21 - Low Pressure Core Spray 1N27-Feedwater