ML120260090

Relief Request No. IP2-ISI-RR-14, Code Case N-770-1, RCS Cold Leg Nozzle Weld Inspection Frequency Extension

ML120260090
Person / Time
Site:	Indian Point
Issue date:	02/02/2012
From:	Wilson G A Plant Licensing Branch 1
To:	Entergy Nuclear Operations
Boska J P, NRR/DORL/LPLI-1, 301-415-290
References
TAC ME6801
Download: ML120260090 (9)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 2, 2012 Vice President, Operations Entergy Nuclear Operations, Inc. Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 INDIAN POINT NUCLEAR GENERATING UNIT NO.2 -RELIEF REQUEST NO. IP2-ISI-RR-14, CODE CASE N-770-1, REACTOR COOLANT SYSTEM COLD LEG NOZZLE WELD INSPECTION FREQUENCY EXTENSION (TAC NO. ME6801)

Dear Sir or Madam:

By letter dated August 3, 2011, (Agencywide Documents Access and Management System (ADAMS) Accession Number ML 11224A026), as supplemented by letter dated November 8, 2011, (ADAMS Accession Number ML 11319A216 and ML 11319A217), Entergy Nuclear Operations, Inc. (Entergy or the licensee) submitted relief request 14 (RR-14) to the Nuclear Regulatory Commission (NRC). RR-14 pertains to certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) associated with inspection of reactor pressure vessel (RPV) inlet cold leg nozzle to safe-end dissimilar metal butt welds at Indian Point Generating Unit No.2 (lP2). Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(a)(3)(i), the licensee requested to use the proposed alternative on the basis that the alternative provides an acceptable level of quality and safety. The NRC staff concludes that the licensee provided sufficient technical basis to demonstrate that compliance with the requirements of 10 CFR 50.55a(g)(6)(ii)(F) would cause an unnecessary burden on the licensee without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii), the proposed alternative provides reasonable assurance of structural integrity and leak tightness, and is in compliance with the Code of Federal Regulation's requirements.

Therefore, in accordance with 10 CFR 50.55a(a)(3)(ii) the NRC staff authorizes the extension of the weld inspections until the spring 2014 refueling outage. All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

-2 If you have any questions, please contact the Indian Point project manager, John Boska, at (301) 415-2901.

Sincerely, . George A Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-247

Enclosure:

As stated cc w/encl: Distribution via Listserv UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST NO. 14 ENTERGY NUCLEAR OPERATIONS.

INC. INDIAN POINT NUCLEAR GENERATING UNIT NO.2 DOCKET NO. 50-247

1.0 INTRODUCTION

By letter dated August 3,2011, (Agencywide Documents Access and Management System (ADAMS) Accession Number ML 11224A026), as supplemented by letter dated November 8, 2011, (ADAMS Accession Number ML 11319A216 and ML 11319A217), Entergy Nuclear Operations, Inc. (Entergy or the licensee) submitted relief request 14 (RR-14) to the Nuclear Regulatory Commission (NRC). RR-14 pertains to certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) associated with inspection of reactor pressure vessel (RPV) inlet cold leg nozzle to safe-end dissimilar metal (OM) butt welds at Indian Point Generating Unit No.2 (lP2). 2.0 REGULATORY REQUIREMENTS The inservice inspection (lSI) of ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," of the ASME Code and applicable editions and addenda as required by Title 10, Code of Federal Regulations, Part 50.55a(g)

[10 CFR 50.55a(g)], except where specific written relief has been granted by the Commission.

10 CFR 50.55a(g)(6)(ii) states that the Commission may require the licensee to follow an augmented lSI program for systems and components for which the Commission deems that added assurance of structural reliability is necessary.

10 CFR 50.55a(g)(6)(ii)(F) requires, in part, augmented inservice volumetric inspection of class 1 piping and nozzle OM butt welds of pressurized-water reactors (PWRs) in accordance with ASME Code Case N-770-1 I subject to the conditions specified in paragraphs (2) through (10) of 10 CFR 50.55a(g)(6)(ii)(F}.

Alternatives to requirements under 10 CFR 50.55a(g) may be authorized by the NRC pursuant to 10 CFR 50.55a(a)(3)(I) or 10 CFR 50.55a(a)(3)(ii).

In proposing alternatives or requests for relief, the licensee must demonstrate that: (I) the proposed alternatives would provide an acceptable level of quality and safety; or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Enclosure

-2 By letter dated August 3,2011, the licensee proposed an alternative, RR-14, in accordance with 10 CFR 50.55a(a)(3)(i) for the inspection of RPV inlet cold leg nozzle to safe-end OM butt welds at IP2. Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for Inservice Inspection (lSI) of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components.

The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. 3.0 TECHNICAL EVALUATION

3.1 Licensee

Relief Request 3.1.1 Component Identification Weld 21-14A -Loop 21 cold leg nozzle to safe-end weld. Weld 22-14A -Loop 22 cold leg nozzle to safe-end weld. Weld 23-14A -Loop 23 cold leg nozzle to safe-end weld. Weld 24-14A -Loop 24 cold leg nozzle to safe-end weld. 3.1.2 Code Requirements for Which Relief is Requested 10 CFR 50.55a(g)(6)(ii)(F) requires, in part, a volumetric inspection of RPV inlet cold leg nozzle to safe-end OM welds of PWRs in accordance with ASME Code Case N-770-1 "Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities", subject to the conditions specified in paragraphs (2) through (10) of 10 CFR 50.55a(g)(6)(ii)(F).

ASME Code Case N-770-1, Table 1, Inspection Item B requires volumetric examination of essentially 100% of each weld once every second inspection period not to exceed 7 years. 3.1.3 Licensee's Proposed Alternative The licensee proposes a one-time extension to the Code Case N-770-1, Table 1, Inspection Item B, volumetric examinations from a period of "not to exceed 7 years" to a period of "not to exceed 8 years." 3.1.4 Licensee's Duration of Relief Request The licensee requests relief from the regulatory requirement which would require inspection during the scheduled spring 2012 refueling outage and allow the inspection to be performed

-3 during the scheduled spring 2014 refueling outage at IP2. This is a request for a one-time extension of the inspection frequency.

3.1.5 Licensee's

Basis for Relief The licensee stated that relief was requested due to the need to examine the RPV inlet cold leg nozzle to safe-end welds from the inside surface of the weld. This requires access to the cold leg nozzle from inside the reactor vessel to insert automated volumetric inspection equipment to perform the examination.

As such, it would be necessary to remove the core barrel. At IP2, the core barrel is next scheduled to be removed for inspection of the vessel shell welds and vessel internal inspection required by MRP-227 during the spring 2014 refueling outage. Requiring the additional removal of the core barrel during the spring 2012 refueling outage would result in an additional radiological dose to the workers of approximately 1.5 Rem. Additionally, the licensee stated that volumetric inspection of the RPV inlet cold leg nozzle to safe-end welds from the outside surface would be undesirable due to the welds being located in a very restricted cavity known as a sandbox and covered with asbestos insulation.

The licensee estimated that approximately 11 Rem would be accumulated to perform the inspection from the outside surface, including the personnel hazard of dealing with asbestos materials.

The licensee's technical basis for the relief request is based on the temperature dependence of the susceptibility of these welds to primary water stress-corrosion cracking (PWSCC) and the previous inspection history at IP2. The licensee notes that the susceptibility to PWSCC of alloy 82/182 welds, such as those that are the subject of this relief request, is largely a function of time at temperature.

The RPV inlet cold leg nozzle to safe-end welds currently operate at a temperature of about 536 of but previously operated slightly lower for a significant portion of their operating lifetime.

Additionally, the licensee stated that the cold leg welds would be ranked as only moderately susceptible to PWSCC based on the susceptibility formula provided in previously required NRC Order EA-03-009 for the upper RPV head penetration nozzles and welds. The licensee also stated that since PWSCC is temperature dependant, it would be expected that welds exposed to hot leg temperatures, about 598 of, would show evidence of crack initiation before welds exposed to cold leg temperatures, and no evidence of cracking has been identified in either set of welds at IP2. Hot leg temperature welds were inspected most recently in March 2010 and will be inspected again in 2012, as core barrel removal is not required to gain inspection access for these welds. Further, the cold leg temperature welds that are the subject of this relief request were inspected in May of 2006 with both volumetric and eddy current techniques which verified no indications in the weld. In response to an NRC request for additional information, the licensee, by letter dated November 8, 2011, provided a crack growth calculation for a hypothetical flaw that would have initiated just after the May 2006 inspection of a RPV inlet cold leg nozzle to safe-end weld at IP2. The licensee applied the recently created guidelines of MRP-287, "Primary Water Stress Corrosion Cracking (PWSCC) Flaw Evaluation Guidance," in their evaluation.

The analysis showed that a comparison of the depth of the maximum flaw which could have been reasonably missed during the 2006 inspection (i.e. 0.25-inch) with the depth of the maximum flaw (i.e. inch) which would not grow beyond the ASME Code,Section XI limits after 8 years of service results in a margin of conservatism of approximately 5 beyond the margins of safety required by the ASME Section XI Code. As such, the licensee found the technical basis sufficient to ensure public health and safety by extending the inspection frequency of the RPV inlet cold leg nozzle to safe-end DM welds at IP2 from a maximum of 7 years to a new maximum of 8 years. 3.2 NRC Staffs Evaluation The NRC staff notes that the generic rules for the frequency of volumetric examination of dissimilar metal butt welds were established to provide reasonable assurance of the structural integrity of the reactor coolant pressure boundary.

As such, the staff finds that plant-specific analysis could be used to provide a basis for inspection relief if the inspection requirement presents a significant hardship.

As such, the staff reviewed the licensee's proposed alternative under the requirements of 10 CFR 50.55a(a)(3)(ii), such that: "Compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety." As the RPV inlet cold leg DM welds are in sandboxes, and inspection of the welds from the inside of the nozzle would require the licensee to remove the RPV core barrel just for these examinations in the spring 2012 refueling outage, the staff found the licensee had a sufficient basis for a radiological dose hardship.

Therefore, the staff reviewed licensee's deterministic assessment and supporting inspection results to assess the authorization of RR-14. The staff reviewed the licensee's previous inspection results and found they provided a strong basis for the initial flaw size used in the deterministic crack growth analysis.

The initial flaw size is a critical component of a flaw analysis.

The staff found the licensee's data and supporting eddy current inspection results provided a reasonable basis for the initial flaw size assumptions.

The staff reviewed the licensee's stress analysis and found it followed the recommendations of MRP-287 and numerous NRC public meeting discussions with industry since November 19, 2009, on effective weld residual stress calculations to address PWSCC flaw analysis.

Of note, for significant conservatism, a 50% inside surface weld repair 360 0 around the circumference was simulated in the analysis.

The fabrication sequence was simulated based on information provided in the plant specific drawings.

The staff also found that the use of the maximum stress path through the weld, of the three stress paths calculated for hoop stresses, was effective and consistent with NRC staff expectations.

The staff reviewed the final proprietary stress analysis through the thickness of the weld and found both the hoop and axial residual stress curve contours were consistent with analyses using similar geometries and fabrication methods. As such, the staff's review found the licensee's plant-specific stress analysis for these welds to have conservative inputs and assumptions and, therefore, was adequate to be used in the flaw evaluation.

The staff found that the licensee's flaw evaluation used reasonable inputs and industry methodologies to determine maximum end-of-evaluation period flaw sizes for both axial and circumferential flaws. The staff found the licensee's use of the maximum allowable flaw size of 75% of the wall thickness in accordance with the requirements of ASME Code,Section XI, Paragraph IWB-3640, is an adequate approach.

The staff found the licensee's use of the MRP-115, "Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking

-5 (PWSCC) of Alloy 82, 182, and 132 Welds," adequate for the analysis. Further, the staff recognized the licensee's basis of the effect of temperature on the crack growth rates for PWSCC flaws at cold leg operating temperatures.

For example, a PWSCC flaw grown in the same material and same environmental conditions will grow, on average, 7 to 8 times slower at the cold leg operating temperature at IP2 versus a typical operating hot leg temperature at a U.S. PWR. The last component of the staff's review concerned the licensee's flaw analysis results and licensee's conclusions to provide a technical basis to support the relief request. Figures 7-1 and 7-2 of the licensee's November 8,2011, letter provide the PWSCC crack growth curves through the thickness of the welds for both an axial and circumferential flaw, respectively.

The flaw analysis shows that only a flaw 1.2-inch in depth or greater (-50% depth of the weld) could grow to the allowable ASME Code flaw size limit in 8 years. The staff noted that the time-to-failure for both curves is quite large. However, the long time span is a product of the low or compressive calculated residual stresses and the cold leg temperature effect on the crack growth rate. Therefore, the staff found the licensee's flaw analysis evaluation acceptable.

The licensee used the results of the flaw analysis to support the conclusion that since no flaw was identified in the spring 2006 inspection of each weld, the next inspection can be delayed to the spring 2014 refueling outage, while maintaining reasonable assurance of the structural integrity and tightness of each weld. The licensee's 2006 examination included an Appendix VIII demonstrated volumetric examination obtaining essentially 100% coverage, and an eddy current surface examination that found no indications of surface connected flaws. The staff found the inspection techniques and results provided a reasonable basis for the licensee's conclusion that any flaw connected to the wetted surface with a size of 10% in depth or greater should have been identified.

Further, the staff concurred that there is sufficient margin between the initial flaw size of 1.2-inches that would be required to grow to an unacceptable flaw size in 8 years and the maximum flaw size of approximately 10% depth or 0.25-inches which could have been reasonably missed during the 2006 inspection at IP2. Therefore, the staff found that the licensee has provided an adequate technical basis to provide reasonable assurance of structural integrity and leak-tightness for the extended inspection frequency requested in RR-14, which requests to increase the maximum inspection frequency for these welds from 7 to 8 calendar years. Therefore, given the hardship of reaching the location of the RPV inlet cold leg nozzle to end DM welds, either through sandboxes or by removing the RPV core barrel, and the flaw analysis demonstrating a sufficient safety margin, the staff concluded that the licensee has provided adequate technical basis to demonstrate that compliance with the requirements of 10 CFR 50.55a(g)(6)(ii)(F) for the volumetric inspection of the RPV inlet cold leg nozzle to safe-end DM welds at IP2, during the spring 2012 refueling outage, would cause an unnecessary hardship or unusual difficulty on the licensee without a compensating increase in the level of quality and safety given that the volumetric inspections will be performed during the spring 2014 refueling outage at IP2.

4.0 CONCLUSION

S As set forth above, the NRC staff concludes that the licensee provided sufficient technical basis to demonstrate that compliance with the requirements of 10 CFR 50.55a(g)(6)(ii)(F) would cause an unnecessary burden on the licensee without a compensating increase in the level of quality

-and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii), the proposed alternative provides reasonable assurance of structural integrity and leak tightness, and is in compliance with the 10 CFR requirements.

Therefore, in accordance with 10 CFR 50.55a(a){3)(ii) the NRC staff authorizes the licensee's proposed alternative, RR-14, as supplemented by letter dated November 8,2011, at IP2, for the spring 2012 refueling outage until the spring 2014 refueling outage. All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

Principal Contributor:

J. Collins Date: February 2,2012

-If you have any questions, please contact the Indian Point project manager, John Boska, at (301) 415-2901.

Sincerely, /raJ George A Wilson, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-247

Enclosure:

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