ML14199A444
| ML14199A444 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 08/04/2014 |
| From: | Benjamin Beasley Plant Licensing Branch 1 |
| To: | Entergy Nuclear Operations |
| Pickett D | |
| References | |
| TAC MF3346 | |
| Download: ML14199A444 (8) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 August 4, 2014 Vice President, Operations Entergy Nuclear Operations, Inc.
Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249
SUBJECT:
INDIAN POINT NUCLEAR GENERATING UNIT NO.3- SAFETY EVALUATION FOR RELIEF REQUEST IP3-ISI-RR-07 FOR REACTOR VESEL COLD LEG NOZZLE TO SAFE-END WELD EXAMINATIONS (TAC NO. MF3346)
Dear Sir or Madam:
By letter dated January 13, 2014, as supplemented by letter dated February 4, 2014, Entergy Nuclear Operations, Inc., the licensee for Indian Point Nuclear Generating Unit No. 3 (IP3),
submitted relief request IP3-ISI-RR-07 as an alternative to certain requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(6)(ii)(F) and the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) associated with inservice inspection of reactor pressure vessel inlet cold leg nozzle to safe-end dissimilar metal butt welds at IP3. The licensee requested authorization to use the proposed alternative pursuant to 10 CFR 50.55a(a)(3)(ii) on the basis that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Based on the information submitted, the U.S. Nuclear Regulatory Commission (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) and the ASME Code 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 use of the licensee's proposed alternative, IP3-ISI-RR-07, at IP3, through the spring 2019 refueling outage.
All other ASME Code,Section XI and 10 CFR 50.55a(g)(6)(ii)(F) requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear lnservice Inspector.
Vice President, Operations Please feel free to contact Douglas Pickett at 301-415-1364 or Douglas.Pickett@nrc.gov if you have any questions Sincerely,
~JJ~
Benjamin G. Beasley, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-286
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 REQUEST FOR RELIEF IP3-ISI-RR-07 REGARDING ENTERGY NUCLEAR OPERATIONS, INC INDIAN POINT NUCLEAR GENERATING UNIT NO.3 DOCKET NUMBER 50-286
1.0 INTRODUCTION
By letter dated January 13, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14017A054), as supplemented by letter dated February 4, 2014, (ADAMS Accession No. ML14051A166), Entergy Nuclear Operations, Inc., the licensee for Indian Point Nuclear Generating Unit No. 3 (IP3), submitted relief request IP3-ISI-RR-07 (RR-07), as an alternative to certain requirements of Title 10 of the Code of Federal Regulations (1 0 CFR) 50.55a(g)(6)(ii)(F) and 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 IP3. The licensee requested authorization to use the proposed alternative pursuant to 10 CFR 50.55a(a)(3)(ii) on the basis that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
2.0 REGULATORY EVALUATION
The inservice inspection (lSI) of ASME Code Class 1, 2 and 3 components is to be performed in accordance with Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," of the ASME Code and applicable editions and addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the Nuclear Regulatory Commission (NRC or Commission).
Paragraph 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 DM butt welds of pressurized water reactors 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,Section XI, Division 1' (ASME Approval Date: December 25, 2009)." subject to the conditions specified in paragraphs (2) through (1 0) 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: (1) the proposed alternatives would provide an acceptable level of quality and safety; or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Paragraph 10 CFR 50.55a{g){6)(ii)(F) requires licensees of existing operating pressurized water reactors to implement the requirements of ASME Code Case N-770-1, subject to the conditions specified in paragraphs (g)(6)(ii)(F)(2) through (g)(6)(ii)(F)(1 0) of this section, by the first refueling outage after August 22, 2011.
Based on analysis of the regulatory requirements, the NRC staff finds that the regulatory authority exists to authorize the licensee's proposed alternative on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the staff has reviewed and evaluated the licensee's request pursuant to 10 CFR 50.55a(a)(3)(ii).
3.0 TECHNICAL EVALUATION
3.1 Licensee Relief Request The affected components are as follows:
Weld 1-4100-16(DM) Loop 31 cold leg nozzle to safe-end weld Weld 1-4200-16(DM) Loop 32 cold leg nozzle to safe-end weld Weld 1-4300-16(DM) Loop 33 cold leg nozzle to safe-end weld Weld 1-4400-16(DM) Loop 34 cold leg nozzle to safe-end weld Paragraph 10 CFR 50.55a(g)(6)(ii)(F) requires, in part, a volumetric inspection of RPV inlet cold leg nozzle to safe-end DM welds of pressurized water reactors in accordance with ASME Code Case N-770-1 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 percent of each weld once every second inspection period not to exceed 7 years.
The licensee proposes a onetime 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 10 years."
The licensee requests relief from the regulatory requirement which would require inspection during the scheduled March 2015 refueling outage and allow the inspection to be performed during the scheduled spring 2019 refueling outage at IP3. This is a one-time extension inspection frequency request.
The licensee stated that the relief request was 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 lower portion of the RPV to insert automated volumetric inspection equipment to perform the examination. As such, it would be necessary to remove the core barrel and other RPV internals. At IP3, the core barrel is scheduled to be removed for inspection of the vessel shell welds and vessel internal inspection required by Electric Power Research Institute report,
MRP-227A, during the spring 2019 refueling outage. Requiring the additional removal of the core barrel and other internals during the March 2015 refueling outage would result in an additional radiological personnel dose.
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 inside a sandbox and covered with insulation. The sandbox was installed during original plant construction after all welding was completed.
The licensee's technical basis for the relief request is based on the temperature dependence on the susceptibility of these welds to primary water stress corrosion cracking (PWSCC) and the previous inspection history at IP3. 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 and temperature. The RPV inlet cold leg nozzle to safe-end welds operate at a temperature of less than 541.1 degrees Fahrenheit for a significant portion of their operating lifetime. Additionally, the licensee claims that the welds would be ranked as 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 states that since PWSCC is temperature dependent, it would be expected that hot leg temperature welds would show evidence of crack initiation before cold leg temperature welds, and no evidence of cracking has been identified in either hot leg and cold leg welds at IP3. Further, the cold leg temperature welds that are the subject of this relief request were inspected in March of 2009 with volumetric techniques which verified no indications in the welds.
The licensee provided a crack growth calculation for a hypothetical flaw that would have initiated just after the March 2009 inspection of a RPV inlet cold leg nozzle to safe-end weld at IP3. The licensee applied the recently created guidelines of Electric Power Research Institute report, MRP-287, "Primary Water Stress Corrosion Cracking (PWSCC) Flaw Evaluation Guidance," in their evaluation. The licensee stated that their analysis showed significant margin to ensure that ASME Section XI flaw size limits would not be exceeded during the extended period of inspection frequency. 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 OM welds at IP3 from a maximum of 7 years to a new maximum of 10 years.
3.2 NRC Staff's Evaluation The NRC staff notes that the generic rules for the frequency of volumetric examination of OM butt welds were established to provide reasonable assurance of the structural integrity of the reactor coolant pressure boundary. As such, the staff finds that a 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.
The NRC staff reviewed the licensee's basis for hardship. As the RPV inlet cold leg DM welds are located in sandboxes, and inspection of the welds would require the licensee to remove the RPV core barrel just for these examinations in the March 2015 refueling outage, the staff found the licensee had a sufficient basis for a radiological dose hardship.
Given this hardship, the licensee's basis is that the proposed alternative provides reasonable assurance of structural integrity and leak tightness. The licensee also states that no flaw of a size that could have been potentially missed during the 2009 refueling outage inspection could reasonably grow to an unacceptable size during the period of increased inspection frequency.
Therefore, the NRC staff reviewed the licensee's inspection results and flaw analysis to assess the acceptability of RR-07.
The NRC staff reviewed the licensee's previous inspection methods and results to assess the licensee's basis for a maximum hypothetical initial flaw size assumption during the 2009 outage.
The licensee's 2009 examination included an Appendix VIII demonstrated volumetric examination obtaining essentially 100 percent coverage that found no indications of surface connected flaws. The staff found the licensee's qualified inspection techniques provide a reasonable basis that any flaw connected to the wetted surface with a size of 10 percent in depth or greater should have been identified. Also, the staff found the licensee's data and supporting inspection results provided a reasonable basis for their initial flaw size assumptions.
The NRC staff reviewed the licensee's flaw analysis, which consisted of a stress analysis and a flaw evaluation. 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, the licensee simulated a 50 percent inside surface weld repair 360 degrees around the circumference in its analysis. The licensee also simulated the fabrication sequence based on information provided in the plant specific drawings. The staff 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 staff expectations. The staff reviewed the final plant-specific proprietary stress analysis through the thickness of the weld and found both the hoop and axial residual stress curve contours were consistent with generic analyses using similar geometries and fabrication methods as that of IP3. As such, the staff 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 NRC 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 noted that the licensee's use of the maximum allowable flaw size of 75 percent 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 (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, approximately 7 times slower at the cold leg operating temperature at IP3 versus a typical operating hot leg temperature at a U. S. pressurized water reactor plant.
The last component of the NRC staffs review concerned the licensee's flaw analysis results and the licensee's conclusions to provide a technical basis to support the relief request.
Figures 7-1 and 7-2 of the licensee's letter dated February 4, 2014, provide the PWSCC crack growth curves through the thickness of the welds for both an axial and circumferential flaw, respectively. The licensee's flaw analysis shows that a flaw would be required to be at least 1-inch in depth or greater (-40 percent depth of the weld) in order to grow to the allowable ASME Code flaw size limit (75 percent through wall) in 10 years. The licensee used these results to support the conclusion that since no flaw was identified in the March 2009 inspection of each weld, the next inspection can be delayed to the spring 2019 refueling outage, while maintaining reasonable assurance of the structural integrity and leak tightness of each weld. The staff assessed the licensee's conclusion by performing a series of flaw evaluations. As stated above, the staff found that the maximum flaw size which could have been reasonably missed during the 2009 inspection at IP3 was only 10 percent in depth or 0.25-inches. Therefore, since no flaws were identified by the licensee during the 2009 inspection, the staff's flaw evaluations postulated an initial flaw size with a depth of 10 percent of the pipe wall thickness. The staff's flaw evaluations find that there is sufficient margin between the hypothetical maximum size of the postulated flaw after 10 years of growth and the code allowable flaw size to support the licensee's conclusion. Therefore, the staff finds that the licensee has provided an adequate technical basis to support reasonable assurance of structural integrity and leak tightness for the extended inspection frequency requested in RR-07.
Therefore, given the hardship of the location of the RPV inlet cold leg nozzle to safe-end OM welds in sandboxes and the flaw analysis demonstrating a sufficient safety margin, the NRC staff concludes 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 OM welds at IP3, during the March 2015 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 2019 refueling outage at IP3.
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) and the ASME Code would cause an unnecessary burden on the licensee without a compensating increase in the level of quality and safety. Accordingly, the 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 use of the licensee's proposed alternative, RR-07, at IP3, through the spring 2019 refueling outage.
All other ASME Code,Section XI and 10 CFR 50.55a(g)(6)(ii)(F) requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear lnservice Inspector.
Vice President, Operations Please feel free to contact Douglas Pickett at 301-415-1364 or Douglas. Pickett@nrc.gov if you have any questions Sincerely, IRA!
Benjamin G. Beasley, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-286
Enclosure:
As stated cc w/encl: Distribution via Listserv DISTRIBUTION:
PUBLIC LPL 1-1 R/F ABurritt, R1 RidsNrrDorl RidsRgn1 MaiiCenter SVitto, EPNB RidsNrrDorllpl1-1 RidsNrrPMindianPoint RidsNrrDeEpnb RidsNrrLAKGoldstein RidsAcrsAcnw_MaiiCTR RidsNrrDoriDpr JCollins, EPNB ADAMS ACCESSION NO* .. ML14199A444 OFFICE LPL1-1/PM LPL1-1/LA EPNB/BC(A) LPL 1-1/BC NAME DPickett KGoldstein JTsao by email BBeasley dated DATE 08/01/2014 08/ 01 /2014 07/17/2014 08/04/2014 OFFICIAL RECORD COPY