ML112710328

Grand Gulf Nuclear Station, Unit 1 - Request for Alternative GG-ISI-013 from Examination Requirements for Reactor Pressure Vessel Weld Inspections for Third 10-Year Inservice Inspection Interval (TAC ME5990)

ML112710328
Person / Time
Site:	Grand Gulf
Issue date:	11/04/2011
From:	Markley M T Plant Licensing Branch IV
To:	Entergy Operations
Wang, A B, NRR/DORL/LPLIV, 415-1445
References
TAC ME5990
Download: ML112710328 (10)

Text

UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555*0001 November 4, 2011 Vice President, Operations Entergy Operations, Inc. Grand Gulf Nuclear Station P.O. Box 756 Port Gibson, MS 39150 GRAND GULF NUCLEAR STATION, UNIT 1 -RELIEF REQUEST GG-ISI-013 BASED ON CODE CASE N-702 (TAC NO. ME5990) Dear Sir or Madam: By letter dated April 6, 2011, Entergy Operations, Inc. (Entergy, the licensee), requested U.S. Nuclear Regulatory Commission (NRC) approval of changes to the inservice inspection (lSI) program for the third 1 O-year lSI interval for Grand Gulf Nuclear Station (GGNS), Unit 1. This Request for Alternative GG-ISI-013 would revise the inspection requirements for certain reactor pressure vessel (RPV) nozzle-to-vessel welds and nozzle inner radii from those based on the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI to an alternative based on ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds," without using the Code Case-specified visual (VT-1) examination. Based on the enclosed safety evaluation, the NRC staff concludes that the licensee's proposed alternative provides an acceptable level of quality and safety and applies to all requested GGNS, Unit 1 RPV nozzles, with the exception of feedwater nozzles and control rod drive return nozzles. The NRC staff also concludes that the licensee's adoption of ASME Code Case N-648-1, "Alternative Requirements for Inner Radius Examination of Class 1 Reactor Vessel Nozzles,Section XI, Division 1," which is consistent with the NRC's position stipulated in Regulatory Guide 1.147, Revision 15, "Inservice Inspection Code Case Acceptability ASME Section XI, Division 1," October 2007, provides reasonable assurance of structural integrity of the nozzles' inner radii. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in paragraph 50.55a(a)(3)(i) of Title 10 of the Code of Federal Regulations and is in compliance with the ASME Codes' requirements. Therefore, the NRC authorizes the licensee's proposed alternative for inspection of the RPV nozzle-to-vessel shell welds and nozzle inner radii sections listed in Attachment 1 of the licensee's April 6, 2011, submittal, with the exception of feedwater nozzles and control rod drive return nozzles, for GGNS, Unit 1 through the end of the third 10-year lSI interval, which ends in June 2017.

-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. Sincerely, Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. Safety cc w/encl: Distribution via Listserv UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST GG-ISI-013 FOR FACILITY OPERATING LICENSE NO. NPF-29 GRAND GULF NUCLEAR STATION, UNIT 1 ENTERGY OPERATIONS, INC. DOCKET NO. 50-416 1.0 INTRODUCTION By letter dated April 6, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 110960459), Entergy Operations, Inc. (Entergy, the licensee), requested changes to the inspection program for the third 10-year inservice inspection (lSI) interval for Grand Gulf Nuclear Station (GGNS), Unit 1. The proposed changes in Relief Request GG-ISI-013 would revise the inspection requirements for certain reactor pressure vessel (RPV) nozzle-to-vessel welds and nozzle inner radii from those based on American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, to an alternative based on ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds," without using the Code Case-specified visual (VT-1) examination. 2.0 REGULATORY EVALUATION Inservice inspection (lSI) of the ASME Code Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code and applicable addenda as a way to detect anomaly and degradation indications so that structural integrity of these components can be maintained. This is required by Title 10 of the Code ofFederal Regulations (10 CFR) 50.55a(g), except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). The regulations in 10 CFR 50.55a(a)(3) state that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if: (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. For all RPV nozzle-to-vessel shell welds and nozzle inner radii, ASME Code,Section XI, requires 100 percent inspection during each 10-year lSI interval. However, ASME Code Case N-702 proposes an alternative which reduces the inspection of RPV nozzle-to-vessel shell welds and nozzle inner radius areas from 100 percent to 25 percent of the nozzles for each Enclosure nozzle type during each 10-year interval. In its safety evaluation (SE) dated December 19, 2007 (ADAMS Accession No. ML073600374), the NRC approved the BWRVIP-108 report, "BWRVIP-108: BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radii," which contains the technical basis supporting ASME Code Case N-702. The NRC staff's SE for the BWRVIP-108 report specified plant-specific requirements which must be satisfied by applicants who propose to use ASME Code Case N* 702. Similar applications have been approved for several plants, including Dresden Nuclear Power Station, Units 2 and 3, Hope Creek Generating Station, and Cooper Nuclear Station. TECHNICAL EVALUATION The following plant-specific requirements are specified in the NRC staffs SE for the BWRVIP-108 report supporting use of the ASME Code Case N-702: [E]ach licensee should demonstrate the plant-specific applicability of the BWRVIP-108 report to their units in the relief request by showing that all the following general and nozzle-specific criteria are satisfied: the maximum RPV heatup/cooldown rate is limited to less 115 For recirculation inlet nozzles (pr/t)/CRPV < 1.15 p = RPV normal operating r = RPV inner t = RPV wall thickness, CRPV = 19332 ... [p(r/ + r?)/ (ro2 -rj2)]/CNOZZLE < 1.15 p = RPV normal operating r0 = nozzle outer rj = nozzle inner radius, CNOZZLE = 1637 ... For recirculation outlet nozzles (pr/t)/CRPV < 1.15 p = RPV normal operating r = RPV inner t = RPV wall thickness, CRPV= 16171 ... ;

-3 p =RPV normal operating ro =nozzle outer rj =nozzle inner radius, CNOZZLE = 1977 This plant-specific information was required by the NRC staff to ensure that the probabilistic fracture mechanics (PFM) analysis documented in the 8WRVIP-108 report applies to the RPV of the applicant's plant. 3.1 Licensee Component(s) for which Alternative is Requested (ASME Code Class Reactor Recirculation Inlet Nozzles -N2A. N2B, N2C, N2D, N2E, N2F, N2G, N2H, N2J, N2M, and N2N Main Steam Nozzles -N3A, N38, N3C, and N3D Core Spray Nozzles -N5A and N5B RPV Head Nozzles -N6A, N6B, N6C, N7 and N8 Jet Pump Instrumentation Nozzles -N9A and N98 Note that the feedwater nozzles and control rod drive return nozzles were not included in the licensee's request. Examination Category B-D, Full Penetration Welded Nozzles in Vessels Examination Item Number 83.90, "Nozzle-to-Vessel Welds" and 83.100, "Nozzle Inside Radius Section" ASME Code Requirement for which Alternative is Requested (as stated by the licensee) [The 2001 Edition 2003 Addenda (The applicable lSI Code of Record for the third 10-year lSI interval for GGNS, Unit 1)] of ASME [Code,] Section XI, Table IW8-2500-1, "Examination Category 8-D, Full Penetration Welds of Nozzles in Vessels -Inspection Program 8": Item 83.90 -Requires a volumetric examination of Reactor Vessel Vessel Welds. Item 83.100 -Requires a volumetric examination of Reactor Vessel Nozzle Inside Radius Sections.

Proposed Alternative to the ASME Code (as stated by the licensee) Pursuant to 10 CFR 50.55a(a)(3)(i), Entergy requests an alternative from performing the ASME Code required examinations on 100% of the vessel assemblies and nozzle inner radius sections identified in Tables 1 and 2[1), respectively. Specifically, Entergy proposes to adopt ASME Code Case N-702, which allows examination of a minimum of 25% of the nozzle-to-vessel welds and nozzle inner radius sections, including at least one nozzle from each system and nominal pipe size. For each of the identified nozzle assemblies, both the inner radius and the nozzle-to-vessel weld will be examined. ASME Code Case N-702 stipulates that the VT-1 visual examination method may be used in lieu of the volumetric examination method for the inner radius sections (Item No. B3.1 00). GGNS adopted ASME Code Case N-648-1, Alternative Requirements for Inner Radius Examination of Class 1 Reactor Vessel Nozzles, with the provisions stipulated in Regulatory Guide 1.147 [Revision 15, "I nservice Inspection Code Case Acceptability, ASME Section XI, Division 1," October 2007 (ADAMS Accession No. ML072070419)] in the GGNS Third Interval lSI Program Plan. [ASME] Code Case N-648-1 contains a similar allowance; therefore, GGNS may perform examinations on inner radius sections with either the VT-1 or the volumetric examination method. Bases for Alternative In Section 5.0, "Plant Specific Applicability," of the NRC staff's SE for the BWRVIP-1 08 report, the NRC stated, in part, that Licensees who plan to request relief from the ASME Code,Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRVIP-108 report as the technical basis for the use of ASIV1E Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability for the BWRVI P-1 08 report to their units in the relief request by showing that all the following general and specific criteria. [Criteria listed in Section 3.0 above.] Entergy performed this demonstration in Attachment 2121 Criterion l' the maximum RPV heatup/cooldown rate is less than 115 OF/hour, Maximum heatup and cooldown rates are limited to s 100 OF in any hour period, in accordance with GGNS Technical Specification Surveillance Requirement 3.4.11.1. Refers to Tables 1 and 2 of the licensee's April 6, 2011, submittal. Tables 1 and 2 are not included in this SE. Refers to Attachment 2 of the licensee's April 6, 2011, submittaL Attachment 2 is not included in this SF 2

-5 Criteria 2 and 3: for recirculation inlet nozzles, (pr/t)/CRPV < 1.15; the calculation for GGNS, Unit 1 recirculation inlet nozzles results in 0.9296 which is less than 1.15 and satisfies Criterion 2. [p(ro2 +rj2)/(ro2-rj2)]/CNOZZLE < 1.15, the calculation for GGNS, Unit 1 nozzles results in 0.9598 which is less than 1.15 and satisfies Criterion 3. Criteria 4 and 5: for recirculation outlet nozzles, (pr/t)/CRPV < 1.15, the calculation for GGNS, Unit 1 recirculation outlet nozzles results in 1.104 which is less than 1.15 and satisfies Criterion 4. [p(ro2 +rj2)/(ro2-rj2)]/CNOZZLE < 1.15, the calculation for the GGNS, Unit 1 nozzles results in 0.977 which is less than 1.15 and satisfies Criterion 5. Based upon the above information, the licensee concluded that it has established applicability of BWRVIP-108 to GGNS, Unit 1 and the proposed use of ASME Code Case N-702 provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(a)(3)(i) for all applicable RPV nozzle-to-vessel weld and nozzle inside radius, excluding the N04 nozzles (feedwater) and N10 nozzles (control rod drive return nozzles), listed in Tables 1 and 2 of the request. Period of Application Third 1 O-year inspection interval (June 2008 to June 2017). NRC Staff Evaluation The NRC staff's SE for the BWRVIP-108 report specified five plant-specific criteria that licensees must meet to demonstrate that the BWRVI P-1 08 report results apply to their plants. The five criteria are related to the driving force of the PFM analyses for the recirculation inlet and outlet nozzles. It was stated in the NRC staff's SE for the BWRVIP-108 report that the nozzle material fracture toughness-related reference temperature (RT NDT) used in the PFM analyses were based on data from the entire fleet of BWR RPVs. Therefore, the BWRVIP-108 report PFM analyses are bounding with respect to fracture resistance, and only the driving force of the underlying PFM analyses needs to be evaluated. It was also stated in the NRC staff's that, except for the RPV heatup/cooldown rate, the plant-specific criteria are for the recirculation inlet and outlet nozzles only because the probabilities of failure, P(FIE)s, for other nozzles are an order of magnitude lower. The plant-specific heatup/cooldown rate that the NRC staff established in Criterion 1 regards the rate under the plant's normal operating condition, which is limiting. Events with excursions of heatup/cooldown rates exceeding 115 degrees Fahrenheit per hour (OF/hour) are considered as transients. According to the NRC staff's SE for the BWRVIP-108 report, the PFM results with a very severe low temperature overpressure transient is not limiting, largely because the event frequency for that transient is 1 x1 0-3 as opposed to 1.0 for the normal operating condition.

-6 In its submittal dated April 6, 2011, the licensee provided Entergy's plant-specific data for the GGNS, Unit 1 RPV and its evaluation of the five driving-force factors, or ratios, against the criteria established in the NRC staff's SE for the BWRVIP-10S report. The staff verified the licensee's evaluation, which indicated that all criteria are satisfied. As a result, the reduced inspection requirements in accordance with ASME Code Case N-702 apply to all proposed GGNS, Unit 1 RPV nozzles, and the NRC staff concluded that the licensee's proposed alternative for all GGNS, Unit 1 RPV nozzles included in this application (see Section 3.1 of this SE) provides an acceptable level of quality and safety The NRC staff notes that the RPV feedwater nozzles and control rod drive return line nozzles are outside the scope of ASME Code Case N-702 and, accordingly, are outside the scope of this application. ASME Code Case N-702 permits a VT-1 visual examination of the nozzle inner radius without performing a sensitivity demonstration of detecting a 1-mil width wire or crack. This is not consistent with the NRC position established in Regulatory Guide, Revision 15 regarding ASME Code Case N-64S-1, ""Alternative Requirements for Inner Radius Examination of Class 1 Reactor Vessell\lozzles,Section XI, Division 1." However, since the licensee's proposed alternative indicated that Entergy is currently using ASME Code Case N-64S-1, subject to the conditions provided in RG 1.147, Revision 15, for examinations of all nozzle inner radii, the inconsistency between ASME Code Case N-702 and the NRC position regarding VT-1 is not an issue in this application and is, therefore, acceptable. 4.0 CONCLUSIOI\I The NRC staff has reviewed the submittal regarding the licensee's evaluation of the five plant-specific criteria specified in the December 19,2007, SE for the BWRVIP-10S report, which provides technical bases for use of ASME Code Case N-702, to examine RPV nozzle-to-vessel welds and nozzle inner radii at GGNS, Unit 1. Based on the evaluation in Section 3.2 of this SE, the NRC staff concluded that the licensee's proposed alternative provides an acceptable level of quality and safety and applies to all requested GGNS, Unit 1 RPV nozzles, with the exception of feedwater nozzles and control rod drive return nozzles. The NRC staff also concludes the licensee's adoption of ASME Code Case N-64S-1 consistent with the NRC position stipulated in R.G. 1.147 provides reasonable assurance of structural integrity of the nozzles' inner radii. 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)(i) and is in compliance with the ASME Codes' requirements. Therefore, the NRC authorizes the licensee's proposed alternative for inspection of the RPV nozzle-to-vessel shell welds and nozzle inner radii sections listed in Attachment 1 of the licensee's April 6, 2011, submittal, with the exception of feedwater nozzles and control rod drive return nozzles, for GGNS, Unit 1 through the end of the third 1 O-year lSI interval, which ends in June 2017.

7 All other ASME Code,Section XI requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector. Principal Contributor: Simon Sheng Date: November 4, 2011

-2 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 Docket No. 50-416 Safety cc w/encl: Distribution via LPLIV RidsAcrsAcnw MailCTR RidsNrrDeEvib RidsNrrDorlDpr RidsNrrDorlLpl4 RidsNrrLAJBurkhardt RidsNrrPMGrandGulf RidsRgn4MailCenter JMcHale, EDO SSheng, ADAMS Accession No. ML112710328 Sincerely, IRAI Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation *SE email dated ,*.*.," OFFICE NRR/LPL4/PM NAME AWang JBurkhardt 10/19/11 10/17/11 OFFICIAL