ML14148A262

Grand Gulf Nuclear Station, Unit 1, Relief Request GG-ISI-017, Alternative to Reactor Vessel Internal Structure Examination, Use of BWRVIP Guidelines, Third 10 -Year Inservice Inspection Interval (TAC No. MF2357)

ML14148A262
Person / Time
Site:	Grand Gulf
Issue date:	06/30/2014
From:	Broaddus D A Plant Licensing Branch IV
To:	Entergy Operations
Wang A B
References
TAC MF2357
Download: ML14148A262 (13)

Text

Vice President, Operations Entergy Operations, Inc. Grand Gulf Nuclear Station P.O. Box 756 Port Gibson, MS 39150 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 30, 2014 SUBJECT: GRAND GULF NUCLEAR STATION, UNIT 1 -REQUEST FOR RELIEF GG-ISI-017, ALTERNATIVE TO USE BOILING WATER REACTOR VESSEL AND INTERNALS PROJECT GUIDELINES IN LIEU OF SPECIFIC ASME CODE REQUIREMENTS (TAC NO. MF2357) Dear Sir or Madam: By letter dated June 27, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 13179A041), as supplemented by letters dated January 29, 2014 and April14, 2014 (ADAMS Accession Nos. ML 14029A592 and ML 14105A456, respectively), Entergy Operations, Inc. (the licensee), submitted Relief Request GG-ISI-017 for its third 1 0-year interval in service inspection (lSI) program plan for its reactor vessel internals components at Grand Gulf Nuclear Station, Unit 1 (GGNS). In this Relief Request, the licensee proposed to use Boiling Water Reactor Vessel and Internals Project (BWRVIP) guidelines as an alternative to certain requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) for lSI of reactor pressure vessel interior surfaces, attachments, and core support structures. This Relief Request is requested pursuant to paragraph 50.55(a)(3)(i) of Title 10 of the Code of Federal Regulations (1 0 CFR). The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the GG-ISI-017 and concludes, as set forth in the enclosed safety evaluation (SE), that the alternatives proposed by the licensee, as summarized in the Attachment to the SE, will ensure that the integrity of the reactor pressure vessel interior surfaces, attachments, and core support structures is maintained with an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the licensee's proposed alternative for GGNS is authorized by the NRC staff for the third 10-year lSI interval at GGNS, which ends on July 1, 2017. All other requirements of the ASME Code,Section XI for which an alternative has not been specifically requested remain applicable, including third-party review by the Authorized Nuclear In-service Inspector. Any ASME Code,Section XI, RVI components that are not included in this request for alternative are to continue to be inspected in accordance with the ASME Code,Section XI requirements. The NRC staff notes that if the licensee intends to take exceptions to, or deviations from, the NRC staff-approved BWRVIP inspection guidelines (specifically, those inspection requirements listed in the Attachment to the SE), this will require the licensee to revise and re-submit this request for alternative. The licensee shall obtain staff approval for such exceptions prior to

-2-implementing the revised inspection guidelines for the GGNS unit's reactor pressure vessel interior surfaces, attachments, and core support structures. The NRC staff's safety evaluation is enclosed. If you have any questions, please contact Mr. Alan Wang at 301-415-1445 or via e-mail at Alan.Wang@nrc.gov. Docket No. 50-416 Enclosure: Safety Evaluation cc w/encl: Distribution via Listserv Sincerely, Douglas A. Broaddus, Chief Plant Licensing IV-2 and Decommissioning Transition Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 RELIEF REQUEST GG-ISI-017 TO ALLOW USE OF BOILING WATER REACTOR VESSEL AND INTERNALS PROJECT GUIDELINES 1.0 INTRODUCTION ENTERGY OPERATIONS. INC. GRAND GULF NUCLEAR STATION. UNIT 1 DOCKET NO. 50-416 By letter dated June 27, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 13179A041), as supplemented by letters dated January 29, 2014 and April14, 2014 (ADAMS Accession Nos. ML 14029A592 and ML 14105A456, respectively), Entergy Operations Inc., (the licensee) submitted Relief Request GG-ISI-017 for its third 1 0-year interval inservice inspection (lSI) program plan for its reactor vessel internals (RVI) components at Grand Gulf Nuclear Station, Unit 1 (GGNS). In this safety evaluation (SE), the term "RVI components" includes reactor pressure vessel interior surfaces, attachments, and core support structures. In the Relief Request, the licensee proposed to use Boiling Water Reactor Vessel and Internals Project (BWRVIP) guidelines as an alternative to certain requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) for lSI of reactor pressure vessel interior surfaces, attachments, and core support structures. 2.0 REGULATORY REQUIREMENTS The lSI of ASME Code Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code as required by 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 10 CFR 50.55a(g) may be used, when authorized by the U.S. Nuclear Regulatory Commission 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. Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, "Rules for lSI of Nuclear Power Plant Components," to the extent practical within the limitations of design, Enclosure

-2-geometry, and materials of construction of the components. The regulations require that lSI examination of components and system pressure tests conducted during each 1 0-year interval 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. The applicable ASME Code of record for the third 1 0-year lSI interval for GGNS is ASME Code,Section XI, 2001 Edition through the 2003 Addenda. 3.0 LICENSEE'S EVALUATION The Components for Which an Alternative is Requested The ASME Code,Section XI, Class 1, Examination Categories B-N-1 and B-N-2, Code Item Numbers B13.1 0, Vessel Interior; B13.20, Interior Attachments within Beltline Region; B13.30, Interior Attachments Beyond Beltline Region; and B 13.40, Core Support Structure. Examination Requirements from Which an Alternative is Requested The ASME Code,Section XI requires a visual examination (VT) of certain RVI components. These examinations are included in Table IWB-2500-1, Categories B-N-1 and B-N-2, and identified with the following item numbers:

• B13.1 0 -Examine accessible areas of the reactor vessel (RV) interior during each period using a technique which meets the requirements for a VT-3 examination, as defined in paragraph IWA-2213 of Section XI of the ASME Code.

• B13.20 -Examine interior attachment welds within the RV beltline region during each interval using a technique which meets the requirements for a VT-1 examination, as defined in paragraph IWA-2211 of Section XI of the ASME Code.

• B13.30-Examine interior attachment welds outside of the beltline region during each interval using a technique which meets the requirements for a VT-3 examination, as defined in paragraph IWA-2213 of Section XI of the ASME Code.

• B13.40-Examine surfaces of core support structures during each interval using a technique which meets the requirements for a VT-3 examination, as defined in paragraph IWA-2213 of Section XI of the ASME Code. These examinations are performed to periodically assess the structural integrity of the reactor pressure vessel interior surfaces, attachments, and core support structures. Licensee's Basis for Requesting an Alternative and Justification for Granting Relief In its letter dated June 27, 2013, the licensee, in lieu of ASME Section XI, Code requirements, submitted an alternative inspection program per the BWRVIP guidelines for Category B-N-1 and B-N-2 reactor pressure vessel interior surfaces, attachments, and core support structures at GGNS. The licensee stated that implementation of the alternative inspection program will maintain an adequate level of quality and safety of the affected welds and components and will

-3-not adversely impact the health and safety of the public. As part of its justification for the relief, the licensee stated that boiling-water reactors (BWRs) now examine the reactor pressure vessel interior surfaces, attachments, and core support structures in accordance with BWRVIP inspection and evaluation (I&E) guidelines in lieu of ASME Code,Section XI criteria. The BWRVIP I&E guidelines are more stringent and comprehensive than the requirements set forth in Section XI of the ASME Code. The proposed alternative includes examination methods, examination volume, frequency, training and successive and additional examinations, flaw evaluations, and reporting. The BWRVIP guidelines were written to address the examination of safety significant RVI components using appropriate methods and reexamination frequencies. Furthermore, the licensee stated that relief from the examinations specified in Table IWB-2500-1 of the ASME Code,Section XI are requested pursuant to 10 CFR 50.55a(a)(3)(i). Alternative Examination In lieu of the requirements specified in Section XI of the ASME Code, the licensee proposed to examine the GGNS RVI components in accordance with BWRVIP I&E guideline requirements. In its submittal, the licensee included only those RVI components that are categorized under the jurisdiction of Section XI of the ASME Code. The following BWRVIP reports contain the relevant I&E guidelines for RVI surfaces, attachments, and core support structures. Furthermore, the licensee clarified that not all RVI components listed in the following BWRVIP reports are ASME Code,Section XI components.

• BWRVIP-03, "BWRVIP Reactor Pressure Vessel and Internals Examination Guidelines"

• BWRVIP-18, Revision 1, "BWRVIP Core Spray Internals Inspection and Flaw Evaluation Guidelines"

• BWRVIP-25, "BWRVIP Core Plate Inspection and Flaw Evaluation Guidelines"

• BWRVIP-26-A, "BWRVIP Top Guide Inspection and Flaw Evaluation Guidelines"

• BWRVIP-27-A, "BWRVIP BWR Standby Liquid Control System/Core Plate Delta P Inspection and Flaw Evaluation Guidelines"

• BWRVIP-38, "BWRVIP Shroud Support Inspection and Flaw Evaluation Guidelines"

• BWRVIP-41, Revision 3, "BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines"

• BWRVIP-42, Revision 1, "Low Pressure Coolant Injection (LPCI) Coupling Inspection and Flaw Evaluation Guidelines"

• BWRVIP-47-A, "BWR Lower Plenum Inspection and Flaw Evaluation Guidelines"

• BWRVIP-48-A, "VesseiiD [Inside Diameter] Attachment Weld Inspection and Flaw Evaluation Guidelines"

-4-* 8WRVIP-76, Revision 1, "8WR Core Shroud Inspection and Flaw Evaluation Guidelines"

• 8WRVIP-94, Revision 2, 8WRVIP Program Implementation Guide"

• 8WRVIP-100-A, "Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for 8WR Core Shroud" The licensee stated that the inspection services from an Authorized Inspection Agency will be applied to the proposed alternative. The licensee further indicated that the 8WRVIP has established reporting protocol for examination results and deviations that are consistent with the guidelines of 8WRVIP-94. The licensee clarified that its use of any revision to a 8WRVIP report would satisfy all of the relevant portions of the prior revision unless NRC staff approval is obtained prior to its implementation. In Table 1 of Attachment 1 of its submittal dated June 27, 2013, the licensee provided a comparison of the ASME Code,Section XI, examination requirements for Category 8-N-1 and 8-N-2 categories for RVI surfaces, attachments, and core support structures with the previously identified 8WRVIP I&E Guideline documents. In Attachment 2 of the submittal, the licensee provided additional information regarding the 8WRVIP inspection requirements for welds of the RVI surfaces, attachments, and core support structures and their subcomponents. As an example, the following excerpt from the Attachment 2 of the licensee's submittal indicates the applicable ASME Code,Section XI category/item numbers that are applicable to some of the GGNS RVI components:

• Core Spray Piping -Item No. 813.10

• Jet Pump-Item No. 813.20

• Core Shroud-Item No. 813.30

• Core Shroud Support and Core Support Structure-Item No. 813.40 The licensee stated that the above examples demonstrate that the inspection techniques recommended by the 8WRVIP I&E guidelines are superior to the inspection techniques mandated by the ASME Code,Section XI lSI program. Additionally, these examples supported the licensee's assertions that the 8WRVIP I&E guidelines require more frequent inspections of some RVI components than the corresponding ASME Code,Section XI requirements. The licensee stated that by implementing the 8WRVIP I&E guidelines, the aging degradation of the RVI surfaces, attachments, and core support structures can be identified in a timely manner so that proper corrective action can be taken to maintain the integrity of the applicable components. Therefore, the licensee concluded that implementation of the 8WRVIP I&E guidelines for the GGNS RVI surfaces, attachments, and core support structures would provide an acceptable level of quality and safety. The licensee's proposed alternative for the RVI components and subcomponents covered under the scope of this alternative request is summarized in the Attachment 1 to this SE.

-5-4.0 STAFF EVALUATION The NRC staff reviewed the information provided by the licensee in its submittal dated June 27, 2013, as supplemented by letters dated January 29, 2014, and April14, 2014, regarding its proposed alternative to the ASME Code,Section XI lSI requirements and the technical bases for the licensee's proposed alternatives. The staff reviewed the status of each of the referenced BWRVIP guidance documents to provide effective aging management and found application of all of the referenced BWRVIP reports to be acceptable, provided that the conditions associated with each BWRVIP report are implemented. The following paragraphs address the NRC staff's requests for additional information (RAis) dated December 26, 2013 (ADAMS Accession No. ML 13361A027), the licensee's responses, and the staff's evaluation of the licensee's responses to the RAis. In RAI-1, the NRC staff requested that the licensee identify whether there are any sensitized stainless steel vessel attachment welds associated with RVI components at GGNS. Furnace-sensitized stainless steel welds tend to be more susceptible to intergranular corrosion cracking (IGSCC). In its response dated April14, 2014, the licensee stated that some internal attachment welds (e.g., the jet pump riser brace and the steam dryer hold-down bracket attachment welds), were furnace-sensitized at GGNS. The licensee stated that it will continue to comply with the BWRVIP-48-A I&E guidelines to monitor any IGSCC in these welds. Because inspection guidelines for furnace-sensitized stainless steels are addressed in the BWRVIP-48-A report, the NRC staff concludes that the licensee's response is acceptable and this issue is closed. In RAI-2, the NRC staff requested that the licensee provide an explanation for not including the following specific BWRVIP reports, which are used to monitor active aging degradation in jet pumps, the steam dryer, and the top guide. None of the following components are addressed by ASME Code,Section XI requirements:

• BWRVIP-138, Revision 1, "BWRVIP Updated Jet Pump Beam Inspection and Flaw Evaluation Guidelines";

• BWRVIP-139, "BWR Vessel Internals Project, Steam Dryer Inspection and Flaw Evaluation Guidelines"; and

• BWRVIP-183, "BWRVIP, Top Guide Grid Beam Inspection and Flaw Evaluation." In its January 29, 2014, response to RAI-2, the licensee stated that the aforementioned BWRVIP reports are used at GGNS as part of its inspection program for non-ASME Code,Section XI components as follows:

• For jet pump beams, which are, non-ASME Code,Section XI components at GGNS, the licensee stated that BWRVIP-138, Revision 1 is used to monitor aging degradation in jet pump beams.

• For steam dryer assemblies, which are non-ASME Code,Section XI components at GGNS, the licensee stated that BWRVIP-139 is used to monitor aging degradation in steam dryer assemblies.

-6-* For the top guide (code and non-code) the licensee stated that it will perform inspections of the top guide consistent with the guidelines addressed in BWRVIP-183. The NRC staff concludes that the licensee's response to RAI-2 is acceptable because implementation of l&E guidelines addressed in the aforementioned BWRVIP reports for the RVI components is effective in identifying aging effects in a timely manner and, therefore, this issue is closed. In its December 26, 2013 response to RAI-3, the licensee confirmed that it used NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," November 1980 (ADAMS Accession No. ML031600712}, as a basis for its lSI examinations offeedwater nozzles. Six feedwater nozzle inner radii are part of this examination at GGNS. The control rod drive line nozzle was cut and capped and, therefore, no examinations were performed on this nozzle. BWRVIP-48-A is used with enhanced visual testing (EVT-1) to examine the feedwater piping brackets and feedwater sparger tee welds. The NRC staff has concluded that this response is acceptable because the licensee is implementing NUREG-0619 and BWRVIP-48-A for the subject welds and this issue is closed. Attachment 3 of the licensee's June 27, 2013, submittal stated that one indication with characteristics of IGSCC and irradiation-assisted stress-corrosion cracking (IASCC) was observed in core shroud circumferential weld H4. Based on the operating experience thus far, it was established that IGSCC and IASCC are two active aging mechanisms in core shroud welds. For the core shroud, the licensee follows the guidance in BWRVIP-76, Revision 1 at GGNS. Therefore, the licensee is monitoring the growth of this indication in accordance with this guideline. The NRC staff concludes that this response is acceptable because the licensee is following the BWRVIP-76 guidelines and this issue is closed. Nickel alloy welds fabricated with Alloy 182 are more susceptible to IGSCC compared to stainless steel welds. In its January 29, 2014, response to RAI-4, the licensee identified Alloy 182 welds in the following RVI components at GGNS: (1) guide rod brackets; (2) core spray brackets; (3) steam dryer support brackets; (4) feedwater sparger brackets; (5) shroud support (weld H9); and (6) shroud support legs (weld H12). In its response to RAI-4, the licensee stated that it inspected these RVI components in accordance with BWRVIP-38 and BWRVIP-48-A. To date, no cracking has been identified in any of the associated Alloy 182 welds at GGNS. Based on the above, the NRC staff concludes that this response is acceptable and this issue is closed. The NRC staff reviewed the previous inspection results for GGNS for the various RVI components that were included in Attachment 3 of the licensee's June 27, 2013, submittal. In addition, in RAI-5, the staff requested that the licensee address how the aging degradation is monitored in the most susceptible weld connections in the RVI components for the (1) top guide; (2) core spray line and spargers; (3) low pressure coolant injection (LPCI) coupling; (4) jet pumps; (5) in-core dry tubes; and, (6) control rod drive (CRD) housings. In its January 29, 2014, response to RAI-5, the licensee stated that the top guide was inspected in accordance with BWRVIP-183 in 2012 and following that guidance, it will be inspected with EVT -1 every 6 years.

-7-For the core spray line and spargers, the LPCI coupling, the jet pumps, the in-core dry tubes, and the CRD housings, the licensee stated that these components are outside of the scope of the subject relief request because they are not included in the ASME Code,Section XI lSI program. The licensee stated that the relevant BWRVIP I&E guideline reports (BWRVIP-18, -41, -42, and -47) for these items will be used to monitor relevant aging effects. With respect to the in-core dry tubes, the licensee stated that all four in-core dry tubes were replaced in 2012. The NRC staff concludes that the licensee's response to RAI-5 for these RVI components is acceptable because they were inspected in accordance with BWRVIP I&E guidelines, which were previously approved by the staff and, therefore, this issue is closed. In its January 29, 2014, response to RAI-6, the licensee addressed its implementation of On-line Noble Chemical Addition (OLNC) at GGNS, which includes hydrogen water chemistry (HWC) and noble metal chemical addition (NMCA). To support its statement that implementation of OLNC is effective, the licensee provided measured values of electrochemical potential (ECP) for the OLNC-treated stainless steel coupon and the amount of platinum deposition for a test coupon. The NRC staff reviewed the ECP values and confirmed that implementation of HWC and NMCA is consistent with BWRVIP-62, Revision 0, "BWR Vessel and Internals Project, Technical Basis for Inspection Relief for BWR Internal Components with Hydrogen Injection." The staff is currently reviewing BWRVIP-62, Revision 1, which provides the technical basis for application of OLNC. Therefore, the staff has not completed its review of the licensee's information related to platinum loading. Even though the staff did not review the licensee's platinum loading value, the staff considers that OLNC implementation is effective because: (1) the ECP value is a primary essential variable to measure the effectiveness of HWC and NMCA and the licensee's low ECP value confirms that the implementation of OLNC is effective at GGNS; (2) the low ECP value also confirms adequate availability of hydrogen and platinum in RVI components; and (3) lower values of platinum loading are expected for OLNC compared to NMCA and the licensee's measured platinum deposition for a test coupon indicates that the ECP values are lower than the maximum allowed value addressed in BWRVIP-62, Revision 0. This indicates that ONLC will be effective in mitigating stress-corrosion cracking in some of the RVI components. Based on the above, the NRC staff concludes that the response to RAI-6 is acceptable and, therefore, this issue is closed. The Attachment to this SE includes attributes related to inspection techniques and frequency of inspections for various RVI components at GGNS. A comparison of the required ASME Code,Section XI, Category B-N-1 and B-N-2 examination requirements with the current BWRVIP I&E Guideline requirements that are applicable to GGNS is also included in the Attachment. Attachment 3 of the licensee's June 27, 2013, submittal addressed previous inspection results of RVI components. Based on the information provided, the NRC staff concludes that most RVI component inspections did not reveal the presence of any indications at GGNS. For those RVI components where indications were found, the licensee implemented corrective action program measures as recommended by the relevant BWRVIP reports to provide reasonable assurance that the aging degradation mechanisms will be adequately monitored during the third lSI interval at GGNS. The BWRVIP I&E guidelines require more frequent inspections than ASME Code,Section XI criteria for RVI components that are susceptible to aging degradation mechanisms. Therefore, subsequent inspections of the RVI components per the relevant BWRVIP I&E guidelines will provide reasonable assurance that any emerging aging effects will be identified in

-8-a timely manner. In addition, frequent inspections in accordance with the BWRVIP I&E guidelines will enable the licensee to effectively monitor existing aging degradation in RVI surfaces, attachments, and core support structures. Therefore, the NRC staff concludes that known active aging degradation mechanisms in the GGNS RVI components are adequately monitored by the appropriate BWRVIP I&E guidelines at GGNS. Consistent with the determination in the NRC staff's SEs that approved a majority of the previously-cited BWRVIP guideline documents, as supplemented by the NRC staff-approved inspection guidelines for the feedwater nozzle and sparger welds, the licensee's proposed alternative is expected to identify aging degradation of the RVI components in a timely manner. Therefore, the NRC staff concludes that the implementation of the inspection parameters specified in the licensee's proposed alternative will continue to ensure that the integrity of the RVI components is maintained with an acceptable level of quality and safety. 5.0 CONCLUSION Based on the information provided in the licensee's submittals, the NRC staff concludes that the alternatives proposed by the licensee as summarized in the Attachment to this SE, will ensure that the integrity of the RVI surfaces, attachments, and core support structures is maintained with an acceptable level of quality and safety. The NRC staff reviewed the status of each of the referenced BWRVIP guidance documents to determine if they will provide effective aging management and concludes that the application of all of the referenced BWRVIP reports (included in the proposed alternative) to be acceptable for this purpose, provided that the conditions associated with each BWRVIP report are implemented. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the licensee's proposed alternative for GGNS is authorized for the third 1 0-year lSI interval, which ends on July 1, 2017. All other requirements of the ASME Code,Section XI for which an alternative has not been specifically requested remain applicable, including third-party review by the Authorized Nuclear In-service Inspector. Any ASME Code,Section XI, RVI components that are not included in this request for alternative are to continue to be inspected in accordance with the ASME Code,Section XI requirements. The NRC staff notes that if the licensee intends to take exceptions to, or deviations from, the NRC staff-approved BWRVIP inspection guidelines (specifically, those inspection requirements listed in Attachment to this SE), this will require the licensee to revise and re-submit this request for alternative. The licensee shall obtain staff approval for such exceptions prior to implementing the revised inspection guidelines for the GGNS unit's reactor pressure vessel interior surfaces, attachments, and core support structures. Principal Contributor: G. Cheruvenki Date: Attachment: Comparison of ASME Category B-N-1 and B-N-2 Requirements with BWRVIP Guidance Requirements ASME Item No. Table IWB-2500-1 813.10 813.20 B13.30 -*-*-----------ATTACHMENT Comparison of ASME Category B-N-1 and B-N-2 Requirements With BWRVIP Guidance Requirements <1> ASME Exam ASME ASME Applicable BWRVIP BWRVIP BWRVIP Component Scope Exam Frequency Document Exam Scope Exam BWRVIP Frequency Reactor Vessel Interior Accessible VT-3 Each period BWRVI P-18-Revision BWRVIP examinations satisfy ASME Code, Areas 1, 25, 26-A, 27-A, 38,Section XI, VT-3 inspection requirements. (Non-41, Revision3, 42, specific) Revision 1 , 4 7 -A, 48-A, 76 Revision 1 Interior Attachments Within BWRVIP-48-A, Table Riser Brace EVT-1 1 00% in first 12 years; Beltline-Jet Pump Riser Accessible Each 10-3-2 Attachment 25% during each Braces Welds VT-1 year interval subsequent 6 years Lower Surveillance BWRVIP-48-A, Table Bracket VT-1 Each 1 0-year Interval Specimen Holder Brackets 3-2 Attachment Guide Rod Brackets BWRVIP-48-A Table Bracket VT-3 Each 1 0-year Interval 3-2 Attachment Steam Dryer Support BWRVIP-48-A Table Bracket EVT-1 Each 1 0-year Interval Brackets 3-2 Attachment Steam Dryer hold-down BWRVIP-48-A Table Bracket VT-3 Each 1 0-year Interval Brackets Accessible VT-3 Each 10-3-2 Attachment Feedwater Sparger Welds year interval BWRVIP-48-A Table Bracket EVT-1 Each 1 0-year Interval Brackets 3-2 Attachment Core Spray Piping primary BWRVIP-48-A Table Bracket EVT-1 Each 4 re-fueling and supplemental Brackets 3-2 Attachment cycles Upper Surveillance BWRVIP-48-A Table Bracket VT-3 Each 1 0-year Interval Specimen Holder Brackets 3-2 Attachment Shroud Support (Weld H9) BWRVIP-38, 3.1.3.2 Weld H9<2> EVT-1 or Maximum of 6 years including gussets where Figures 3-2 and UT for one sided EVT -1 , applicable 3-5 Maximum of 1 0 years for UT Weld H12 Shroud Support BWRVIP-38 3.2.3 Weld H12 Per When accessible Legs BWRVIP-Rarely VT-3 Each 10-38 NRC Accessible year interval SER (7-24-200}, inspect with appropriate method<3> Upper Surveillance BWRVIP-48-A Table Bracket VT-3 Each 1 0-year Interval older Brackets 3-2 Attachment

-10-ASME ASME Item No. Exam ASME ASME Applicable BWRVIP BWRVIP BWRVIP Table JWB-2500-1 Component Scope Exam Frequency Document Exam Scope Exam BWRVIP Frequency Welded Core Support-BWRVIP-38 3.1.3.2, Shroud EVT-1 or Based on as found Shroud Support Figures 3-2 and 3-5 Support UT conditions, to a H8/H91eg Maximum 6 years for welds one-sided EVT -1 , 1 0 including years for UT where gussets as accessible applicable Shrouded Vertical Welds BWRVIP-76 R1, Vertical and EVT-1 or Maximum 6 years for Section 2.3 Figure 3-3 Ring UT one-sided EVT -1 , 1 0 Accessible Each 10-Segment years for UT B13.40 VT-3 Welds as Surfaces year interval applicable Shroud Horizontal Welds BWRVIP-76 2.2, Welds H1-H7 EVT-1 or Based on as found Figure 2.2 to 2.5 as applicable UT conditions to a maximum 10 years for UT when inspected from both sides of the welds Shroud Repairs (4> BWRVIP-76, R1 Tie-Rod VT-3 Per Designer Section 3.5 Repair recommendations per BWRVIP-71 R1 NOTES: (1) This Table provides only an overview of the requirements. For more details, refer to the ASME Code,Section XI, Table IWB-2500-1, and the appropriate BWRVIP document. (2) In accordance with Appendix A of BWRVIP-38, a site-specific evaluation will determine the minimum required weld length to be examined. (3) When inspection tooling and methodologies are available, they will be used to establish a baseline inspection of these welds. (4) No repairs have been performed on the shroud.

-2 -implementing the revised inspection guidelines for the GGNS unit's reactor pressure vessel interior surfaces, attachments, and core support structures. The NRC staff's safety evaluation is enclosed. If you have any questions, please contact Mr. Alan Wang at 301-415-1445 or via e-mail at Alan.Wang@nrc.gov. Docket No. 50-416 Enclosure: Safety Evaluation cc w/encl: Distribution via Listserv DISTRIBUTION: PUBLIC LPL4-2 r/f RidsAcrsAcnw_MaiiCTR Resource RidsNrrDeEvib Resource RidsNrrDoriDpr Resource RidsNrrDoriLpl4-2 Resource RidsNrrLAJBurkhardt Resource RidsNrrPMGrandGulf Resource RidsRgn4MaiiCenter Resource GCheruvenki, DE/EVIB ADAMS Accession No ML 14148A262 Sincerely, IRA/ Douglas A. Broaddus, Chief Plant Licensing IV-2 and Decommissioning Transition Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation NRR-028 *SE via email OFFICE NRR/DORLILPL4-2/PM NRR/DORLILPL4-2/LA NRR/DE/EVIB/BC NRR/DORLILPL4-2/BC NAME A Wang JBurkhardt SRosenberg* DBroaddus DATE 6/3/14 6/3/14 5/23/14 6/30/14 OFFICIAL AGENCY RECORD