ML14127A327
| ML14127A327 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 05/30/2014 |
| From: | Doug Broaddus Plant Licensing Branch IV |
| To: | Entergy Operations |
| Wang A | |
| References | |
| TAC MF1867 | |
| Download: ML14127A327 (14) | |
Text
,i Vice President, Operations Entergy Operations, Inc.
River Bend Station 5485 US Highway 61 N St. Francisville, LA 70775 UNITED STATES Nl.JCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 30, 2014
SUBJECT:
RIVER BEND STATION, UNIT 1-REQUEST FOR RELIEF NO. RBS-ISI-019, ALTERNATIVE TO USE BOILING WATER REACTOR VESSEL AND INTERNALS PROJECT GUIDELINES IN LIEU OF ASME CODE, SECTION XI REQUIREMENTS FOR THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL (TAC NO. MF1867)
Dear Sir or Madam:
By letter dated May 16, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13141A257), as supplemented by letter dated December 17, 2013 (ADAMS Accession No. ML13361A086), Entergy Operations, Inc. (Entergy, the licensee),
submitted Relief Request No. RBS-ISI-019 for the fourth 1 0-year interval in service inspection (lSI) program plan for its reactor vessel internals components at River Bend Station, Unit 1 (RBS). In its request, the licensee proposed to use the 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 (ASME) Code for lSI of reactor pressure vessel interior surfaces, attachments, and core support structures. This relief request is requested pursuant to Title 10 of the Code of Federal Regulations ( 10 CFR) paragraph 50.55a(a)(3)(i).
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the relief request and as set forth in the enclosed safety evaluation (SE), concludes 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 acc;eptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the licensee's proposed alternative for RBS is authorized by the NRC staff for the fourth 1 0-year lSI interval at RBS.
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 implementing the revised inspection guidelines for the RBS unit's reactor pressure vessel interior surfaces, attachments, and core support structures.
All other requirements of the ASME Code,Section XI for which an alternative has not been specifically requested remair) applicable, including third-party review by the Authorized Nuclear lnservice Inspector. Any ASME Code,Section XI, reactor vessel internals components that are not included in this request for alternative will continue to be inspected in accordance with the ASME Code,Section XI requirements.
If you have any questions, please contact Alan Wang at 301-415-1445 or via e-mail at Alan.Wang@nrc.gov.
Docket No. 50-458
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 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST RBS-ISI-019. ALTERNATIVE TO USE BWRVIP GUIDELINES IN LIEU OF ASME CODE. SECTION XI REQUIREMENTS FOR THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL
1.0 INTRODUCTION
ENTERGY OPERATIONS. INC.
RIVER BEND STATION, UNIT 1 DOCKET NO. 50-458 By letter dated May 16, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13141A257), as supplemented by letter dated December 17, 2013 (ADAMS Accession No. ML13361A086), Entergy Operations, Inc. (Entergy, the licensee),
submitted Relief Request No. RBS-ISI-019 for the fourth 10-year interval inservice inspection (lSI) program plan for its reactor vessel internals (RVI) components at River Bend Station, Unit 1 (RBS). In this relief request, the licensee proposed to use the 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 (ASME) Code for lSI of reactor pressure vessel interior surfaces, attachments, and core support structures. In this safety evaluation (SE), the term "RVI components" include reactor pressure vessel interior surfaces, attachments, and core support structures.
2.0 REGULATORY REQUIREMENTS lnservice inspection of ASME Code Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code as required by Title 10 of the Code of Federal Regulations (1 0 CFR) paragraph 50.55a(g), except where specific relief has been granted by tt'le U.S. Nuclear Regulatory Commission (NRC) 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 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. The licensee has requested this relief request pursuant to 10 CFR 50.55a(a)(3)(i).
Enclosure 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, geometry, and materials of construction of the components. The regulations require that lSI examination of components and system pressure tests conducted during the first 1 0-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. The applicable ASME Code of record for the fourth 1 0-year lSI interval for RBS is ASME Code,Section XI, 2001 Edition through the 2003 Addenda.
3.0 TECHNICAL EVALUATION
3.1 Licensee's Evaluation The Components for Which an Alternative is Requested ASME Code,Section XI, Class 1, Examination Categories B-N-1 and B-N-2, Code Item Numbers 813.10, Vessel Interior, 813.20, Interior Attachments within Beltline Region, 813.30, Interior Attachments Beyond Beltline Region, and 813.40, Core Support Structure.
Examination Requirements from Which an Alternative is Requested ASME Code,Section XI requires the 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:
B 13.10 - Examine accessible areas of the RV interior each period using a technique which meets the requirements for a VT-3 examination, as defined in paragraph IWA-2213 of the ASME Code,Section XI.
813.20- Examine interior attachment welds within the beltline region each interval using a technique which meets the requirements for a VT-1 examination as defined in paragraph IWA-2211 of the ASME Code,Section XI.
813.30- Examine interior attachment welds beyond the beltline region each interval using a technique which meets the requirements for a VT-3 examination, as defined in paragraph IWA-2213 of the ASME Code,Section XI.
813.40- Examine surfaces of the core support structure each interval using a technique which meets the requirements for a VT-3 examination, as defined in paragraph IWA-2213 of the ASME Code,Section XI.
These examinations are performed to assess the structural integrity of the reactor pressure vessel interior surfaces, attachments, and core support structures.
Basis for Requesting an Alternative and Justification for Granting Relief In its letter dated May 16, 2013, the licensee submitted an alternative inspection program per the BWRVIP guidelines for B-N-1 and B-N-2 reactor pressure vessel interior surfaces, attachments, and core support structures, in lieu of ASME Section XI, Code requirements for RBS. 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 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 proposed alternative includes examination methods, examination volume, frequency, training and successive and additional examinations, flaw evaluations, and reporting. These guidelines have been written to address the examination of safety significant RVI components using appropriate methods and reexamination frequencies. Furthermore, the licensee stated that relief from examinations 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 of the applicable Edition and Addenda of the ASME Code,Section XI, the licensee proposed to examine theRBS's RVI components. in accordance with BWRVIP guidelines. The licensee included only the RVI components (code components) that are categorized under the jurisdiction of the ASME Code,Section XI. The following BWRVIP reports include I&E guidelines for the ASME Code,Section XI, reactor pressure vessel interior 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 Guideline" 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" BWRVIP-76, Revision 1, "BWR Core Shroud Inspection and Flaw Evaluation Guidelines" SWRVIP-94, Revision 2, "SWRVIP Program Implementation Guide" SWRVIP-1 00, Revision 1, "Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Core Shroud" The licensee stated that inspection services of an Authorized Inspection Agency will be applied to the proposed alternative. The licensee further indicated that the SWRVIP has established reporting protocol for examination results and deviation that are consistent with the requirements of SWRVIP-94 report. The licensee clarified that any revised version of a SWRVIP report will meet I&E guidelines of its original version, and if it does not meet this criteria, NRC staff approval is mandatory prior to its implementation.
In Table 1, Attachment 1 of the its submittal dated May, 16, 2013, the licensee provided a comparison of the ASME Code,Section XI, examination requirements for B-N-1 and S-N-2 Categories of the reactor pressure vessel interior surfaces, attachments, and core support structures with the above current SWRVIP I&E guidelines. As an example, in Attachment 2 of the submittal dated May 16, 2013, the licensee provided additional information regarding the SWRVIP inspection requirements for the following welds of the reactor pressure vessel interior surfaces, attachments, and core support structures and their subcomponents representing each of the aforementioned ASME Code,Section XI category/item numbers (Item Numbers S13.1 0, S13.20, S13.30, and S13.40):
Core Spray Piping--S 13.10 Jet Pump--813.20 Core Shroud--S 13.30 Core Shroud Support and Core Support Structure--S 13.40.
The licensee claimed that these examples demonstrated that the inspection techniques that are recommended by the BWRVIP inspection guidelines are superior to the inspection techniques mandated by the ASME Code,Section XI lSI program. Additionally, these examples proved that the SWRVIP inspection guidelines require more frequent inspections of some RVI components than the corresponding ASME Code,Section XI lSI program. The licensee stated that by implementing.the BWRVIP inspection guidelines the aging degradation of the reactor pressure vessel interior surfaces, attachments, and core support structures can be identified in a timely manner so that proper corrective action can be taken to restore the integrity of the applicable component. Therefore, the licensee concluded that implementation of the BWRVIP inspection guidelines for the RSS reactor pressure vessel interior 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 to this SE.
3.2
NRC Staff Evaluation
The NRC staff reviewed the information provided by the licensee in its submittal dated May 16, 2013, as supplemented by letter dated December 17, 2013, regarding its proposed alternatives 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 which provide effective aging management program (AMP) and found all of the referenced BWRVIP reports to be acceptable, with any additional conditions associated with the implementation of the subject BWRVIP reports outlined in the corresponding NRC staff SE for that report. The staff did, however, identify some issues. which required additional clarification by the licensee. The following addresses the NRC staff's request for additional*
information (RAI) dated November 14, 2013 (ADAMS Accession No. ML13319A109), the licensee's RAI responses, and the staff's evaluation of the RAI responses.
In RAI-1, the NRC staff requested that the licensee identify whether there are any furnace-sensitized stainless steel vessel attachment welds associated with the RVI components at RBS.
Furnace-sensitized stainless steel welds tend to be more susceptible to intergranular stress-corrosion cracking (IGSCC). In its response dated December 17, 2013, the licensee stated that some internal attachment welds (e.g., jet pump riser brace and steam dryer hold down bracket attachment welds) were furnace-sensitized at RBS. The licensee stated that it will continue to comply with BWRVIP-48-A I&E guidelines to monitor for any IGSCC in these welds.
Based on the above, the NRC staff concludes that the licensee's response is acceptable and, therefore, this issue is closed.
In RAI:-2, the NRC staff requested that the licensee provide an explanation for not including the following BWRVIP reports which are used to monitor active aging degradation in the jet pump, steam dryer, and top guide-non-ASME Code, Section x*l RVI components.
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."
BWRVIP-183, "BWRVIP, Top Guide Grid Beam Inspection and Flaw Evaluation."
By letter dated December 17, 2013, the licensee stated that the aforementioned BWRVIP reports are used at RBS as part of its inspection program for non-ASME Code,Section XI components (i.e., jet pump, steam dryer and top guide).
For jet pump beams, which are part of non-ASME Code,Section XI component at RBS, the licensee stated that BWRVIP-138, Revision 1, report is used to monitor aging degradation in jet pump beams.
For steam dryer assemblies, which are part of non-ASME Code,Section XI component at RBS, the licensee stated that BWRVIP-139 report is used to monitor aging degradation in steam dryer assemblies.
In its response to RAI-2, the licensee stated that it will perform inspections of the top guide consistent with the guidelines addressed in BWRVIP-183 report.
Based on the above, the NRC staff concludes that the implementation of I&E guidelines addressed in the aforementioned BWRVIP reports for the RVI components (stated above) is effective in identifying aging effects in a timely manner. The NRC staff concludes that the licensee's response is acceptable and, therefore, this issue is closed.
By letter dated December 17, 2013, in its response to RAI-3, the licensee confirmed that it used NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking,"
for monitoring the feedwater sparger piping, spacer brackets, flow holes and sparger tee welds.
Based on the above, the NRC staff concludes that the licens.ee's response is acceptable and, therefore, this issue is closed.
According to BWRVIP-76, when enhanced visual testing (EVT-1) is used for inspecting the core shroud horizontal welds, the inspection frequency is to be maintained at every 6 years. Table 1 of of the licensee's letter dated May 16, 2013, does not address the inspection frequency of every 6 years for EVT-1 examination. In RAI-4, the NRC staff requested the licensee to clarify the inspection frequency for EVT-1 for the core shroud horizontal welds. In its response dated December 17, 2013, the licensee indicated that it performed ultrasonic testing (UT) on core shroud horizontal welds and therefore, the inspection frequency of every 10 years as addressed in Table 1 is consistent with BWRVIP-76. Because the licensee is in compliance with the BWRVIP-76 guidelines, the NRC staff concludes that the licensee's response is acceptable and, therefore, this issue is closed.
The NRC staff noted that lnconel welds fabricated with Alloy 182 welding electrode are more prone to IGSCC than stainless steel (308/316) welds. By letter dated December 17, 2013, in response to RAI-5, the licensee addressed this issue in which it identified Alloy 182 welds in the following RVI components at RBS: (1) guide rod brackets; (2) core spray bracket; (3) steam dryer support bracket; (4) feedwater sparger bracket; (5) shroud support (weld H9); and (6) shroud support legs. As part of its response to RAI-5, the licensee stated that it has inspected these RVI components per I&E guidelines addressed in BWRVIP-38 and BWRVIP-48-A and thus far no cracking has been identified in the Alloy 182 welds. Based on the above, the NRC staff concludes that the licensee's response is acceptable and, therefore, this issue is closed The NRC staff reviewed the previous inspection results for the various RVI components that are
. addressed in the Attachment 3 of the licensee's letter dated May 16, 2013, and, in RAI-6, requested that the licensee address how the aging degradation is monitored in the most susceptible areas of the weld connections in the RVI components: (1) top guide; (2) core spray and core spargers; (3) low pressure coolant injection (LPCI) coupling; (4) jet pumps; (5) in-core dry tubes; (6) control rod drive (CRD) housing; and (7) core shroud.
In its response to RAI-6 dated December 17, 2013, the licensee stated that the top guide will be inspected per BWRVIP-183 in 2015, and, hence no inspection data is availablefor this RVI component. For RVI components listed in items 2 through 6 stated above, the licensee stated that these components are outside of the scope of the subject relief request because they are not included in ASME Code,Section XI lSI program. With respect to the inspections on horizontal core shroud H4, the licensee stated that 9 percent of the weld had flaws, and as a part of its corrective action, the licensee concluded that the 1 0-year re-inspection schedule for H4 weld is acceptable. The licensee determined that 1 0-year re-inspection frequency is also applicable to H3, H6, and H7 core shroud welds. Because the RVI components addressed above were inspected per the NRC staff's approved I&E.guidelines of relevant BWRVIP reports, the staff concludes that the licensee's response is acceptable and, therefore, this issue is closed.
By letter dated December 17, 2013, in response to RAI-7, the licensee addressed its implementation of On-line Noble Chemical Addition (OLNC) which includes hydrogen water chemistry (HWC) and noble metal chemical addition (NMCA) at RBS. To substantiate that the implementation of OLNC is effective, the licensee provided measured value of electrochemical potential (ECP) of the OLNC treated stainless steel coupon and amount of platinum deposited on a coupon collected during the period of operation. The NRC staff reviewed the ECP value and confirmed that the implementation of HWC and NMCA is consistent with the NRC staff approved BWRVIP-62, Revision 0, "BWR Vessel and Internals Project, Technical Basis for Inspection Relief for BWR Internal Components with Hydrogen Injection." The licensee also provided a measured value of the platinum loading which is consistent with the requirements of OLNC application. The NRC staff is currently reviewing the technical basis for application of OLNC which is addressed in BWRVIP-62, Revision 1. Therefore, the NRC staff did not review the information related to platinum loading value. Even though the staff did not review the platinum loading value, the staff considers that OLNC implementation is effective based on the following reasons:
(1) the ECP value is an essential primary variable to measure the effectiveness of HWC and NMCA. Low ECP value confirms that the implementation of OLNC is effective at RBS; (2) the low ECP value confirms adequate availability of hydrogen and platinum in RVI components; and (3) the lower values of platinum loading is expected for OLNC as compared to classical NMCA. RBS's measured platinum loading value on a coupon suggests that even at this lower value, the ECP values comply with the required value in BWRVIP-62, Revision 0.
Based on the above, the NRC staff concludes that the licensee's response is acceptable and, therefore, the issue related to RAI-7 is closed.
The Attachment to this SE includes attributes related to inspection techniques and frequency of inspections for various RVI components in the RBS unit. A comparison of the required ASME Code,Section XI, Category B-N-1 and B-N-2 examination requirements with the current BWRVIP guideline requirements that are applicable to the RBS unit is included in the Attachment to this SE.
Attachment 3 of the licensee's letter dated May 16, 2013, addressed previous inspection results of RVI components and, based on the information provided, the NRC staff concludes that with the exception of some RVI components, no indications were found in the majority of.RVI
'components at RBS. The licensee implemented a corrective action program as recommended by the relevant BWRVIP reports thereby providing reasonable assurance that the aging degradation mechanisms in RVI components are adequately monitored during 'the fourth 1 0-year lSI interval at RBS. 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 adequate assurance that any emerging aging effects will be identified in a timely manner. In addition; frequent inspections per these guidelines will enable the licensee to effectively monitor existing aging degradation in reactor pressure vessel interior surfaces, attachments, and core support structures. Based on the above, the NRC staff concludes that the active aging degradation mechanisms in the ASME Code,Section XI, RVI components is adequately monitored with the implementation of the appropriate BWRVIP I&E guidelines at RBS.
Consistent with the determination that was made in the NRC staff's SEs that approved each of the cited BWRVIP inspection requirements, as supplemented by the staff-approved inspection guidelines for the feedwater nozzle and sparger welds, the NRC staff concludes.that the licensee's proposed alternative will identify aging degradation of the RVI components in a timely manner. Therefore, the NRC staff concludes that the implementation of the inspection requirements specified in the licensee's proposed alternative will ensure that the integrity of the RVI components will be maintained with an acceptable level of quality and safety.
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 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 RBS's reactor pressure vessel interior surfaces, attachments, and core support structures.
5.0 CONCLUSION
Based on the above, 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 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 RBS is authorized for the fourth 10-year lSI interval at RBS unit.
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 lnservice Inspector. Any ASME Code,Section XI, RVI components that are not included in this request for alternative will continue to be inspected in accordance with the ASME Code,Section XI requirements.
Principal Contributor: Ganesh Cheruvenki Date: May 30, 2014 Attachment Comparison of ASME Category 8-N-1 and 8-N-2 Requirements with 8WRVIP Guidance Requirements
ASME Item No.
Table IWB-2500~1 B13.10 B13.20 B13.30 Comparison of ASME Category B-N-1 and B-N-2 Requirements With BWRVIP Guidance Requirements <11 BWRVIP ASME Exam ASME ASME..
Applicable BWRVIP.
,:.' ~WRVIP Frequency*
Component Scope Exam Fr~quency *
. Document Ex~111 S~ope Exalll* '
Reactor Vessel Interior Accessible VT-3 Each period BWRVIP-18-A, 25, BWRVIP examinations satisfy ASME Code,Section XI, Areas 26-A, 27 -A, 38, 4 7 -A VT-3 inspection requirements.
(Non-specific) 48-A, 76 Revision 1, and 138, Revision 1 Interior Attachments Within Beltline-BWRVIP-48-A, Riser. Brace EVT-1 100% in first 12 years, Jet Pump Riser Braces Table 3-2 Attachment 25% during each Accessible VT-1 Each 1 0-year subsequent 6 years Welds Interval Lower Surveillance Specimen Holder BWRVIP-48-A, Bracket VT-1 Each 10-year Interval Brackets
\\
Table 3-2 Attachment Guide Rod Brackets BWRVIP-48-A, Bracket VT-3 Each 1 0-year Interval Table 3-2 Attachment Steam Dryer Support Brackets BWRVIP-48-A, Bracket EVT-1 Each 1 0-year Interval Table 3-2 Attachment Steam Dryer hold-down Brackets Accessible BWRVIP-48-A, Bracket VT-3 Each 10-year Interval Welds Table 3-2 Attachment Feedwater Sparger Brackets BWRVIP-48-A, Bracket EVT-1 Each 10-year Interval Table 3-2 Attachment Core Spray Piping B'rackets BWRVIP-48-A, Bracket EVT-1 Each 4 re-fueling cycles Table 3-2 Attachment
- Upper Surveillance Specimen Holder Rarely BWRVIP-48-A, Bracket VT-3 Each 10-year Interval Brackets Accessible Each 1 0-year Table 3-2 Attachment Shroud Support (Weld H9) including VT-3 Interval BWRVIP-38,<4l Weld H9<2l EVT-1 or Maximum of 6 years for gussets where applicable 3.1.3.2 Figures 3-2
. UT one sided EVT-1, and 3-5 Maximum of 10 years for UT Weld H12 Shroud Support Legs BWRVIP-38,<4l 3.2.3 Weld H12 Per When accessible BWRVIP-38<3l NRC safety evaluation report (7-24-200),
inspect with appropriate method (4)
I I
Attachment ASME Item No.
Tab!e IWB-ASMEExam ASME ASME Applicable BWRVIP BV\\,IRVIP BWRVIP 2500~1 Comporyenf.
Scope Exam Frequency Document
.. Exam Scope Exam BWRVIP Frequency Welded core support --Shroud BWRVI P-38, <
4 Shroud EVT-1 or Based on as found Support 3.1.3.2, Figures 3-2 Support UT conditions, to a Maximum and 3-5 H8/H91eg 6 years for one-sided welds EVT-1, 10 years for UT including where accessible gussets as applicable Shroud Vertical Welds BWRVIP-76 R1, Vertical and EVT-1 or Maximum 6 years for one-Section 2.3 Ring Segment UT sided EVT-1, 10 years for B13.40
- Accessible VT-3 Each 1 0-year Figure 3-3 Welds as UT Surfaces Interval applicable Shroud horizontal Welds BWRVIP-76, 2.2, Welds H1-H7 EVT-1 or Based on as found Figure 3-3 as applicable UT conditions to a maximum 1 0 years for UT when inspected from both sides of the welds Shroud Repairs (S)
BWRVIP-76, R1 Tie-Rod
'VT-3 Per designer Section 3.5 Repair recommendations per BWRVIP-76 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)
Deviation to BWRVIP-38 was submitted to extend the 10-year ultrasonic re-inspection frequency by additional6 months, starting 10/31/2014 and ending when the subject re-inspections are performed in re-fueling outage 18, currently scheduled for 2015.
(5)
No repairs have been performed on the shroud; however, if shroud repairs are performed in the future then this submittal request includes use of BWRVIP guidance for this examination.
not included in this request for alternative will continue to be inspected in accordance with the ASME Code,Section XI requirements.
If you have any questions, please contact Alan Wang at 301-415-1445 or via e-mail at Alan.Wang@nrc.gov.
Docket No. 50-458
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 RidsNrrPMRiverBend Resource RidsRgn4MaiiCenter Resource GCheruvenki, NRR/DE/EVIB ADAMS Accession No. ML14127A327 Sincerely,
/RAJ 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 dated **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 05/15/14 05/13/14 04/25/14 05/30/14 OFFICIAL AGENCY RECORD