PLA-7126, Response to Request for Additional Information on Fourth Ten-Year Inservice Inspection Interval Proposed Relief Requests
| ML14031A081 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 01/31/2014 |
| From: | Franke J Susquehanna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| PLA-7126 | |
| Download: ML14031A081 (16) | |
Text
I AN a 1 2014 Jon A. Franke Site Vice President U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PPL Susquehanna, LLC 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.2904 Fax 570.542.1504 jfranke@ pplweb.com SUSQUEHANNA STEAM ELECTRIC STATION RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON FOURTH TEN-YEAR INSERVICE INSPECTION INTERVAL PROPOSED RELIEF REQUESTS PLA-7126 Docket No 50-387 and No. 50-388
References:
- 1. PPL Letter (PLA-7052), titled "Proposed Relief Requests for the Fourth Ten-Year Inservice Inspection Interval for Susquehanna Units 1 and 2," dated August 30, 2013.
- 2.
NRC Letter, titled "Request for Additional Information Regarding Relief Requests for the Fourth 10-Year Inservice Inspection Interval," dated December 19, 2013 (Accession No. ML13247A167).
The purpose of this letter is for PPL Susquehanna, LLC (PPL) to provide a response to the request for additional information contained in Reference 2. The NRC discussed a draft of the requested information with PPL on December 19, 2013, to ensure the questions were understood. PPL agreed to provide this response by January 31, 2014, in a subsequent discussion PPL held with the NRC. The information supports a review of the proposed alternative requests in Reference 1, specifically for relief requests 4RR-02 and 4RR-06.
The requested information is in Attachment 1 to this letter. Attachment 2 contains a revised 4RR-02, Revision 1 using the additional information.
There are no regulatory commitments associated with this response.
If you have any questions or require additional information, please contact Mr. Duane L. Filchner (570) 542-6501.
J. A.
anke Attachments: Response to Request for Additional Information Relief Request 4RR-02, Revision 1 Copy:
NRC Region I Mr. J. Greives, NRC Sr. Resident Inspector Mr. J. Whited, NRC Project Manager Mr. L. Winker, PA DEP/BRP Document Control Desk PLA-7126 to PLA-7126 Response to Request for Additional Information to PLA-7126 Page 1 of7 Response to Request for Additional Information By letter dated August 30, 2013,<1) PPL Susquehanna, LLC (PPL, the licensee), submitted a set of Relief Requests from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI for the Fourth Ten-Year Inservice Inspection (lSI)
Interval at Susquehanna Steam Electric Station (SSES) Units 1 and 2.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Part 50.55a(a)(3)(i), the licensee requested relief from certain inspection requirements regarding lSI of reactor pressure vessel (RPV) circumferential welds, consistent with the guidance provided in Generic Letter 98-05, "Boiling Water Reactor Licensee Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Welds," dated November 10, 1998, and the Nuclear Regulatory Commission (NRC) staffs safety evaluation (SE) for the BWRVIP-05 report issued on July 28, 1998. The licensee's submittal stated that Relief Request 4RR-02 was being provided as an administrative placeholder because this relief request was submitted in the Second Ten-Year lSI Interval as 2RR-22 by letter dated November 7, 2000,<2) and was subsequently approved by the NRC staff in a safety evaluation dated February 28, 2001, (3) until the end of the initial license for both units, which includes the Fourth 10-Year lnservice Inspection Interval. Relief Request 4RR-02 is essentially identical to Relief Request 2RR-22. However, because the initial license period ends at midnight on July 17, 2022 for SSES, Unit 1 and Midnight on March 24, 2024 for SSES, Unit 2, while the 4th ten-year lSI interval ends June 1, 2024, the staff determined that a part of the 4th 1 0-year interval is not covered by the relief request. Therefore, the licensee requested that the staff review and disposition relief request 4RR-02 for the Fourth Ten-Year lSI Interval.
To complete its review of Relief Request 4RR-02, the NRC staff requests responses to the questions below.
RAI4RR021:
Confirm the requested duration for Relief Request 4RR-02.
PPL's RESPONSE:
The requested duration for the Request for Alternative 4RR-02 is to the end of the Fourth Ten-Year lSI Interval, (June 1, 2024 for each Unit). See Attachment 2 of this letter for a revised Relief Request 4RR-02.
(1) PPL letter PLA-7052, "Proposed Relief Requests for the Fourth Ten-Year Inservice Inspection Interval for Susquehanna Units 1 and 2," dated August 30,2013 (Accession No. ML13247A167).
(2) PPL letter PLA-5251, "Proposed Relief Request No. RR-22 Request for Alternative to 10 CFR 50.55a Examination Requirements of Category B1.11 Reactor Pressure Vessel Welds," dated November 7, 2000 (Accession No. ML003769393).
(3) NRC Safety Evaluation, "Relief Request No. 22 (RR-22)," dated February 28, 2001, (TAC Nos.
MB0484 and MB0485), (Accession No. ML010330383).
to PLA-7126 Page 2 of7 Response to Request for Additional Information RAJ 4RR-02 2:
Both relief request 2RR-22 and 4RR-02 determined the adjusted reference temperature for the limiting circumferential weld based on a peak RPV inner diameter neutron fluence of 0.078 x 1019 n/cm2 associated with 32 effective full power years (EFPY) of operation. Since the Fourth Ten-Year lSI for SSES Unit 1, and SSES, Unit 2 extends beyond the end of the initial license period for both units, confirm that the neutron fluence at the end of the Fourth Ten-Year lSI Interval for both units will remain bounded by the value for 32 EFPY. If the neutron fluence is not bounded, provide a revised evaluation of the adjusted reference temperature of the limiting circumferential weld for SSES at the end of the Fourth Ten-Year lSI Interval.
PPL's RESPONSE:
Refer to the revised 4RR-02 (Revision 1) in Attachment 2 for a revised evaluation of the adjusted reference temperature at 54 EFPY, which is beyond the end of the Fourth Ten-Year lSI Interval.
RAJ 4RR-02 3:
The "Code Requirement" section of 4RR-02 states, in part, that:
10 CFR 50.55a(g)(6)(ii)(A)(2) requires volumetric of RPF shell welds to be performed completely, once, as an augmented examination requirement. These examinations are required to be performed using the 1989 Edition of the ASME Code Section XL These examinations are required during the inspection interval when the regulation was approved or the first period of the next inspection interval.
The "Relief Requested" section of 4RR-02 states, in part, that:
Approval of this alternative examination is requested in accordance with 10 CFR 50.55a(a)(3)(i) and 10 CFR 50.55a(g)(6)(ii)(A)(5) for permanently excluding volumetric examination of circumferential RPV welds.
The staff notes that the requirements of 10 CFR 50.55a(g)(6)(ii)(A) as stated above have been removed and replaced by "[Reserved]" in the current version of 10 CFR 50.55a. However, the staff also notes that ASME Code,Section XI editions more recent than 1989 contain the requirement to perform volumetric examination of essentially 100% of RPV shell welds.
The staff requests the licensee revise 4RR-02 accordingly to reflect the current version of 10 CFR 50.55a and the version of the ASME Code,Section XI applicable for the Fourth Ten-Year lSI Interval for SSES, Units 1 and 2.
PPL's RESPONSE:
Accordingly, a revised 4RR-02 (Revision 1) is in Attachment 2.
to PLA-7126 Page 3 of7 Response to Request for Additional Information 4RR-06:
Specifically, pursuant to 10 CFR 50.55a(a)(3)(i) the licensee requested an alternative from the requirements of ASME Code Section XI, Subsection IWF for the visual inspection of snubber attachments, on the basis that the alternatives provide an acceptable level of quality and safety.
To complete its review of Relief Request 4RR-06, the NRC staff requests responses to the questions below.
Based on information in the relief request, two different ASME Codes will be used. Verify that for snubbers (pin-to-pin), ASME OM Code 2004 Edition with 2005 and 2006 Addenda and for the snubber supports and attachments, ASME Section XI 2007 Edition with 2008 addenda will be used.
PPL's RESPONSE:
For snubber attachments (pressure boundary attachment to building attachment), visual examinations will be handled under the PPL Snubber Program which will utilize ASME OM Code 2004 Edition through the 2006 Addenda. For non-snubber supports (pressure boundary attachment to building attachment), visual examinations will be handled under the PPL lSI Program which will use the ASME Section XI Code, 2007 through 2008 Addenda. Under both programs, the individuals performing the visual examinations will be VT-3 qualified examiners.
Most of the plants, snubber inservice examination is implemented by a separate snubber program. Please explain and provide details how two different programs (1) snubbers and their supports including attachments; and (2) supports including attachments (without snubber), will be maintained and implemented.
PPL's RESPONSE:
- 1) Snubbers and associated attachments will be administered under the PPL Snubber Program which is committed to the ASME OM Code 2004 Edition through the 2006 Addenda.
- 2) Supports (without snubbers) will be administered under the PPL lSI Program which is committed to the ASME Section XI Code 2007 Edition through the 2008 Addenda.
While under two different programs, visual inspections for both programs will be conducted by VT-3 qualified individuals.
to PLA-7126 Page 4 of7 Response to Request for Additional Information RAI 4RR-06 3:
Clarify that while using Figure 2 of relief request 4RR-06 as the new boundary for inspection of snubbers, and their supports and attachments, if the support attachment is connected to insulated pipe, then the insulation will be removed for inspection of the attachment/support.
PPL's RESPONSE:
For those snubber attachments that are covered by insulation, the insulation will be removed prior to the VT-3 Visual Examination.
Please verify that for supports included under this relief request (Figure 2) all items will be examined as specified in IWF-2500.
PPL's RESPONSE:
The following items are listed in IWF-2500:
(a) Mechanical connections to pressure retaining components and building structures (b) Weld connections to building structure (c) Weld and mechanical connections at intermediate joints in multiconnected integral and nonintegral supports (d) Clearances of guides and stops, alignment of supports, and assembly of support items (e) Hot or cold settings of spring supports and constant load supports (f) Accessible sliding surfaces PPL will examine the above items within the boundary identified in Figure 2 of the Request for Alternative 4RR-06 as they pertain to the snubber attachment configuration. For the snubber itself, ISTD-4200 will be followed as applicable.
IWF-2430(a) states, in part, that component supports examination performed in accordance with Table IWF-2500-1 that reveal flaws or relevant conditions exceeding the acceptance standards of IWF-3400, and that requires corrective measures, shall be extended, during the current outage, to include the component supports immediately adjacent to flawed supports. Please explain how this situation (to include supports adjacent to flawed supports) will be considered, because this relief request only considers supports with snubbers. (Note: the adjacent supports could be without snubber) to PLA-7126 Page 5 of7 Response to Request for Additional Information PPL's RESPONSE:
The requirement found in IWF-2430 to examine the supports immediately adjacent to a snubber that is found exceeding the acceptance standards, and that requires corrective measures, will be examined regardless of whether the adjacent support includes a snubber.
RAJ 4RR-06 6:
In the "Proposed Alternative and Basis for Use" section of Relief Request 4RR-06, it is noted that ASME OM Code Case OMN-13 requires 100% of safety related snubbers to be examined and evaluated at least once every 10 years. This exceeds the requirements of the ASME Section XI, Table IWF-2500-1, which only requires 15% of Class 1, 15% of Class 2, and 10% of Class 3 supports/attachment over a 10-year interval. Please respond to the following:
(a) ASME OM Code Case OMN-13 allows to extend the snubber visual examination once every 10 years, and can be implemented after the requirements of ISTD-4251 and ISTD-4252 have been satisfied and the previous examination per Table ISTD 4252-1 was performed satisfactory at a maximum interval of two fuel cycles. (a) Please specify that the ISTD-4251 and ISTD-4252 requirements are met at Susquehanna and Table ISTD 4252-1 requirements have been satisfied to use Code Case OMN-13; and (b) Please elaborate on how the proposed snubber examination extended up to 10 years can be aligned with the 10-year lSI interval of the support and attachments (containing the snubber) for inspection.
(b) While using OMN-13 for snubbers during the extended interval of 10-years, if the number of unacceptable snubbers (pin-to-pin) exceeds the ISTD-4252-1 limits (these unacceptable snubbers can be found during non-inspection activities such as walkdowns or any other events i.e. water hammer, etc.), what action will be taken and how will these findings align with the supports and attachments inspection.
PPL's RESPONSE:
(a) (a) PPL implemented the use of Code Case OMN-13 during its third ten-year inspection interval after satisfactorily meeting the requirements of ISTD-4251 and ISTD-4252.
The last visual examinations were performed in 2012 (Unit 2) and 2013 (Unit 1). In both cases the number of snubber failures as part of the visual examination program were below the required threshold for reduction of snubber visual examination frequency.
(b) The "interval" for the PPL Snubber Program is the same as the PPL lSI Program.
Alignment of the two programs is not necessary. The beginning of the Fourth Ten-year Inspection Interval for both the PPL Snubber and PPL lSI Programs is June 1, 2014.
to PLA-7126 Page 6 of7 Response to Request for Additional Information (b) Should the number of unacceptable snubbers exceed the limits as prescribed in ISTD-4252-1, a reduction in the frequency of visual examinations will occur. Only snubbers and their associated attachments will be under the requirements of this reduced visual inspection frequency.
Verify that for snubbers and their supports and attachments, VT-3 will be performed as defined in IW A-2213 (i.e. IW A-2213(a) thru IW A-2213(g)).
PPL's RESPONSE:
Visual examination will be carried out in accordance with ASME Section XI, IW A-2213(a) thru IW A-2213(g). Visual examinations will be carried out by VT-3 qualified examiners.
Please explain how the proposed inspection of supports containing snubbers (once every 10 years) will be aligned with inspection scheduled for others supports (without snubbers), as defined in Table IWF-2410-1.
PPL's RESPONSE:
As stated above in response to Question 6, aligning the examination of snubbers and other supports (without snubbers) is not necessary. The purpose for the request for this alternative is to separate them into two different programs.
What action will be taken if a snubber/support/attachment failed during the proposed combined inspection interval? Please provide separate responses for snubbers and supports/attachments.
(Note: Also see RAI 4RR-06 6(b) above for snubbers and IWF-2430, "Additional Examination,"
for supports)
PPL's RESPONSE:
a) Under the PPL Snubber Program, should a snubber or its associated attachment fail its visual examination, then the snubber will be removed for functional testing. The visual failure and the reason for the failure will also be entered into the SSES Corrective Action Program (CAP) system. Should the snubber fail its functional test, it is then considered a visual failure and the requirements of ASME OM Code ISTD-4240 then apply.
b) A failure of a support (without a snubber) will follow the requirements of IWF-2430.
to PLA -7126 Page 7 of7 Response to Request for Additional Information RAI 4RR-06 10:
How will the newly added support(s) with snubber(s) and support(s) without snubber(s) be inspected during the 10-year interval as per this relief request? Please provide your response based on proposed relief request and IWF-2410(c).
PPL's RESPONSE:
New snubbers added will be examined in accordance with the ASME OM Code. New supports (without snubbers) will be examined in accordance with ASME Section XI, IWF-2410(c).
to PLA-7126 Relief Request 4RR-02, Revision 1
REUEFREQUEST4RR~2 REVISION 1 In response to NRC RAis, the revised Relief Request 4RR-02 is provided below. to PLA-7126 Page 1 of 5 SYSTEM/COMPONENT(S) FOR WHICH RELIEF IS REQUESTED Examination Category B-A, Item Number, Bl.11 Welds on Unit 1 and 2: Weld IDs AA, AB, AC, AD, and AE.
CODE REQUIREMENT The ASME Section XI Code, 2007 Edition with the 2008 Addenda, Table IWB-2500-1, Examination Category B-A, Item Number B 1.11, describing volumetric examination of RPV circumferential welds, requires examination of essentially 100 percent of the weld length.
RELIEF REQUESTED PPL requests approval of an alternative RPV examination for SSES Units 1 and 2. Approval of this alternative examination is requested in accordance with 10 CFR 50.55a(a)(3)(i) for permanently excluding volumetric examination of circumferential RPV welds. PPL also requests approval to implement the alternative RPV examination in lieu of the lSI requirements for circumferential welds in the ASME Code,Section XI 2007 Edition through the 2008 Addenda Table IWB-2500-1, Examination Category B-A, Item Number Bl.ll volumetric examination of RPV circumferential welds. The Code of record for the fourth inspection interval is the ASME Code,Section XI, 2007 Edition through the 2008 Addenda.
BASIS FOR RELIEF In Generic Letter 98-05, the NRC stated that the estimated failure frequency of the BWR RPV circumferential welds is well below the acceptable core damage frequency (CDF) and large early release frequency (LERF) criteria discussed in Regulatory Guide 1.174, "An Approach for using Probabilistic Risk Assessment in Risk Informed Decisions On Plan-Specific Changes to the Licensing Basis." Furthermore, the NRC indicated that the estimated frequency of RPV circumferential weld failure bounds the corresponding CDF and LERF that may result from a RPV weld failure. On this basis, the NRC concluded the proposal in the BWRVIP-05 report, as modified by two criteria, was acceptable and that BWR licensees may request permanent relief from the lSI requirements of 10CFR50.55a for the volumetric examination of circumferential reactor welds by demonstrating the two criteria discussed below. The generic letter states that licensees still need to perform their required inspections of "essentially 100 percent" of all axial welds.
Generic Letter 98-05 Criterion 1 At the expiration of the license, the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the staffs July 28, 1998 safety evaluation (of GL 98-05 Permitted Action).
PPL Response REUEFREQUEST4RR~2 REVISION 1 to PLA-7126 Page 2 of 5 At end of the Fourth Ten-Year lSI Interval, which is beyond the original 40-year license of both Units, the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the staff's July 28, 1998 safety evaluation (of GL 98-05 Permitted Action).
Table 1 illustrates that SSES Units 1 and 2 RPVs have additional conservatism in comparison to Table 2.6-4 for the Limiting Plant-Specific Analysis (32 EFPY) from the staff's July 28, 1998 Safety Evaluation of the BWRVIP-05 report. The chemistry factor, ilRTNDr, RTNDT(U) and Mean RTNDT are determined in accordance with the guidelines of Regulatory Guide 1.99, Rev. 2.
Table 1 SSES Units 1 and 2 Comparative Parameters at NRC Limiting Plant Specific Parameter 54 EFPY for the Bounding Description Circumferential Weld Analyses Parameters at 32 Heat/Lot EFPY SER Table 2.6-4 624263/E204A27 A*
Cu, wt%
0.06 0.10 Ni, wt%
0.89 0.99 CF 82 109.5 EOL ID Fluence, x1019 n/cm2 0.118 0.51 ilRTNDr, °F 30.9 109.5 RTNDT(U)
-20
-65 Mean RTNDr, °F 41.9 44.5
- Unit 2 data: Unit 1 data is enveloped by this data.
The chemistry factors for the SSES Units 1 and 2 limiting circumferential welds are lower than the NRC's Limiting Plant-Specific Analyses (32 EFPY), and the End of Life (EOL) fluence for 54 EFPY is lower than the NRC's limit for 32 EFPY. The resulting shift in the reference temperature, ilRTNor, and the Mean RTNorat 54 EFPY for the SSES Units 1 and 2limiting circumferential welds remain bounded by the NRC's evaluation of the BWRVIP-05 technical bases at 32 EFPY. Since 54 EFPY, which coincides with 60 years of plant operation, conservatively envelopes the Fourth Ten-Year lSI Interval, scheduled to complete on May 31, 2024, the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the staff's July 28, 1998 safety evaluation.
BASIS FOR RELIEF Generic Letter 98-05 Criterion 2 RELIEF REQUEST 4RR-02 REVISION 1 to PLA -7126 Page 3 of 5 Licensees have implemented operator training and established procedures that limit the frequency of cold over-pressure events to the amount specified in the staff's July 28, 1998 safety evaluation.
PPL Response PPL has in place procedures which monitor and control reactor temperature and water inventory during all aspects of cold shutdown which would minimize the likelihood of a Low Temperature Over-Pressurization (LTOP) event. Additionally, these procedures are reinforced through operator training.
The System Leakage Test and the System Hydrostatic Test (as modified by ASME Code Case N-498-1), which have been used at SSES, have sufficient procedural guidance to prevent a cold overpressurization event. The System Leakage Test is performed at the conclusion of each refueling outage, while the System Hydrostatic Test is performed once each Ten-Year Inspection Interval. Briefings for these tests generally detail the anticipated testing evolution with special emphasis on conservative decision making, plant safety awareness, the process in which the test would be aborted if plant systems responded in an adverse manner, and lessons learned from similar in-house or industry operating experiences. Specific attention is devoted to avoidance of rapid overpressurization by an inadvertent SCRAM at test pressure (in the manner of Clinton Power Station LER 89-016). Vessel temperature and pressure are required to be monitored throughout these tests to ensure compliance with the Technical Specification 3.4.10 pressure-temperature curve. The procedures for these tests prescribe the designation of a test director (on a shift basis) for the duration of the test who is a single point of accountability, responsible for the coordination of testing from initiation to closure and for maintaining shift and line management cognizant of the status of the test. Additionally, the Shift Supervisor provides an oversight function during the test.
Additionally, to ensure a controlled, deliberate pressure increase, the rate of pressure increase is administratively limited throughout the performance of the test. If the pressurization rate exceeds this limit, direction is provided to remove the Control Rod Drive (CRD) pumps, which are used for pressurization, from service.
With regard to inadvertent system injection resulting in an LTOP condition, the high pressure make-up systems (High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems), as well as the normal feedwater supply (via the Reactor Feedwater Pumps) at SSES are all steam driven. During reactor cold shutdown conditions, no reactor steam is available for the operation of these systems. Therefore, it is not possible for these systems to contribute to an over-pressure event while the unit is in cold shutdown. Although auxiliary steam is used to test the associated turbines while the plant is shutdown, the pump is uncoupled from the turbine during the actual test which would prevent an LTOP condition.
RELIEF REQUEST 4RR-02 REVISION 1 BASIS FOR RELIEF (Continued) to PLA -7126 Page 4 of5 Procedural control is also in place to respond to an unexpected or unexplained rise in reactor water level which could result from a spurious actuation of an injection system. Actions specified in this procedure include preventing condensate pump injection, securing ECCS injection, tripping CRD pumps, terminating all other injection sources and lowering RPV level via the Reactor Water Cleanup (RWCU) system.
In addition to procedural barriers, Licensed Operator Training is in place which further reduces the possibility of the occurrence of LTOP events. During Initial Licensed Operator Training the following topics are covered: Brittle fracture and vessel thermal stress; Technical Specification training, including Section 3.4.10 "RCS Pressure and Temperature (Pff) Limits," and Simulator Training of plant heatup and cooldown including performance of surveillance tests which ensure pressure-temperature curve compliance. In addition, operator training has been provided on the expectations for procedural compliance as provided in the operations standards manual.
During plant outages, the work control processes assure that the outage schedule and changes to the schedule receive a thorough shutdown risk assessment review to ensure defense-in-depth is maintained. Work activities are reviewed by Station Management and Operations Management to ensure safe operation and that plant mode can support the schedule work.
During outages, work is coordinated through the Outage Control Center and the Ops Work Control Center which provides an additional level of Operations oversight. In the Control Room, the Shift Supervisor is required, by procedure, to maintain cognizance of any activity that could potentially affect reactor level or decay heat removal during refueling outages. The Control Room Operators are required to provide positive control of reactor water level within the specified bands, and promptly report when operating outside the specified band, including restoration of actions being taken.
In addition to the above, ongoing review of industry operating plant experiences is conducted to ensure the PPL procedures consider the impact of actual events, including LTOP events.
Appropriate adjustments to the procedures and associated training are then implemented, to preclude similar situations from occurring at SSES.
Summary The BWRVIP-05 report provides the technical basis for eliminating inspection of BWR RPV circumferential shell welds. The BWRVIP-05 report concludes that the probability of failure of the BWR RPV circumferential shell welds is orders of magnitude lower than that of that axial shell welds. Based on an assessment of the materials in the circumferential weld in the beltline of the SSES Unit 2 RPV, the conditional probability ofRPV failure should be less than or equal to that estimated in the NRC's analysis. Based on operator training and established procedures that have been implemented, the probability of cold over-pressure transients will limit the frequency of cold over-pressure events to the amounts specified in the NRC's June 30, 1998 safety evaluation.
REFERENCES RELIEF REQUEST 4RR-02 REVISION 1 to PLA-7126 Page 5 of 5
- 1. NRC Generic Letter 98-05, "Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds," dated November 10, 1998.
- 2. EPRI TR 105697, BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05), September 1995.
- 3. NRC Letter from Gus C. Lainas, Acting Director, Division of Engineering, Office of Nuclear Reactor Regulation, to Carl Terry, BWRVIP chairman, Niagara Mohawk Company, July 28, 1998.
ALTERNATE EXAMINATIONS PPL proposes to perform inspections of essentially 100 percent of the longitudinal seam welds in the RPV shell and essentially zero percent of the RPV circumferential seam welds, which will result in partial examination (i.e., approximately two to three percent) of the circumferential welds at their points of intersection with the longitudinal welds. These inspections are being proposed as an alternative to the augmented examinations specified in the lSI requirements for circumferential welds in the ASME Code,Section XI 2007 Edition through the 2008 Addenda.
IMPLEMENTATION SCHEDULE This relief will remain in effect for the duration of the Fourth 10 year interval of the lSI Program for SSES Units 1 and 2 (May 31, 2024).
Note: This relief has previously been authorized by the NRC in a safety evaluation dated February 28, 2001, (Accession ML010330383).