ML13267A097
| ML13267A097 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 09/30/2013 |
| From: | Travis Tate Plant Licensing Branch III |
| To: | Pacilio M Exelon Generation Co |
| Mozafari B | |
| References | |
| TAC ME9668, TAC ME9669, TAC ME9670, TAC ME9671, TAC ME9672, TAC ME9673, TAC ME9674 | |
| Download: ML13267A097 (40) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 30, 2013 Mr. Michael J. Pacilio Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 - SAFETY EVALUATION IN SUPPORT OF REQUEST FOR RELIEF ASSOCIATED WITH THE FIFTH 10 YEAR INTERVAL INSERVICE INSPECTION PROGRAM (TAC NOS. ME9668, ME9669, ME9670, ME9671, ME9672, ME9673, ME9674, ME9675, ME9676, ME9677, ME9678, ME9679, ME9680, ME9681)
Dear Mr. Pacilio:
By letter dated September 28,2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12275A070), as supplemented by letters dated November 28, 2012, January 10, 2013, and May 30, 2013 (ADAMS Accession Nos. ML12333A262, ML13010A456 and ML13151A107 respectively), Exelon Generation Company, LLC (the licensee), submitted Relief Requests 15R-01, 15R-02, 15R-03, 15R-04, 15R-06, 15R-09 and 15R-10 to the U.S. Nuclear Regulatory Commission (NRC). The licensee proposed alternatives to or requested relief from certain inservice inspection (lSI) requirements of the American Society of Mechanical Engineers (ASME) Code,Section XI, "Rules for In-service Inspection (lSI) of Nuclear Power Plant Components," at Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, for the fifth 10-year lSI program interval.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(a)(3)(i), the licensee requested to use the proposed alternatives in Relief Requests 15R-02, 15R-04, 15R-06, 15R-09, and 15R-10 on the basis that the alternatives provide an acceptable level of quality and safety. Pursuant to 10 CFR Part 50, Section 50.55a(a)(3)(ii), the licensee requested to use the proposed alternatives in 15R-01 and 15R-03 on the basis that the proposed alternatives provide reasonable assurance of quality and safety of the subject components, and compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the levels of quality and safety.
The NRC staff has reviewed the subject requests and concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(i) for requests 15R-02, 15R-04, 15R-06, 15R-09, and 15R-10. Further, pursuant to 10 CFR 50.55a(a)(3)(ii) for requests 15R-01 and 15R-03, the NRC staff concludes as set forth in the enclosed safety evaluation that the licensee is In compliance with the ASME Code requirements. Therefore, the NRC staff authorizes alternative requests 15R-01, 15R-02. 15R-03, 15R-04, 15R-06, 15R-09. and 15R-10. at QCNPS Units 1 and 2, for the fifth 1 O-year lSI program interval, which commenced on April 2, 2013, and scheduled to be completed by April 1,2023.
All other ASME Code requirements for which relief was not specifically requested and approved in the subject requests remain applicable.
M. Pacilio
- 2 If you have any questions on this action, please contact the NRC Project Manager, Brenda Mozafari, at (301) 415-2020.
Sincerely, Travis L. Tate, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-254 and 50-265
Enclosures:
- Safety Evaluation - Relief Request 15R-01 - Safety Evaluation - Relief Request 15R-02 - Safety Evaluation - Relief Request 15R-03 - Safety Evaluation - Relief Request 15R-04 - Safety Evaluation - Relief Request 15R-06 - Safety Evaluation - Relief Request 15R-09 - Safety Evaluation - Relief Request 15R-10 cc w/encl: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUESTS FOR RELIEF 15R-01 REGARDING STANDBY LIQUID CONTROL NOZZLE EXELON GENERATION COMPANY, LLC QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265
1.0 INTRODUCTION
By letter dated November 28, 2012, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12333A262) Exelon Generation Company, LLC (the licensee) superseded Relief Request 15R-01 of the original request dated September 28,2012, (ADAMS Accession No. ML12275A070) and proposed an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),
Section XI, for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. The licensee provided additional information in a supplement dated May 30, 2013 (ADAMS Accession No. ML13151A107).
The request for an alternative proposed in Relief Request 15R-01 covers the section of the standby liquid control (SBLC) nozzle inner radius where the licensee was unable to essentially achieve a 100 percent inspection coverage due to the design of the subject component.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(a)(3)(ii), the licensee requested alternative requirements for inservice examination or the basis that the ASME Code requirement presents a hardship without a compensating increase in quality and safety.
As an alternate examination, the licensee proposes to perform a visual (VT-2) examination of the subject nozzles at QCNPS, Units 1 and 2, each refueling outage.
2.0 REGULATORY EVALUATION
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 preservice examination requirements, as set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 1 O-year interval and subsequent intervals comply with ENCLOSURE 1 (15R-01)
- 2 the requirements in the latest edition and addenda of Section XI of the ASME Code, which was 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.
Section 50.55a(a)(3) of 10 CFR states that proposed alternatives to the requirements of Paragraphs (c), [d), (e), (f), (g), and (h) of this section, or portions thereof, may be used when authorized by the Director of the Office of Nuclear Reactor Regulation or the Director of the Office of New Reactors, as appropriate. Any proposed alternatives must be submitted and authorized prior to implementation.
Section 50.55a(a)(3) of 10 CFR further states that alternatives to the requirements of Paragraph (g) may be used, when authorized by the NRC, if: (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The licensee has requested an alternative from ASME Code requirements pursuant to 10 CFR 50.55a(a)(3)(ii). The ASME Code of Record for QCNPS, Units 1 and 2, fifth 10-year inservice inspection (lSI) interval program, is the 2007 Edition through the 2008 Addenda of Section XI.
Based on the above, and subject to the following safety evaluation, the U.S. Nuclear Regulatory Commission (NRC) staff finds that regulatory authority exists for the licensee to request and the NRC to grant the relief requested.
3.0 NRC STAFF TECHNICAL EVALUATION FOR RELIEF REQUEST 15R-01 The information provided by the licensee in support of the alternative to the ASME Code requirements has been evaluated and the bases for disposition are documented below.
Request for Relief 15R-1 ASME Code,Section XI. Examination Category B-D, Item B3.100 ASME Code Requirement for Full Penetration Welded Nozzles in Vessels QCNPS, Units 1 and 2. Request for Relief 15R-1 ASME Code Class:
1
Reference:
IWB-2500, Table IWB-2500-1, Figure IWB-2500-7 Examination Category:
B-D Item Number:
B3.100
==
Description:==
Inspection of Standby Liquid Control Nozzle Inner Radius Component Number:
QCNPS, Unit 1: N10 QCNPS, Unit 2: N10 Drawing Number:
QCNPS, Unit 1: M-3106, Sheet 1 QCNPS, Unit 1: M-3116, Sheet 1 ASME Code Requirement ASME Code,Section XI, Table IWB-2500-1, Figure IW8-2500-7, requires a volumetric examination is required to be performed on the inner radius section of all reactor pressure vessel nozzles each inspection interval. This includes nozzles with full penetration welds-to
3 vessel shell (head) cast nozzles, but excludes man ways and holes either welded to or integrally cast in vessel.
Licensee's Basis for Relief Request (As stated)
Pursuant to 10 CFR 50.55a(a)(3)(ii), relief is requested on the basis that conformance with the [ASME] Code requirements impose hardship without a compensating increase in the level of quality and safety.
The Standby Liquid Control (SBLC) nozzle, as shown in Figure 15R-01.1 [provided in the submittal], is designed with an integral socket to which the boron injection piping is fillet welded. The SBLC nozzle is located near the bottom of the vessel in an area which is inaccessible for ultrasonic examinations from the inside of the vessel. Therefore, ultrasonic [UT] examinations would need to be performed from the outside diameter of the vessel. As shown in Figure 15R-01.1, the [UT] scan would need to travel through the full thickness of the vessel into a complex cladding/socket configuration. These geometric and material reflectors inherent in the design prevent a meaningful examination from being performed on the inner radius of the SBLC nozzle.
In addition, the inner radius socket attaches to piping which injects boron at locations far removed from the nozzle. Therefore, the SBLC nozzle inner radius is not subjected to turbulent mixing conditions that are a concern at other nozzles.
Compliance with the applicable [ASME] Code requirements would require an [UT]
examination to be performed on the outside diameter of the reactor pressure vessel.
Geometric and material reflectors would prevent a meaningful examination, resulting in inaccurate data. Based on this, the [ASME] Code requirements impose hardship without a compensating increase in the level of quality and safety in accordance with 10 CFR 50.55a(a)(3)(ii).
Additional information was provided by the licensee in its response dated May 30, 2013, to the NRC staff's request for additional information (RAI). The NRC staff requested that the licensee discuss why its proposed alternative provides reasonable assurance of structural integrity or leak tightness of the subject components.
The licensee's response (As stated)
The proposed alternative verifies structural integrity and leak tightness by ensuring no leakage at nominal operating pressure during the [ASME Code] Class 1 System Leak test. This test is a [ASME] Code required exam that verifies no through wall leakage for the entire [ASME Code] Class 1 boundary. It is performed prior to start-up and unit operation every outage to provide confidence there is no leakage. The Standby Liquid Control nozzle is included as part of this test. Passing this test with no identified through wall leakage at the nozzle provides reasonable assurance of structural integrity and leak tightness of the Standby Liquid Control nozzle. During normal plant operation, the Standby Liquid Control nozzle is inaccessible for direct visual observation based on its location in the drywell. This does not preclude the nozzle from being indirectly monitored. Leakage from the nozzle would be collected in the drywell sumps, prompting
-4 action from the station. No leakage from this nozzle has ever been observed at QCNPS from this location.
Licensee's Proposed Alternative Examination (As stated)
As an alternate examination, Exelon Generation Company, LLC will perform a VT-2 visual examination of the subject nozzles at QCNPS, Units 1 and 2 each refueling outage in conjunction with the Class 1 System Leakage Test.
NRC Staff Evaluation
The ASME Code requires 100 percent volumetric examination to be performed on the inner radius section of all reactor pressure vessel nozzles each inspection interval. This test is an ASME Code-required exam that verifies no through wall leakage for the entire ASME Code, Class 1, boundary. This includes nozzles with full penetration welds-to-vessel shell (head) cast nozzles, but excludes man ways and holes either welded to or integrally cast in vessel.
However, the NRC staff determined that the design configurations of the subject welds and the proximity of surrounding area limit access for ultrasonic test (UT) scanning. In order to effectively increase the examination coverage, the nozzle-to-vessel welds would require design modifications and removal of adjacent components. Thus, the NRC staff determines that 100 percent ASME Code-required volumetric examinations are considered a hardship without a compensating increase in safety.
The SBLC nozzle, as shown in Figure 15R-01.1, provided in the licensee's submittal, is designed with an integral socket to which the boron injection piping is fillet welded. The subject SBLC nozzle is located near the bottom of the vessel in an area which is inaccessible for UT examinations from the internal side of the vessel. The NRC staff determined that UT examinations would need to be performed from the outside diameter of the vessel. As a result of the configuration, in order for the licensee to perform a UT examination, the UT scan would be required to travel through the full thickness of the vessel into a complex cladding/socket configuration. The NRC staff determined that the licensee would not be able to perform a meaningful examination on the inner radius of the SBLC nozzle due to these geometric and material reflectors inherent in the design.
The SBLC nozzle inner radius is not subjected to turbulent mixing conditions that are a concern at other nozzles due to subject nozzle attached to piping which injects boron at locations far removed from the nozzle. The NRC staff determined that based on the above, the ASME Code requirement to require the licensee to volumetrically examine the subject nozzle would be a hardship without a compensating increase in safety. The licensee also noted that during normal plant operation, the SBLC nozzle is inaccessible for direct visual observation based on its location in the drywell. This does not preclude the nozzle from being indirectly monitored.
Leakage from the nozzle would be collected in the drywell sumps, prompting action from the station. No leakage from this nozzle has ever been observed at QCNPS, Units 1 and 2, from this location. The licensee's proposed alternative to perform ASME Code-required system leakage tests prior to start-up and unit operation every outage provides reasonable assurance of structural integrity, leak tightness, and no leakage at nominal operating pressure of the subject nozzle, associated piping, and components.
- 5
4.0 CONCLUSION
As set forth above, the NRC staff has determined that authorizing the licensee's proposed alternative pursuant to 10 CFR 50.55a(a)(3)(ii) is authorized by law and will not endanger life or property, or the common defense and security, and is otherwise in the public interest given due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Furthermore, the NRC staff concluded that the proposed alternative will provide reasonable assurance of leak tightness of the subject component. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii). Therefore, the NRC staff authorizes the licensee's proposed alternative contained in Relief Request 15R-01, Revision 1, for QCNPS, Units 1 and 2, fifth 10-year lSI interval.
All other ASME Code,Section XI, requirements for which relief was not speCifically requested and approved in the subject requests remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555"()001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST NO. 15R-02 REGARDING A RISK-INFORMED INSERVICE INSPECTION PROGRAM EXELON GENERATION COMPANY, LLC QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-254 and 50-265
1.0 INTRODUCTION
By letter dated September 28, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12275A070), as supplemented by letter dated May 30, 2013 (ADAMS Accession No. ML13151A107), Exelon Generation Company, LLC (the licensee),
requested U.S. Nuclear Regulatory Commission (NRC) authorization to extend the risk-informed inservice inspection (RI-ISI) program plan for Quad Cities Nuclear Power Station (QCNPS),
Units 1 and 2, into the fifth 1 O-year inservice inspection interval. The QCNPS RI-ISI program was initially submitted to the NRC by letter dated November 30, 2000 (ADAMS Accession No. ML003776493), and was approved by the NRC for use in the third 10-year lSI interval by letter dated February 5, 2002 (ADAMS Accession No. ML020180003). The use of the QCNPS RI-ISI program was requested for the fourth 10-year interval by the licensee in a letter dated January 17,2003 (ADAMS Accession No. ML030290324), and subsequently approved by the NRC staff by letter dated January 28, 2004 (ADAMS Accession No. ML033560386).
The licensee has considered relevant information since the development of the original program and has reviewed and updated the RI-ISI program. The current licensee submittal proposed the continuation of the updated RI-ISI program during the fifth 10-year lSI interval.
2.0 REGULATORY EVALUATION
Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(g), American Society of Mechanical Engineers Bolier and Pressure Vessel Code (ASME Code) Class 1,2, and 3, components (including supports) shall meet the requirements, "except design and access provisions and pre service examination requirements" set forth in the ASME Code to the extent practical within the limitations of design, geometry, and materials of construction of the components. As stated in 10 CFR 50.55a(g), lSI of the ASME Code, Class 1, 2, and 3, components is to be performed in accordance with Section XI of the ASME Code and applicable addenda, except where specific relief has been granted by the NRC.
ENCLOSURE 2 (15R-02)
~ 2 ~
The regulations also require during the first 1 O~year lSI interval and during subsequent intervals, that the licensee's lSI program complies with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference into 10 CFR 50.55a(b), 12 months prior to the start of the 120-month interval, subject to the conditions listed therein. QCNPS is currently in its fifth 10-year lSI interval.
Pursuant to 10 CFR 50.55a(g), a certain percentage of ASME Code Category 8-F, 8-J, C-F-1 and C~F-2, pressure retaining piping welds must receive an lSI during each 1 O-year interval.
The ASME Code requires 100 percent of all 8-F welds and 25 percent of all 8-J welds greater than 1-inch nominal pipe size be selected for volumetric or surface examination, or both, on the basis of existing stress analyses. For Categories C-F-1 and C-F-2 piping welds, 7.5 percent of non-exempt welds are selected for volumetric or surface examination, or both. According to 10 CFR 50.55a(a)(3), the NRC may authorize alternatives to the requirements of 10 CFR 50.55a(g), if an applicant demonstrates that the proposed alternatives would provide an acceptable level of quality and safety, or that compliance with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The NRC staff finds that there is a regulatory basis for the licensee to request, and the NRC to authorize, this alternative pursuant to the technical evaluation that follows. The information provided by the licensee in support of the request has been evaluated by the NRC staff and the bases for disposition are documented below.
The NRC staff has developed the following documents to evaluate proposed RI-ISI programs:
Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis" (ADAMS Accession No. ML023240437).
RG 1.178, "An Approach For Plant-Specific Risk-Informed Decisionmaking - Inservice Inspection of Piping" (ADAMS Accession No. ML032510128).
RG 1.200, Revision 1, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (ADAMS Accession No. ML070240001).
RG 1.174 provides guidance on the use of probabilistic risk analysis (PRA) findings and risk insights in support of licensee requests for changes to a plant's licensing basis. RG 1.178 describes an RI-ISI program as one that incorporates risk insights that can focus inspections on more important locations, while at the same time maintaining or improving public health and safety. RG 1.200 describes one acceptable approach for determining whether the quality of the PRA, in total or the parts, used to support an application is sufficient to provide confidence in the results, such that the PRA can be used in regulatory deCision-making.
- 3 3.0 NRC STAFF TECHNICAL EVALUATION FOR RELIEF REQUEST 15R-02 Licensee's Proposed Alternative to ASME Code The licensee proposed to continue use of the QCNPS RI-ISI program plan approved by the NRC staff in the third and fourth 10-year intervals and for the fifth 1 O-year interval. The QCNPS RI-ISI program plan will be used as an alternative to the current ASME Code,Section XI, 2007 Edition through the 2008 Addenda, examination requirements for Class 1 Examination Category 8-F and 8-J piping welds and Class 2 Examination Category C-F-1 and C-F-2 piping welds. The proposed alternative is sought for the QCNPS fifth 10-year lSI interval which began on April 2, 2013, and is scheduled to end April 1, 2023.
The licensee's process used to develop the initial RI-ISI program was based on Electric Power Research Institute, Inc. (EPRI) Topical Report (TR)-112657, "Revised Risk-Informed Inservice Inspection Evaluation Procedure," Revision 8-A (Reference 1). The alternative will continue to use the same two enhancements proposed and approved in the previous intervals RI-ISI program.
The licensee stated that in lieu of the evaluation and sample expansion requirements in Section 3.6.6.2, "RI-ISI Selected Examinations" of Reference 1, the requirements of Subarticle-2430, "Additional Examinations," contained in ASME Code Case N-578-1 (Reference 3), will be used as the first enhancement. The second enhancement proposed by the licensee is to use Table 1, Examination Category R-A, "Risk-Informed Piping Examinations," contained in Code Case N-578-1 as an alternative to the requirements listed in Table 4-1, "Summary of Degradation-Specific Inspection Requirements and Examination Methods" of Reference 1.
NRC Staff Evaluation
The NRC staff has reviewed and evaluated the licensee's proposed RI-ISI program, including those portions related to the applicable methodology and processes, based on guidance and acceptance guidelines provided in RGs 1.174 and 1.178, in Standard Review Plan (SRP) 3.9.8, and in the EPRI-TR-112657, Revision 8-A. An acceptable RI-ISI program plan is expected to meet the five key principles discussed in RGs 1.174 and 1.178, SRP 3.9.8, and the EPRI-TR, as stated below:
- 1. The proposed change meets the current regulations unless it is explicitly related to a requested exemption or rule change.
- 2. The proposed change is consistent with the defense-in-depth philosophy.
- 3. The proposed change maintains sufficient safety margins.
- 4. When proposed changes result in an increase in core damage frequency (CDF) or risk, the increases should be small and consistent with the intent of the NRC's Safety Goal Policy Statement.
- 5. The impact of the proposed change should be monitored by using performance measurement strategies.
The NRC staff determined that the first principle is met in this relief request because an alternative lSI program may be authorized pursuant to 10 CFR 50.55a(3)(i) and, therefore, a relief request is not required.
-4 The second and third principles require assurance that the alternative program is consistent with the defense-in-depth philosophy and that sufficient safety margins are maintained, respectively.
Assurance that the second and third principles are met is based on the application of the approved methodology and not on the particular inspection locations selected. The methodology used to develop the fifth interval RI-ISI program is unchanged from the methodology approved for use in the third and fourth interval. The licensee did state in the May 30, 2013, submittal, that the augmented inspection programs associated with intergranular stress corrosion cracking, service water integrity, flow accelerated corrosion and high energy line breaks remain unaffected by the fifth RI-ISI interval program. The approved methodology was applied to the piping added in accordance with the 2007 Edition through the 200B Addenda of ASME Code,Section XI. The NRC staff determined that since the methodology used to develop the RI-ISI program for the fifth 1 O-year interval is unchanged from the methodology approved for development of the RI-ISI program used in the third and fourth 10-year lSI intervals, augmented programs remain unchanged and the second and third principles are met.
The fourth principle, in that any increase in CDF and risk are small and consistent with the NRC's Safety Goal Policy Statement, requires an estimate of the change in risk. The change in risk estimate is dependent on the location of inspections in the proposed lSI program compared to the location of inspections that would be performed using the requirements of ASME Code,Section XI. The NRC staff has previously determined that it is not necessary to develop a new deterministic ASME Code program for each new 1 O-year interval but, instead, it is acceptable to compare the new proposed RI-ISI program with the last deterministic ASME Code program. The licensee states that a new risk impact analysis was performed. The fifth interval update of the risk impact assessment provided in the response to NRC an RAI dated May 30, 2013 (ADAMS Accession No. ML13151A107), represents a change of 6.51 E-1 0 for Unit 1 and 1.74E-09 for Unit 2, with regards to CDF and 1.51 E-09 for Unit 1, and 2.0BE-09 for Unit 2, with regard to large early release frequency (LERF). The NRC staff determined that these values satisfy the acceptance criteria of RG 1.174 and EPRI TR-112657 when compared to the last deterministic Section XI inspection program. Thus, the NRC staff finds that the licensee's analysis provides assurance that the fourth key principle is met.
The fourth principle also requires demonstration of the technical adequacy of the PRA. As discussed in RGs 1.17B and 1.200, an acceptable change in risk evaluation (and risk-ranking evaluation used to identify the most risk significant locations) requires the use of a PRA of appropriate technical quality that models the as-built and as-operated plant. A review of the QCNPS PRA was conducted under the auspices of the BOiling Water Reactor Owners' Group peer review in February 2000. The licensee states that there were no significant level A facts and observations (F&Os) from the peer review and all significant level B F&Os were addressed and closed out with the completion of a QCNPS PRA model update in 2005. The QCNPS PRA was updated in 2005, and 2010, and assessments of the status of gap analysis relative to the new models and the requirements in Addendum B of the ASME PRA standard were completed after each update.
In the submittal, the licensee refers to EPRI TR-1021467, "Nondestructive Evaluation:
Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs" (Reference 2), which received a safety evaluation (SE) from NRC in January 2012 (ADAMS Accession No. ML11325A340). EPRI TR-1021467 provides guidance on determining the technical adequacy of PRAs used to develop a RI-ISI program that utilizes the traditional methodology as described in EPRI TR-112657, Revision B-A. A licensee that
- S satisfies the Capability Categories (CCs) as described in this topical report need not provide further justification on the adequacy of its PRA for RI-ISI to the NRC staff for review. In the submittal for the relief request, the licensee states that the QCNPS PRA contains two gaps that are related to CC II of the ASME/American Nuclear Society RA-Sa-2009. In response to an NRC PRA, the licensee states that two gaps originally presented in the submittal are met with CC I and identifies two new gaps. According to NRC's SE for EPRI TR-1021467. the two gaps identified in the original submittal only require CC I and do not require CC II. Therefore. the NRC staff determined that those supporting requirements meet the EPRI TR-1021467 CC assignment guidelines. The two new gaps identified in the response to the RAI, however. do not meet the minimum EPRI TR-1 021467 CC assignment guidelines. The disposition of the first gap demonstrates that the licensee's PRA model estimates the number of demands for standby equipment for calculating the failure probabilities based on the operating experience.
surveillance, or maintenance test procedures. The disposition of the second gap demonstrates that the surveillance test procedures are judged to address the appropriate failure modes with respect to the estimated number of demands. The NRC staff agrees that these dispositions provide adequate justification that the quality of licensee's PRA model for those supporting requirements is comparable to EPRI TR-1021467 CC assignment guidelines and finds that any difference in the results from failing to fully comply with EPRI TR-1021467, as described by the licensee will not affect the proposed program. Therefore, the NRC staff finds the QCNPS PRA model suitable for use in this RI-ISI application.
The fifth principle of risk-informed decision making requires that the impact of the proposed change be monitored by using performance measurement strategies. The RI-ISI program is a living program and, as such, is subject to periodic reviews. The licensee indicates that, the consequence evaluation, degradation mechanism assessment, risk ranking, element selection and risk impact assessment steps encompass the living program process applied to the QCNPS RI-ISI program. The May 30, 2013, submittal, stated that examination locations for the fifth interval had changed based on PRA model revisions, plant modifications, and optimize code examination coverage. Therefore, the NRC staff finds that the licensee's proposed alternative provides assurance the fifth principle is met.
Based on the above discussion, the NRC staff concludes that the five key principles of risk informed decision making are ensured by the licensee's proposed fifth 10-year RI-ISI program, and, therefore, the proposed program for the fifth 10-year lSI interval is acceptable pursuant to 10 CFR SO.SSa(a)(3)(i).
4.0 CONCLUSION
S As set forth above, the NRC staff has determined that authorizing the licensee's proposed alternative pursuant to 10 CFR SO.SSa(a)(3)(i) is authorized by law and will not endanger life or property. or the common defense and security, and is otherwise in the public interest given due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Furthermore, the NRC concluded that the proposed alternative will provide reasonable assurance of leak tightness of the subject component. Accordingly. the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements.
- 6 All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
5.0 REFERENCES
- 1. EPRI TR-112657, Revision B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure, Final Report," December 1999.
- 2. EPRI TR-1021467, "Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs,"
June 18, 2012.
- 3. ASME Code Case N-578-1, Risk-Informed Requirements for Class 1,2, or 3 Piping, Method B,Section XI Division 1, ASME, New York, New York, March 28, 2000.
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ALTERNATIVE 15R-03 REGARDING PRESSURE TESTING OF REACTOR PRESSURE VESSEL HEAD FLANGE SEAL LEAK DETECTION SYSTEM EXELON GENERATION COMPANY. LLC QUAD CITIES NUCLEAR POWER STATION. UNITS 1 and 2 DOCKET NOS. 50-254 AND 50-265
1.0 INTRODUCTION
By letter dated September 28,2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12275A070), as supplemented by letter dated January 10, 2013 (ADAMS Accession No. ML13010A456), Exelon Generation Company, LLC (the licensee) submitted Relief Request 15R-03 for the U.S. Nuclear Regulatory Commission (NRC) approval.
The licensee proposed an alternative to certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI. Relief Request 15R-03 relates to system leakage testing of the reactor pressure vessel head (RPVH) flange seal leak detection (Ieakoff) system. Relief Request 15R-03 is requested for the fifth 10-year inservice inspection (lSI) interval of the Quad Cities Nuclear Power Station (QCNPS),
Units 1 and 2, which commenced on April 2, 2013, and will end on April 1, 2023.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR). Section 50.55a(a)(3)(ii), the licensee proposed alternative system leakage test for the RPVH flange seal leak detection system, on the basis that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
2.0 REGULATORY EVALUATION
Section 10 CFR 50.55a(g)(4) of 10 CFR specifies that ASME Code Class 1, 2, and 3, components (including supports) must 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 In-service Inspection (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 inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the ENCLOSURE 3 (15R-03)
- 2 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 conditions listed therein.
Section 50.55a(a)(3) of 10 CFR states, in part, that alternatives to the requirements of Paragraph (g), or portions thereof, may be used when authorized by the NRC, if the licensee demonstrates that: (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.
Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request, and the NRC to authorize, the alternative requested by the licensee.
3.0 NRC STAFF TECHNICAL EVALUATION FOR RELIEF REQUEST 15R-03 ASME Code Components Affected (As stated)
ASME Code Class:
Class 2 Examination Category:
C-H, Table IWC-2500-1 Item No.:
C7.10 Component:
Flange Seal Leak Detection Line Pressure Retaining Components System:
Reactor Pressure Vessel Head Flange Seal Leak Detection System The component, for which RFA 15R-03 is applicable, is listed below:
Plant Drawing Quad Cities, Unit 1 M-35 Sh. 1 Quad Cities, Unit 2 M-77 Sh. 1 Applicable Code Edition and Addenda (As stated)
The [ASME] Code of Record for the fifth 10-year lSI interval at QCNPS, Units 1 and 2, is the 2007 Edition through 2008 Addenda of the ASME Code,Section XI.
Applicable Code Requirement (As stated)
The ASME Code,Section XI, Table IWC-2500-1, Examination Category C-H, Item No. C7.10, requires that all Class 2 pressure retaining components are subjected to a system leakage test with a visual (VT -2) examination in accordance with IWC-5220. This pressure test is to be conducted once each inspection period.
Reason for Request (As stated)
The RPVH flange leak detection line is separated from the reactor pressure boundary by one passive membrane, a silver plated O-ring located on the vessel flange. A second O-ring is
- 3 located on the opposite side of the tap in the vessel flange. This line is required during plant operation in order to indicate failure of the inner flange seal O-ring. Failure of the O-ring would result in the annunciation of a High Level Alarm in the control room. On this annunciation, control room operators would quantify the leakage rate from the O-ring and then isolate the leak detection line from the drywell sump by closing the valve. This action is taken in order to prevent steam cutting the O-ring and the vessel flange. Failure of the inner O-ring is the only condition under which this line is pressurized.
The configuration of this system precludes manual testing while the vessel head is removed, because the odd configuration of the vessel tap combined with the small size of the tap and the high test pressure requirement (1000 pounds per square inch gage (psig) minimum) prevents the tap in the flange from being temporarily plugged. The opening in the flange is only 3/16 inch in diameter and smooth walled, making a high pressure temporary seal very difficult. Failure of this seal could possibly cause ejection of the device used for plugging into the vessel.
A pneumatic test performed with the head installed is precluded due to the configuration of the top head. The top head of the vessel contains two grooves that hold the O-rings. The O-rings are held in place by a series of retainer clips spaced 15 degrees apart. The retainer clips are contained in a recessed cavity in the top head. If a pressure test was performed with the head on, the inner O-ring would be pressurized in a direction opposite to what it would see in normal operation. This test pressure would result in a net inward force on the O-ring that would tend to push it into the recessed cavity that houses the retainer clips. The O-ring material is only 0.050 inch thick with a silver plating thickness of 0.004 inch to 0.006 inch, and could very likely be damaged by this deformation into the recessed areas on the top head.
In addition to the problems associated with the O-ring design that preclude this testing, it is also questionable whether a pneumatic test is appropriate for this line. Although the line will initially contain steam if the inner O-ring leaks, the system actually detects leakage rate by measuring the level of condensate in a collection chamber. This would make the system medium water at the level switch. Finally, the use of a pneumatic test performed at a minimum of 1000 psig would represent an unnecessary risk in safety for the inspectors and test engineers in the unlikely event of a test failure, due to the large amount of stored energy contained in air pressurized to 1000 psig.
Proposed Alternative and Basis for Use (As stated)
The VT-2 examination will be performed on the RPVH flange seal leak detection line during vessel flood-up during a refueling outage. The static head developed due to the water above the vessel flange will allow for the detection of any gross leakage in the line. This examination will be performed with the frequency specified in Table IWC-2500-1 for the system leakage test, once each inspection period.
Duration of Relief (As stated)
Relief Request 15R-03 is applicable for the fifth 10-year lSI interval of QCNPS, Units 1 and 2, which commenced on April 2, 2013, and ends on April 1, 2023.
-4 4.0 STAFF EVALUATION The NRC staff has evaluated Relief Request 15R-03 pursuant to 10 CFR 50.55a(a)(3)(ii), and evaluated whether the compliance with the specified requirements of 10 CFR 50.55a(a)(g), or portions thereof, would result in hardship or unusual difficulty. The ASME Code,Section XI, Table IWC-2500-1, Examination Category C-H, requires that all Class 2 pressure retaining components are subjected to a system leakage test and a VT-2 examination in accordance with IWC-5220 each inspection period. For the system leakage test of the RPVH flange seal leak detection (Ieakoff) line system, the licensee proposed alternative (i.e., pressure) to the above requirements (Le., testing at system operating pressure) on the basis that the specified requirements would result in hardship or unusual difficulty. The proposed alternative is to perform the system leak test using static pressure head developed by water above the vessel flange when the vessel is flooded for refueling during the outage. The frequency of examination will be once each inspection period in accordance with the Table IWC-2500-1 requirement.
Within the context of Relief Request 15R-03, the licensee specified the limitations that precluded the IWC-5220 system leak test. The licensee stated that the leakoff line is designed such that the inner O-ring seal has to be failed in order to perform the required leak test. The licensee has considered performing a pneumatic pressure test with the RPVH on. The licensee stated the issue with this type of test is that the inner O-ring is pressurized in the direction that is opposite to the direction that is pressurized during normal operation. The resulting net inward force from the pressure tends to push the O-ring into the recessed cavity that houses the retainer clips holding the O-ring in place. The unusual forced deformation of the O-ring into the recessed cavity would very likely damage the O-ring and the examination WOUld, in all likelihood be unsuccessful. The licensee has also considered temporarily plugging the tap in the flange and pressurizing the line manually with the RPVH off. The licensee stated that the issue with this type of test is that the leakoff line opening in the flange is smooth walled making a high pressure temporary seal very challenging. The temporary seal would very likely fail causing ejection of the plugging device into the vessel. In addition, the use of a pneumatic test would expose the personnel conducting the leak test to an unnecessary risk in the unlikely event of a test failure due to presence of large amount of stored energy in the pressurized air. The NRC staff finds that the limitations provided by the licensee and discussed above constitute a justifiable hardship if the leak test were to be performed in accordance with the pressure requirement of IWC-5220.
By letter to the NRC staff dated January 10, 2013, the licensee provided an estimate for personnel radiation exposure if the above mentioned manual pressurization of the RPVH flange leak detection line would be performed. The licensee stated that for performance of the examination and resulting corrective maintenance, the personnel would be exposed up to 75 millirem per hour radiation field with total station exposure estimated to be up to 7.3 rem radiation dose. The licensee stated that this estimate to radiation exposure is based on a single RPVH disassembly and reassembly for refueling in the previous outages. The licensee stated that exposure to the above mentioned levels of radiation is not in accordance with industry principles of maintaining personnel radiation exposure as low as reasonably achievable (ALARA). The NRC staff notes that a large part of this radiation dose would be incurred during a routine refueling outage, if the test were conducted prior to initial removal of the reactor vessel head, significant personnel safety issues would occur while the plant is at normal operating temperature. The NRC staff finds that personnel exposure levels of radiation above ALARA, and potential personnel safety impacts constitute a justifiable hardship.
- S In addition, the NRC staff notes that during normal operating condition over the life of the plant operation, the likelihood that the RPVH flange leakoff line piping components be exposed to environments and operational conditions that cause potential materials degradation is very low, because this line would have only been used in the event of inner O-ring seal failure and leak.
In summary, the NRC staff finds that the licensee's proposed alternative leak test (i.e., using static pressure head developed by water above the vessel 'flange when the vessel is flooded for refueling), accompanied with a VT -2 examination each inspection period, is adequate to detect leakage in the RPVH flange seal leak detection line system and provides reasonable assurance of structural integrity or leak tightness of the system. The NRC staff also determined that complying with the requirements would result in hardship or unusual difficulty due to potential personnel safety, equipment damage, and personnel exposure to radiation. Therefore, the staff authorizes Relief Request ISR-03.
S.O CONCLUSION As set forth above, the NRC staff determines that the proposed alternative provides reasonable assurance of structural integrity or leak tightness of the subject components and complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR SO.SSa(a)(3)(ii).
Therefore, the NRC staff authorizes the use of Relief Request ISR-03 for the fifth 10-year lSI interval of QCNPS, Units 1 and 2, which commenced on April 2, 2013, and end on April 1, 2023.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including the third party review by the Authorized Nuclear In-service Inspector.
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 15R-04 REGARDING CONTINUOUS PRESSURE MONITORING OF THE CONTROL ROD DRIVE SYSTEM ACCUMULATORS EXELON GENERATION COMPANY, LLC QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265
1.0 INTRODUCTION
By letter dated September 28. 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12275A070). Exelon Generation Company. LLC (the licensee) submitted proposed alternative Relief Request 15R-04, "Request for Relief for Continuous Pressure Monitoring of the Control Rod Drive (CRD) System Accumulators," for U.S. Nuclear Regulatory Commission (NRC) review and authorization. The licensee superseded Relief Request 15R-04 by letter dated January 24, 2013 (ADAMS Accession number MI13025A 161), Specifically, the licensee requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) for performing a system leakage test, visual (VT-2) examination of the nitrogen side of the control rod drive (CRD) accumulators, including the attached piping, at Quad Cities Nuclear Power Station (QCNPS). Units 1 and 2. The licensee states that continuous monitoring of CRD accumulator pressure functions as a pressure decay-type test and that technical speCification (TS) surveillance requirement (SR) exceed the ASME Code requirement for a VT -2 examination. The licensee requested authorization to use the proposed alternative pursuant to Title 10 of the Code of Federal Regulations Part 50 (10 CFR).
Section 55a(a)(3)(i), on the basis that the proposed alternative provides an acceptable level of quality and safety.
2.0 REGULATORY EVALUATION
Pursuant to 10 CFR 50.55a(g)(4), Inservice Inspection [lSI] Requirements, 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 Inservice Inspection 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 inservice examination of components and system pressure tests conducted during the first ENCLOSURE 4 (15R-04)
- 2 10-year inspection 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 inspection interval, subject to the conditions listed therein.
Section 55a(a)(3) of 10 CFR 50 states, in part, that alternatives to the requirements of 10 CFR 50.55a(g) may be used when authorized by the NRC, if: (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The NRC staff finds that the regulatory authority exists to authorize the licensee's proposed alternative to the ASME Code requirement on the basis that the proposed alternative provides an acceptable level of quality and safety. Accordingly, the NRC staff has reviewed and evaluated the licensee's request pursuant to 10 CFR 50.55a(a)(3)(i).
3.0 NRC STAFF TECHNICAL EVALUATION FOR RELIEF REQUEST 15R-04 Components for which Relief is Being Requested Nitrogen side of the CRD accumulators including the attached piping, Examination Category C-H, Item No. C7.10.
ASME Code Requirements The [ASME Code] of record for QCNPS, Units 1 and 2, for the fifth 10-year lSI interval that commenced on April 2, 2013, and ends on April 1, 2023, is the 2007 Edition through the 2008 Addenda of the ASME Code,Section XI.
Paragraph IWC-2500, Table IWC-2500-1, Examination Category C-H, Item No. C7.10, requires that all Class 2 pressure retaining components be subject to a system leakage test in accordance with IWC-5220 with a VT-2 examination once each inspection period.
Licensee's Proposed Alternative As an alternate to the VT-2 examination requirements, the licensee proposes to perform continuous pressure monitoring of the nitrogen side of the CRD accumulators.
The licensee has cited the following precedents in support of the proposed alternative:
Dresden Nuclear Power Station, Units 2 and 3, fourth lSI Interval, Relief Request 14R
- 07. (ADAMS Accession No. ML032370480).
LaSalle County Nuclear Power Station, Units 1 and 2, third lSI Interval, Relief Request 13R-09. (ADAMS Accession No. ML073610587).
Quad Cities Nuclear Power Station, Units 1 and 2, fourth lSI Interval, Relief Request 14R-06. (ADAMS Accession No. ML033560386).
- 3 Licensee's Basis for Requesting Relief As required by the TSs at QCNPS, Units 1 and 2, the CRD system accumulator pressure must be greater than or equal to 940 pounds per square inch gauge (psig) to be considered operable. During normal operation, the accumulators are isolated from the source of make-up nitrogen and the accumulator pressure is continuously monitored by system instrumentation. Should accumulator pressure fall below 1000 psig, an alarm is received in the control room. Since the accumulators are isolated from the source of make-up nitrogen, continuous monitoring of pressure functions as a pressure decay-type test.
If an alarm is received in the control room, the pressure drop for the associated accumulator is recorded and the accumulator is recharged in accordance with QCNPS procedures. If an accumulator requires charging more than twice in a 30-day period, then a leak check is performed to determine the cause of the pressure loss. When leakage is detected, corrective actions are taken to repair the leaking component as required by QCNPS procedures. An additional VT-2 examination performed once per inspection period would not provide an increase in quality and safety Relief is requested from the VT-2 examination requirement specified in Table IWC-2500 1, for the nitrogen side of the CRD system accumulators, including attached piping, on the basis that QCNPS TS SR requirements exceed the ASME Code requirement for a VT-2 examination.
NRC Staff Evaluation
In the ASME Code a VT-2 examination is required to be performed on all Class 2 pressure retaining components once during each inspection period. The VT-2 examination is performed at normal operating system pressure with the fluid in the system serving as the pressurizing medium. In order to perform a VT-2 examination of the nitrogen side of the CRD accumulators and the attached piping, it is necessary to apply a soap solution to all surfaces of the subject components and visually examine the surfaces for soap bubbles that would indicate leakage. If a leak occurs between successive visual examinations and continuous pressure monitoring is not employed, the accumulator pressure could drop below the pressure required by the plant TSs for operability for an extended time before it is detected.
As an alternative to the ASME Code-required VT-2 examinations, the licensee has proposed to utilize the continuous on-line pressure monitoring, currently used to confirm pressure, is consistent with required plant TSs requirements. If the nominal pressure falls below 1000 psig, an alarm is triggered in the plant control room and plant corrective action is taken to re-charge the specific accumulator to maintain its function. If pressure decay below 1000 pSig is observed more than two times in any 30-day period, plant corrective action is to perform a leak check to determine the cause of the pressure loss and, when leakage is detected. perform corrective actions to repair the leaking component.
The proposed alternative requires that the pressure of each accumulator be continuously monitored by system instrumentation. Since each accumulator is isolated from the
-4 source of make-up nitrogen, the NRC staff finds that continuous monitoring of the CRD accumulator pressure functions as a pressure decay-type test and any leakage from the accumulator would be detected by normal system instrumentation. The NRC staff also finds that continuous pressure monitoring ensures immediate detection of large leaks and long-term continuous monitoring results in high detection sensitivity for even small leaks. Furthermore, continuous monitoring ensures that the CRD system accumulator minimum pressure of 940 psig, required plant TSs for operability. is also continuously monitored and maintained. The NRC staff finds that continuous on-line monitoring of the CRD system accumulator pressure and the required corrective actions when leakage is detected provide reasonable assurance that adequate pressure for CRD actuation is maintained. As such. the licensee's proposed alternative provides an acceptable level of quality and safety.
4.0 CONCLUSION
As set forth above, the NRC staff determines that Relief Request 15R-04, "Request for Relief for Continuous Pressure Monitoring of the Control Rod Drive (CRD) System Accumulators," provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(i) and authorizes use of the proposed alternative at QCNPS, Units 1 and 2, during the fifth 10-year lSI interval which commenced on April 2, 2013, and ends on April 1, 2023.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized in the subject proposed alternative remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST ISR-06 REGARDING REACTOR PRESSURE VESSEL INTERNALS AND COMPONENTS INSPECTIONS REQUIREMENTS AT BOILING-WATER REACTORS EXELON GENERATION COMPANY, LLC QUAD CITIES NUCLEAR POWER STATION! UNITS 1 AND 2 DOCKET NOS. SO-2S4 AND SO-26S
1.0 INTRODUCTION
By letter dated September 28, 2012, as supplemented by letter dated November 28, 2012, Exelon Generation Company, LLC (the licensee) submitted Relief Request 1SR-06 for its fifth 10-year inservice inspection (lSI) interval program plan for its reactor vessel internals (RVI) components at Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. In this safety evaluation (SE), the term "RVI components" include 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 the lSI of reactor pressure vessel interior surfaces, attachments, and core support structures.
2.0 REGULATORY EVALUATION
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 Title 10 Code of Federal Regulations (10 CFR). Section SO.SSa(g), except where specific relief has been granted by the U.S. Nuclear Regulatory Commission (NRC) pursuant to 10 CFR SO.SSa(g)(6)(i). 10 CFR SO.SSa(a)(3) states that alternatives to the requirements of 10 CFR SO.SSa(g) may be used when authorized by the 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 SO.SSa(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-ENCLOSURE S (ISR-06)
- 2 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 state that lSI examination of components and system pressure tests conducted during the first 10-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 Codes of record for the fifth 10-year lSI interval for QCNPS, Units 1 and 2, is ASME Code,Section XI, 2007 Edition through the 2008 Addenda.
3.0 NRC STAFF TECHNICAL EVALUATION OF RELIEF REQUEST 15R-06 The Components for Which an Alternative is Requested In ASME Code,Section XI, Class 1, Examination Categories 8-N-1 and 8-N-2, Code Item Nos. 813.10, Vessel Interior; 813.20, Interior Attachments within 8eltline Region; 813.30, Interior Attachments 8eyond 8eltline Region; and, 813.40, Core Support Structure.
Examination Requirements from Which an Alternative is Requested In ASME Code,Section XI, it requires the visual (VT) examination of certain RVI components and are included in Table IW8-2500-1, Categories 8-N-1 and 8-N-2, and identified with the following item numbers:
813.10 - 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.
Licensee's 8asis for Requesting an Alternative and Justification for Granting Relief In relief request 15R-06, the licensee, in lieu of ASME Code,Section XI, requirements, submitted an alternative inspection program per the 8WRVIP guidelines for 8-N-1 and 8-N-2 reactor pressure vessel interior surfaces, attachments, and core support structures at QCNPS, Units 1 and 2. The licensee stated that implementation of the alternative inspection program will
- 3 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 of the reactor pressure vessel interior surfaces, attachments, and core support structures in accordance with BWRVIP guidelines. 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). The licensee stated that by letter dated April 30, 2008, the NRC staff issued an SE for the implementation of inspection and evaluation (I&E) guidelines addressed in the relevant BWRVIP reports in lieu of the ASME Code,Section XI, lSI requirements for the reactor pressure vessel interior surfaces, attachments, and core support structures at QCNPS, Units 1 and 2, for the licensee's fourth lSI interval.
Alternative Examination In lieu of the requirements of the applicable Edition and Addenda of the ASME Code,Section XI, the licensee proposed to examine the QCNPS, Units 1 and 2, RVI components in accordance with BWRVIP guideline requirements. 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-47-A, "BWR Lower Plenum Inspection and Flaw Evaluation Guidelines" BWRVIP-48-A, "Vessel 10 Attachment Weld Inspection and Flaw Evaluation Guidelines" BWRVIP-76, Revision 1, "BWR Core Shroud Inspection and Flaw Evaluation Guidelines" BWRVIP-94, Revision 2, BWRVIP Program Implementation Guide" BWRVIP-138, Revision 1, "BWRVIP Updated Jet Pump Beam Inspection and Flaw Evaluation Guidelines" BWRVIP-183, "BWRVIP, Top Guide Grid Beam Inspection and Flaw Evaluation" The license stated that inspection services by an authorized inspection agency will be applied to the proposed alternative. The licensee further indicated that the BWRVIP has established reporting protocol for examination results and deviation that are consistent with the requirements of BWRVIP-94 report. The licensee clarified that a revised version of a BWRVIP report will meet I&E guidelines of its original version, and if it does not, NRC staff approval is mandatory prior to its implementation.
-4 The licensee stated that for the RVI code components at QCNPS, Units 1 and 2, a conditional authorization (dated April 30, 2008) for the fourth lSI interval was approved by NRC staff. The licensee further stated that Inspections that occurred during May 2006, outage for QCNPS, Unit 1, and April 2006, outage for QCNPS, Unit 2, revealed no indications or flaws were detected in the reactor pressure vessel interior surfaces, attachments, and core support structures.
The licensee, in Table 1 of its submittal dated September 28,2012, provided a comparison of the ASME Code,Section XI, examination requirements for B-N-1 and B-N-2 categories of the reactor pressure vessel interior surfaces, attachments, and core support structures with the above current BWRVIP I&E guidelines. As an example, in Attachment 1 of the submittal, the licensee provided additional information regarding the BWRVIP 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 Nos. B13.10, B13.20, B13.30, and B13.40, addressed in page 2 of this SE).
Core Spray Piping Jet Pump Core Shroud Core Shroud Support and Core Support Structure B13.10 B13.20 B13.30 B13.40 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 BWRVIP inspection guidelines require more frequent inspections of some RVI components than the corresponding ASME Code,Section XI, lSI program. The licensee claimed 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 QCNPS, Units 1 and 2, reactor pressure vessel interior surfaces, attachments, and core support structures would provide an equivalent 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 Attachment 1 of this SE.
4.0 STAFF EVALUATION The NRC staff reviewed the information provided by the licensee in its submittal dated September 28,2012, as supplemented by letter dated November 28,2012, regarding its proposed alternatives to the ASME Code,Section XI, lSI requirements and the technical bases for the licensee's proposed alternatives. The NRC staff reviewed the status of each of the referenced BWRVIP guidance documents 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 NRC staff did, however, identify some issues which required additional clarification by the licensee.
- 5 In a request for additional information (RAI) 15R-06-02, the NRC staff stated that: Section 4.1, Item 5, of the BWRVIP-100-A report, "Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Core Shrouds," states that fracture toughness values of stainless steel materials that are exposed to a neutron fluence value greater than 1 x 1021 n/cm2 (E> 1 MeV) are lower than those used in Appendix C of the BWRVIP-76 report, "BWR Core Shroud Inspection and Flaw Evaluation Guidelines." The NRC staff requested the licensee to identify whether the core shroud welds and base materials will be exposed to a neutron fluence value greater than 1 x 1021 n/cm2 (E > 1 MeV) during the current lSI interval. Since the inspection frequency in the BWRVIP-76 report is based on fracture toughness values which are not consistent with the BWRVIP-100-A report, the NRC staff requested that the licensee address the following issue:
The inspection frequency and strategy that are specified in Section 3 of the BWRVIP-76 report require further evaluation taking into account the lower fracture toughness values that are specified in BWRVIP-100-A report.
The licensee, by letter dated May 30, 2013, stated that it will take into account the lower fracture toughness values specified in BWRVIP-100-A report for establishing the inspection frequency for the core shroud components. Therefore, the NRC staff considers this issue to be closed. of this SE includes attributes related to inspection techniques and frequency of inspections for various RVI components in the QCNP, Units 1 and 2. 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 QCNPS, Units 1 and 2, is included in Attachment 1 of this SE.
On April 30, 2008, the NRC staff issued an SE for the QCNPS, Units 1 and 2, fourth lSI interval, which allowed the licensee to implement the BWRVIP I&E guidelines in lieu of the ASME Code,Section XI, lSI requirements for reactor pressure vessel interior surfaces, attachments, and core support structures. In this SE, the NRC staff imposed a condition which stated that the licensee is required to continue to implement the ASME Code,Section XI, lSI requirements for the jet pump code components. I&E guidelines for the jet pump code components are addressed in BWRVIP 41, "BWRVIP Jet Pump Assembly Inspection and Flaw Evaluation Guidelines." The NRC staff imposed the condition in the SE dated April 30, 2008, because the licensee did not include BWRVIP-41 in its submittal for the fourth lSI interval. Similarly, for the fifth lSI interval, the licensee in its submittal dated September 28, 2012, did not include the BWRVIP-41 as part of the lSI program for the RVI components. Therefore, the licensee is required to inspect the jet pump code components per ASME Code,Section XI, criteria. According to Appendix A of NUREG-6 1796, "Safety Evaluation Report, Related to the License Renewal of the Dresden Nuclear Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station Units 1 and 2," for QCNP, Units 1 and 2, the licensee was to implement BWRVIP-41 for the jet pump components. Consistent with this requirement, compliance with BWRVIP-41 for ASME Code and non-code jet pump components at QCNPS, Units 1 and 2, as becomes necessary.
In request 15R-06 for QCNPS, Units 1 and 2, the licensee stated that no indications or flaws were detected in RVI Code components since May 2006, outage for QCNPS, Unit 1, and April 2006, outage for QCNPS, Unit 2. Based on this information, the staff determined that the active aging degradation mechanisms in the ASME Code,Section XI, RVI components may be
-6 stabilized or arrested. 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.
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 NRC staff-approved inspection guidelines for the feedwater nozzle and sparger welds, the licensee's proposed alternative will identify aging degradation of the RVI components in a timely manner. Therefore, the 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.
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 reactor pressure vessel interior surfaces, attachments, and core support structures is maintained with an acceptable level of quality and safety. However, this does not include the requested alternatives which apply to the inspection of jet pump assembly components based on the provisions of BWRVIP-41. Alternatives based on this topical report are not approved.
Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the licensee's proposed alternative for the QCNP, Units 1 and 2, are authorized, with the condition stated above, for the fifth 10-year lSI intervals at QCNPS, Units 1 and 2. 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 will continue to be inspected in accordance with the ASME Code,Section XI requirements. Consistent with the requirements addressed in Appendix A of NUREG-1796 for QCNPS, Units 1 and 2, I&E guidelines addressed in the relevant BWRVIP reports should be implemented for the non-ASME Code,Section XI, RVI components.
ATTACHMENT (to Enclosure 5)
Comparison of ASME Category B-N-1 and B-N-2 Requirements With BWRVIP Guidance Requirements (1)
ASME Item No.
Table IWB-2500 1
Component ASME Exam Scope ASME Exam ASME Frequenc y
Applicable BWRVIP Document BWRVIP Exam Scope BWRVIP Exam BWRVIP Frequency r---
B13.10 Reactor Vessel Interior Accessible Areas (Non specific)
VT-3 Each period BWRVIP-18-A, 25, 26-A, 27-A, 38, 47-A, 48-A, 76 Revision 1, and 138, Revision 1 BWRVIP examinations satisfy ASME Code,Section XI, VT-3 inspection requirements.
B13.20 Interior Attachments Within Beltline - Jet Pump Riser Braces Accessible Welds VT-1 Each 10 year Interval BWRVIP-48-A Table 3-2 Riser Brace AUachmen t
EVT-1 100% in first 12 years (with 50% to be inspected in the first 6 years); 25% during each subsequent 6 years Lower Surveillance Specimen BWRVIP-48-A Bracket VT-1 Each 10-year Interval Holder Brackets Table 3-2 Atlachmen t
ENCLOSURE 5 (15R-06)
- 2 ASME Item No.
Component ASME Exam ASME ASME Applicable BWRVIP BWRVIP BWRVIP Frequency Table IWB-2S00 Scope Exam Frequenc BWRVIP Exam Exam 1
y Document Scope B13.30 Guide Rod Brackets Accessible VT-3 Each 10 BWRVIP-48-A Bracket VT-3 Each 10-year Welds year Table 3-2 Attachmen Interval Interval t
,-:~~ ~
~ ~-~ ~
Steam Dryer Support BWRVIP-48-A Bracket VT_3\\Oi Each 1 O-year Brackets Table 3-2 Attachmen Interval t
I Feedwater Sparger Brackets BWRVIP-48-A Bracket VT_3,::l)
Each 10-year Table 3-2 Attachmen Interval t
~
Core Spray Piping Brackets BWRVIP-48-A Bracket VT-3 Each 10-year Table 3-2 Attachmen Interval t
Upper Surveillance BWRVIP-48-A Bracket VT-3 Each 1 O-year I
Specimen Holder Brackets Table 3-2 Attachmen Interval Rarely t
Shroud Support (Weld H9)
Accessible BWRVIP-38, Weld H9,L)
EVT-1 or Maximum of 6 years 3.1.3.2 Figures UT for one sided EVT 3-2 and 1, Maximum of 10 3-S
},:ears for UT Weld H12 Shroud Support
-SWRVIP-38, Weld H12 Per When accessible Legs 3.2.3 BWRVIP-38 NRCSER (7-24-200),
inspect with appropriate I
method (4)
- 3 ASME Item No.
Component ASME Exam ASME ASME Applicable BWRVIP BWRVIP BWRVIP Frequency Table IWB-2S00 Scope Exam Frequency BWRVIP Exam Exam 1
Document Scope B13.40 Shroud Support Weld H10 Accessible Surfaces VT-3 Each 10 year Interval BWRVIP-38 3.1.3.2, Figures 3-2 and 3-S Shroud Support H10 weld and Leg EVT-1or UT Based on as found conditions, to a Maximum 6 years for one-sided EVT-1, 10 years for UT where accessible Shroud Vertical Welds c--~ ~ ~
~ -----~
BWRVIP-76 R1,3.3, Figures 3-1 and 3-3
~~~~
~--~~
Vertical and Ring Segment Welds as applicable EVT-1or UT Maximum 6 years for one-sided EVT-1, 10 years for UT BWRVIP-76 EVT-1or Maximum 6 years for 2.3, Figure 3 UT one-sided EVT-1, 10 3
years for UT Shroud Repairs (3)
_c_~~~
BWRVIP-76, R1 Section 3.S Tie-Rod Repair VT-3 Per designer recommendations per BWRVIP-76 R1 I
Note (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.
Note (2) In accordance with Appendix A of BWRVIP-38, a site specific evaluation will determine the minimum required weld length to be examined.
Note (3) Shroud repairs are currently installed at QCNP, units 1 and 2.
Note (4) When inspection tooling and methodologies are available, they will be utilized to establish a base line inspection of these welds.
Note (5) QCNP, Units 1 and 2 do not have furnace sensitized stainless steel (E309/E308, or E308L1E309L) or Alloy 182 weld, therefore, the ASME Code,Section XI required VT-3 examinations will be performed in accordance with BWRVIP-48-A, Table 3-2.
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF 15R-09 REGARDING USE OF ASME CODE CASE N-532-4 EXELON GENERATION COMPANY, LLC QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265
1.0 INTRODUCTION
By letter dated September 28, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12275A070), Exelon Generation Company, LLC (the licensee) requested an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, "Rules for Inservice Inspection [lSI]
of Nuclear Power Plant Components," 2007 Edition through the 2008 Addenda, Subsection IWA-2441 (b), for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(a)(3)(i), the licensee requested approval of Relief Request 15R-09, to allow the use of ASME Code Case N-532-4 (Reference 1), for the fifth i0-year interval at QCNPS.
2.0 REGULATORY EVALUATION
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 preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for In service Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of components. Section 50.55a(g)(4)(i), requires that inservice examination of components and system pressure tests conducted during the first 10-year inspection interval comply with the requirements of the latest edition and addenda of Section XI of the ASME Code incorporated by reference in Section 10 CFR 50.55a(b )(2), 12 months prior to the issuance of the operating license, subject to the conditions listed therein. Section 50.55a(g)(4)(ii) requires that inservice examination of components and system pressure tests conducted during subsequent 10-year inspection intervals comply with the requirements of the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b)(2), 12 months prior to the start of the 120-month inspection interval, subject to the conditions listed therein.
ENCLOSURE 6 (15R-09)
- 2 In 10 CFR 50.55a(a)(3) it states that alternatives to the requirements of paragraph (g) may be used, when authorized by the U.S. Nuclear Regulatory Commission (NRC), if (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. 10 CFR 50.55a(g)(5)(iii) states that if the licensee has determined that conformance with certain code requirements is impractical for its facility, the licensee shall notify the Commission and submit, as specified in 10 CFR 50.4, to support the determinations.
The NRC staff finds that there is regulatory basis for the licensee to request, and the NRC to authorize this alternative, pursuant to the technical evaluation that follows. The information provided by the licensee in support of the request has been evaluated by the NRC staff and the bases for disposition are documented below.
3.0 NRC STAFF TECHNICAL EVALUATION FOR RELIEF REQUEST 15R-09 Licensee's Request for Alternative Code Requirements ASME Code Section XI, IWA-2441(b), requires code cases to be used in an lSI plan to be applicable to the edition and addenda specified in the inspection plan.
The Applicability Index for Section XI code cases shows the applicability of ASME Code Case N-532-4 to be the 1981 Edition with the Winter 1983 Addenda through the 2004 Edition with the 2005 Addenda.
The QCNPS lSI program is based on the ASME Code,Section XI, 2007 Edition with the 2008 Addenda.
Licensee's Proposed Alternative The licensee proposes to utilize ASME Code Case N-532-4 with the 2007 Edition with the 2008 Addenda for the fifth 10-year lSI interval at QCNPS, Units 1 and 2.
Basis for Proposed Alternative The NRC has accepted ASME Code Case N-532-4 in Regulatory Guide (RG) 1.147, Revision 16 (Reference 2), as an acceptable alternative to the repair/replacement activity documentation requirements and inservice summary report preparation and submission requirements of Section XI.
NRC Staff Evaluation
The licensee's proposed alternative would allow QCNPS to utilize ASME Code Case N-532-4 with the 2007 Edition through the 2008 Addenda of ASME Section XI. Code Case N-532-4 is listed as acceptable in Table 1 of Reference 2. This indicates that the NRC staff has found the
- 3 alternative requirements of the code case acceptable for licensees to utilize in lieu of the requirements of ASME Section XI. Therefore, the repair/replacement activity documentation requirements and inservice summary report preparation and submission requirements of N-532-4 are acceptable to the NRC staff.
The NRC staff determined that the Applicability Index shows the latest code applicable to N 532-4 being the 2004 Edition with the 2005 Addenda is a timing issue. The 2004 Edition with the 2005 Addenda was the latest code approved by the ASME Code committee when they approved N-532-4. Footnote 1 to N-532-4 states, in part, that all references to IWA-4000 and IWA-6000 used in the case refer to the 2004 Edition with 2005 Addenda of Section XI and Table 3 is provided to provide accurate references for earlier ASME code editions and addenda.
The NRC staff reviewed changes made to the paragraphs/subparagraphs of Section XI referenced in N-532-4 in editions and addenda from 2004 Edition with 2005 Addenda up to the 2008 Addenda. This review found that there were no changes in IWA-6210{c), (d), (e) and (f),
IWA-6220, IWA-6230(b), (c) and (d), IWA-6240(b), and IWA-6350, that would impactthe requirements spelled out in N-532-4. Therefore, the NRC staff finds that the licensee can utilize Code Case N-532-4 with the 2007 Edition with the 2008 Addenda of Section XI.
Based on the above, the NRC staff has determined that the licensee's proposed alternative will not impact the repair/replacement activity documentation requirements and inservice summary report preparation and submission requirements of Code Case N-532-4. Therefore, the NRC staff finds that the licensee's proposed alternative wilt provide an acceptable level of quality and safety.
4.0 CONCLUSION
As set forth above, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a, and the proposed alternative Relief Request 15R-09 provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3){i), and is in compliance with the ASME Code requirements.
Therefore, the NRC staff authorizes use of Relief Request 15R-09 for the fifth 10-year lSI interval at QCNPS, Units 1 and 2. All other ASME Code,Section XI, reqUirements for which relief was not specifically requested and approved in the subject request for relief remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
5.0 REFERENCES
- 1. Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 16, October 2010. (ADAMS Accession No. ML101800536)
- 2. American Society of Mechanical Engineers (ASME) Code Case N-532-4, "Repair/Replacement Activity Documentation and Inservice Summary Report Preparation and Submission,Section XI, Division 1," April, 19,2006.
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR RELIEF REQUEST 15R-10 REGARDING REPAIR OF CLASS 2 AND 3 PIPING USING ASME CODE CASE N-661-1 QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2 EXELON GENERATION COMPANY, LLC DOCKET NOS. 50-254 AND 50-265
1.0 INTRODUCTION
By letter dated September 28, 2012 (Agencywide Documents and Access Management System, (ADAMS) Accession No. ML12275A070), Exelon Generation Company, LLC (the licensee) submitted the proposed fifth Interval10-year inservice inspection (lSI) program for the Quad Cities Nuclear Power Station (QCNPS). Units 1 and 2. In accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, the lSI program is required to comply with the latest edition and addenda of the ASME Code incorporated by reference in Title 10, Code of Federal Regulations. Part 50 (10 CFR). Section 50.55a. 12 months prior to the start of the interval in accordance with 10 CFR 50.55a(g)(4)(ii).
The fifth lSI interval for QCNPS. Units 1 and 2 commenced on April 2, 2013, and ends by April 1. 2023.
The licensee will adopt the 2007 Edition through the 2008 Addenda of the ASME Code,Section XI. as the Code of Record at QCNPS, Units 1 and 2, for the upcoming fifth 10-year lSI interval program. Within the proposed lSI program, the licensee proposed Relief Request 15R-10 which is related to the use of ASME Code Case N-661-1 to repair ASME Class 2 and 3, carbon steel piping for raw water service. This safety evaluation (SE) is specifically related to Relief Request 15R-10.
2.0 REGULATORY EVALUATION
Pursuant to 10 CFR 50.55a(g)(4), ASME Code, Class 1, 2 and 3, components (including supports) must meet the requirements. except the design and access provisions and the preservice examination requirements, set forth in the ASME Code.Section XI. "Rules for Inservice Inspection (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 inservice examination of components and system pressure tests conducted during the first 1 O-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.
ENCLOSURE 7 (15R-10)
- 2 Pursuant to 10 CFR 50.55a(a)(3) alternatives to requirements may be authorized by the NRC if the licensee demonstrates that: (i) the proposed alternatives 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.
Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request and the Commission to authorize the alternative requested by the licensee.
3.0 NRC STAFF TECHNICAL EVALUATION FOR RELIEF REQUEST 15R-10 ASME Code Component(s) Affected The ASME Code, Class 2 and 3, piping applicable under ASME Code Case N-661-1, "Alternative Requirements for Wall Thickness Restoration of Class 2 and 3 Carbon Steel Piping for Raw Water Service."
Applicable Code Edition and Addenda
The ASME Code,Section XI, 2007 Edition through the 2008 Addenda.
Applicable Code Requirement
The IWA-2441 (b) requires code cases be applicable to the edition and addenda specified in the inspection plan. ASME Code Case N-661-1 provides requirements that may be used to restore wall thickness for raw water piping systems that have experienced internal wall thinning.
Reason For Request On April 2,2013, QCNPS, Units 1 and 2, will start its fifth 10-year lSI interval program under the requirements of the 2007 Edition through the 2008 Addenda of the ASME Code,Section XI.
When implementing this edition of the ASME Code,Section XI. Paragraph IWA-2441(b).
requires code cases be applicable to the edition and addenda specified in the lSI Plan.
The ASME Code Case N-661-1 has an applicability limited up to the 2004 Edition through the 2005 Addenda. Since ASME Code Case N-661-1 only applies up to the 2004 Edition through the 2005 Addenda, Paragraph IWA-2441(b) does not allow the use of ASME Code Case N-661-1 for the QCNPS fifth 10-year lSI interval program.
Proposed Alternative and Basis for Use (As stated)
The QCNPS, Units 1 and 2, requests the applicability of ASME Code Case N-661-1 be extended to the 2007 Edition through the 2008 Addenda for use in the plant's fifth lSI interval program. The NRC has accepted the use of ASME Code Case N-661-1 as an acceptable method for restoring wall thickness for raw water piping systems that have experienced internal wall thinning in the latest revision of Regulatory Guide (RG) 1.147, Revision 16.
No technical changes to ASME Code Case N-661-1 are being proposed in this relief request.
This relief request is being submitted to correct a timing situation, which has resulted from the
- 3 application of the 2007 Edition through the 2008 Addenda of ASME Code,Section XI, for QCNPS, Units 1 and 2. Since no technical change is proposed, QCNPS considers this alternative provides an acceptable level of quality and safety, and is consistent with provisions of 10 CFR 50.55a(a)(3)(i).
Duration of Proposed Alternative (As stated)
Relief is requested for the fifth 10-year inspection interval for QCNPS, Units 1 and 2.
NRC Staff Evaluation
The licensee requested relief from IWA-2441 (b) of the 2007 Edition through the 2008 Addenda of the ASME Code,Section XI, which will be the Code of Record for the fifth 10-year lSI interval.
As an alternative, the licensee proposed to extend the applicability of ASME Code Case N-661-1 to the 2007 Edition through the 2008 Addenda of the ASME Code,Section XI, for the repair of Class 2 and 3, carbon steel piping of the raw water service system. However, Code Case N-661-1 limits its applicability to the 2004 Edition through the 2005 Addenda of the ASME Code,Section XI.
The NRC staff notes that 10 CFR 50.55a has incorporated by reference the 2007 Edition through the 2008 Addenda of the ASME Code. Therefore, the licensee is permitted to use the edition and addenda specified in 10 CFR 50.55a. The NRC also notes that 10 CFR 50.55a does not impose conditions on Code Case N-661-1. The NRC staff has conditionally accepted Code Case N-661-1 in RG 1.147, Revision 16, with the two following conditions: (1) If the cause of the [piping] degradation has not been determined, the repair [of the degraded pipe] is only acceptable until the next refueling outage, and (2) when through-wall repairs are made by welding on surfaces that are wet or exposed to water, the weld overlay repair is only acceptable until the next refueling outage. The licensee stated that it will follow Code Case N-661-1 without any deviations). In addition, the licensee did not ask for deviation from the two conditions imposed in RG 1.147, Revision 16.
Although IWA-2441 (b) of the ASME Code,Section XI, requires code cases be applicable to the edition and addenda specified in the lSI program, the NRC staff does not believe the provisions in the 2007 Edition through the 2008 Addenda would in any way invalidate the design, installation, and examination requirements of Code Case N-661-1 or its applicability. Similarly, the NRC staff determines that use of Code Case N-661-1 does not invalidate the provisions of the 2007 edition through 2008 addenda of the ASME Code,Section XI. That is, there is no conflict between the code case and the 2007 edition through the 2008 of the ASME Code,Section XI. In addition, the NRC staff finds no conflict in using Code Case N-661-1 under the 2007 Edition through the 2008 Addenda as opposed to the 2004 Edition through the 2005 Addenda of the ASME Code,Section XI.
Therefore, the NRC staff finds that the licensee's proposed alternative is acceptable and the licensee is permitted to use Code Case N-661-1. Pursuant to 10 CFR 50.55a(a)(3)(i), the NRC staff determines that extending the applicability of Code Case N-661-1 to the 2007 Edition thro~gh 2008 Addenda of the ASME Code,Section XI, provides an acceptable level of quality and safety.
- 4 The NRC staff notes that the ASME Code committees have approved Code Case N-661-2. The NRC has not officially approved Code Case N-661-2 but once the NRC approves Code Case N-661-2 in RG 1.147, via rulemaking of 10 CFR 50.55a, Code Case N-661-2 will supersede N-661-1. The NRC expects that QCNPS, Units 1 and 2, will use Code Case N-661-2 once it is approved.
Therefore, the NRC approves the use of Relief Request 15R-10 for the fifth 10-year lSI interval or until the NRC approves Code Case N-661-2, whichever occurs earlier.
4.0 CONCLUSION
As set forth above, the NRC staff determines that the proposed alternative provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(i),
and is in compliance with the ASME Code requirements. Therefore, the NRC staff authorizes the use of Relief Request 15R-10 at QCNPS, Units 1 and 2, for the fifth 10-year lSI interval which commenced on April 2, 2013, and ends on April 1, 2023, or until the NRC approves Code Case N-661-2 in RG 1.147, Revision 17, whichever occurs earlier. All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Prinicipal Contributors:
T. McLellan J. Poehler M. Reisi-Fard K. Hoffman A. Rezai J. Wallace J. Tsao Date: September 30, 2013
M. Pacilio
-2 If you have any questions on this action, please contact the NRC Project Manager, Brenda Mozafari, at (301) 415-2020.
Sincerely,
/ RA/
Travis L. Tate, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-254 and 50-265
Enclosures:
- Safety Evaluation - Relief Request 15R-01 - Safety Evaluation - Relief Request 15R-02 - Safety Evaluation - Relief Request 15R-03 - Safety Evaluation - Relief Request 15R-04 - Safety Evaluation - Relief Request 15R-06 - Safety Evaluation - Relief Request 15R-09 - Safety Evaluation - Relief Request 15R-10 cc w/encl: Distribution via Listserv DISTRIBUTION PUBLIC RidsOgcRp Resource LPL3-2 R/F RidsNrrDorlLpl3-2 Resource RidsNrrPMQuadCities Resource RidsNrrLASRohrer Resource RidsAcrsAcnw_MailCTR Resource RidsNrrDeEvib Resource RidsRgn3MailCenter Resource RidsNrrDraApla Resource RidsNrrDeEpnb Resource ADAMS ACCESSION NO' ML13267A097 "Memo Dated*
NRR-028
, OFFICE LPL3-2/BC LPL3-2/LA NRRlEVIB/BC NRR/NDE/BC NRRlAPLAlBC LPL3-2PM LPL3-2/PM HHamzehee BMozafari NAME BMozafari SRohrer SRosenberg TLupold TTate DATE 09/26/13 07/29/13 08/13/13 08/13/13 09/30/13 09/30/13 07/10/13 09/26/13 07/25/13 05/2113 02/7/13 I 01123/13
- 01/18/13 I 02/7113 I
I OFFICIAL RECORD COPY
M. Pacilio
- 2 If you have any questions on this action, please contact the NRC Project Manager, Brenda Mozafari, at (301) 415-2020.
Sincerely.
t RAt Travis L. Tate, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-254 and 50-265
Enclosures:
- Safety Evaluation - Relief Request 15R-01 - Safety Evaluation - Relief Request 15R-02 - Safety Evaluation - Relief Request 15R-03 - Safety Evaluation - Relief Request 15R-04 - Safety Evaluation - Relief Request 15R-06 - Safety Evaluation - Relief Request 15R-09 - Safety Evaluation - Relief Request 15R-10 cc wtencl: Distribution via Listserv DISTRIBUTION:
PUBLIC RidsOgcRp Resource LPL3-2 RlF RidsNrrDorlLpl3-2 Resource RidsNrrPMQuadCities Resource RidsNrrLASRohrer Resource RidsAcrsAcnw_MailCTR Resource RidsNrrDeEvib Resource RidsRgn3MailCenter Resource RidsNrrDraApla Resource RidsNrrDeEpnb Resource ADAMS ACCESSION NO' ML13267A097 "Memo Dated NRR-028 OFFICE DATE LPL3-2/PM lri 09/26/13 LPL3-2/LA SRohrer 09/26/13 NRRlEVIB/BC SRosenberg 07/29/13 07/10/13 NRRlNDE/BC TLupold 08/13/13 07/25/13 0512113 0217/13 01/23/13 01/18/13 02/7113 NRR/APLAlBC HHamzehee 08113113 LPL3-2/BC TTate 09/30/13 LPL3-2PM BMozafari 09/30/13 OFFICIAL RECORD COPY