ML20077L339

Proposed Inservice Inspection Alternatives RP5-02 and RI5-02

ML20077L339
Person / Time
Site:	Cooper
Issue date:	03/19/2020
From:	Jennifer Dixon-Herrity Plant Licensing Branch IV
To:	Dent J Nebraska Public Power District (NPPD)
Wengert T, NRR/DORL/LPLIV, 415-4037
References
EPID L-2019-LLR-0063, EPID L-2019-LLR-0064
Download: ML20077L339 (12)

Text

March 19, 2020 Mr. John Dent, Jr.

Vice President-Nuclear and Chief Nuclear Officer Nebraska Public Power District 72676 648A Avenue Brownville, NE 68321

SUBJECT:

COOPER NUCLEAR STATION - PROPOSED INSERVICE INSPECTION ALTERNATIVES RP5-02 AND RI5-02 (EPID L-2019-LLR-0063 AND EPID L-2019-LLR-0064)

Dear Mr. Dent:

By letter dated June 28, 2019, Nebraska Public Power District (the licensee) requested the U.S. Nuclear Regulatory Commission (NRC) approval of Relief Requests (RRs) RP5-02 and RI5-02, Revision 2 for the fifth 10-year inservice inspection (ISI) interval at Cooper Nuclear Station (CNS).

Relief Request RP5-02 proposes an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, subparagraph IWB-5222(b). The licensee requested authorization to perform alternative system leakage testing of various ASME Code Class 1 piping segments. The licensee submitted proposed alternative RP5-02 pursuant to Title 10 of the Code of Federal Regulations (10 CFR) paragraph 50.55a(z)(2) on the basis that complying with the specified ASME Code requirements would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety.

In addition, the licensee submitted a revision to its proposed alternative, RR RI5-02, Revision 2 for its fifth 10-year ISI interval for its reactor vessel internals components at CNS. Relief Request RI5-02, Revision 2 changes the specified revision of two of the Boiling Water Reactor

[BWR] Vessel and Internals Project (BWRVIP) topical reports that are used as a basis for the ASME Code alternative authorized in the NRC staffs letter dated July 31, 2018. The applicable BWRVIP guidelines are BWRVIP-41, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines, and BWRVIP-94NP, BWR Vessel and Internals Project, Program Implementation Guide, which are two of the BWRVIP documents referenced in RR RI5-02, Revision 1.

As set forth in the enclosed safety evaluations, the NRC staff has determined that the licensee has demonstrated that the proposed alternative for RR RP5-02 provides reasonable assurance of structural integrity of the subject piping, and that complying with the ASME Code, Section XI, would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. The NRC staff has also determined that the licensee has demonstrated that the proposed alternative for RR RI5-02 provides reasonable assurance of structural integrity of the reactor pressure vessel interior surfaces, interior attachments, and core support structures with acceptable level of quality and safety. Accordingly, the NRC staff concludes that the

J. Dent, Jr. licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1) and (2). Therefore, the NRC staff authorizes the use of the proposed alternatives in RRs RP5-02 and RI5-02 for the duration of the fifth 10-year ISI interval.

All other requirements of the ASME Code, Section XI, for which relief was not specifically requested and approved by the NRC staff remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

If you have any questions, please contact the Cooper Nuclear Station Project Manager, Thomas J. Wengert, at 301-415-4037 or by e-mail to Thomas.Wengert@nrc.gov.

Sincerely,

/RA/

Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-298

Enclosures:

1. Safety Evaluation for RP5-02

2. Safety Evaluation for RI5-02 cc: Listserv

SAFETY EVALUATION BY THE OFFICE NUCLEAR REACTOR REGULATION REQUEST FOR ALTERNATIVE RP5-02 REGARDING SYSTEM LEAKAGE TESTING OF CLASS 1 PIPING NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NO. 50-298

1.0 INTRODUCTION

By letter dated June 28, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19190A092), Nebraska Public Power District (the licensee) proposed 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 of Nuclear Power Plant Components, subparagraph IWB-5222(b), at Cooper Nuclear Station.

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) paragraph 50.55a(z)(2),

Hardship without a compensating increase in quality and safety, the licensee submitted Relief Request RP5-02 to allow alternative system leakage testing of various ASME Code Class 1 piping segments on the basis that complying with the specified ASME Code requirements would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety.

2.0 REGULATORY EVALUATION

Pursuant to 10 CFR 50.55a(g)(4), Inservice inspection standards requirement for operating plants, the ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and preservice examination requirements, 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.

Pursuant to 10 CFR 50.55a(z), Alternatives to codes and standards requirements, alternatives to the requirements of paragraph (g) of 10 CFR 50.55a may be used when authorized by the Director, Office of Nuclear Reactor Regulation. A proposed alternative must be submitted and authorized prior to implementation. The licensee must demonstrate (1) the proposed alternative would provide an acceptable level of quality and safety; or (2) compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Enclosure 1

Based on the above, and subject to the following technical evaluation, the U.S. Nuclear Regulatory Commission (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 TECHNICAL EVALUATION

3.1 Licensees Request for Alternative 3.1.1 ASME Code Components Affected The licensee identified the affected piping segments as all components subject to pressurization during a system leakage test.

The components affected are ASME Code Class 1 piping. In accordance with Subarticle IWB-2500, Examination and Pressure Test Requirements, Table IWB-2500-1, Examination Categories, they are classified as Examination Category B-P, Item Number B15.10.

3.1.2 Applicable Code Edition and Addenda The code of record for the fifth 10-year ISI interval is the 2007 Edition with 2008 Addenda of the ASME Code, Section XI.

3.1.3 Duration of Relief Request The licensee submitted this relief request for the fifth 10-year inservice inspection (ISI) interval which began on April 1, 2016, and will end on February 28, 2026.

3.1.4 Applicable Code Requirement The ASME Code, Section XI, Subarticle IWB-2500, Table IWB-2500-1, Examination Category B-P requires that the system leakage test be conducted according to Sub-subarticle IWB-5220, System Leakage Test, and the associated VT-2 visual examinations according to Sub-subarticle IWA-5240 prior to plant startup following each refueling outage. In accordance with subparagraph IWB-5221(a), the system leakage test shall be conducted at a pressure not less than the pressure corresponding to 100 percent rated reactor power. In accordance with subparagraph IWB-5222(a), the pressure-retaining boundary during the system leakage test shall correspond to the reactor coolant pressure boundary, with all valves in the position required for normal reactor operation startup. The required VT-2 visual examination shall, however, extend to and include the second closed valve at the boundary extremity. In accordance with subparagraph IWB-5222(b), the pressure-retaining boundary during system leakage test conducted at or near the end of each inspection interval shall extend to all Class 1 pressure-retaining components within the system pressure boundary.

3.1.5 Proposed Alternative, Basis for Use, and Reason for Relief The licensee stated in its letter dated June 28, 2019, that in lieu of a system leakage test during reactor startup, as required by subparagraph IWB-5222(a), a system pressure test is performed at the pressure associated with 100 percent rated reactor power.

a) The outboard reactor feedwater (RF) check valves and the High-Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) injection check valves are the Class 1 boundary valves and are closed for this test.

The RF check valves are normally open for reactor startup. The inboard RF check valve (RF-CV-16CV) on one feedwater line is kept open by Reactor Water Cleanup (RWCU) flow. The RWCU system is kept in service during the pressure tests. Thus, the outboard RF check valve and the RCIC injection check valve on this line will be pressurized during this test. The portion of piping between the other two RF check valves, including the HPCI injection line, will not be pressurized.

b) The four outboard Main Steam Isolation Valves (MSIV) will be closed for the system pressure test and the ten-year system pressure test

[subparagraph IWB-5222(b)]. The inboard MSIVs are opened to pressurize the system to the outboard valves. Both Main Steam drain valves are normally open to facilitate for pressure control; however, the outboard Class 1 boundary valve may be closed to provide leakage isolation if needed. The outboard valves are the Class 1 boundary valves.

c) Both HPCI and both RCIC steam supply valves will be closed for the system pressure test following a refueling outage. These valves close automatically on low steam supply pressure. During the ten-year system pressure test

[subparagraph IWB-5222(b)], the system will be pressurized to the outboard valves. The outboard valves are the Class 1 boundary valves.

The position of the valves for the system leakage test as described above and as listed in Tables 1 and 2 [of the licensees letter dated June 28, 2019] are consistent with the intent of IWB-5222(a). Abnormal lineups and installation of jumpers are not required for the system leakage test. The valves described above are normally open during a reactor startup. In order to pressurize the reactor coolant pressure boundary for testing, these valves must be closed.

Except as described above, the Class 1 boundary is pressurized as required by the code. The VT-2 inspection includes the entire reactor coolant pressure boundary.

Since the portions of the piping between the valves described above are operated at or above reactor pressure during normal operation, any through-wall leakage would be detected by the drywell leakage collection system, or by operations personnel on normal rounds.

The licensee further stated in its letter dated June 28, 2019:

Performing a system pressure test at 100 % reactor power would result in a hardship without a compensating increase in quality and safety. At 100 % power primary containment is inerted and radiation levels are high. The proposed alternative provides reasonable assurance of operational readiness of the subject components.

In summary, three of the RF check valves, HPCI injection check valve, the outboard MSIVs, and the HPCI and RCIC steam supply valves will be closed during the system leakage test, but will be included in the VT-2 visual examination. A VT-2 examination will be performed during the system leakage

test at a pressure not less than that associated with 100 % rated reactor power and will provide reasonable assurance of the continued operational readiness of mechanical connections, extending to the Class 1 boundary. In addition, once at or near the end of the inspection interval, the system leakage test shall extend to the Class 1 boundary as required by IWB-5222(b).

3.1.6 Hardship Justification The licensee stated that performing the pressure test with the boundaries stated in subparagraph IWB-5222(a) would impose an unnecessary hardship, without a compensating increase in quality and safety, due to excessive radiation exposure and personnel safety concerns due to temperature levels in the drywell.

3.1.7 NRC Staff Evaluation ASME Code, Section XI, Table IWB-2500-1, Examination Categories B-P, Item Number B15.10 requires that a system leakage test be performed in accordance with the ASME Code, Section XI, Sub-subarticle IWB-5220. Specifically, subparagraph IWB-5222(a) states, in part, that the pressure-retaining boundary during the system leakage test shall correspond to the reactor coolant pressure boundary, with all valves in the position required for normal reactor operation startup.

The NRC staff finds that performing the system leakage test during reactor startup, and with the orientation stated in ASME Code, Section XI, subparagraph IWB-5222(a), would result in a hardship due to the excessive radiation exposure and an inerted atmosphere where elevated temperatures in the drywell would present safety concerns to personnel performing the visual examination.

To determine whether this hardship is outweighed by a compensating increase in quality or safety, the NRC staff evaluated how the licensees proposed alternative testing boundary satisfies the intent of Section XI. The purpose of the system pressure tests is to detect through-wall leakage in the reactor coolant pressure boundary by visual examination. Instead of performing the system leakage test during reactor startup, a system pressure test will be performed at the pressure associated with 100 percent rated reactor power. To achieve and maintain this pressure without the reactor operating at 100 percent power requires multiple valves that are typically open to remain closed and maintain the pressure boundary. All portions of piping between the closed valves are operated at or above reactor pressure during normal operation, and any through-wall leakage would be detected by the drywell leakage collection system or by operations personnel on normal rounds.

Furthermore, to address the piping sections that operate at or above reactor pressure during normal operation but are not at test pressure in the proposed alternative, the licensee described the detection methods in its letter dated June 28, 2019, as follows:

• The temperature alarm subsystem of the leak detection system is comprised of temperature sensing elements installed in the vicinity of residual heat removal system, RWCU system, HPCI system, RCIC, and main steam lines (MS), and temperature switches that actuate annunciators in the Control Room. It is designed to detect leaks in the major steam piping system, especially in remote or enclosed areas such as the steam tunnel. If a steam or water leak occurs, the temperature element would sense a rise in ambient

temperature and cause an alarm in the Control Room. In addition, the continuous temperature signals are transmitted to the Plant Management Information System computer for the Safety Parameter Display System display.

• Control Room operators monitor Main Steam Tunnel temperatures twice per shift [every six hours] and record in [the] Operations log when temperature exceeds 160 degrees Fahrenheit.

• Drywell unidentified and identified leak rates are monitored in accordance with Operations daily surveillance log every eight (8) hours.

The NRC staff finds that the licensees defense-in-depth measures for both the pressurized and non-pressurized components that are covered under this relief request are suitable to provide reasonable assurance that any reactor coolant system leakage will be detected, despite the alternate testing conditions. Additionally, the NRC staff determines that the licensees proposal to perform a VT-2 visual examination during the system leakage test at a pressure not less than that associated with 100 percent rated power, and with systems in their normal lineup to the extent practical, will satisfy the intent of Section XI, paragraph IWB-5222 and will demonstrate structural integrity and leak tightness of the affected piping systems. Finally, the NRC staff finds that performing the system leakage test in accordance with subparagraph IWB-5222(a) would result in a hardship, without a compensating increase in quality and safety.

4.0 CONCLUSION

As set forth above, the NRC staff determines that the proposed alternative provides reasonable assurance of structural integrity and leak tightness of the subject piping segments, 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 NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2). Therefore, the NRC staff authorizes the use of the licensees proposed alternative at Cooper Nuclear Station, for the fifth 10-year ISI interval, which will end on February 28, 2026.

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 Inservice Inspector.

Principal Contributor: B. Fu, NRR Date: March 19, 2020

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PREQUEST FOR ALTERNATIVE RI5-02, REVISION 2 FOR FIFTH 10-YEAR INSERVICE INSPECTION INTERVAL NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NO. 50-298

1.0 INTRODUCTION

By letter dated August 17, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17241A048), Nebraska Public Power District (the licensee) submitted a proposed alternative, Relief Request (RR) RI5-02, Revision 1, for its reactor vessel internals (RVI) components at Cooper Nuclear Station (CNS). In RR RI5-02, Revision 1, 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 (ASME) Boiler and Pressure Vessel (BPV) Code (ASME Code) for inservice inspection (ISI) of reactor pressure vessel interior surfaces, interior attachments, and core support structures. This alternative was requested for the fifth 10-year ISI interval at CNS, which began on April 1, 2016, and will end on February 28, 2026. By letter dated July 31, 2018 (ADAMS Accession No. ML18183A325), the U.S. Nuclear Regulatory Commission (NRC) staff authorized the proposed alternative in RR RI5-02, Revision 1 pursuant to Title 10 of the Code of Federal Regulations (10 CFR) paragraph 50.55a(z)(1) on the basis that the alternative provides an acceptable level of quality and safety.

By letter dated June 28, 2019 (ADAMS Accession No. ML19190A092), the licensee submitted a revision to its proposed alternative, RR RI5-02, Revision 2 for its fifth 10-year ISI interval for its RVI components at CNS. Relief Request RI5-02, Revision 2 changes the specified revision of two of the BWRVIP topical reports that are used as a basis for the ASME Code alternative authorized in the NRC staffs letter dated July 31, 2018. The applicable BWRVIP guidelines are BWRVIP-41, BWR [Boiling Water Reactors] Jet Pump Assembly Inspection and Flaw Evaluation Guidelines, and BWRVIP-94NP, BWR Vessel and Internals Project, Program Implementation Guide, which are two of the BWRVIP documents referenced in RR RI5-02, Revision 1.

2.0 REGULATORY EVALUATION

The ISI of ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI of the ASME Code and applicable edition and addenda as required by Enclosure 2

10 CFR 50.55a(g), Preservice and inservice inspection requirements, except where specific relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(i), Impractical ISI requirements: Granting of relief. Pursuant to 10 CFR 50.55a(z), Alternatives to codes and standards requirements, alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC if (1) the proposed alternatives would provide an acceptable level of quality and safety or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), Inservice inspection standards requirement for operating plants, 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 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 requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(a)(1)(ii),

ASME Boiler and Pressure Vessel Code, Section XI, twelve months prior to the start of the 120-month interval, subject to the conditions listed in 10 CFR 50.55a(b)(2), Conditions on ASME BPV Code, Section XI.

The applicable ASME code of record for the fifth 10-year ISI interval for CNS is the ASME Code, Section XI, 2007 Edition through 2008 Addenda.

3.0 TECHNICAL EVALUATION

3.1 The Components for Which an Alternative is Requested The licensee requested to use alternative inspection criteria for the following components: ASME Code, Section XI, Class 1, Subarticle IWB-2500, Examination and Pressure Test Requirements, Table IWB-2500-1, Examination Categories, Examination Categories B-N-1 and B-N-2, Code Item Numbers B13.10 (Vessel Interior), B13.20 (Interior Attachments within Beltline Region),

B13.30 (Interior Attachments beyond Beltline Region), and B13.40 (Core Support Structure).

3.2 Examination Requirements for Which an Alternative is Requested The ASME Code, Section Xl requires the visual examination (VT) of certain RVI components.

These examinations are included in Table IWB-2500-1, Categories B-N-1 and B-N-2, and identified with the following item numbers:

B13.10 - Examine accessible areas of the reactor vessel interior each period using a technique, which meets the requirements for a VT-3 examination, as defined in paragraph IWA-2213, VT-3 Examination, of the ASME Code, Section XI.

B13.20 - 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, VT-1 Examination, of the ASME Code, Section XI.

B13.30 - 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.

B13.40 - 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, interior attachments, and core support structures.

3.3 Licensees Basis for Requesting an Alternative and Justification for Granting Relief In RR RI5-02, Revision 2, the licensee proposed to replace BWRVIP-41, Revision 3 with Revision 4-A dated December 2018 (ADAMS Accession No. ML19297G484). The licensee also proposed to replace BWRVIP-94NP, Revision 2 with Revision 3 dated September 2018 (ADAMS Accession No. ML11271A058).

3.4 NRC STAFF EVALUATION The NRC staff reviewed the information provided by the licensee in its submittal dated June 28, 2019, regarding its proposed alternatives to the ASME Code, Section XI ISI requirements and the technical bases for the licensees proposed alternatives.

3.4.1 BWRVIP-41, Revision 4-A As stated in the publicly available submittal letter dated October 22, 2019 (ADAMS Accession No. ML19297G503), Topical Report BWRVIP-41, Revision 4-A contains inspection and flaw evaluation guidelines for the BWR jet pump assemblies. In its July 2, 2018 safety evaluation (ADAMS Accession No. ML18130A024), the NRC staff found that the topical report, as modified and clarified to incorporate NRC staff conditions, is acceptable for use with respect to the proposed inspections and flaw evaluation guidelines for the BWR jet pump assemblies. Topical Report BWRVIP-41, Revision 4-A incorporates the NRC staff conditions referenced in the July 2, 2018, safety evaluation and is acceptable for use with respect to the proposed inspections and flaw evaluation guidelines for the BWR jet pump assemblies.

3.4.2 BWRVIP-94NP, Revision 3 Topical Report BWRVIP-94NP, Revision 3 is an administrative document intended to incorporate the NRC-endorsed Nuclear Energy Institute (NEI) guidance document NEI 03-08, Guidelines for the Management of Materials Issues (ADAMS Accession No. ML19079A253), for the BWRVIP Program, and to ensure consistent application of BWRVIP guidelines by BWRVIP utilities. The NRC staff has not evaluated this BWRVIP topical report for generic use; however, the staff finds that the licensees reference to Revision 3 is an acceptable way for the licensee to incorporate the guidance of NEI 03-08 for the subject proposed alternative. The NRC staff also confirmed that the licensees implementation of BWRVIP-94NP, Revision 3 will not impact any of the NRC staffs safety determinations concerning implementation of the applicable BWRVIP guidelines as an alternative to the subject ASME Code, Section XI requirements.

The NRC staff finds that the use of BWRVIP-41, Revision 4-A, dated December 2018 and BWRVIP-94NP, Revision 3, dated September 2018 to be acceptable because BWRVIP-41, Revision 4-A is an NRC-approved topical report and BWRVIP-94NP, Revision 3, will provide reasonable administrative controls for implementation of BWRVIP guidelines for the CNS RVI components.

4.0 CONCLUSION

As set forth above, the NRC staff determines that the licensees 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(z)(1).

Therefore, the NRC staff authorizes the alternative in RR RI5-02, Revision 2 at CNS for the fifth 10-year ISI interval, which began on April 1, 2016 and will end on February 28, 2026.

All other ASME BPV Code, Section XI, requirements for which an alternative was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

Principal Contributor: J. Jenkins, NRR Date: March 19, 2020

ML20077L339 *by e-mail OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DNRL/NPHP/BC* NRR/DNRL/NVIB/BC* NRR/DORL/LPL4/BC NAME TWengert (SLingam for) PBlechman MMitchell HGonzalez JDixon-Herrity DATE 03/19/2020 03/19/2020 01/31/2020 02/10/2020 03/19/2020