ML19092A140

Proposed Inservice Inspection Alternative PR5-02

ML19092A140
Person / Time
Site:	Cooper
Issue date:	04/29/2019
From:	Robert Pascarelli Plant Licensing Branch IV
To:	Dent J Nebraska Public Power District (NPPD)
Wengert T, NRR/DORL/LPLIV, 415-4037
References
EPID L-2018-LLR-0136
Download: ML19092A140 (9)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Ap:-il 29, 2019 Mr. John Dent, Jr.

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

SUBJECT:

COOPER NUCLEAR STATION - PROPOSED INSERVICE INSPECTION ALTERNATIVE PR5-02 (EPID L-2018-LLR-0136)

Dear Mr. Dent:

By letter dated November 5, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18313A092), as supplemented by letter dated November 8, 2018 (ADAMS Accession No. ML18319A095), Nebraska Public Power District (the licensee) requested approval from the U.S. Nuclear Regulatory Commission (NRC) for relief from certain requirements of the American Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel Code,Section XI, IWB-5222(b), at the Cooper Nuclear Station. The licensee requested authorization to perform alternative system leakage testing of various ASME Class 1 piping segments.

The licensee submitted proposed alternative PR5-02 pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 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.

On November 5, 2018, as documented in an NRC e-mail dated November 6, 2018 (ADAMS Accession No. ML18311A319), the NRC staff verbally authorized the use of Relief Request PR5-02. This letter documents the NRC staff's final review of request for alternative PR5-02. As set forth in the enclosed safety evaluation, the NRC staff has determined that the licensee has demonstrated that the proposed alternative 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. 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 proposed alternative in PR5-02 until the conclusion of the Cooper Nuclear Station Refueling Outage 30.

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 lnservice Inspector.

J. Dent 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, Robert J. Pascarelli, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-298

Enclosure:

Safety Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE NUCLEAR REACTOR REGULATION REQUEST FOR ALTERNATIVE PR5-02 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 November 5, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18313A092), as supplemented by letter dated November 8, 2018 (ADAMS Accession No. ML18319A095), Nebraska Public Power District (the licensee) proposed an alternative to the requirements of the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (ASME Code),Section XI, IWB-5222(b), at Cooper Nuclear Station.

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee submitted Relief Request PR5-02 to allow alternative system leakage testing of various ASME 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.

On November 5, 2018, as documented in a U.S. Nuclear Regulatory Commission (NRC) e-mail dated November 6, 2018 (ADAMS Accession No. ML18311A319), the NRC staff verbally authorized the use of Relief Request PR5-02 until the conclusion of the Cooper Nuclear Station Refueling Outage 30.

2.0 REGULATORY EVALUATION

Adherence to Section XI of the ASME Code is mandated by 10 CFR 50.55a(g)( 4), which states, in part, that ASME Code Class 1, 2, and 3 components will meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI.

Paragraph 50.55a(z) of 10 CFR states that alternatives to the requirements of paragraphs (b) through (h) of 10 CFR 50.55a, or portions thereof, 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 that: (1) the proposed alternative would provide an acceptable level of quality and safety, or (2) compliance with the Enclosure

specified requirements of this section 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 the use of an alternative, and the NRC to authorize the proposed alternative.

3.0 TECHNICAL EVALUATION

3.1 Licensee's Request for Alternative 3.1.1 ASME Code Components Affected All ASME Code,Section XI, Class 1, Examination Categories B-P, Item No. B 15.10 components that are subject to pressurization during a system leakage test are affected by this request for alternative, as shown in the supplemental letter dated November 8, 2018, and include:

• Outboard Feedwater Check Valves (RF-CV-13CV, -15CV)

• Inboard Feedwater Check Valves (RF-CV-14CV)

• Outboard Main Steam Isolation Valves (MS-AOV-A086A, B, C, D)

• Inboard High Pressure Coolant Injection (HPCI) Steam Supply (HPCI-MOV-M015)

• Outboard HPCI Steam Supply (HPCI-MOV-M016)

• Inboard Reactor Core Isolation Cooling (RCIC) Steam Supply (RCIC-MOV-M015)

• Outboard RCIC Steam Supply (RCIC-MOV-M016) 3.1.2 Applicable Code Edition The lnservice Inspection (ISi) Program for the fifth 10-year inservice inspection interval is based on the ASME Code,Section XI, 2007 Edition with 2008 Addenda.

3.1.3 Applicable Code Requirement The pressure retaining boundary during the system leakage test shall correspond to the reactor coolant boundary, with all valves in the position required for normal reactor operation startup.

The visual examination shall, however, extend to and include the second closed valve at the boundary extremity.

3.1.4 Reason for Request The licensee requested relief from the ASME Code,Section XI requirements for performing a system leakage test stating that performing the pressure test with the boundaries stated in paragraph 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.

3.1.5 Licensee's Proposed Alternative and Basis for Use In the letter dated November 5, 2018, the licensee stated, in part:

In lieu of a system leakage test during reactor startup, as required by IWB-5222(a), a system pressure test is performed at the pressure associated with 100% [percent] rated reactor power.

a) The outboard Feedwater 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 Feedwater check valves are normally open for reactor startup. The inboard 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 Feedwater 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 Feedwater 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

[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 [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 is 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.

Performing a system pressure test at 100 percent reactor power would result in a hardship without a compensating increase in quality and safety. At 100%

[percent] power primary containment is inserted and radiation levels are high.

The proposed alternative provides reasonable assurance of operational readiness of the subject components.

In summary, three of the Feedwater 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% [percent] 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 provided the following justification for the hardship in the letter dated November 5, 2018:

Pursuant to 10 CFR 50.55a, "Codes and Standards," Paragraph (z)(2), relief is requested from the requirements of ASME Code Section XI requirements for performing a system leakage test using the boundaries stated in Paragraph IWB-5222(a) because performing the pressure test with this boundary would result in a hardship without a compensating increase in quality and safety due to excessive radiation exposure and personnel safety concerns (temperature levels in the drywall).

3.1. 7 Duration of the Proposed Alternative As stated in the letter dated November 8, 2019, the licensee requested that the proposed alternative be authorized until the end of Refueling Outage 30 and not for the entirety of the fifth inservice inspection interval.

3.2 NRC Staff Evaluation The ASME Code,Section XI, Table IWB-2500-1, Examination Categories B-P, Item Number 815.10, requires that a system leakage test be performed in accordance with the ASME Code,Section XI, IWB-5220. Specifically, IWB-5222(a) states, in part, that the pressure retaining boundary during the system leakage test shall correspond to the reactor coolant 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 IWB-5222(a), would result in a hardship due to the excessive radiation exposure, and an inserted atmosphere where elevated temperatures in the drywall 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 licensee's 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 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 the November 8, 2018, supplemental letter, 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 (RHR), reactor water cleanup (RWCU), high pressure coolant injection (HPCI), reactor core isolation cooling (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 (PMIS) for Safety Parameter Display System (SPDS) 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 F [degrees Fahrenheit (°F)].

• Drywell unidentified and identified leak rates are monitored in accordance with operations daily surveillance log every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

The NRC staff finds that the licensee's 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 licensee's 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, IWB-5222, and will demonstrate structural integrity and leaktightness of the affected piping systems. Finally, the NRC staff concludes that performing the system leakage test in accordance with 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 licensee has demonstrated that the proposed alternative provides reasonable assurance of structural integrity of the subject piping segments, and 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.

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 authorizes the use of Relief Request PR5-02 at Cooper Nuclear Station for Refueling Outage 30.

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 lnservice Inspector.

Principal Contributor: A. Young Da~: April 29, 2019

SUBJECT:

COOPER NUCLEAR STATION-PROPOSED INSERVICE INSPECTION ALTERNATIVE PR5-02 (EPID L-2018-LLR-0136) DATED APRIL 29, 2019 DISTRIBUTION:

PUBLIC PM File Copy RidsACRS_MailCTR Resource RidsNrrDorlLpl4 Resource RidsNrrPMCooper Resource RidsNrrDmlrMphb Resource RidsNrrLAPBlechman Resource RidsRgn4MailCenter Resource AYoung, NRR ADAMS Access1on No.: ML19092A140 *b>Y e-ma1*1 OFFICE D0RL/LPL4/PM D0RL/LPL4/LA DMLR/MVIB/BC* D0RL/LPL4/BC NAME TWengert PBlechman SRuffin RPascarelli (LRonewicz for)

DATE 04/26/2019 04/10/2019 01/31/2019 04/29/2019 OFFICIAL RECORD COPY