NLS2019034, 10 CFR 50.55a Relief Request RP5-02 and RI5-02, Revision 2
| ML19190A092 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 06/28/2019 |
| From: | Dent J Nebraska Public Power District (NPPD) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NLS2019034 | |
| Download: ML19190A092 (9) | |
Text
H Nebraska Public Power District Always there when you need us IO CFR 50.55a NLS2019034 June 28, 2019 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001
Subject:
10 CFR 50.55a Relief Request RP5-02 and RI5-02, Revision 2 Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46
Dear Sir or Madam:
The purpose of this letter is for the Nebraska Public Power District (NPPD) to request that the Nuclear Regulatory Commission grant relief from, and authorize alternative to, inservice inspection code requirements for the Cooper Nuclear Station (CNS) pursuant to 10 CFR 50.55a.
The 10 CFR 50.55a requests pertain to inservice examination test requirements in Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code.
Relief Request RP5-02 and RI5-02, Revision 2, attached to this letter, will be used for the duration of the fifth ten-year inservice inspection interval. NPPD requests approval of these requests by June 30, 2020.
This letter contains no regulatory commitments.
Should you have any questions concerning this matter, please contact David Van Der Kamp, Acting Licensing Manager, at (402) 825-2904.
Sincerely, Qf)
- tJ;;f Vice President - Nuclear and Chief Nuclear Officer
/dv
Attachment:
- 1. 10 CFR 50.55a Relief Request RP5-02, Definition of Pressure Retaining Boundary for System Leakage Test
- 2. 10 CFR 50.55a Relief Request RI5-02, Revision 2
~OOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 www.nppd.com
NLS2019034 Page 2 of2 cc: Regional Administrator w/ attachments USNRC - Region IV Cooper Project Manager w/ attachments USNRC - NRR Plant Licensing Branch IV Senior Resident Inspector w/ attachments USNRC-CNS NPG Distribution w/o attachments CNS Records w/ attachments
NLS2019034 Page 1 of 5 10 CFR 50.55a Relief Request RPS-02 Definition of Pressure Retaining Boundary for System Leakage Test Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2)
Hardship without a Compensating Increase in Quality or Safety American Society of Mechanical Engineers (ASME) Code Component(s) Affected Code Class: 1 Examination Category: B-P Item Number: B15.10 Component Numbers: All Components Subject to Pressurization During a System Leakage Test
Applicable Code Edition and Addenda
ASME Code Section XI, 2007 Edition, 2008 Addenda
Applicable Code Requirement
Paragraph IWB-5222(a)
Article IWB-5000, "System Pressure Tests," Sub-subarticle IWB-5220, "System Leakage Test," Paragraph IWB-5222, "Boundaries," states that:
a) 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.
b) The Class 1 pressure retaining boundary which is not pressurized when the system valves are in the position required for normal reactor startup shall be pressurized and examined at or near the end of the inspection interval. This boundary may be tested in its entirety or in portions and testing may be performed during the testing of the boundary of IWB-5222(a).
Table IWB-2500-1, Examination Category B-P, Note 2 states that:
The system leakage test (IWB-5220) shall be conducted prior to plant startup following a reactor refueling outage.
Reason for Request
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
NLS2019034 Page 2 of 5 10 CFR 50.55a Relief Request RP5-02 Definition of Pressure Retaining Boundary for System Leakage Test 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 drywell).
To obtain normal operating pressure with all valves in the position for normal reactor operation startup, the reactor must be in startup with the core critical. However, 10 CFR Part 50, Appendix G requires pressure tests and leak tests of the reactor vessel that are required by ASME Section XI, to be completed before the core is critical.
Proposed Alternative and Basis for Use In lieu of a system leakage test with all valves in the position required for normal reactor operation startup, as required by IWB-5222(a), a system pressure test is performed at the pressure associated with 100% rated reactor power with the following valve positions:
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 I 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 RClC 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 [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 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 RClC 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 positions of the valves for the system leakage test as described above and as listed in Tables 1 and 2 are consistent with the intent ofIWB-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 I boundary is pressurized as required by the code. The VT-2 inspection includes the entire reactor coolant pressure boundary.
NLS2019034 Attachment I Page 3 of 5 10 CFR 50.55a Relief Request RPS-02 Definition of Pressure Retaining Boundary for System Leakage Test For the portions of piping operated at or above reactor pressure during normal operation that are not at test pressure, defense-in-depth for detection of possible through-wall leakage is provided by the following:
- 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 system, 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 and record in 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.
Performing a system pressure test at I 00% reactor power would result in a hardship without a compensating increase in quality and safety. At I 00%, 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 I 00% rated reactor power and will provide reasonable assurance of the continued operational readiness of mechanical connections, extending to the Class I boundary. In addition, once at or near the end of the inspection interval, the system leakage test shall extend to the Class I boundary as required by IWB-5222(b).
Based on the above, Nebraska Public Power District requests relief from the ASME Section XI requirements for performing a system leakage test using the boundaries stated in IWB-5222(a).
Duration of Proposed Alternative This proposed alternative will be applied for the duration of the fifth ten-year inservice inspection interval.
NLS2019034 Page 4 of 5 10 CFR 50.55a Relief Request RP5-02 Definition of Pressure Retaining Boundary for System Leakage Test Precedents PR-02 was previously approved by the Nuclear Regulatory Commission (NRC) for the fourth ten-year interval for Cooper Nuclear Station (CNS) on October 2, 2006. (ML062260195)
PR5-02 was emergently approved by the NRC, for the CNS refueling outage RE-30 only, on November 5, 2018 (MLl 8311A319). Written approval was received April 29, 2019.
References
- 1. Letter to U.S. Nuclear Regulatory Commission from Randall K. Edington (Nebraska Public Power District) dated February 23, 2006, "10 CPR 50.55a Requests for the Fourth Ten-Year Inservice Inspection Interval." (ML060590300)
- 2. Letter to U.S. Nuclear Regulatory Commission from Randall K. Edington (Nebraska Public Power District) dated June 15, 2006, "Revision of Relief Request PR-02." (ML061710101)
- 3. U.S. Nuclear Regulatory Commission letter to Nebraska Public Power District dated October 2, 2006, "Cooper Nuclear Station RE: Fourth 10-Year Interval Inservice Inspection Request for Relief No. PR-02." (ML062260195)
- 4. Letter to U.S. Nuclear Regulatory Commission from John Dent, Jr. (Nebraska Public Power District) dated November 5, 2018, "10 CPR 50.55a Relief Request PR5-02."
- 5. U.S. Nuclear Regulatory Commission email to Nebraska Public Power District dated November 6, 2018, "Cooper Nuclear Station - Verbal Authorization of Relief Request PR5-02." (ML18311A319)
- 6. Letter to U.S. Nuclear Regulatory Commission from John Dent, Jr. (Nebraska Public Power District) dated November 8, 2018, "10 CPR 50.55a Relief Request PR5-02 Supplement."
(MLl 8319A095)
NLS2019034 Page 5 of 5 10 CFR 50.55a Relief Request RPS-02 Definition of Pressure Retaining Boundary for System Leakage Test Table 1: Valves not in position required for normal reactor startup:
Position Required Position During Valve Description for Normal Reactor System Leakage Test Startup Outboard F eedwater RF-CV-13CV Open Closed Check Valve Inboard Feedwater RF-CV-14CV Open Closed Check Valve Outboard F eedwater RF-CV-15CV Open Closed Check Valve Outboard Main Steam MS-AOV-A086A Open Closed Isolation Valve Outboard Main Steam MS-AOV-A086B Open Closed Isolation Valve Outboard Main Steam MS-AOV-A086C Open Closed Isolation Valve Outboard Main Steam MS-AOV-A086D Open Closed Isolation Valve Inboard HPCI Steam HPCI-MOV-M015 Open Closed Supply Outboard HPCI Steam HPCI-MOV-M016 Open Closed Supply Inboard RCIC Steam RCIC-MOV-M015 Open Closed Supply Outboard RCIC RCIC-MOV-M016 Open Closed Steam Supply Table 2: Other valves discussed in Relief Request:
Position Required Position During Valve Description for Normal Reactor System Leakage Test Startup Inboard Main Steam MS-MOV-M074 Open/Closed Open Drain Valve Outboard Main Steam MS-MOV-M077 Open/Closed Open/Closed Drain Valve HPCI Injection Check HPCI-CV-29CV Closed Closed Valve Inboard Feedwater RF-CV-16CV Open Open Check Valve
NLS2019034 Page 1 of2 10 CFR 50.55a Relief Request RI5-02, Revision 2 Revision to Relief Request RI5-02, Revision 1 Associated with Implementation of BWRVIP Documents in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection In Accordance with 10 CFR 50.55a(z)(l)
Reason for Request
Pursuant to 10 CFR 50.55a(z)(l), a revision is requested to the Nuclear Regulatory Commission (NRC) previously approved Cooper Nuclear Station (CNS) relief request RI5-02, Revision 1 (Reference 3) associated with the use of the Boiling Water Reactor Vessel and Internals Project (BWRVIP) guidelines in lieu of specific American Society of Mechanical Engineers (ASME)
Code requirements on Reactor Pressure Vessel internals and components inspection. Since the issuance of the NRC Safety Evaluation Report, revisions to BWRVIP documents have occurred.
BWRVIP-41, Revision 4-A (Reference 1) and BWRVIP-94NP, Revision 3 (Reference 2) have been issued. Nebraska Public Power District (NPPD) is requesting approval to use the latest NRC/BWRVIP approved documents in place of the document revisions cited in the Safety Evaluation Report on the basis that these approved BWRVIP documents provide an acceptable level of quality and safety.
The BWRVIP guidelines have recommended aggressive specific inspection by Boiling Water Reactor (BWR) operators to identify material condition issues with BWR components. A wealth of inspection data has been gathered during these inspections across the BWR industry. These guidelines focus on specific and susceptible components, specify appropriate inspection methods capable of identifying anticipated degradation mechanisms, and require re-examination at conservative intervals. In contrast, the code inspection requirements were prepared before the BWRVIP initiative and have not evolved with BWR inspection experience.
Proposed Revision NPPD submitted relief request RI5-02, Revision 1 proposing to use the BWRVIP guidelines as an alternative to the requirements of Section XI of the ASME Code for the inservice inspection of the Reactor Pressure Vessel interior surfaces, attachments, and core support structures. RI5-02, Revision 1 has been approved by the NRC with specific revisions of BWRVIP documents listed. The Safety Evaluation Report restricts the use of the relief request benefits to the BWRVIP document revisions specifically addressed within the relief request submittal. In the time since the staffs approval ofNPPD's proposed alternative, BWRVIP-41 has been revised by the BWRVIP and approved by the NRC. BWRVIP-94NP is an administrative document that has also been revised and approved by the BWRVIP Executive Committee. NPPD requests that the latest NRC approved revision ofBWRVIP-41 and the latest BWRVIP Executive Committee approved revision ofBWRVIP-94 be used as an alternative to the revisions currently listed in the referenced Safety Evaluation Report.
NLS2019034 Attachment 2 Page 2 of2 10 CFR 50.55a Relief Request RI5-02, Revision 2 References
- 1. BWRVIP-41, Revision 4-A: BWR Vessel and Internals Project BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines. EPRI, Palo Alto, California: 3002014254, dated December 2018.
- 2. BWRVIP-94NP, Revision 3, BWR Vessel and Internals Project Program Implementation Guide. EPRI, Palo Alto, California: 3002013101, dated September 2018.
- 3. U.S. Nuclear Regulatory Commission letter to Nebraska Public Power District dated July 31, 2018, "Cooper Nuclear Station - Requests for Relief Associated with the Fifth IO-Year Inservice Inspection Interval Program." (ML18183A325)
Precedents
- 1. Letter to U.S. Nuclear Regulatory Commission from James Barstow (Exelon Generation Company, LLC) dated February 19, 2019, "Revision to Relief Requests Associated with the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection." (ML19050A363)
- 2. U.S. Nuclear Regulatory Commission letter to Exelon Generation Company, LLC, dated April 30, 2019, "Clinton Power Station, Unit No. 1; Dresden Nuclear Power Station, Units 2 and 3; James A. Fitzpatrick Nuclear Power Plant; LaSalle County Station, Units 1 and 2; Limerick Generating Station, Units 1 and 2; Peach Bottom Atomic Power Station, Units 2 and 3; and Quad Cities Nuclear Power Station, Units 1 and 2 - Revision to Approved Alternatives to Use Boiling Water Reactor Vessel and Internals Project Guidelines."
(MLl 9098A034)
Table 1 Updated BWRVIP Revisions CNS Safety Evaluation Listed Requested Listed Requested ADAMS Accession No. BWRVIP-41 BWRVIP-41 BWRVIP-94NP BWRVIP-94NP Revision Revision Revision Revision
---+--------+-----------I ML18183A325 3 4-A 2 3