ML20044D072
| ML20044D072 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/11/2020 |
| From: | V Sreenivas Plant Licensing Branch 1 |
| To: | Bryan Hanson Exelon Generation Co |
| Sreenivas V | |
| References | |
| EPID L-2019-LLA-0157 | |
| Download: ML20044D072 (13) | |
Text
0 March 11, 2020 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
R. E. GINNA NUCLEAR POWER PLANT - ISSUANCE OF AMENDMENT NO. 137 TO REVISE TECHNICAL SPECIFICATION 3.7.1, MAIN STEAM SAFETY VALVES (MSSVs) (EPID L-2019-LLA-0157)
Dear Mr. Hanson:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 137 to Renewed Facility Operating License No. DPR-18 for the R. E. Ginna Nuclear Power Plant in response to your application dated July 23, 2019.
The amendment revises Technical Specification 3.7.1, Main Steam Safety Valves (MSSVs),
Surveillance Requirement 3.7.1.1 to increase the allowable as-found main steam safety valves lift setpoint tolerance from +1 percent, -3 percent to +1.4 percent, -4 percent for valve numbers 3508, 3509, 3510, 3511, 3512, and 3515.
A copy of the safety evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
V. Sreenivas, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244
Enclosures:
- 1. Amendment No. 137 to Renewed DPR-18
- 2. Safety Evaluation cc: Listserv
EXELON GENERATION COMPANY, LLC DOCKET NO. 50-244 R. E. GINNA NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 137 Renewed License No. DPR-18
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Exelon Generation Company, LLC (the licensee), dated July 23, 2019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-18 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 137, are hereby incorporated in the renewed license. Exelon Generation shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: March 11, 2020
ATTACHMENT TO LICENSE AMENDMENT NO. 137 R. E. GINNA NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 DOCKET NO. 50-244 Replace the following page of Renewed Facility Operating License No. DPR-18 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Insert 3
3 Replace the following page of the Appendix A, Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Insert 3.7.1-2 3.7.1-2 (b)
Exelon Generation pursuant to the Act and 10 CFR Part 70, to possess and use four (4) mixed oxide fuel assemblies in accordance with the RG&E's application dated December 14, 1979 (transmitted by letter dated December 20, 1979), as supplemented February 20, 1980, and March 5, 1980; (3)
Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required.
( 4)
Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)
Exelon Generation pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:
(1)
(2)
(3)
Maximum Power Level Exelon Generation is authorized to operate the facility at steady-state power levels up to a maximum of 1775 megawatts (thermal).
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 137, are hereby incorporated in the renewed license.
Exelon Generation shall operate the facility in accordance with the Technical Specifications.
Fire Protection Exelon Generation shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee's amendment request dated March 28, 2013, supplemented by letters dated December 17, 2013; January 29, 2014; February 28, 2014; September 5, 2014; September 24, 2014; December 4, 2014; March 18, 2015; June 11, 2015; August 7, 2015; and as approved in the safety evaluation report dated November 23, 2015.
Except where NRC approval for changes or deviations is required R. E. Ginna Nuclear Power Plant Amendment No. 137
MSSVs 3.7.1 SURVEILLANCE REQUIREMENTS SR3.7.1.1 SURVEILLANCE FREQUENCY
-NOTE -
Only required to be performed in MODES 1 and 2.
Verify each MSSV lift setpoint specified below in In accordance accordance with the INSERVICE TESTING PROGRAM.
with the Following testing, lift settings shall be within+/- 1 %.
INSERVICE VALVE NUMBER 3509 3511 3515 3513 3508 3510 3512 3514 LIFT SETTING 1140 (psig + 1.4%, -4%)
1140 (psig + 1.4%, -4%)
1140 (psig + 1.4%, -4%)
1085 (psig + 1%, -3%)
TESTING PROGRAM R.E. Ginna Nuclear Power Plant 3.7.1-2 Amendment 80, 124,137
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 137 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 EXELON GENERATION COMPANY, LLC R. E. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244
1.0 INTRODUCTION
By application dated July 23, 2019 (Agencywide Documents Access and Management System Accession No. ML19204A349) Exelon Generation Company, LLC (Exelon or the licensee) submitted a license amendment request (LAR) for the R. E. Ginna Nuclear Power Plant (Ginna). The amendment would revise Technical Specification (TS) 3.7.1, Main Steam Safety Valves (MSSVs), Surveillance Requirement (SR) 3.7.1.1 to increase the allowable as-found MSSV lift setpoint tolerance from +1 percent, -3 percent to +1.4 percent, -4 percent for valve numbers 3508, 3509, 3510, 3511, 3512, and 3515.
The proposed change would reduce unnecessarily-restrictive TS SR MSSV surveillance test criteria by accounting for setpoint drift.
2.0 REGULATORY EVALUATION
2.1 Regulatory Requirements and Guidance Technical Specification Requirements Under Title 10 of the Code of Federal Regulations (10 CFR) Section 50.92(a), determinations on whether to grant an applied-for license amendment are to be guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Both the common standards for licenses in 10 CFR 50.40(a), and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public.
The U.S. Nuclear Regulatory Commission (NRC or the Commission) regulatory requirements related to the content of the TSs are set forth in 10 CFR 50.36, Technical specifications.
The regulation in 10 CFR 50.36(a)(1) states that:
Each applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications.
Pursuant to 10 CFR 50.36(c), TSs are required to include items in the following categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) SRs; (4) design features; and (5) administrative controls. The regulation does not specify the requirements to be included in a plants TSs.
Per 10 CFR 50.36(b), TSs must be derived from the analyses and evaluation included in the safety analysis report and amendments thereto.
Per 10 CFR 50.36(c)(2), LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility.
As discussed in 10 CFR 50.36(c)(3), SRs are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that LCOs will be met.
Other Regulations and Guidance The regulation in 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, establishes, in part, standards for the calculation of emergency core cooling system (ECCS) accident performance and acceptance criteria for that calculated performance.
The regulation in 10 CFR 50.55a(f)(4), Inservice testing standards requirement for operating plants, states:
Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, pumps and valves that are within the scope of the ASME
[American Society of Mechanical Engineers] OM [Operation and Maintenance]
Code must meet the inservice test requirements (except design and access provisions) set forth in the ASME OM Code and addenda that become effective subsequent to editions and addenda specified in paragraphs (f)(2) and (3) of this section and that are incorporated by reference in paragraph (a)(1)(iv) of this section, to the extent practical within the limitations of design, geometry, and materials of construction of the components.
The applicable OM Code edition and addenda that the Ginna inservice testing (IST) program is based on is the 2012 Edition of the ASME OM Code.
NUREG-0800, Revision 2, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, dated March 2007, provides guidance to NRC staff in performing safety reviews of construction permit or operating license applications (including requests for amendments) under 10 CFR Part 50.
3.0 TECHNICAL EVALUATION
3.1 Background
Ginna has four MSSVs for each steam line. The first valve lifts at 1,085 pounds per square inch gauge (psig), and the remaining three valves are set to lift at 1,140 psig. The MSSVs basic design function is to limit the secondary system pressure to less than 110 percent of design pressure in accordance with the ASME OM. Reactor coolant system heat removal/
overpressure protection is an additional MSSV design function that provides a heat sink for removal of reactor coolant system energy if the preferred (but non-safety-related) condenser heat sink is not available. The MSSV design includes staggered setpoints so that only the needed valves will actuate. Staggered setpoints reduce the potential for valve chattering that is due to steam pressure insufficient to fully open all valves following a turbine/reactor trip.
The licensee is requesting a change to the as-found acceptance range of the MSSVs set to lift at 1,140 psig to minimize unnecessary maintenance and testing.
3.2 Staff Evaluation SR 3.7.1.1 verifies the operability of the MSSVs by the verification of each MSSV lift setpoint in accordance with the IST program. The ASME OM Code requires that safety and relief valve tests be performed in accordance with Appendix I of ASME OM Code-1998. The following is a list of the required tests:
- a. Visual examination,
- b. Seat-tightness determination,
- c. Setpoint pressure determination (lift setting),
- d. Compliance with owners seat-tightness criteria, and
- e. Verification of the balancing device integrity on balanced valves.
The ASME standard requires that all valves be tested every 5 years, and a minimum of 20 percent of the valves be tested every 24 months. The ASME OM Code specifies the activities and frequencies necessary to satisfy the requirements.
SR 3.7.1.1 currently requires a (+1 percent, -3 percent) tolerance for all as-found lift settings. In the LAR, the licensee requests an increased range to (+1.4 percent, -4 percent) for the six MSSVs that are set at 1,140 psig by using available margin in Ginnas design analyses. The tolerance for the two MSSVs that are set at 1,085 psig will remain unchanged. The valves are reset to +/-- 1 percent during the surveillance to allow for drift. This change does not alter the SR 3.7.1.1 as-left setting requirement of (+1 percent, -1 percent). As explained below, the staff found that the modified SR setpoints would continue to assure that the MSSVs would still function when needed, that the amended SR would provide reasonable assurance that the facility operation would be within the safety limits, and that the limiting conditions for operation would be maintained.
Table 1 of the LAR provides the Ginna Updated Final Safety Analysis Report (UFSAR)
Chapter 15 accident analyses and the value of the MSSV tolerance used in the analyses. The table shows that nearly all the analyses bound the high side ( >=) of the proposed as-found tolerance (+1.4 percent) or do not model the MSSV with one exception, the small break loss-of-coolant (SBLOCA) analysis. The NRC staff reviewed all UFSAR Chapter 15 events that model MSSVs and confirmed that the only analysis that does not bound the high side tolerance is the SBLOCA analysis that applies 1 percent tolerance for the MSSV setpoints.
The licensee described its review of the SBLOCA analyses in the LAR. As described in the LAR for Ginnas SBLOCA, the 1,140 psig setpoint valves are not predicted to lift for the limiting 2-inch break. For non-limiting SBLOCA events (3 inches and larger), slight event timing changes (on the order of seconds) may occur, but the changes would not result in these breaks becoming more limiting than the 2-inch break. The licensee concluded there would be no impact on the limiting peak-cladding temperature case (2-inch break). The NRC staff determined that the change to proposed as-found high-side tolerance (+1.4 percent) of the MSSVs would have a negligible impact on the overall peak cladding temperature response based on their experience with similar changes to Westinghouse 2-loop pressurized-water reactors. Additionally, there is significant margin to the 10 CFR 50.46 acceptance criteria for SBLOCA (> 1,000 degrees Fahrenheit °F), which provides additional assurance that the analysis would continue to meet that acceptance criteria with an MSSV opening at the as-found high-side tolerance. Therefore, the staff determined that this change to the MSSVs is acceptable for the SBLOCA analysis and will meet the 10 CFR 50.46 acceptance criteria. It is noted that the MSSVs are not modeled in the large break loss-of-coolant accident analysis in Westinghouse 2-loop pressurized-water reactors.
The licensee stated that by increasing the low-side tolerance to -4 percent, it is possible that the MSSVs could lift earlier than with the existing -3 percent margin. This results in a potential for removing heat earlier in events than previously. This change does not alter the SR 3.7.1.1 as-left setting requirement of (+1 percent, -1 percent). Ginna targets as-left settings that are biased low (but within the normal +/-1 percent) because the MSSVs have significantly more as-found margin on the low side. The as-found tolerance is expanded here to (-4 percent) to avoid testing failures from normal drift, given that Ginna attempts to leave the valves in the low end of the (+1 percent, -1 percent) as-left range.
The staff determined that removing heat earlier results in an improvement in protection against secondary system overpressure and an improvement in reactor coolant system temperatures.
Therefore, in terms of heat removal, the NRC staff finds this change acceptable.
The licensee stated that steam release for dose for steam generator tube rupture (SGTR) events is not affected by MSSVs potentially lifting at a lower setting, because the atmospheric relief valves are credited for SGTR events. Only the first set of MSSVs (1,085 psig valves) would open for SGTR events. Since the 1,085 psig valve tolerance is not changing, the NRC staff determined that the SGTR analysis will not be impacted.
The licensee discussed the potential effect of the change on cyclic thermal and pressure loads associated with design transients in the LAR and determined that there would be ~1 °F increase in steam temperature. The staff reviewed the design transient cyclic loading discussed in the Ginna UFSAR Sections 3.9.1 and 5.1.5. The staff determined that this steam temperature change is reasonable, given the small change in the MSSV tolerance range. Based on the NRC staffs experience with steam changes of this magnitude, the staff determined that there would be a negligible effect on the design transient analyses.
ASME OM Code Requirements The proposed change would revise the allowable as-found MSSV lift setpoint tolerance from
+1 percent/- 3 percent to +1.4 percent/-4 percent. The NRC staff reviewed the licensees proposed amendment as it relates to the ASME OM Code requirements imposed by 10 CFR 50.55a(f). The applicable OM Code edition and addenda that the Ginna IST program is based on is the 2012 Edition of the ASME OM Code.
The ASME OM Code, Mandatory Appendix I, requires that:
For each valve tested for which the as-found set-pressure (first test actuation) exceeds the greater of either the +/-tolerance limit of the Owner-established set-pressure acceptance criteria of subparagraph I-1310(e) or +/-3% of valve nameplate set-pressure, two additional valves shall be tested from the same valve group.
The LAR indicates that Ginna is establishing new owner-established set-pressure acceptance criteria, which would allow the proposed lower bound setpoint tolerance of
-4 percent to exceed the +/-3 percent acceptance criteria. The guiding principles for owner-defined acceptance criteria are defined in paragraph I-1310(e) of the ASME OM Code, which states that:
The Owner, based upon system and valve design basics or technical specification, shall establish and document acceptance criteria for tests required by this Appendix.
Based on a review of the contents of the LAR and the proposed TS changes, the licensee has established and documented the acceptance criteria for the tests required by Appendix I of the ASME Code. Therefore, the staff finds the proposed allowable as-found MSSV lift setpoint tolerance from +1 percent/-3 percent to +1.4 percent/-4 percent to be acceptable based on the inservice testing requirements in the ASME OM Code.
In summary, the proposed change would revise SR 3.7.1.1 to increase the allowable as-found MSSV lift setpoint tolerance from +1 percent/-3 percent to +1.4 percent/-4 percent for the six MSSVs that are set at 1,140 psig. This change would reduce an unnecessarily restrictive SR.
The NRC staff has reviewed the Ginna UFSAR and the evaluations and justifications provided by the licensee in its LAR and determined that the proposed change will not impact the reliability of the MSSVs or adversely impact their ability to perform their safety function. The licensee demonstrated that SR 3.7.1.1 will continue to provide reasonable assurance that the necessary quality of systems and components is maintained and meets the regulatory requirements of 10 CFR 50.36(c)(3) and 10 CFR 50.55a(f). Therefore, the NRC staff concludes that the proposed change to increase the MSSV lift setpoint tolerance from
+1 percent/-3 percent to +1.4 percent/- 4 percent is acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the New York State official was notified of the proposed issuance of the amendment on February 10, 2020. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes requirements with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes SRs.
The NRC staff has determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on September 10, 2019 (84 FR 47547).
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
F. Forsaty I. Tseng M. Hamm Date: March 11, 2020
- by memorandum
- by e-mail OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DSS/STSB/BC(A)* NRR/DSS/SNSB/BC(A)*
NAME VSreenivas LRonewicz SKrepel JBorromeo DATE 02/13/2020 02/13/2020 02/13/2020 02/11/2020 OFFICE NRR/DEX/EMIB/BC(A)** OGC - NLO**
NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME TScarbrough RAugustus JDanna VSreenivas DATE 02/11/2020 03/03/2020 03/11/2020 03/11/2020