ML17139C372
| ML17139C372 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 06/19/2017 |
| From: | Kimberly Green Plant Licensing Branch III |
| To: | Hamilton D FirstEnergy Nuclear Operating Co |
| Green K, NRR/DORL/LPLIII, 415-1627 | |
| References | |
| CAC MF8760, L-16-179 | |
| Download: ML17139C372 (13) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 19, 2017 Mr. David B. Hamilton Site Vice President FirstEnergy Nuclear Operating Company Mail Stop A-PY-A290 P.O. Box 97, 10 Center Road Perry, OH 44081-0097
SUBJECT:
PERRY NUCLEAR POWER PLANT, UNIT NO. 1 - ISSUANCE OF AMENDMENT CONCERNING REDUCTION OF STEAM DOME PRESSURE VALUE SPECIFIED IN TECHNICAL SPECIFICATION 2.1.1 (CAC NO. MF8760)
(L-16-179)
Dear Mr. Hamilton:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 176 to Facility Operating License No. NPF-58 for Perry Nuclear Power Plant, Unit No. 1. The amendment consists of changes to the technical specifications (TSs) in response to your application dated November 1, 2016.
The amendment revises TS 2.1.1, "Reactor Core Safety Limits," to reduce the reactor steam dome pressure value specified in TS 2.1.1.1 and TS 2.1.1.2 from 785 pounds per square inch gauge (psig) to 686 psig.
A copy of our safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Docket No. 50-440
Enclosures:
- 1. Amendment No. 176 to NPF-58
- 2. Safety Evaluation cc w/encls: Distribution via Listserv Sincerely, Kimberly J. Green, Senior Project Manager Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FIRSTENERGY NUCLEAR OPERATING COMPANY FIRSTENERGY NUCLEAR GENERATION, LLC DOCKET NO. 50-440 PERRY NUCLEAR POWER PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 176 License No. NPF-58
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment filed by FirstEnergy Nuclear Operating Company (the licensee, FENOC), dated November 1, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 1 O 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 Commission's 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 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-58 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 176, are hereby incorporated into the license. FENOC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of its date of its issuance and shall be implemented within 60 days of the date of issuance.
Attachment:
Changes to the Facility Operating License No. NPF-58 and Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION 6J9.'l/~
David J. Wrona, Chief Plant Licensing Branch 111 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance:
June 1 9, 2 O 1 7
ATTACHMENT TO LICENSE AMENDMENT NO. 176 PERRY NUCLEAR POWER PLANT. UNIT NO. 1 FACILITY OPERATING LICENSE NO. NPF-58 DOCKET NO. 50-440 Replace the following pages of the Facility Operating License and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE INSERT License NPF-58 License NPF-58 2.0-1 2.0-1 C.
This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level FENOC is authorized to operate the facility at reactor core power levels not in excess of 3758 megawatts thermal (100% power) in accordance with the conditions specified herein.
(2)
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 176, are hereby incorporated into the license. FENOC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Antitrust Conditions
- a. FirstEnergy Nuclear Generation, LLC Amendment No. 176
2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs SLs 2.0 2.1.1.1 With the reactor steam dome pressure < 686 psig or core flow< 10%
rated core flow:
THERMAL POWER shall bes; 23.8% RTP.
2.1.1.2 With the reactor steam dome pressure ;::: 686 psig and core flow ;::: 10%
rate core flow:
The Minimum Critical Power Ratio (MCPR) shall be ;::: 1.10 for two recirculation loop operation or<:: 1.13 for single recirculation loop operation.
2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.
2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall bes; 1325 psig.
2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.
PERRY - UNIT 1 2.0-1 Amendment No. 176
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 176 TO FACILITY OPERATING LICENSE NO. NPF-58 FIRSTENERGY NUCLEAR OPERATING COMPANY FIRSTENERGY NUCLEAR GENERATION. LLC PERRY NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-440
1.0 INTRODUCTION
By application dated November 1, 2016, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16307A029), FirstEnergy Nuclear Operating Company (the licensee or FENOC) requested changes to the technical specifications (TSs) for the Perry Nuclear Power Plant, Unit 1 (PNPP or Perry). The proposed amendment would reduce the reactor steam dome pressure specified in TS 2.1.1, "Reactor Core SLs, from 785 pounds per square inch gauge (psig) to 686 psig.
The proposed change addresses a condition with the potential to momentarily violate the reactor core safety limits (SLs) during a pressure regulator failure maximum demand (open)
(PRFO) transient. This condition was identified by GE Energy-Nuclear in a March 29, 2005, notification1 to the U.S. Nuclear Regulatory Commission (NRC or Commission) under Title 1 O of the Code of Federal Regulations (1 O CFR) Part 21, "Reporting of Defects and Noncompliance" (GE Part 21 notification).
2.0 REGULATORY EVALUATION
2.1 Background
When the steam dome pressure is less than 785 psig or core flow is less than 10 percent of rated core flow, PNPP TS 2.1.1.1 currently requires that thermal power shall be less than or equal to 23.8 percent of rated thermal power (RTP). The GE Part 21 notification identified using newer computer analysis codes that a PRFO transient could result in a condition where the reactor steam dome pressure momentarily decreases below 785 psig while thermal power is above the plant-specific thermal power limit specified in the TS 2.1.1.1. This condition would violate the reactor core SL in TS 2.1.1.1.
1GE Energy-Nuclear, "10CFR21 Reportable Condition Notification: Potential to Exceed Low Pressure Technical Specification Safety Limit," dated March 29, 2005 (ADAMS Accession No. ML050950428).
Also identified as GE Part 21 report SCOS-03.
Initially the Boiling Water Reactor Owners' Group (BWROG) attempted to generically resolve the issue identified in the GE Part 21 notification. On July 18, 2006, the Technical Specifications Task Force (TSTF) and the BWROG submitted TSTF-495, Revision 0, "Bases Change to Address GE Part 21 SC05-03, dated July 18, 2006 (ADAMS Accession No. ML061990227), proposing a modification to the Standard Technical Specifications (STSs) bases for boiling-water reactors (BWRs). This change proposed to clarify that the SL did not apply to momentary depressurization transients by revising the "Applicable Safety Analysis,
portion of the STS bases for the reactor core SL (Section B 2.1.1 ). By letter dated August 27, 2007 (ADAMS Accession No. ML072340113), the NRG staff denied TSTF-495, Revision 0, because the proposed change to the STS bases would modify the corresponding TSs by providing an exception to the explicit SL. The NRG staff's safety evaluation (SE) enclosed with the letter stated in part:
The staff agrees with the applicant's position that the PRFO transient does not threaten fuel cladding integrity, since the margin to [the SL minimum critical power ratio (MCPR)] increases with decreasing reactor pressure. However, the staff is concerned that in some depressurization events which occur at or near full power, there may be enough bundle stored energy to cause some fuel damage. If a reactor scram does not occur automatically, the operator may have insufficient time to recognize the condition and to take the appropriate actions to bring the reactor to a safe configuration.
Consequently, the BWROG discontinued its effort to resolve the issue generically.
Subsequently, affected BWR licensees have proposed resolution of the GE Part 21 issue on a plant-specific basis by submittal of license amendment requests (LARs) that lower the reactor steam dome pressure SL value in the TSs. This approach takes advantage of the fact that some advanced fuel designs have a critical power correlation with a lower-bound pressure significantly below the reactor steam dome pressure currently specified in TS 2.1.1.
Currently, TS 2.1.1 for PNPP specifies a reactor steam dome pressure of 785 psig. The licensee proposes in the LAR to reduce the reactor steam dome pressure consistent with the lower-bound pressure of the critical power correlations for the fuel designs which currently comprise the PNPP core.
2.2 Proposed Change The licensee's proposed changes would reduce the reactor steam dome pressure specified in TS 2.1.1.1 and TS 2.1.1.2 from 785 psig to 686 psig for PNPP.
The licensee's application also provided revised TS Bases pages to be implemented with the associated TS changes. These pages were provided for information only. Changes to the TS Bases would be made in accordance with the TS Bases Control Program.
2.3 Applicable Regulatory Requirements The NRG staff considered the following regulatory requirements and guidance documents in its review of the proposed license amendment.
Section 50.36 of 1 O CFR, "Technical specifications, establishes the regulatory requirements related to the content of TSs. As stated in 10 CFR 50.36(c)(1 )(i)(A):
Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. If any safety limit is exceeded, the reactor must be shut down. The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. Operation must not be resumed until authorized by the Commission.
Appendix A, "General Design Criteria (GDC) for Nuclear Power Plants," to 1 O CFR Part 50 establishes the minimum requirements for the principal design criteria for water-cooled nuclear power plants. For PNPP, Section 3.1, "Conformance with NRC General Design Criteria," of the updated safety analysis report (USAR) evaluates the plant design basis against the GDC or draft GDC, as appropriate. The USAR evaluation concludes that PNPP fully satisfies and complies with the GDC. The licensee states in its application that the proposed amendment satisfies the requirements of GDC 10, "Reactor design," regarding acceptable fuel design limits.
GDC 10 was considered in the NRC staff's review of the proposed amendment. GDC 1 O requires the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences (AOOs).
NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," provides guidance on the acceptability of the reactivity control systems, the reactor core and fuel system design. Specifically, Section 4.2, "Fuel System Design, (ADAMS Accession No. ML070740002) specifies the criteria for evaluation of fuel damage and whether fuel designs meet the SAFDLs. Section 4.4, "Thermal and Hydraulic Design, (ADAMS Accession No. ML070550060) provides guidance on the review of thermal-hydraulic design in meeting the requirement of GDC 1 O and the fuel design criteria established in Section 4.2. It states that the critical power ratio (CPR) is to be established such that at least 99.9 percent of fuel rods in the core would not be expected to experience departure from nucleate boiling, or onset of transition boiling, during normal operation or AOOs.
3.0 TECHNICAL EVALUATION
Each fuel vendor has developed critical power correlations valid over specified pressure and flow ranges (mass flow rates). These critical power correlations have become increasingly fuel design dependent as advanced fuel designs evolved. The critical power correlations for some advanced fuel designs have received NRC approval, or were developed using NRG-approved methodologies, down to a lower pressure than those approved previously. The lower bound of the extended pressure ranges for these advanced fuel designs can be used to justify a lower reactor steam dome pressure than specified in the TSs for previous fuel designs. As such, a wider pressure range would be available for transients to demonstrate compliance with MCPR limits. The licensee proposes to reduce the reactor steam dome pressure specified in TS 2.1.1.1 and TS 2.1.1.2 from 785 psig to 686 psig at PNPP based on the lower-bound pressure for the critical power correlation for the fuel currently used in the reactor core for the facility.
In its application, the licensee stated that PNPP currently has a mixed core of GE14 and GNF2 fuel produced by Global Nuclear Fuel - Americas, LLC (GNF). The CPR calculations for GE14 and GNF2 fuel use the critical power correlations known as GEXL 14 and GEXL 17, respectively.
The GEXL 14 correlation is documented in GNF report NEDC-32851 P-A, "GEXL 14 Correlation for GE14 Fuel," Revision 5, dated April 2011 (ADAMS Package Accession No. ML111290540).
The GEXL 17 correlation is documented in GNF report NEDC-33292P, "GEXL 17 Correlation for GNF2 Fuel," Revision 3, dated June 2009.2 As discussed in these reports, the GEXL 14 and GEXL 17 correlations are used in the core design process to determine the expected thermal margin for the operating cycle. In the safety analysis process, the correlations are used to determine the change in CPR during postulated transients and to determine the MCPR SL.
The acceptability of the GEXL 14 and GEXL 17 correlations is associated with the NRG-approved GNF Licensing Topical Report (L TR) NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (referred to as GESTAR II). This LTR provides generic information relative to the fuel design and analyses of BWRs that use the GE and GNF fuel designs. This LTR consists of a description of the fuel licensing criteria and fuel thermal-mechanical, nuclear, and thermal-hydraulic analyses bases. In accordance with TS 5.6.5, "Core Operating Limits Report (COLA)," PNPP may use the analytical methods in versions of GESTAR II which have been previously reviewed and approved by the NRC to determine the core operating limits.
GESTAR II includes a methodology for development of critical power correlations, and also contains criteria for when NRC approval of new critical power correlations is needed. The GEXL 14 correlation report was approved by the NRC, and a copy of the associated NRC SE is included with NEDC-32851 P-A. The GEXL 17 correlation report did not require NRC approval.
The acceptability of the GEXL 17 correlation is based on the NRC staff's approval of Amendment 33 to GESTAR 11.3 As such, the GEXL 17 correlation for GNF2 fuel is approved for use per GESTAR II by reference. In a letter dated March 5, 2010 (ADAMS Package Accession No. ML100700464), GNF submitted proposed Amendment No. 33 to GEST AR II for NRC review and approval. The letter also provided GNF report NEDC-33270P, "GNF2 Advantage Generic Compliance with NEDE-24011-P-A (GEST AR II)," Revision 3, dated March 201 O.
NEDC-33270P documented the completion of the requirements for the new GNF2 fuel design per the criteria in GESTAR II, including an explanation of how the development of the GEXL 17 correlation complies with the GESTAR II methodology. Based on this, the NRC staff considers the use of GEXL 14 and GEXL 17 correlations for GE14 and GNF2 fuel, respectively, to be acceptable for use in CPR calculations at PNPP.
The GEXL 14 and GEXL 17 correlation reports discuss the pressure range over which the critical power correlations are valid for the GE14 fuel and GNF2 fuel, respectively. As discussed in Section 3.0 of the application, the lower bound pressure limit for the GEXL 14 and GEXL 17 correlations is 700 pound per square inch absolute (psia), which is equivalent to (700 psia -14.7 psia) 685.3 psig. The licensee's application proposed to reduce the reactor steam dome pressure specified in PNPP TS 2.1.1.1 and TS 2.1.1.2 to 686 psig. The proposed 686 psig value falls inside the pressure range over which the critical power correlations are valid for both the GE14 fuel and GNF2 fuel. Therefore, the NRC staff determined that the proposed 686 psig limit for TS 2.1.1.1 and TS 2.1.1.2 is acceptable for the fuel in the PNPP core.
2This proprietary report was submitted to the NRC by letter dated June 30, 2009 (ADAMS Accession No. ML091830614). A public version of the report was included with the submittal and is available at ADAMS Accession No. ML091830624.
3Amendment No. 33 was incorporated in Revision 17 to NEDE-24011-P-A by GNF letter dated September 22, 201 O (ADAMS Package Accession No. ML102660094). A copy of the NRC staff's approval and SE for Amendment No. 33 is included in NEDE-24011-P-A.
The proposed TS 2.1.1.1 requires thermal power to be less than or equal to 23.8 percent RTP when the reactor steam dome pressure is less than 686 psig or core flow is less than 1 O percent rated core flow. The proposed 2.1.1.2 specifies MCPR limits when the reactor steam dome pressure is greater than or equal to 686 psig and core flow is greater than or equal to 1 O percent rated core flow. Thus, the proposed change offers a greater pressure margin in TS 2.1.1.1 for the PRFO transient than what is currently available such that the reactor pressure remains above the proposed low pressure SL of 686 psig.
Technical Conclusion Based on the above, the NRC staff concludes that as long as the core pressure and flow are within the range of validity of the critical power correlation applicable to the current PNPP reactor core, the proposed changes to TS 2.1.1.1 and TS 2.1.1.2 provide reasonable assurance that 99.9 percent of the fuel rods in the core are not expected to experience onset of transition boiling during normal operation or AOOs. As such, the SLs will continue to ensure that SAFDLs are not exceeded during normal operation or AOOs, consistent with the requirements in GDC 10. Furthermore, the staff concludes that the proposed changes establish reactor core SLs, reasonably certain to protect the integrity of the fuel cladding barrier and guard against an uncontrolled release of radioactivity, consistent with the requirements in 1 O CFR 50.36(c)(1 ).
Therefore, the NRC staff concludes that the proposed amendments are acceptable.
The NRC staff notes that if PNPP transitions to a new fuel design in the future where the lower bound of the fuel's CPR correlation has not been approved for use down to the reactor steam dome pressure specified in the TS reactor core SLs, NRC approval would be required prior to transitioning to that fuel design. As long as the lower bound pressure associated with the correlation for the new fuel design is less than or equal to the TS 2.1.1.1 and TS 2.1.1.2 reactor steam dome pressure, then an LAR would not be required.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Ohio State official was notified on May 15, 2017, of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 1 O CFR Part 20 or changes the surveillance requirements. 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 (81 FR 92868, dated December 20, 2016). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 1 O 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 Commission's 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 Contributor: M. Razzaque, NRR Date of issuance: June 1 9, 2O1 7
- via Email OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA N RR/DSS/SRXB/BC NAME KGreen SRohrer EOesterle DATE 5/22/17 5/22/17 5/8/17 OFFICE OGC-NLO NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME RNorwood DWrona KGreen DATE 6/13/17 6/19/17 6/19/17