ML16270A038

Issuance of Amendment No. 200 Safety Limit Minimum Critical Power Ratio Change

ML16270A038
Person / Time
Site:	Hope Creek
Issue date:	10/13/2016
From:	Carleen Parker Plant Licensing Branch 1
To:	Sena P Public Service Enterprise Group
Parker C, NRR/DORL/LPLI-II, 415-1603
References
CAC MF7793
Download: ML16270A038 (12)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 13, 2016 Mr. Peter P. Sena, Ill President PSEG Nuclear LLC - N09 P.O. Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

HOPE CREEK GENERATING STATION - ISSUANCE OF AMENDMENT RE: SAFETY LIMIT MINIMUM CRITICAL POWER RATIO CHANGE (CAC NO. MF7793)

Dear Mr. Sena:

The Commission has issued the enclosed Amendment No. 200 to Renewed Facility Operating License No. NPF-57 for the Hope Creek Generating Station (HCGS). This amendment consists of changes to the technical specifications (TSs) in response to your application dated June 8, 2016.

The amendment incorporates a revised safety limit minimum critical power ratio for single recirculation loop operation. The change results from a cycle-specific analysis performed to support the operation of HCGS in the upcoming Cycle 21.

A copy of our safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sp; Carleen J. Par ~ Project Manager Plant Licensing ranch 1-2 Division of Ope ting Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-354

Enclosures:

1. Amendment No. 200 to Renewed License No. NPF-57

2. Safety Evaluation cc w/enclosures: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PSEG NUCLEAR LLC DOCKET NO. 50-354 HOPE CREEK GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 200 Renewed License No. NPF-57

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment filed by PSEG Nuclear LLC dated June 8, 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 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 Commission's regulations set forth in 10 CFR Chapter I; 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 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 Renewed Facility Operating License No. NPF-57 is hereby amended to read as follows:

Enclosure 1

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 200, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. The license amendment is effective as of its date of issuance and shall be implemented prior to startup from the fall 2016 refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION v~~ Douglas A. Broaddus, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. NPF-57 and the Technical Specifications Date of Issuance: CX:tober 13, a:l16 Enclosure 1

ATTACHMENT TO LICENSE AMENDMENT NO. 200 HOPE CREEK GENERATING STATION RENEWED FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 Replace the following page of the Renewed Facility Operating License with the revised page.

The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert Page 3 Page 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 Page 2-1 Page 2-1

reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4) PSEG Nuclear LLC, 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; (5) PSEG Nuclear LLC, 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 (6) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. Mechanical disassembly of the GE14i isotope test assemblies containing Cobalt-60 is not considered separation.

(7) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Part 30, to intentionally produce, possess, receive, transfer, and use Cobalt-60.

C. This renewed 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 PSEG Nuclear LLC is authorized to operate the facility at reactor core power levels not in excess of 3840 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 200, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Renewed License No. NPF-57 Amendment No. 200

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 24% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 24% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

THERMAL POWER, High Pressure and High Flow 2.1.2 With reactor steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow:

The MINIMUM CRITICAL POWER RATIO (MCPR) shall be 2 1.08 for two recirculation loop operation and shall be 2 1.11 for single recirculation loop operation.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With reactor steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow and the MCPR below the values for the fuel stated in LCO 2.1.2, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

HOPE CREEK 2-1 Amendment No. 200

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 200 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-57 PSEG NUCLEAR LLC HOPE CREEK GENERATING STATION DOCKET NO. 50-354

1.0 INTRODUCTION

By letter dated June 8, 2016 (Agencywide Document Access and Management System (ADAMS) Accession Nos. ML16181A193 and ML16181A194), PSEG Nuclear LLC (PSEG, or the licensee) requested changes to the Hope Creek Generating Station (HCGS) technical specifications (TSs). The proposed changes would revise the safety limit minimum critical power ratio (SLMCPR) for single recirculation loop operation (SLO). The proposed change results from a cycle-specific analysis performed to support the operation of HCGS in upcoming Cycle 21.

2.0 REGULATORY EVALUATION

The U.S. Nuclear Regulatory Commission (NRC or the Commission) staff reviewed the proposed TS changes against the regulatory requirements and guidance listed below to ensure that there is reasonable assurance that the systems and components affected by the proposed TS changes will perform their safety functions.

2.1 Requlatorv Requirements The NRC staff identified the following regulatory requirements as applicable to the proposed amendment.

2.1.1 Applicable TS Regulations The Commission's regulatory requirements related to the content of the TSs are set forth in Title 1O of the Code of Federal Regulations (1 O CFR) Section 50.36, "Technical specifications."

This regulation requires that the TSs include items in the following five specific categories:

(1) safety limits (SLs), limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in plant TSs.

Enclosure 2

As discussed in 10 CFR 50.36(c)(1 ), SLs for nuclear reactors are limits upon important process variables that protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity. The fuel cladding is one of the physical barriers that separate radioactive materials from the environment. The SLMCPR is required to be in the TSs to ensure that fuel design limits are not exceeded. The SLMCPR limit is contained in HCGS TS 2.1.2, and the parameter on which it is based can vary from cycle to cycle.

2.1.2 General Design Criteria General Design Criterion (GDC) 10, "Reactor design," of Appendix A to 10 CFR Part 50 states that "[t]he reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences." The purpose of the SLMCPR is to ensure that specified acceptable fuel design limits (SAFDLs) are not exceeded during steady state operation and analyzed transients.

2.2 Regulatory Guidance The NRC staff identified the following regulatory guidance as being applicable to the proposed amendment.

• Guidance on the acceptability of the reactivity control systems, the reactor core, and fuel system design is provided in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR

[Light-Water Reactor] Edition (SRP)."

o SRP Section 4.2, "Fuel System Design," Revision 3, March 2007 (ADAMS Accession No. ML070740002), specifies all fuel damage criteria for evaluation of whether fuel designs meet the SAFDLs.

o SRP Section 4.4, "Thermal and Hydraulic Design," Revision 2, March 2007 (ADAMS Accession No. ML070550060) provides guidance on the review of thermal-hydraulic design in meeting the requirement of GDC 1O and the fuel design criteria established in SRP 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 boiling transition during normal operation or anticipated operational occurrences.

3.0 TECHNICAL EVALUATION

3.1 Background Fuel design limits can be exceeded if the core exceeds critical power. Critical power is a term used for the power at which the fuel departs from nucleate boiling and enters a transition to film boiling. For boiling-water reactors, the critical power is predicted using a correlation known as the General Electric (GE) critical quality boiling length correlation, better known as the GEXL

correlation. Due to core-wide and operational variations, the margin to boiling transition is most easily described in terms of a CPR, which is defined as the rod critical power, as calculated by GEXL, divided by the actual rod power. The greater a CPR value exceeds 1.0, the greater the margin is to boiling transition. The SLMCPR is calculated using a statistical process that takes into account operating parameters and uncertainties. The operating limit minimum critical power ratio (OLMCPR) is equal to the SLMCPR plus a CPR margin for transients. At the OLMCPR, at least 99.9 percent of the rods avoid boiling transition during steady-state operation and transients caused by a single operator error or equipment malfunction. The OLMCPR is required to be established and documented in the core operating limits report for each reload cycle.

3.2 Proposed Changes For Operating Cycle 21, the HCGS calculated SLMCPR will change from greater than or equal to (~) 1.10 to ~ 1.11 for SLO. The SLMCPR value for two recirculation loop operation (TLO) will remain the same (~ 1.08). Accordingly, PSEG proposes to revise HCGS TS 2.1.2 to read as follows (change in bold):

2.1.2 With reactor steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow:

The MINIMUM CRITICAL POWER RATIO (MCPR) shall be~ 1.08 for two recirculation loop operation and shall be ~ 1.11 for single recirculation loop operation.

3. 3 Cycle 21 Core The absolute value of SLMCPR tends to vary cycle-to-cycle, typically due to the introduction of improved fuel bundle types, changes in fuel vendors or applicable computer codes, and changes in core loading pattern. Fresh fuel bundles generally dominate the SLMCPR calculation.

Cycle 21 will be the first reload of Global Nuclear Fuel (GNF) GNF2 fuel at HCGS. This license amendment request supports the core design for the upcoming HCGS Cycle 21, which will start after the fall 2016 refueling outage.

3.4 Methodology GNF performed the HCGS Cycle 21 SLMCPR calculation using the following NRC-approved methodologies and uncertainties:

• NED0-24011-A "General Electric Standard Application for Reactor Fuel (GESTAR II, Main)," Revision 22, November 2015 (ADAMS Accession No. ML15324A148).

• NED0-32601-A, "Methodology and Uncertainties for Safety Limit MCPR Evaluations," Revision O, August 1999 (ADAMS Accession No. ML14093A216).

• NEDC-32694P-A, "Power Distribution Uncertainties for Safety Limit MCPR Evaluations," August 1999 ("Acceptance for Referencing of Licensing Topical

Reports NEDC-32601 P, Methodology and Uncertainties for Safety Limit MCPR Evaluations; NEDC-32694P, Power Distribution Uncertainties for Safety Limit MCPR Evaluation; and Amendment 25 to NEDE-24011-P-A on Cycle-Specific Safety Limit MCPR," dated March 11 , 1999 (ADAMS Accession No. ML993140059)).

• NED0-32505P-A, "A-Factor Calculation Method for GE 11, GE 12 and GE 13 Fuel," Revision 1, July 1999 (ADAMS Accession No. ML060520636).

These methodologies were used for HCGS Cycle 21 SLMCPR calculations. The NRC staff reviewed the proposed change to ensure that the generic methods were appropriately applied to HCGS. The HCGS Cycle 21 core will be the first full reload of GNF2 fuel assemblies, and no plant hardware or operational changes are required for this proposed change.

The A-Factor is an input into the GEXL correlation used to describe the local pin-by-pin power distribution and the fuel assembly and channel geometry on the fuel assembly critical power.

The A-Factor uncertainty analysis includes an allowance for power peaking modeling uncertainty, manufacturing uncertainty, and channel bow uncertainty. NEDC-32505P-A is the generic A-Factor methodology report that describes the methodology that was adopted after part length rods were introduced. The NRC staff's safety evaluation approving NEDC-32505P-A includes a requirement that the applicability of the A-Factor methodology is confirmed when a new fuel type is introduced. By letter dated March, 14, 2007, GNF submitted FLN-2007-011, "GNF2 Advantage Generic Compliance with NEDE-24011-P-A (GESTAR II), NEDC-33270P, March 2007, and GEXL 17 Correlation for GNF2 Fuel, NEDC-33292P, March 2007" (ADAMS Accession No. ML070780335). This letter confirmed that the A-factor methodology of NEDC-32505P-A is applicable to GNF2, and that all of the criteria defined in NED0-24011-A have been met for the GNF2 fuel design. As part of an NRC audit related to this report, the GNF2 fuel design was verified to have been evaluated in accordance with the NRC approved methodologies listed above. This was documented in an audit report dated September 25, 2008 (ADAMS Accession No. ML081630579).

On the basis of the analysis performed by GNF using the NRG-approved methodologies described above, the licensee has proposed to amend the HCGS TS Section 2.1.2 to revise the SLMCPR for the Operating Cycle 21. This information regarding requested changes to the HCGS TS SLMCPR is based on the core full rated power, and at minimum core flow of 94.80 percent. For Cycle 21 , the minimum core flow SLMCPR calculation performed at 94.8 percent core flow and rated core power condition was limiting as compared to the rated core flow and rated core power condition.

The current required SLMCPR values in HCGS TS is 1.08 for TLO and 1.10 for SLO.

Calculations performed by GNF for HCGS Cycle 21 resulted in a minimum calculated value of SLMCPR to stay at 1.08 for TLO, and 1.11 for SLO. GNF's calculation of the revised plant-specific SLMCPR numeric values for HCGS Cycle 21 was performed as part of the reload licensing analysis for HCGS Cycle 21, and is based upon NRG-approved methods, therefore it is acceptable. No departures from NRG-approved methodologies, or deviations from NRC-approved calculational uncertainties, were identified in the HCGS, Cycle 21, SLMCPR calculations.

3.5 Technical Conclusion Based on the above findings, the NRC staff concludes that the proposed changes will continue to meet the applicable regulatory requirements and guidance, and that the analysis performed to calculate the HCGS Cycle 21 SLMCPR numeric values are based upon NRG-approved methodologies. The NRC staff concludes that the new SLMCPR value for SLO will continue to provide assurance that 99.9 percent of the fuel rods in the core will not exceed the CPR, and that fuel cladding integrity will be maintained under conditions of normal operation and with appropriate margin for anticipated operational occurrences.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New Jersey State Official was notified 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 installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. 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 50748 dated August 2, 2016). 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 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: F. Forsaty Date: CX:tober 13, 2016

ML16270A038 *b>Y memo d ate d OFFICE DORL/LPL 1-2/PM DORL/LPL 1-2/LA DSS/SRXB/BC* DSS/STSB/BC NAME CParker LRonewicz and EOesterle AKlein PBlechman DATE 9/27/2016 9/29/2016 9/20/2016 9/29/2016 OFFICE OGG DORL/LPL 1-2/BC DORL/LPL 1-2/PM NAME VHoang DBroaddus CParker DATE 10/11/2016 10/13/2016 10/13/2016