ML17317A605

Issuance of Amendment, License Amendment Request to Adopt Technical Specifications Task Force (TSTF) Traveler TSTF-535

ML17317A605
Person / Time
Site:	Hope Creek
Issue date:	12/13/2017
From:	Lisa Regner Plant Licensing Branch 1
To:	Sena P Public Service Enterprise Group
REGNER, L, NRR/DORL/LPLI
References
CAC MF951, EPID L-2017-LLA-0183
Download: ML17317A605 (11)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 13, 2017 Mr. Peter P. Sena, Ill President and Chief Nuclear Officer PSEG Nuclear LLC - N09 P.O. Box236 Hancocks Bridge, NJ 08038

SUBJECT:

HOPE CREEK GENERATING STATION - ISSUANCE OF AMENDMENT TO ADOPT TECHNICAL SPECIFICATIONS TASK FORCE (TSTF) TRAVELER TSTF-535 (CAC NO. MF9501; EPID L-2017-LLA-0183)

Dear Mr. Sena:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 208 to Renewed Facility Operating License No. NPF-57 for the Hope Creek Generating Station. This amendment consists of changes to the technical specifications in response to your application dated March 27, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17086A071 ). The amendment adopts Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-535, "Revise Shutdown Margin Definition to Address Advanced Fuel Designs," Revision 0, dated August 8, 2011 (ADAMS Accession No. ML112200436).

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

Lisa M. Regner, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-354

Enclosures:

1. Amendment No. 208 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 RENEWED FACILITY OPERATING LICENSE Amendment No. 208 Renewed License No. NPF-57

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

A. The application for amendment filed by PSEG Nuclear LLC, on March 27, 2017, 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 Vvithout endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance Vvith the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment Vvill not be inimical to the common deimse 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 paiagraph 2.C.(2) of 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. 208, 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 within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION GCM,,/U~

Jaml~ G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed License and Technical Specifications Dateoflssuance: December 13, 2017

ATTACHMENT TO LICENSE AMENDMENT NO. 208 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 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 1-7 1-7

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 GE 14i 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. 208, 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. 208

DEFINITIONS SECONDARY CONTAINMENT INTEGRITY 1.39 SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All secondary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, or deactivated automatic valve or damper, as applicable secured in its closed position, except as provided in Table 3.6.5.2-1 of Specification 3.6.5.2.

b. All secondary containment hatches and blowout panels are closed and sealed.

c. The filtration, recirculation and ventilation system is in compliance with the requirements of Specification 3.6.5.3.

d. For double door arrangements, at least one door in each access to the secondary containment is closed.

e. For single door arrangements, the door in each access to the secondary containment is closed, except for normal entry and exit.

f. The sealing mechanism associated with each secondary containment penetration, e.g., welds, bellows or 0-rings, is OPERABLE.

g. The pressure within the secondary containment is less than or equal to the value required by Specification 4.6.5.1.a.

SHUTDOWN MARGIN (SOM) 1.40 SOM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:

a. The reactor is xenon free;

b. The moderator temperature is~ 68°F, corresponding to the most reactive state; and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SOM.

SITE BOUNDARY 1.41 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled, by the licensee.

HOPE CREEK 1-7 Amendment No. 208

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 208 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 March 27, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17086A071 ), PSEG Nuclear LLC (the licensee) requested changes to the Hope Creek Generating Station (Hope Creek) Technical Specifications (TSs).

Specifically, the licensee requested to adopt U.S. Nuclear Regulatory Commission (NRC or the Commission)-approved Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-535, Revision 0, "Revise Shutdown Margin Definition to Address Advanced Fuel Designs" (ADAMS Accession No. ML112200436), dated August 8, 2011.

The proposed change would revise the TS definition of shutdown margin (SDM) to require calculation of SDM at the reactor moderator temperature corresponding to the most reactive state throughout the operating cycle (68 degrees Fahrenheit (°F) or higher). The purpose is to address newer boiling-water reactor (BWR) fuel designs, which may be more reactive at shutdown temperatures above 68 °F.

The licensee stated that it did not propose any significant variations or deviations from the TS changes as described in TSTF-535 or the applicable parts of the NRC staff's model safety evaluation. The availability of this TS improvement was announced in the Federal Register on February 26, 2013 (78 FR 13100), as part of the consolidated line item improvement process.

2.0 REGULATORY EVALUATION

2.1 Background In water-moderated reactors, water is used to slow down, or moderate, high energy fast neutrons to low energy thermal neutrons through multiple scattering interactions. The low energy thermal neutrons are much more likely to cause fission when absorbed by the fuel.

However, not all of the thermal neutrons are absorbed by the fuel; a portion of them are instead absorbed by the water moderator. The amount of moderator and fuel that is present in the core heavily influences the fractions of thermal neutrons that are absorbed.

Enclosure 2

Water-moderated reactors are designed such that they tend to operate in what is known as an under-moderated condition. In this condition, the ratio of the moderator-to-fuel in the core is small enough that the overall effectiveness of water as a moderator decreases with increasing temperature. As temperature increases, fewer neutrons are absorbed in the moderator due to the decrease in its density (i.e., lost to the reaction). At the same time, this increase in reactivity is overshadowed by fewer neutrons being moderated (i.e., slowed enough to cause fissions),

which reduces overall reactivity. The result is a decrease in power and temperature, which produces a negative reactivity feedback effect so that the reactor becomes self-regulating.

However, if the amount of moderator becomes too large with respect to the amount offuel, the reactor can enter an over-moderated condition. In this condition, the o\erall effectiveness of water as a moderator increases with the temperature. The number of neutrons moderated (i.e., slowed down) enough to cause fission outweighs the loss of neutrons being absorbed in the moderator. This causes an increase in power that leads to a further increase in temperature, creating a potentially dangerous positive reactivity feedback cycle.

As practical examples in support of the proposed changes to the definition of SOM, TSTF-535 discussed SDM with regard to fuel assembly-GE14 (GE14) and Global Nuclear Fuel-2 (GNF2) fuels. TSTF-535 indicated that for historical fuel products leading up to and including GE 14, the maximum reactivity condition for SOM always occurred at a moderator temperature of 68 °F.

These fuel products were designed so that the core is always under-moderated when all control rods are inserted, except for the single most reactive rod. In cores with GNF2 fuel, TSTF-535 projected that the maximum reactivity condition at the beginning of cycle will remain at 68 °F, but later in the cycle, the most limiting SOM may occur at a higher temperature. Thus, the GNF2 fuel design could potentially cause an over-moderated condition in the core.

2.2 Technical Specification Changes The licensee's adoption of TSTF-535 for Hope Creek proposes to revise the TS definition of SOM to require calculation of SOM at the reactor moderator temperature corresponding to the most reactive state throughout the operating cycle (68 °F or higher).

The curr.ent definition of SOM in Section 1.0, "Definitions," of Hope Creek TSs is:

SHUTDOWN MARGIN 1.40 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68°F; and xenon free.

The licensee proposed to delete the current definition of SOM stated above and replace it with the following definition in accordance with TSTF-535:

SHUTDOWN MARGIN (SOM) 1.40 SOM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:

a. The reactor is xenon free;

b. The moderator temperature is~ 68 °F, corresponding to the most reactive state; and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SOM.

The licensee also proposed to add "(SOM)" next to the title, 'SHUTDOWN MARGIN," to show it is now an abbreviated term.

2.3 Regulatory Review Title 10 of the Code of Federal Regulations ( 10 CFR), Part 50 Appendix A, General Design Criteria for Nuclear Power Plants" (GDC) 26, "Reactivity control system redundancy and capability," and GDC 27, "Combined reactivity control systems capability," respectively state that reactivity within the core is to be controllable to ensure subcriticality is achievable and maintainable under cold conditions. Reactivity must also have appropriate margin for stuck rods and it must be controllable to assure that under postulated accident conditions, the capability to cool the core is maintained.

In addition, 10 CFR 50.36(c )(2)(ii) requires the establishment of a limiting condition for operation (LCO) for a process variable, design feature, or operating restriction that is an initial condition of a design-basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The TS definition of SOM and the LCOs placed on SOM serve, in part, to satisfy GDCs 26 and 27. They ensure there is always sufficient negative reactivity worth available to offset the positive reactivity worth of changes in moderator and fuel temperature, the decay of fission product poisons, the failure of a control rod to insert, and reactivity insertion accidents. Given this margin, the core can be held subcritical for conditions of normal operation, including anticipated operational occurrences.

The NRC's guidance for the format and content of the licensee's TSs can be found in NUREG-1433, Revision 4, "Standard Technical Specifications General Electric Plants (BWR/4)"

(STSs).

Revision 3 of NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition," Section 4.3, "Nuclear Design,"

Revision 3, dated March 2007 (ADAMS Accession No. ML070740003), provides the procedures concerning the review of control systems and SOM to help ensure compliance with GDCs 26 and 27.

3.0 TECHNICAL EVALUATION

3.1 Current Definition of Shutdown Margin In BWR plants, the control rods are used to hold the reactor core subcritical under cold conditions. The control rod negative reactivity worth must be sufficient to ensure the core is subcritical by a margin known as the SOM. It is the additional amount of negative reactivity worth needed to maintain the core subcritical by offsetting the positive reactivity worth that can occur during the operating cycle due to changes in moderator and fuel temperature, the decay of

fission product poisons, the failure of a control rod to insert, and reactivity insertion accidents.

Specifically, Section 1.0, "Definitions," of the STSs defines SOM as the amount of reactivity by which the reactor is subcritical or would be subcritical, assuming that the reactor is (1) xenon free, (2) the moderator is 68 °F, and (3) all control rods are fully inserted except for the rod of highest worth, which is assumed to be fully withdrawn.

The three criteria provided in the definition help exemplify what has traditionally been the most reactive design condition for a reactor core. Xenon is a neutron poison produced by fission product decay, and its presence in the core adds negative reactivity worth. Assuming the core is xenon free removes a positive reactivity offset and is representative of fresh fuel at the beginning of cycle.

The minimum temperature the reactor moderator is anticipated to experience is 68 °F, making it the point at which the moderator will be at its densest and, therefore, capable of providing the highest positive reactivity worth. By assuming the highest \f\.Orth rod is fully withdrawn, the core can be designed with adequate SOM to ensure it remains safely shut down, even in the event of a stuck control rod, as specified in GDCs 26 and 27.

Determination of the SOM under the aforementioned conditions yields a conservative result that, along with the requirements set forth in Hope Creek's TS 3/4.1.1, "Shutdown Margin,"

LCO 3.1.1, helps ensure:

(a) the reactor can be made subcritical from all operating conditions and transients and design-basis events, (b) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (c). the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

3.2 Proposed Definition of Shutdown Margin The specified moderator temperature of 68 °F facilitates the maximum reactivity condition only if the core exists in an under-moderated condition. In addition to burnable poisons, many modem fuel designs also incorporate partial length rods for increased neutron economy, which are employed in order to extend the operating cycle. Both of these affect the ratio of moderator to fuel. The strong local absorption effects of the burnable poisons in fresh fuel make the core under-moderated. As burnable poisons are depleted during the fuel cycle, the core becomes less under-moderated, potentially leading to a slightly over-moderated condition wherein the core will be more reactive at a moderator temperature higher than the 68 °F specified in the SOM definition. Thus, the maximum core reactivity condition and the most limiting SDM may occur later in the fuel cycle at a temperature greater than 68 °F. Consequently, calculation of the SOM at the currently defined moderator temperature of 68 °F may not accurately determine the available margin.

Consistent with TSTF-535, the licensee proposed a change to the definition of SDM to enable calculation of the SOM at a reactor moderator temperature of 68 °F or a higher temperature corresponding to the most reactive state throughout the operating cycle. SDM would be calculated using the appropriate limiting conditions for all fuel types at any time in core life.

In support of the proposed change, TSTF-535 cited the requirements for SOM as specified in Topical Report NED0-24011-A, Revision 18, "General Electric Standard Application for Reactor Fuel (GESTAR II, Main)," dated April 2011 (ADAMS Accession No. ML111120046).

Section 3.2.4.1, "Shutdown Reactivity," of GESTAR II states, in part:

The core must be capable of being made subcritical, with margin, in the most reactive condition throughout the operating cycle with the most reactive control rod fully withdrawn and all other rods fully inserted.

TSTF-535, Revision 0, also cites SRP Section 4.3, which states, in part, the following, concerning the review of control systems and SOM:

The adequacy of the control systems to assure that the reactor can be returned to and maintained in the cold shutdo'Ml condition at any time during operation. The applicant shall discuss shutdown margins (SOM). Shutdown margins need to be demonstrated by the applicant throughout the fuel cycle.

Although the licensing basis requirements for SOM in GESTAR II are only applicable for cores licensed with GNF methods, they are consistent with the review procedures set forth in the SRP, which are provided to help ensure compliance with GDCs 26 and 27.

While the SRP does not prescribe the temperature at which the minimum SOM should be determined, the requirement of shutting down the reactor and maintaining it in a shutdo'Ml condition "at any time during operation" suggests that considering a range of thermal and exposure conditions would be appropriate in the determination of the minimum SOM. Because newer fuel designs employ elements such as partial length rods and burnable absorbers that may cause the maximum core reactivity conditions and the most limiting SOM to occur later in the fuel cycle at a temperature greater than 68 °F, the NRC staff agrees with the TSTF-535 assessment in this regard. Additionally, the NRC staff finds that allowing calculation of the SOM at the most limiting core reactivity condition is prudent with respect to ensuring compliance with GDCs 26 and 27 and concludes that the proposed changes to the Hope Creek TSs are acceptable.

The drive for TSTF-535 was to provide for a more broadly applicable SOM definition in recognition of modern fuel designs for which the core may not be in its most reactive condition at 68 °F. The proposed language would require the licensee to consider all temperatures equal to or exceeding 68 °F and all times in the operating cycle. This change places an additional responsibility on any implementing licensee to identify the most limiting time-in-cycle and temperature, a change that is effectively more restrictive than the current definition. Therefore, the change can be considered acceptable for any facility that currently adheres to the current definition of SOM that is contained in the STSs. Based on these considerations, the NRC staff concludes that the change is applicable to any BWR, regardless of its fuel design. The NRC staff also concludes that the revised definition is consistent with the 10 CFR 50.36 requirements pertaining to LCOs, because it ensures that the LCOs for SOM consider a broadly conservative range of potential initial conditions in the anticipated operational occurrence analyses.

3.3 Technical Evaluation Summary The NRC staff has reviewed the licensee's implementation of the TSTF-535 proposed revisions to the definition of SOM. Based on the considerations discussed above, the NRC staff concludes that the proposed revisions are acceptable and will provide a conservative and

improved approach to the calculation of SOM that ensures use of the appropriate limiting conditions for all fuel types at any time in the life of the core. The NRC staff concludes the proposed revisions serve to satisfy the requirements set forth in GDCs 26 and 27, as discussed in SRP, Section 4.3. Additionally, the NRC staff concludes the proposed changes to the definition of SOM would require the licensee to calculate SOM in consideration of the most limiting conditions in the core. Therefore, the revised SOM definition is acceptable for BWR facilities using any current fuel design, and specifically, for Hope Creek.

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 on November 13, 2017. 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 im.olves 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 im.olves no significant hazards consideration, and there has been no public comment on such finding (May 9, 2017; 82 FR 21560). Accordingly, the amendment meets the eligibilitycriteria 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 deense and security or to the health and safety of the public.

Principal Contributor: Caroline Tilton Date: December 13, 2017

SUBJECT:

HOPE CREEK GENERATING STATION - ISSUANCE OF AMENDMENT TO ADOPT TECHNICAL SPECIFICATIONS TASK FORCE (TSTF) TRAVELER TSTF-535 (CAC NO. MF9501; EPID L-2017-LLA-0183) DATED DECEMBER 13, 2017 DISTRIBUTION:

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DATE 11/15/2017 12/07/2017 11/7/2017 \

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DATE 12/05/2017 12/07/2017 12/13/2011J OFFICIAL RECORD COPY