ML15287A371

Issuance of Amendments for the Adoption of Technical Specifications Task Force Standard Technical Specifications Change Traveler TSTF-535 (CNL-15-029)

ML15287A371
Person / Time
Site:	Browns Ferry
Issue date:	12/04/2015
From:	Farideh Saba Plant Licensing Branch II
To:	James Shea Tennessee Valley Authority
References
CAC MF5823, CAC MF5824, CAC MF5825
Download: ML15287A371 (26)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 4, 2015 Mr. Joseph W. Shea Vice President, Nuclear Licensing Tennessee Valley Authority 3R-C Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3 - ISSUANCE OF AMENDMENTS FOR ADOPTION OF TECHNICAL SPECIFICATION TASK FORCE TRAVELER-535 (CAC NOS. MF5823, MF5824, AND MF5825)

Dear Mr. Shea:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment Nos. 291, 316, and 274 to Renewed Facility Operating Licenses Nos. DPR-33, DPR-52, and DPR-68 for the Browns Ferry Nuclear Plant, Units 1, 2, and 3, respectively. These amendments are in response to your application dated March 9, 2015, as supplemented by letter dated July 10, 2015.

These amendments adopt NRG-approved Technical Specification Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-535, "Revise Shutdown Margin Definition to Address Advanced Fuel Designs" (Agencywide Documents Access and Management System Accession No. ML112200436), Revision 0, dated August 8, 2011, revising the Technical Specification definition of shutdown margin (SOM) to require calculation of SOM 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 the Browns Ferry Nuclear Plant, Units, 1, 2, and 3, boiling-water reactor fuel designs, which may be more reactive at shutdown temperatures above 68 °F.

A copy of the related Safety Evaluation (SE) is enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

J. Shea If you have any questions concerning this letter and the SE, contact me at 301-415-1447 or e-mail farideh.saba@nrc.gov.

Sincerely, Fd:::. Seni~ci~r Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-259, 50-260, and 50-296

Enclosures:

1. Amendment No. 291 to DPR-33

2. Amendment No. 316 to DPR-52

3. Amendment No. 27 4 to DPR-68

4. Safety Evaluation cc w/encls: Distribution via listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT. UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 291 Renewed License No. DPR-33

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

A. The application for amendment by the Tennessee Valley Authority (the licensee),

dated March 9, 2015, as supplemented by letter dated July 10, 2015, 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, as amended, 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 license 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 Enclosure 1

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. DPR-33 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 291, are hereby incorporated in this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance, and shall be implemented no later than 60 days from the date of its issuance.

FOR THE NUCLEAR REGULA TORY COMMISSION

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Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-33 and Technical Specifications Date of Issuance: December 4, 2015

ATIACHMENT TO LICENSE AMENDMENT NO. 291 RENEWED FACILITY OPERATING LICENSE NO. DPR-33 DOCKET NO. 50-259 Replace the following page of the Renewed Facility Operating License No. DPR-33 with the attached revised page. The changed area is identified by a marginal line.

REMOVE INSERT Page 3 Page 3 Replace the following page of Appendix A, Technical Specifications, with the attached revised pages. The revised page is identified by amendment number. The changed areas are identified by marginal lines.

REMOVE INSERT Page 1.1-7 Page 1.1-7

(3) 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) 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 equipment and instrument calibration or associated with radioactive apparatus or components; (5) 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 renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:

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; 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 The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 3458 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 291, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

For Surveillance Requirements (SRs) that are new in Amendment 234 to Facility Operating License DPR-33, the first performance is due at the end of the first surveillance interval that begins at implementation of the Amendment 234. For SRs that existed prior to Amendment 234, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the surveillance was last performed prior to implementation of Amendment 234.

BFN-UNIT 1 Renewed License No. DPR-33 Amendment No. 291

Definitions 1.1 1.1 Definitions (continued)

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

a. The reactor is xenon free;

b. The moderator temperature is C?: 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.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

(continued)

BFN-UNIT 1 1.1-7 Amendment No. ~. 291

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 316 Renewed License No. DPR-52

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

A. The application for amendment by the Tennessee Valley Authority (the licensee), dated March 9, 2015, as supplemented by letter dated July 10, 2015, 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, as amended, 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 license 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.

Enclosure 2

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 Renewed DPR-52 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 316, are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance, and shall be implemented no later than 60 days from the date of its issuance.

FOR THE NUCLEAR REGULA TORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-52 and Technical Specifications Date of Issuance: December 4, 2015

ATTACHMENT TO LICENSE AMENDMENT NO. 316 RENEWED FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260 Replace the following page of the Renewed Facility Operating License No. DPR-52 with the attached revised page. The changed area is identified by a marginal line.

REMOVE INSERT Page 3 Page 3 Replace the following page of Appendix A, Technical Specifications, with the attached revised page. The revised page is identified by amendment number. The changed areas are identified by marginal lines.

REMOVE INSERT Page 1.1-6 Page 1.1-6

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) 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 equipment and instrument calibration or associated with radioactive apparatus or components; (5) 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 renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:

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; 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 The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 3458 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 316, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

For Surveillance Requirements (SRs) that are new in Amendment 253 to Facility Operating License DPR-52, the first performance is due at the end of the first surveillance interval that begins at implementation of the Amendment 253. For SRs that existed prior to Amendment 253, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the surveillance was last performed prior to implementation of Amendment 253.

3) The licensee is authorized to relocate certain requirements included in Appendix A and the former Appendix B to licensee-controlled documents.

Implementation of this amendment shall include the relocation of these requirements to the appropriate documents, as described in the licensee's BFN-UNIT 2 Renewed License No. DPR-52 Amendment No. 316

Definitions 1.1 1.1 Definitions (continued)

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a. Described in Section 13.10, Refueling Test Program; of the FSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to (RTP) the reactor coolant of 3458 MWt.

SHUTDOWN MARGIN SOM shall be the amount of reactivity by which the (SOM) 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.

(continued)

BFN-UNIT 2 1.1-6 Amendment No. ~. 316

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT. UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 27 4 Renewed License No. DPR-68

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

A. The application for amendment by the Tennessee Valley Authority (the licensee), dated March 9, 2015, as supplemented by letter dated July 10, 2015, 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, as amended, 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 license 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.

Enclosure 3

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 Renewed DPR-68 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 274, are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance, and shall be implemented no later than 60 days from the date of its issuance.

FOR THE NUCLEAR REGULA TORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-68 and Technical Specifications Date of Issuance: December 4, 2015

ATTACHMENT TO LICENSE AMENDMENT NO. 274 RENEWED FACILITY OPERATING LICENSE NO. DPR-68 DOCKET NO. 50-260 Replace the following page of the Renewed Facility Operating License No. DPR-68 with the attached revised pages. The changed area is identified by a marginal line.

REMOVE INSERT Page 3 Page 3 Replace the following page of Appendix A, Technical Specifications, with the attached revised page. The revised page is identified by amendment number. The changed areas are identified by marginal lines.

REMOVE INSERT Page 1.1-6 Page 1.1-6

(3) 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) 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 equipment and instrument calibration or associated with radioactive apparatus or components; (5) 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 renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:

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; 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 The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 3458 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 274, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

For Surveillance Requirements (SRs) that are new in Amendment 212 to Facility Operating License DPR-68, the first performance is due at the end of the first surveillance interval that begins at implementation of the Amendment 212. For SRs that existed prior to Amendment 212, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the surveillance was last performed prior to implementation of Amendment 212.

BFN-UNIT 3 Renewed License No. DPR-68 Amendment No. 274

Definitions 1.1 1.1 Definitions (continued)

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a. Described in Section 13.10, Refueling Test Program; of the FSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

RA TED THERMAL POWER RTP shall be a total reactor core heat transfer rate to

{RTP) the reactor coolant of 3458 MWt.

SHUTDOWN MARGIN SOM shall be the amount of reactivity by which the (SOM) 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.

(continued)

BFN-UNIT 3 1.1-6 Amendment No. ~.274

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 291 TO FACILITY OPERATING LICENSE NO. DPR-33 AMENDMENT NO. 316 TO FACILITY OPERATING LICENSE NO. DPR-52 AND AMENDMENT NO. 274 TO FACILITY OPERATING LICENSE NO. DPR-68 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2. AND 3 DOCKET NOS. 50-259, 50-260, AND 50-296

1.0 INTRODUCTION

By application dated March 9, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15068A407) as supplemented by letter dated July 10, 2015 (ADAMS Accession No. ML15197A123), Tennessee Valley Authority (TVA or the licensee) requested changes to the Technical Specifications (TSs) for Browns Ferry Nuclear Plant, Units 1, 2, and 3 (BFN). The supplement dated July 10, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register (FR) on June 23, 2015 (80 FR 35985).

The proposed change would revise the TS definition of shutdown margin (SOM) to require calculation of SOM 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 fuel designs, which may be more reactive at shutdown temperatures above 68 °F.

The licensee stated that the license amendment request (LAR) is consistent with NRG-approved Technical Specification Task Force (TSTF) Improved Standard Technical Specifications Change Traveler-535, Revision 0. The availability of this TS improvement was announced in the FR 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.

Enclosure 4

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 in each.

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; fewer neutrons are absorbed in the moderator due to the decrease in its density, but this is overshadowed by the reduction in the number of neutrons that moderate from high fission energy to the lower energy level needed to cause fission. The result is a decrease in power and temperature: a negative reactivity feedback effect where the reactor becomes self-regulating. However, if the amount of moderator becomes too large with respect to the amount of fuel, the reactor can enter an over-moderated condition. In this condition, the overall effectiveness of water as a moderator increases with increasing temperature; the reduction in the number of neutrons absorbed in the moderator outweighs the loss in neutrons reaching lower energies. 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, Revision 0, discussed SOM with regards to GE14 and GNF2 fuels. TSTF-535, Revision 0, indicated that for historical fuel products through GE14, the maximum reactivity condition for SOM always occurred at a moderator temperature of 68 °F because 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, Revision 0, stated that it is expected that the maximum reactivity condition at beginning of cycle will remain at 68 °F, but that later in cycle the most limiting SOM may occur at a temperature greater than this, indicating that with this fuel design the core could potentially achieve an over-moderated condition.

2.2 Technical Specification Changes TVA's adoption of TSTF-535, Revision 0, for BFN 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 current definition of SOM in Section 1.1, "Definitions," of BFN TSs is:

SOM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that:

a. The reactor is xenon free;

b. The moderator temperature is 68°F; 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 proposes the following changes {shown in bold) to the definition of SOM in accordance with TSTF-535, Revision 0:

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.

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

In addition, 10 CFR 50.36(c)(2)(ii)(B) 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 GDC 26 and 27 by ensuring 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."

Since BFN is a pre-GDC plant, the NRC staff's June 17, 2013, request for additional information (ADAMS Accession No. ML15161A392) asked the licensee to provide plant-specific information equivalent to the GDC discussed above so it could determine if the requested TS change and model SE for TSTF-535 were applicable to BFN. In its July 10, 2015, response (ADAMS Accession No. ML15197A123), the licensee stated as follows:

The BFN Units 1, 2 and 3 equivalent of the referenced GDC are the Atomic Energy Commission (AEC) Proposed General Design Criteria of November 1965 (Units 1 and 2) and July 1967 (Unit 3), as discussed in Appendix A of the BFN Updated Final Safety Analysis Report (UFSAR).

AEC Draft Criteria 27, 28, and 29 address information similar to current 10 CFR 50 Appendix A GDCs 26 and 27. BFN UFSAR Appendix A contains an evaluation discussing the plant design bases conformance to the intent of the AEC Draft Criteria. UFSAR Table A.0-5 contains references to other sections of the UFSAR which demonstrates conformance to the intent of the criteria. These U FSAR references are contained in the parentheses after the discussion of each AEC Draft Criterion.

As related to the proposed change, GDC-26, "Reactivity control system redundancy and capability," requires that two independent reactivity control systems of different design principles be provided, and that one of the reactivity control systems shall be capable of holding the reactor core subcritical under cold conditions with appropriate margin for stuck rods. AEC Draft Criteria 27, 28, and 29 provide the equivalent requirements.

AEC Draft Criterion 27, "Redundancy of Reactivity Control," states, "At least two independent reactivity control systems, preferably of different principles, shall be provided" (UFSAR 1.5, 3.4, 3.8, 7.7). AEC Draft Criterion 28, "Reactivity Hot Shutdown Capability," states, "At least two of the reactivity control systems provided shall independently be capable of making and holding the core subcritical from any hot standby or hot operating condition, including those resulting from power changes, sufficiently fast to prevent exceeding acceptable fuel damage limits" (UFSAR 1.5, 3.4, 3.6, 3.8, 7.7, 14.0). AEC Draft Criterion 29, "Reactivity Shutdown Capability," states, "At least one of the reactivity control systems provided shall be capable of making the core subcritical under any conditions (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits. Shutdown margins greater than the maximum worth of the most effective control rod when fully withdrawn shall be provided" (UFSAR 1.5, 3.4, 3.6, 7.2, 14.0).

As related to the proposed change, GDC-27, "Combined reactivity control systems capability," requires that the reactivity control systems have a combined capability, in conjunction with poison addition by the emergency core cooling systems (ECCS), to reliably control reactivity changes under postulated accident conditions, with appropriate conditions for stuck rods and that reactivity within the core be controllable to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained. As quoted above, AEC Draft Criteria 28 and 29 provide the equivalent requirements.

After reviewing the licensee's response, the staff agrees that the plant specific requirements of AEC Draft Criteria 27, 28, and 29 are essentially identical to the 10 CFR 50, Appendix A, GDCs 26 and 27 related to the TVA LAR, since the requirements assure the protection of fuel design limits and the capability of maintaining the core subcritical under any conditions.

3.0 TECHNICAL EVALUATION

3.1 Current Definition of Shutdown Margin For BFN, 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 1.1, "Definitions," of BFN's TSs 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 the 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 worth rod is fully withdrawn, the core can be designed with adequate shutdown margin to ensure it remains safely shutdown even in the event of a stuck control rod, as required by the plant-specific requirements of AEC Draft Criteria 27, 28, and 29.

Determination of the SOM under the aforementioned conditions yields a conservative result that, along with the requirements set forth in Section 3.1.1, "Shutdown Margin," of BFN's TSs, 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 modern 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 SOM 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.

The licensee, therefore, proposed a change to the definition of SOM 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. SOM would be calculated using the appropriate limiting conditions for all fuel types at any time in core life.

In support of the proposed change, the licensee cited the NRG-approved TSTF-535, Revision 0, which in turn, cited Topical Report NED0-24011-A, Revision 18, "General Electric Standard Application for Reactor Fuel (GESTAR II)," dated April 2011 (ADAMS Package Accession No. ML111120046). Section 3.2.4.1 of GESTAR II states:

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 also cited standard review plan (SRP) Section 4.3, which states 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 shutdown 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 provisions for SOM in GESTAR II are only applicable for cores licensed with Global Nuclear Fuels methods, they are consistent with the review procedures set forth in the SRP, which are provided to help ensure compliance with GDC 26 and 27 and, therefore, with the equivalent plant-specific requirements of AEC Draft Criteria 27, 28, and 29.

TSTF-535, Revision 0, stated that 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 shutdown 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, which 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

found the assessment methodology described in TSTF-535, Revision 0, to be acceptable.

Additionally, the NRC staff found that allowing calculation of the SOM at the most limiting core reactivity condition is prudent with respect to ensuring compliance with the plant-specific requirements of AEC Draft Criteria 27, 28, and 29. Accordingly, the staff concludes that the proposed changes to the BFN TSs listed above, are acceptable.

The impetus for TSTF-535, Revision 0, 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 will 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 the licensee to identify the most limiting time-in-cycle and temperature, a change that is more conservative than the current definition and will ensure the licensee maintains adequate SOM as required by their current licensing basis. Therefore, the change is acceptable for BFN units. The NRC staff also finds 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 Summary The NRC staff has reviewed the licensee's implementation of TSTF-535, Revision 0, 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 finds the proposed revisions serve to satisfy the requirements set forth in the plant-specific requirements of AEC Draft Criteria 27, 28, and 29. Additionally, the NRC staff concludes the proposed changes to the definition of SOM will require the licensee to calculate SOM in consideration of the most limiting conditions in the core. Therefore, the revised SOM definition is acceptable for use with any current fuel design.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Alabama State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change 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 amendments involve 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 amendments involve no significant hazards consideration, which was published in the FR on June 23, 2015 (80 FR 35985), and there has been no public comment on such finding. Accordingly, the amendments meet 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 amendments.

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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: Caroline E. Tilton Date: December 4, 2015

J. Shea If you have any questions concerning this letter and the SE, contact me at 301-415-1447 or e-mail farideh.saba@nrc.gov.

Sincerely, IRA AHon for/

Farideh E. Saba, Senior Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-259, 50-260, and 50-296

1. Amendment No. 291 to DPR-33

2. Amendment No. 316 to DPR-52

3. Amendment No. 274 to DPR-68

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