ML14345A895

Issuance of Amendments Regarding Revised Shutdown Margin

ML14345A895
Person / Time
Site:	Hatch
Issue date:	01/06/2015
From:	Martin R Plant Licensing Branch II
To:	Pierce C Southern Nuclear Operating Co
Martin R
References
TAC MF3627, TAC MF3628
Download: ML14345A895 (20)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 6, 2015 Mr. C. R. Pierce Regulatory Affairs Director Southern Nuclear Operating Company, Inc.

Post Office Box 1295, Bin - 038 Birmingham, AL 35201-1295

SUBJECT:

EDWIN I. HATCH NUCLEAR PLANT, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING REVISED SHUTDOWN MARGIN (TAG NOS. MF3627 AND MF3628)

Dear Mr. Pierce:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 273 to Renewed Facility Operating License DPR-57 and Amendment No. 217 to Renewed Facility Operating License NPF-5 for the Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, in response to the license amendment application (LAR) dated March 17, 2014. The LAR requests the adoption of Technical Specification Task Force traveler TSTF-535, Revision 0, "Revise Shutdown Margin Definition to Address Advanced Fuel Designs," which is an approved change to the Standard Technical Specifications. The changes modify the Technical Specification definition of "Shutdown Margin" (SDM) to require calculation of the SDM at a reactor moderator temperature of 68°F or a higher temperature that represents the most reactive state throughout the operating cycle.

A *copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

~CLJ~);C;(

Robert Martin, Senior Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-321 and 50-366 r*

Enclosures:

1. Amendment No. 273 to DPR-57

2. Amendment No. 217 to NPF-5

3. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION

'WASHINGTON, D.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 273 Renewed License No. DPR-57

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

A. The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit No. 1 (the facility) Renewed Facility Operating License No. DPR-57 filed by Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated March 17, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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.

Enclosure 1

2. Accordingly, the license is hereby amended by page 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-57 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.

273 are hereby incorporated in the license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented prior to reactor startup following Unit 1 refueling_ outage 1R27 (spring 2016).

FOR THE NUCLEAR REGULATORY COMMISSION Robert Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-57 and the Technical Specifications Date of Issuance: January 6, 2015

ATTACHMENT TO LICENSE AMENDMENT NO. 273 RENEWED FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the following pages of the License and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages License License DPR-57, Page 4 DPR-57, Page 4 TSs TSs 1.1-5 1.1-5 Ii.

I

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for sample analysis or instrument calibration, or associated with radioactive apparatus or components; (6) Southern Nuclear, 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 license shall be deemed to contain and is subject to the conditions specified In the following Commission regulatiohs in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions specified or incorporated below:

(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at steady state reactor core power levels not in excess of 2804 megawatts thermal.

(2) Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B), as revised through Amendment No. 273 are hereby incorporated in the renewed license. Southern Nuclear shalr operate the* facility in accordance with the Technical Specifications and the Environmental Protection Plan.

The Surveillance Requirement (SR) contained in the Technical Specifications and listed below, is not required to be performed .

immediately upon implementation of Amendment No. 195. The SR listed below shall be successfulfy demonstrated before the time and condition specified:

SR 3.8.1.18 shall be successfully demonstrated at its next

. regularly scheduled performance.

(3) Fire Protection Southern Nuclear shall Implement and maintain in effect.all provisions of the fire protection program, which is referenced in the Updated Final Safety Analysis Report for the facility, as contained in the updated Fire Hazards Analysis and Fire Protection Program for the Edwin I. Hatch Nuclear Plant, Units 1 and 2, which was originally submitted by letter dated July 22, 1986. Southern Nuclear may make changes to the fire protection program without prior Commission approval only if the changes Renewed License No. DPR-57 Amendment No:213. *1

Definitions 1.1 1.1 Definitions (continued)

SHUTDOWN SDM shall be the amount of reactivity by which the reactor is subcritical MARGIN (SDM) 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 ~xcept 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 SDM.

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

TURBINE BYPASS .The TURBINE BYPASS SYSTEM RESPONSE TIME consists of tWo SYSTEM components:

RESPONSE

TIME a. The time from initial movement of the main turbine stop valve or control valve until 80% of the turbine bypass capacity is established; and

b. The time from initial movement of the main turbine stop valve or control valve until initial movement of the turbine bypass valve.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

HATCH UNIT 1 1.1-5 Amendment No. 2731

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY, INC:

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-366 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO.2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 217 Renewed License No. NPF-5

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

A. The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit No. 2 (the facility) Renewed Facility Operating License No. NPF-5 filed by Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated March 17, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 s.uch 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.

Enclosure 2

2. Accordingly, the license is hereby amended by page 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-5 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.

217 are hereby incorporated in the license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications a,_nd the Environmental Protection Plan. *

3.

• This license amendment is effective as of its date of .issuance and shall be implemented prior to reactor startup following Unit 2 refueling outage 2R23 (spring 2015).

FOR THE NUCLEAR REGULATORY COMMISSION Robert Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Offic_e of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating. License No. NPF-5 and the Technical Specifications Date of Issuance: January 6, 2015

• ATTACHMENT TO LICENSE AMENDMENT NO. 217 RENEWED FACILITY OPERATING LICENSE NO. NPF-5 DOCKET NO. 50-366 Replace the following pages of the License and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages License License NPF-5, Page 4 NPF-5, Page 4 TSs TSs

. 1.1-5 1.1-5

.J (6) Southern Nuclear; 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 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, Section 50.54 of Part 50, and Section 70.32 of Part 70; all appiicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions 2 specified or incorporated below:

(1-) Maximum Power Level Southern Nuclear is authorized to operate the facility at steady state reactor core power l~vels not in excess of 2,804 megawatts thermal, in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B); as revised through Amendment No:: 217 are hereby incorporated in the renewed license. Southern Nuclear snarl operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the issuance of the renewed license or within the operational restrictions

• indicated. The removal of these conditions shall be made by an amendment to the license supported by a favorable evaluation by the Commission.

(

(a) Fire Protection.

Southern Nuclear shall implement and maintain in effect all provisions of the fire protection program, which is referenced in the Updated Final Safety Analysis Report for the facility, as contained 2 The original licensee authorized to possess, use, and operate the facility was Geprgia, I Power Company (GPC). Consequently, certain historical references to GPC remain in certain license conditions.

• Renewed License ,No. NPF*5 Amendment No: 21 7

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 Chapter 14, Initial Tests and Operation, of the FSAR;

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

c. Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL RTP shall be a total reactor core heat transfer rate to the reactor POWER (RTP) coolant of 2804 MWt.

REACTOR The RPS RESPONSE TIME shall be that time interval from when the PROTECTION monitored parameter. exceeds its RPS trip setpoint at the channel SYSTEM (RPS) sensor until de-energization of the scram pilot valve solenoids. The RESPONSE TIME response time may be measured by meaQS of any series of sequential, overlapping, or total steps so that the entire response time is measured.

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SHUTDOWN SDM shall be the amount of reactivity by which the reactor is subcritical MARGIN (SDM) or would be sybcritical 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 SDM.

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

(continued)

HATCH UNIT 2 1.1-5 Amendment No.217 1

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 273 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-57 AND AMENDMENT NO. 217 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-5 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

EDWIN I. HATCH NUCLEAR PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-321 AND 50-366 1.0. INTRODUCTION By license. amendment request (LAR) dated March 17,2014 (Agencywide Documents Access and Management System (ADAMS), Accession No. ML14076A141) Southern Nuclear Operating Company, Inc. (SNC, the licensee), requested an amendment to the Technical Specifications (TS) for the Edwin I. Hatch Nuclear Plant, Units 1 and 2 (HNP). The LAR requests the adoption of approved Technical Specification Task Force (TSTF) traveler TSTF-535, Revision 0, "Revise Shutdown Margin Definition to Address Advanced Fuel Designs,"

which is an approved change to the Standard Technical Specifications (STS). The proposed changes modify the TS definition of "Shutdown Margin" (SDM) to require calculation of the SDM at a reactor moderator temperature of 68°F or a higher temperature that represents the most reactive state throughout the operating cycle.

The application states, "Southern Nuclear Operating Company (SNC) has reviewed the model safety evaluation, dated February 19, 2013, as part of the Federal Register Notice of Availability. This review included a review of the NRC staff's evaluation, as well as the information provided in TSTF-535. As described in the subsequent paragraphs, SNC has concluded that the justifications presented in the TSTF-535 proposal and the model safety evaluation prepared by the NRC staff are applicable to Edwin I. Hatch Nuclear Plant (HNP), Unit 1 and Unit 2, and justify this amendment for the incorporation of the changes to the HNP Unit 1 and Unit 2 TS."

Enclosure 3

The Staff's evaluation of the licensee's proposed changes is provided below.

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 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 SDM, TSTF-535 discussed s*oM with regards to GE14 and GNF2 fuels. TSTF-535 indicated that for historical fuel products through GE14, the maximum reactivity condition for SDM 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 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 SDM 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 The licensee's adoption of TSTF-535 for HNP's TSs proposes to revise the TS definition of SDM to require calculation of SDM at the reactor moderator temperature corresponding to the most reactive state throughout the operating cycle (68°F or higher). The current definition of SDM in Section 1.1, "Definitions," of the HNP Units 1 and 2 TSs is:

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

a. The reactor is xenon free

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b. The moderator temperature is 68°F, and

c. All control rods are fully inserted except for the single control rod of the 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 SDM.

The licensee proposes the following changes (shown in bOld) to the HNP, Units 1 and 2, TSs definition of SDM in accordance with TSTF-535:

SDM 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, and corresponding to the most reactive state; and

c. All control rods are fully inserted except for the single control rod of the 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 SDM.

2.3 Regulatory Review Title 10 of the Code of Federal Regulations (1 0 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.

Among other things, 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 SDM and the LCOs placed on SDM serve, in part, to satisfy GDCs 26 and 27 by ensuring there is always suffi~ient negative reactivity worth available to offset the positive reactivityworth 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 following provides the licensee's explanation regarding meeting the GDCs discussed above.

• The licensee's Evaluation Against Criterion 26 and 27 for HNP Units 1 and 2:

The licensee states,,

"The Traveler and model Safety Evaluation (SE) discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). HNP Unit 1 was not licensed to the 10 CFR 50, Appendix A, GDC. The HNP Unit 1 construction permit was received under the 70 general design criteria issued for comment in July 1967, as discussed in section F.3 of the UFSAR. (Appendix F has since beeri designated as historical). The HNP Unit 1 licensing basis criteria which are equivalent to the referenced GDCs in the model SE are:

• 10 CFR 50 Appendix A Criterion 26, "Reactivity Control System Redundancy and Capability" Two independent reactivity control systems of different qesign principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.

HNP Unit 1 10 CFR 50 Appendix A Criterion 26 Reconciliation:

The HNP Unit 1 Updated Final Safety Analysis Report (UFSAR) chapter that discusses the reactivity control system redundancy and capability is Chapter 3.0, "Reactor." This chapter of the Unit 1 UFSAR refers to Chapter 4.0, "Reactor," of the Unit 2 UFSAR in its entirety. As discussed in Chapter 4.0 of the Unit 2 UFSAR, the two independent reactivity control systems of different design ,

principles provided for HNP Unit 1 and Unit 2 are by the control rods and standby liquid control system (SBLC). HNP Unit 2 was licensed under the 10 CFR 50 Appendix A GDC. The UFSAR licensing basis for HNP Unit 1 also meets 10 CFR 50 Appendix A Criterion 26.

• 10 CFR 50 Appendix A Criterion 27, "Combined Reactivity Control Systems Capability" The reactivity control systems shall be designed to have a combined capability, in conjunctionwith poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions

and with appropriate margin for stuck rods the capability to cool the core is maintained.

HNP Unit 1 10 CFR 50 Appendix A Criterion 27 Reconciliation:

As stated above, the HNP Unit 2 reactor (including the control rods and SBLC) was licensing under the current GDC. Given that the UFSAR licensing basis for the reactivity control systems (control rods and SBLC) applies to both Unit 1 and Unit 2, Unit 1 is therefore licensed to this Criterion.

, The application further states, "Therefore, the original licensing basis for HNP Unit 1 does not alter the conclusion that the proposed change is applicable to Unit 1.

As previously stated, Unit 2 was licensed under the current GDC."

The NRC staff, in part, based upon the licensing basis mentioned above, granted the full term operating license to HNP Unit 1 on October 13, 1974. The NRC staff, therefore, concludes that the HNP1 licensing bases, i.e., UFSAR, documents that the HNP1 plant-specific criteria requirements are an acceptable equivalent to the 10 CFR Part 50, Appendix A GDCs 26 and 27.

Based on its review of the licensee's UFSAR, Revision 28, Section 3.0, "Design Of Structures,*

Components, Equipment, And Systems," the NRC staff finds that HNP Unit 2's compliance with Criteria 26 and 27 has been adequately evaluated by the licensee.

3.0 TECHNICAL EVALUATION

3.1 Current Definition of Shutdown Margin In BWR plants, such as, HNP Units 1 and 2, 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 SDM. 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 the STS defines SDM 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 worth rod is fully withdrawn, the core cari be designed with adequate shutdown margin to

ensure it remains safely shutdown even in the event of a stuck control rod, as required by GDCs 26 and 27.

Determination of the SDM under the aforementioned conditions yields a conservative result that, alon,g with the requirements set forth in Section 3.1.1 of the TS, 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 reactorwill 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 SDM 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 SDM at the currently defined moderator temperature of 68°F may not accurately determine the available margin.

TSTF-535 therefore proposed a change to the definition of SDM to enable calculation of the a

SDM at a reactor moderator temperature of 68°F or 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 SDM as specified in Topical Report NED0-24011-A, Revision 18, "General Electric Standard Application for Reactor Fuel (GESTAR II)," dated April 20, 19, and 30, 2011 (ADAMS Accession Nos. ML111120040, ML111120044, ML111120046, and ML111120047, respectively). Section 3.2.4.1 of the in GESTAR II report (ADAMS Accession No. ML111120046) states:

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_)

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.

The Traveler also cited SRP Section 4.3, which states the following concerning the review of control systems and SDM:

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 (SDM). Shutdown margins need to be demonstrated by the applicant throughout the fuel cycle.

Although the licensing basis requirements for SDM 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 GDCs 26 and 27.

TSTF-535 stated that while the SRP does not prescribe the temperature at which the minimum SDM 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 SDM.

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 SDM 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 SDM at the most limiting core reactivity condition is prudent with respect to ensuring compliance with GDCs 26 and 27 or their plant-specific equivalent, and concludes that the proposed changes to the HNP, Units 1 and 2, TSs are acceptable.

The impetus for TSTF-535 was to provide for a more broadly applicable SDM 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 SDM as required by their current licensing basis. Therefore, the change is acceptable for HNP 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 SDM consider a broadly conservative range of potential initial conditions in the anticipated operational occurrence analyses.

4.0 Summary The NRC staff has reviewed the licensee's implementation of TSTF-535 proposed revisions to the definition of SDM. 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 SDM that ensures use of the appropriate limiting conditions for all fuel types at any time in the life of the core. Additionally, the NRC staff concludes the proposed changes to the definition of SDM will require the licensee to calculate SDM In consideration of the most limiting conditions in the core. Therefore, the revised SDM definition is acceptable for use with any current fuel design.

5.0 STATE CONSULTATION

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

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the 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 considerations, and there has been no public comment on the finding (79 FR 42552, July 22, 2014). 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.

7.0 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:

Ravinder Grover, NRR/DSS/STSB Date: January 6, 2015

ML14345A895 *SE transmitted by memo dated OFFICE NRR/LPL2-1/PM NRR/LPL2'-1/LA NRR/DSS/STSB/BC NAME RMartin SFigueroa REIIiott DATE 12/16/14 01/05/15 11/25/14*

OFFICE OGC NRR/LPL2-1/BC NRR/LPL2-1/PM NAME Jlindell RPascarelli RMartin (SWilliams for)

DATE 12/24/14 01/06/15 01/06/15