ML120670065

Issuance of Amendments Reactivity Anomalies Surveillance

ML120670065
Person / Time
Site:	Limerick
Issue date:	03/14/2012
From:	Peter Bamford Plant Licensing Branch 1
To:	Pacilio M Exelon Nuclear
Bamford, Peter J., NRR/DORL 415-2833
References
TAC ME6348, TAC ME6349
Download: ML120670065 (21)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555"()001 Mlrch 14, 2012 Mr. Michael J. Pacilio President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

LIMERICK GENERATING STATION, UNITS 1 AND 2 -ISSUANCE OF AMENDMENTS RE: REACTIVITY ANOMALIES SURVEILLANCE (TAC NOS.

ME6348 AND ME6349)

Dear Mr. Pacilio:

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed Amendment No. 207 to Facility Operating License No. NPF-39 and Amendment No.168 to Facility Operating License No. NPF-85 for Limerick Generating Station (LGS), Units 1 and 2, respectively. The amendments are in response to your application dated June 2, 2011, as supplemented by letter dated November 10,2011.

The amendments revise the Technical Specifications for each unit by changing the method of calculating core reactivity for the purpose of performing the reactivity anomaly surveillance at Limerick Generating Station, Units 1 and 2. The change allows performance of the surveillance based on a comparison of predicted to actual (monitored) core reactivity. The reactivity anomaly verification was previously determined by a comparison of predicted versus actual control rod density.

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

s~~~

Peter Bamford, Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-352 and 50-353

Enclosures:

1. Amendment No.207 to License No. NPF-39

2. Amendment No. 168 to License No. NPF-85

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY. LLC DOCKET NO. 50-352 LIMERICK GENERATING STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment NO..207 License No. NPF-39

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

A. The application for amendment by Exelon Generation Company, LLC (the licensee), dated June 2,2011, as supplemented by letter dated November 10, 2011, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-39 is hereby amended to read as follows:

-2 (2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 207, are hereby incorporated into this license.

Exelon Generation Company 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 within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION lyyj. \( ~g,~

Meena Khanna, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of Issuance: rv6rch 14, 2012

ATTACHMENT TO LICENSE AMENDMENT NO. 207 FACILITY OPERATING LICENSE NO. NPF-39 DOCKET NO. 50-352 Replace the following page of the Facility Operating License with the revised page. The revised page is identified by amendment number and contains marginal lines indicating the area of change.

Remove Page 3 Page 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contains marginal lines indicating the areas of change.

Remove ii ii 1-7 1-7 3/4 1-2 3/4 1-2

-3 (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, 40,70, 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 (5) 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, and to receive and possess, but not separate, such source, byproduct, and special nuclear materials as contained in the fuel assemblies and fuel channels from the Shoreham Nuclear Power Station.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I (except as exempted from compliance in Section 2.0. below) 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 Exelon Generation Company is authorized to operate the facility at reactor core power levels not in excess of 3515 megawatts thermal (100% rated power) in accordance with the conditions specified herein and in Attachment 1 to this license. The items identified in Attachment 1 to this license shall be completed as specified. Attachment 1 is hereby incorporated into this license.

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 2JJ7 ,are hereby incorporated into this license.

Exelon Generation Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. 92, .:we, -tJ+, 447, ~ 2JJ7

Q"W}JD.."JQ~N~S==

SECTION DEFT NITIONS (Conti nued) PAGE 1.20a LOW (POWER) TRIP SETPOINT (LTSP) ............................. 1-4 1.2l (DELETED) .................................................... 1-4 1.22 r~EMBER(S) OF THE PUBLIC ...................*..............*... 1-4 1.22a MAPFAC(F) - (MAPLHGR FLOW FACTOR) ..........*......*.......... 1-4 1.22b MAPFAC(P) (POWER DEPENDENT MAPLHGR MULTIPLIER) *.**.....*..* 1-4 1.23 MINIMUM CRITICAL POWER RATIO (MCPR) .......................... 1-4 1.24 OFFSITE DOSE CALCULATION MANUAL ...................*.......... 1-4 1.25 OPERABLE - OPERABILITY ...................................**.. 1-4 1.26 OPERATIONAL CONDITION - CONDITION .......*.......*............ 1-5 1.27 PHYSICS TESTS .....*.....*.....................*..*****.....*. 1-5

1. 28 PRESSURE BOUNDARY LEAKAGE ....................*.*..*.....**.*. 1- 5 1.29 PRIMARY CONTAINMENT INTEGRITy ................................ 1-5 1.30 PROCESS CONTROL PROGRAM ..*..............*...............*.... 1-5 1.31 PURGE - PURGING *..**...*................*........***.*....... 1-6 1.32 RATED THERMAL POWER .......................................... 1-6 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITy ............ 1-6 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME ...................... 1-6 1.35 RECENTLY IRRADIATED FUEL ..................................... 1-6 1.36 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITy ..........**.. 1-6

1. 37 REPORTABLE EVENT ....*.............................*....**.... 1-7

1. 37a RESTRICTED AREA *...**..*...*............................**.** 1-7 1.38 (DELETED) *...***..*.*.....**............................*..*. 1-7 1.39 SHUTDOWN MARGIN .............................................. 1-7 1.40 SITE BOUNDARy .................................*.............. 1-7 1.41 SOURCE CHECK *..*....*.............................*..*....*.. 1-7 LIMERICK - UNIT 1 ii Amendment No. JJ,4S,~,ee,~,207

REFUELING FLOOR SECONDARY CQNTAJN~ENT INTEGRITY (Continued)

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

2. Closed by at least one manual valve. blind flange, slide gate damper. or deactivated automatic valve secured in its closed pOSition. except as provided by Specification 3.6.5.2.2.

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

c. The standby gas treatment system is in compliance with the requirements of specification 3.6.5.3.

d. At least one door in each access to the refueling floor secondary containment is closed.

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

f. The pressure within the refueling floor secondary containment is less than or equal to the value required by Specification 4.6.5.1.2a.

REPORTABLE EVENT 1.37 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

RESTRICTED AREA 1.37a RESTRICTED AREA means an area, access to which is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials. RESTRICTED AREA does not include areas used as residential quarters, but separate rooms in a residential building may be set apart as a RESTRICTED AREA.

1. 38 WeI eted)

SHUTDOWN t1ARGIN 1.39 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.

SITE BOUt'WARY 1.40 The SITE BOUNDARY shall be that line as defined in Figure 5.1.3 la.

1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

LIMERICK . UNIT 1 1-7

REACTIVITY CONTROL SYSTEMS 3/4,1.2 REACTIVITY ANOMALIES 3.1.2 The reactivity difference between the actual core keN and the predicted core Keff shall not exceed 1% ;'.k/k.

APPLICABILITY: OPERATIONAL CONDITION 1 and 2.

fleT ION:

With the reactivity difference exceeding 1% ~~/k:

a. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> perform an analysis to determine and explain the cause of the reactivity difference; operation may continue if the difference is explained and corrected.

b. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.1.2 The reactivity difference between the actual core k~f and the predicted core k~f shall be verified to be less than or equal to 1% ~k/k:

a. During the first startup following CORE ALTERATIONS, and

b. At least once per 31 effective full power days during POWER OPERATION.

c. The provisions of Specification 4.0.4 are not applicable.

LIMERICK - UNIT 1 3/4 1-2 Amendment No. B, '2fJ7

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 EXELON GENERATION COMPANY, LLC DOCKET NO. 50-353 LIMERICK GENERATING STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.168 License No. NPF-85

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

A. The application for amendment by Exelon Generation Company, LLC (the licensee), dated June 2,2011, as supplemented by letter dated November 10, 2011, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the pUblic; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical SpeCifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-85 is hereby amended to read as follows:

-2 (2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 168 , are hereby incorporated into this license.

Exelon Generation Company 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 within 60 days of issuance.

FOR THE NUCLEAR,~LATORY COMMISSION I..../"y\r\, \ C./'l~

Meena Khanna, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of Issuance: Mlrch 14, 2012

ATTACHMENT TO LICENSE AMENDMENT NO.168 FACILITY OPERATING LICENSE NO. NPF-85 DOCKET NO. 50-353 Replace the following page of the Facility Operating License with the revised page. The revised page is identified by amendment number and contains marginal lines indicating the area of change.

Remove Page 3 Page 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contains marginal lines indicating the areas of change.

Remove ii ii 1-7 1-7 3/4 1-2 3/4 1-2

-3 (4) Pursuant to the Act and 10 CFR Parts 3D, 40, 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 (5) 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, and to receive and possess, but not separate, such source, byproduct, and special nuclear materials as contained in the fuel assemblies and fuel channels from the Shoreham Nuclear Power Station.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I (except as exempted from compliance in Section 2.0. below) 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 Exelon Generation Company is authorized to operate the facility at reactor core power levels of 3515 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 168 ,are hereby incorporated into this license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Fire Protection (Section 9.5, SSER-2, -4)*

Exelon Generation Company shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Updated Final Safety AnalYSis Report for the facility, and as approved in the NRC Safety Evaluation Report dated August 1983 through Supplement 9, dated August 1989, and Safety Evaluation dated November 20, 1995, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

• The parenthetical notation following the title of license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

Amendment No.4, 2:7, &+, 98,~,.tOO, 493168

DEFINITIQNS (Continued) 1.20a LOW (POWER) TRIP SETPOINT (LTSP) ......***..**..*......**.**.* 1-4

.21 (DELETED} ...........*...*....*.**..*...*..*............**.*.. 1-4

.22 ME~1BER(S) OF THE PUBLiC **......*..*.*..............*..*....* 1*4 1.22a MAPFAC(F) - (MAPLHGR FLOW FACTOR) *..**..*....*.***...***...* 14 1.22b MAPFAC(P) - (POWER DEPENDENT MAPLHGR MULTIPLIER) ............ 1-4 1.23 MINIMUM CRITICAL POWER RATIO (MCPR) ......**...*..*..***.**** 1*4 1.24 OFFSITE DOSE CALCULATION MANUAL *******....**.*.***........** 1-4 1.25 OPERABLE OPERABILITY ...*..**.**.***.****....*....*...**..* 1-4 1.26 OPERATIONAL CONDITION CONDITION ....*.*.*.*...*..****..*.** 1-5 1.27 PHYSICS TESTS **...........*....****.*.........*.........*.** 1-5 1.28 PRESSURE BOUNDARY LEAKAGE ..*..*......**.*...*....*.........* 1-5 1.29 PRIMARY CONTAINMENT INTEGRITy .*.*******.**.*....*..********* 1-5 1.30 PROCESS CONTROL PROGRAM ....*.****.*....*.....*..*..*....**.* 1 5 1.31 PURGE - PURGING ............................................. 1 6 1.32 RATED THERMAL POWER **.****.....*..*******...........****.... 1-6 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITy ........... 1-6

1. 34 REACTOR PROTECTION SYSTEM RESPONSE TIME ...*..*............** 1 6 1.35 RECENTLY IRRADIATED FUEL .................................... 1-6 1.36 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITy ***.....*.... 1-6

~.37 REPORTABLE EVENT ............................................ 1-7 1.37a RESTRICTED AREA ***.*.....**.*.***.*..*.*.*....**...........* 1 7 1.38 (DELETED) ****..**..*****...*.**.**.****.**..**.*..*......*.* 1-7

1. 39 SHUTDOWN MARGI N............................................. 1- 7 1.40 SITE BOUNDARy .**..*.*...*******.....*.****.......*........*. 1 7 1.41 SOURCE CHECK .......*.........**.*......*..**..*.........***. 17 LIMERICK - UNIT 2 ii Amendment No. +/-+/-,+&,4@ ,168

REFUELING FLQOR SECQNDARY CONTAINMENT INTEGRITY (Continued)

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

2. Closed by at least one manual valve, blind flange, slide gate damper or deactivated automatic valve secured in its closed position, except as provided by ification 3.6.5.2.2.

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

c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.

d. At least one door in each access to the refueling floor secondary containment is closed.

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

f. The pressure within the refueling floor secondary containment is less than or equal to the value required by Specification 4.6.5.1.2a.

REPORTABLE EVENT 1.37 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

RESTRICTED AREA 1.37a RESTRICTED AREA means an area, access to which is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials. RESTRICTED AREA does not include areas used as residential quarters, but separate rooms in a residential building may be set apart as a RESTRICTED AREA.

1. 38 (Deleted)

SHUTDOWN MARGIN 1.39 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 for the single control rod of highest reactivity worth which is assumed to fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68°F; and xenon free.

SITE BOUNDARY 1.40 The SITE BOUNDARY shall be that line as defined in Figure 5.1.3-1a.

SOURCE CHECK 1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

LIMERICK UNIT 2 1 7 Amendment No. +/-+/-,4B,e9,l4&,+/-48,168

REACTiVITY CONTROL SYSTEMS 3/4.1,2 REACTIVITY ANOMALIES 3.1.2 The reactivity difference between the actual core kerf and the predicted core keff sha 11 not exceed 1% uk/k.

APPLICABILITY; OPERATIONAL CONDITION 1 and 2.

ACTION:

With the reactivity difference exceeding 1% uk/k:

a. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> perform an analysis to determine and explain the cause of the reactivity difference; operation may continue if the diffe rence is explained and corrected.

b. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4. L 2 The reactivity difference between the actual core keff and the predi cted core keff shall be verified to be less than or equal to 1% t.k/k:

a. During the first startup following CORE ALTERATIONS, and

b. At least once per 31 effective full power days during POWER OPERATION.

c. The provisions of Specification 4.0.4 are not applicable.

LIMERICK UNIT 2 3/4 1-2 Amendment No. 168

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 2!J7 TO FACILITY OPERATING LICENSE NO. NPF-39 AND AMENDMENT NO. 168 TO FACILITY OPERATING LICENSE NO. NPF-85 EXELON GENERATION COMPANY, LLC LIMERICK GENERATING STATION, UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353

1.0 INTRODUCTION

By application dated June 2, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML111540122), as supplemented by letter dated November 10, 2011, (ADAMS Accession No. ML113180238) Exelon Generation Company, LLC (Exelon, the licensee), requested changes to the Technical SpeCification (TS) 3/4.1.2, "Reactivity Anomalies," for Limerick Generating Station (LGS), Units 1 and 2. The requested change seeks to modify the method used to perform the reactivity anomaly check. The change allows performance of the surveillance based on a comparison of predicted to actual (monitored) core reactivity. The reactivity anomaly verification was previously determined by a comparison of predicted versus actual control rod density. The acceptance criteria for the surveillance remains at a reactivity level of 1% L\klk.

The supplement dated November 10, 2011, provided additional information that clarified the application, did not expand the scope of the application, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register on August 9, 2011 (76 FR 48911).

2.0 REGULATORY EVALUATION

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," GDC-28, "Reactivity limits," and GDC-29, "Protection against anticipated operational occurrences," require that reactivity within the core be controllable to ensure that subcriticality is achievable and maintainable under cold conditions (most reactive conditions). In addition, these GDC also specify that applicable fuel design limits must not be exceeded during normal operations and anticipated operational occurrences. According to the site Updated Final Safety Analysis Report, Section 3.1, LGS satisfies the requirements of GDC-26, GDC-28 and GDC-29.

The reactivity anomaly check required by TS 3/4.1.2 serves, in part, to satisfy the above GDC by comparing the observed reactivity behavior of the core (at hot operating conditions) to the expected reactivity behavior that was calculated during the core design process. This ensures Enclosure

-2 that certain assumptions in the design-basis accident and transient safety analyses remain valid. Any difference between these two observations is compared to the TS 3/4.1.2 acceptance criterion of 1% liklk and if the criterion is not met, the action required by the TS is then taken.

The U.S. Nuclear Regulatory Commission's (NRC's) regulatory requirements related to the content of the TSs are contained in 10 CFR 50.36. Paragraph (c)(2)(i) of 10 CFR 50.36 states that limiting conditions for operation (LCOs) " ... are the lowest functional capability or performance levels of equipment required for safe operation of the facility." Paragraph (c)(3) of 10 CFR 50.36 states that surveillance requirements are" ... requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met."

3.0 TECHNICAL EVALUATION

3.1 Current Method for Reactivity Anomaly Check As described in the submittal dated June 2, 2011, the licensee currently assesses reactivity anomalies at LGS, Units 1 and 2, by using rod density to provide a convenient representation for the effective core multiplication factor (keff)' This method is utilized primarily because early core monitoring systems did not directly calculate core keff values for comparison to design values. The LGS, Unit 1 and 2 TSs currently require that the reactivity anomaly check be done by comparing a predicted control rod density (calculated prior to the start of operation for a particular cycle) to an actual control rod density. The comparison is done at the frequencies specified by Surveillance Requirement (SR) 4.1.2.

Specifically, comparison of predicted control rod density to actual control rod density is done via a set of reactivity anomaly curves. Development of the curves begins with predicted critical core keff values, which have been calculated for projected operating states and conditions throughout the life of the cycle, and their associated derived control rod patterns. A calculation is made of the number of notches inserted in these rod patterns and also the number of average notches required to make a change of 1% liklk around the predicted critical core keff values. The notches are converted to control rod density and plotted as a function of cycle exposure to produce a predicted control rod density curve with upper and lower bounds that represent the 1% liklk TS acceptance criterion. As a result, the comparison is still based on critical keff' but with a "translation" of acceptance criteria to control rod density.

Under the current method, an anomaly would be the difference between the predicted and measured control rod density in the reactor under the existing conditions, e.g., time in cycle, power level and control rod pattern. The observed anomaly is then translated into a reactivity difference between the two values (the measured versus the predicted control rod density) for comparison to the TS limit of 1% liklk. If the limit is exceeded, the licensee has 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to perform an analysis to determine and explain the cause of the reactivity difference. Operation may continue if the difference is explained and corrected, otherwise the plant must be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The licensee stated that, while being a convenient measurement of core reactivity, the control rod density method has limitations, such as differing impacts on reactivity from deeply inserted

- 3 central control rods versus control rods on the outer edge of the core, or shallowly inserted rods.

The licensee indicated that it is not uncommon for reactivity anomaly concerns to arise during operation simply because of greater use of near-edge or shallow-inserted control rods than anticipated, when in fact no true anomaly exists.

3.2 Proposed Method for Reactivity Anomaly Check The proposed change to the TS would eliminate the translation of core keff into control rod density. Instead, the revised method for evaluating a potential reactivity anomaly would compare the measured core keff and the predicted core keff directly. The proposed TS change will not change the frequency or any condition within the SR.

According to the licensee, advances in computational methods and computer technology support the proposed TS amendment. The licensee utilizes the Global Nuclear Fuels (GNF) three dimensional (3D) core monitoring system, 3D MONICORE, at LGS, Units 1 and 2, which incorporates the 3D core simulator code, PANACEA Version 11 (PANAC11). The system allows for a direct comparison of predicted core keff to monitored core keff. Measured core keff is calculated by PANAC11 using plant operating data provided by 3D MONICORE. The predicted core keff' as a function of cycle exposure, is developed using PANAC11 during the core design process. The PANAC11-computed measured core keff behavior from the previous cycle is used as the starting point for the calculation. Any fuel vendor recommended adjustments due to planned changes in fuel design, core design, or operating strategy for the upcoming cycle are also incorporated into the development of the predicted core keff.

By letter dated March 11, 1999 (ADAMS Accession No. ML993140059), the NRC approved the power distribution uncertainty for the 3D-MONICORE core surveillance system by accepting NEDC-32694P, "Power Distribution Uncertainties for Safety Limit MCPR [minimum critical power ratiO] Evaluation," with limitations, for referencing in license applications. Further, by letter dated November 10, 1999 (ADAMS Accession No. ML993230184), the NRC staff documented an evaluation of a version of the PANACEA core simulator code, referred to as PANAC11. In that evaluation, the NRC staff concluded that a proposed improvement in General Electric (GE) steady-state methods (reflected in PANAC11) was acceptable and appropriate for inclusion into the GE Licensing Topical Report for core design, NEDE-24011-P-A. Therefore, the NRC staff concludes that the use of 3D-MONICORE and PANAC11 as the basis for the proposed changes to the reactivity anomaly surveillance are appropriate. The NRC staff agrees with the licensee that use of actual (measured) to predicted core keff, instead of rod density, eliminates the limitations described in Section 3.1 of this safety evaluation; provides for a technically superior comparison; is a simple and straightforward approach utilizing appropriate current computer codes and methods; and is therefore, acceptable.

The NRC staff notes that the licensee's proposed TS change is similar to the Boiling Water Reactor (BWR)/6 Standard TS for reactivity anomalies, in that it performs the reactivity difference comparison using core keff. Although LGS is a BWRl4 plant, it has the hardware and software in place (3D MONICORE, PANAC11) to allow direct comparison of predicted keff to measured keff as described in the basis for the comparable BWRl6 surveillance (NUREG-1434, Volume 2, Revision 3.0, "Standard Technical Specifications General Electric Plants, BWRl6 Bases").

-4 The licensee also assessed the impact of this request on the LGS transient and accident analyses, and determined that the proposed changes will not affect any of the transient and accident analyses. The licensee stated that this is because only the method of performing the reactivity anomaly surveillance is changing, and the proposed method will provide an adequate acceptable comparison as discussed above. Furthermore, the anomaly check will continue to be performed at the current required frequency. The NRC staff agrees with this assessment and therefore concludes that the proposed surveillance will continue to ensure that the assumptions in the transient and accident analysis regarding core reactivity remain valid with this change.

3.3 Technical Specification Changes Included in the licensee's application dated June 2, 2011, were proposed changes to the LGS TSs to reflect the new method of performing the reactivity anomaly check. The initial version of the proposed change to SR 4.1.2 did not retain the 1% Aklk comparison criterion present in the original TS. In response to a request for additional information from the NRC staff, the licensee modified the original application with a letter dated November 10, 2011. This revised submittal included the 1% Aklk comparison criterion in the revised SR 4.1.2. The NRC staff concludes that the surveillance, as proposed in the licensee's letter dated November 10, 2011, provides acceptance criteria consistent with the LCO, and thus meets the requirements of 10 CFR 50.36(c)(3). Therefore, the proposed change is acceptable. The licensee's proposed LCO 3.1.2 and SR 4.1.2 refer to the reactivity difference between the "actual" core ketr and the predicted core ketr. The NRC staff also notes that, consistent with the discussion in Section 3.2 of this safety evaluation, the "actual" core keff refers to the monitored core reactivity as indicated by the 3D-MONICORE system.

3.4 Summary The NRC staff has reviewed the licensee's request to revise LGS, Units 1 and 2, TS 3/4.1.2.

Based on this review, as described above, the NRC staff concludes that these revisions will provide an improved approach for the determination of reactivity anomalies, and are acceptable.

4.0 STATE CONSULTATION

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

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. 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, and there has been no public comment on such finding (76 FR 48911). 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),

- 5 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) 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 Contributors: A. K. Heller P. Bamford Date: M3rch 14, 2012

March 14, 2012 Mr. Michael J. Pacilio President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

LIMERICK GENERATING STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: REACTIVITY ANOMALIES SURVEILLANCE (TAC NOS.

ME6348 AND ME6349)

Dear Mr. Pacilio:

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed Amendment No.207 to Facility Operating License No. NPF-39 and Amendment No. 168 to Facility Operating License No. NPF-85 for Limerick Generating Station (LGS), Units 1 and 2, respectively. The amendments are in response to your application dated June 2,2011, as supplemented by letter dated November 10, 2011.

The amendments revise the Technical Specifications for each unit by changing the method of calculating core reactivity for the purpose of performing the reactivity anomaly surveillance at Limerick Generating Station, Units 1 and 2. The change allows performance of the surveillance based on a comparison of predicted to actual (monitored) core reactivity. The reactivity anomaly verification was previously determined by a comparison of predicted versus actual control rod density.

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

Sincerely, IRA!

Peter Bamford, Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-352 and 50-353

Enclosures:

1. Amendment No. 207 to License No. NPF-39

2. Amendment No.168 to License No. NPF-85

3. Safety Evaluation cc w/enclosures: Distribution via Listserv DISTRIBUTION:

PUBLIC RidsNrrDorlLpl1-2 Resource RidsNrrPMLimerick Resource RidsNrrlAABaxter Resource RidsNrrDorlDpr Resource LPL 1-2 RlF RidsRgn1 MailCenter Resource RidsAcrsAcnw_MailCTR Resource AKHelier, NRR RidsOgcRp Resource RidsNrrDssSnpb Resource ADAMS Accession No* ML120670065 *viamemo OFFICE LPLI-2/PM LPLI-2/LA DSS/STSB/BC DSS/SNPB/BC OGC LPLI-2/BC NAME PBamford ABaxter RElliott AMendiola* CKanatas MKhanna DATE 3/1/2012 3/7/2012 3/9/2012 02110/2012 3113/2012 3/14/2012 Official Record Copy