ML060540065

License Amendment Request Re Single Loop Operation Safety Limit Minimum Critical Power Ratio (SLO SLMCPR) Change

ML060540065
Person / Time
Site:	Limerick
Issue date:	12/14/2005
From:	Cowan P Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML060540065 (25)

Text

Exelon Nuclear 200 Exelon Way www.exeloncorp.com Nuclear Kennett Square, PA 19348 10 CFR 50.90 December 14,2005 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 Limerick Generating Station, Unit 1 Facility Operating License No. NPF-39 NO. 5 0 - a

SUBJECT:

License Amendment Request Single Loop Operation Safety Limit Minimum Critical Power Ratio (SLO SLMCPR) Change

Dear Sir/Madam:

Pursuant to 10 CFR 50.90 Exelon Generation Company, LLC (Exelon), hereby requests the following amendment to the Technical Specifications (TS), Appendix A of Operating License No.

NPF-39 for Limerick Generating Station (LGS), Unit 1. This proposed change will revise Technical Specification (TS) Section 2.1. This Section will be revised to incorporate a revised Single Loop Operation Safety Limit Minimum Critical Power Ratio (SLO SLMCPR) due to the cycle specific analysis performed by Global Nuclear Fuel for LGS, Unit 1, Cycle 12. The two-loop SLMCPR will not change.

In order to support the upcoming refueling outage at LGS, Unit 1, Exelon requests approval of the proposed amendment by March 1,2006. Once approved, this amendment shall be implemented within 30 days of issuance. Additionally, there are no commitments contained within this letter.

This proposed change has been reviewed by the Plant Operations Review Committee, and approved by the Nuclear Safety Review Board.

Information supporting this License Amendment Request is contained in Enclosure 1 to this letter, and the proposed marked up pages and final camera ready pages are contained in Attachments 2 and 3,respectively. Enclosure 4 (letter from M. J. Mneimneh {Global Nuclear Fuel} to J. Tusar {Exelon Generation Company, LLC}, dated November 9, 2005) specifies the new SLO SLMCPR for LGS, Unit 1. Enclosure 4 contains information proprietary to Global Nuclear Fuel. Global Nuclear Fuel requests that the document be withheld from public disclosure in accordance with 10 CFR 2.390(a)(4). An affidavit supporting this request is also contained in Enclosure 4. Enclosure 5 contains a non-proprietaryversion of the Global Nuclear Fuel document.

LGS Unit 1 License Amendment December 14,2005 Page 2 We are notifying the Commonwealth of Pennsylvania of this application for changes to the Technical Specifications by transmitting a copy of this letter and its attachments to the designated State Official.

If any additional information is needed, please contact Tom Loomis at (610) 765-5510.

I declare under penalty of perjury that the forgoing is true and correct.

Respectfully, Director, Licensing & Regulatory Affairs Exelon Generation Company, LLC

Enclosures:

1-Evaluation of Proposed Change 2-Markup of Technical Specification Pages 3-Camera Ready Technical Specification Pages 4-Proprietary Version of Global Nuclear Fuel Letter 5-Non-Proprietary Version of Global Nuclear Fuel Letter cc: S. J. Collins, Administrator, USNRC Region I S. Hansell, USNRC Senior Resident Inspector, LGS G. Wunder, Project Manager, USNRC R. R. Janati, Commonwealth of Pennsylvania

ENCLOSURE 1 LIMERICK GENERATING STATION UNIT 1 DOCKET NO. 50-352 LICENSE NO. NPF-39 LICENSE AMENDMENT REQUEST EVALUATION OF PROPOSED CHANGE

ENCLOSURE 1 CONTENTS

SUBJECT:

Revision to Single Loop Operation Safety Limit Minimum Critical Power Ratio 1.o DESCRIPTION

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirementdcriteria

7.0 REFERENCES

- License Amendment Request Docket No. 50-352 Evaluation of Proposed Change License No. NPF-39 December 14,2005 Page 1 of 5 1.o DESCRIPTION Exelon Generation Company, LLC (Exelon) Licensee under Facility Operating License No. NPF-39 for Limerick Generating Station (LGS), Unit 1, requests that the Technical Specifications (TS) contained in Appendix A to the Operating License be amended to revise TS 2.1 to reflect a change in the Single Loop Operation Safety Limit Minimum Critical Power Ratio (SLO SLMCPR) due to the cycle specific analysis performed by Global Nuclear Fuel for LGS, Unit 1, Cycle 12. The two-loop SLMCPR will not change.

The marked up Technical Specification page and camera ready Technical Specification page are contained in Enclosures 2 and 3, respectively. Also included in Enclosures 2 and 3 are the associated Bases changes, which are being supplied to you for your information. Enclosure 4 (letter from M. J. Mneimneh (Global Nuclear Fuel) to J. Tusar (Exelon Generation Company, LLC), dated November 9, 2005) specifies the new SLO SLMCPR for LGS, Unit 1, Cycle 12.

2.0 PROPOSED CHANGE

This proposed change will revise Technical Specification (TS) Section 2.1. This Section will be revised to incorporate a revised Single Loop Operation Safety Limit Minimum Critical Power Ratio (SLO SLMCPR) due to the cycle specific analysis performed by Global Nuclear Fuel for LGS, Unit 1, Cycle 12. The new SLO SLMCPR at LGS, Unit 1, Cycle 12 is 1.09. The two-loop operation SLMCPR will remain the same (1.07) as shown in Enclosure 4. Additional information regarding the 1.09 cycle specific SLO SLMCPR for LGS, Unit 1 Cycle 12 is contained in the Enclosure 4 letter.

3.0 BACKGROUND

The proposed change involves revising the Single Loop Operation Safety Limit Minimum Critical Power Ratio (SLO SLMCPR) value contained in TS 2.1 for single recirculation loop operation. The SLO SLMCPR value is being revised for LGS, Unit 1 based on the reload core design for Cycle 12, which uses the GE-14 fuel product line. GE-14 fuel has previously been loaded into the Limerick Generating Station, Unit 1 core. The SLO SLMCPR has been determined in accordance with NRC approved methodology described in General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-15 (GESTAR-II), and U. S. Supplement, NEDE-24011-P-A-15-US, September, 2005, which includes Amendment 25. Amendment 25 provides the methodology for determining the cycle specific MCPR safety limits that replace the former generic fuel type dependent values. Amendment 25 is being used for determining the upcoming Cycle 12 SLO SLMCPR. Future SLO SLMCPRs determined in accordance with Amendment 25 will not need prior NRC approval for each cycle unless the TS value changes. The NRC safety evaluation approving Amendment 25 is contained in a letter from the NRC to General Electric Company, dated March 11, 1999 (F. Akstulewicz (NRC) to G. A. Watford (GE), Acceptance for Referencing of Licensing Topical Reports NEDC-32601P, Methodology and Uncertainties for Safety Limit MCPR Evaluations; NEDC-32694P, Power Distribution Uncertainties for Safety Limit MCPR Evaluation; and Amendment 25 to NEDE-24011-P-A on Cycle-Specific Safety Limit MCPR, (TAC Nos.

M97490. M99069 and M97491)).

- License Amendment Request Docket No. 50-352 Evaluation of Proposed Change License No. NPF-39 December 14,2005 Page 2 of 5 Global Nuclear Fuel has designed GE-14 fuel to be in compliance with Amendment 22 included in General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-15 (GESTAR-II), and U. S. Supplement, NEDE-24011-P-A-15-US, September, 2005.

4.0 TECHNICAL ANALYSIS

The proposed TS change will revise TS 2.1 to reflect the cycle specific analysis performed by Global Nuclear Fuel for LGS, Unit 1, Cycle 12, which includes the use of the GE-14 fuel product line.

The new SLO SLMCPR is calculated using NRC approved methodology described in General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-15 (GESTAR-II), and U. S. Supplement, NEDE-24011-P-A-15-US, September, 2005, which includes Amendment 25. Amendment 25 is being used for determining the upcoming Cycle 12 SLO SLMCPR. Future SLO SLMCPRs determined in accordance with Amendment 25 will not need prior NRC approval for each cycle unless a TS value changes. The NRC safety evaluation approving Amendment 25 is contained in a letter from the NRC to General Electric Company, dated March 11, 1999.

Global Nuclear Fuel has designed GE-14 fuel to be in compliance with Amendment 22 to General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-15 (GESTAR-II), and U. S. Supplement, NEDE-24011-P-A-15-US, September, 2005.

The SLO SLMCPR analysis establishes SLO SLMCPR values that will ensure that greater than 99.9% of all fuel rods in the core avoid transition boiling if the limit is not violated. The SLO SLMCPRs are calculated to include cycle specific parameters which include: 1) the actual core loading, 2) conservative variations of projected control blade patterns, 3) the actual bundle parameters (e.g., local peaking), and 4) the full cycle exposure range. The Cycle 11 calculated SLO SLMCPR is 1.084 and the Cycle 12 calculated SLO SLMCPR is 1.089. These results indicate that the calculated SLO SLMCPR results for both Cycles 11 and 12 differ by +0.005. This +0.005 increase in the SLO SLMCPR is attributed to the statistical variations in the Monte Carlo analysis.

The increase in SLO SLMCPR of 0.01 is attributed to rounding down the Cycle 11 result to 1.08, and rounding up the Cycle 12 result to 1-09. The new SLO SLMCPR at LGS, Unit 1, Cycle 12 is 1.09. The two-loop operation SLMCPR will not change as shown in Enclosure 4. Additional information regarding the cycle specific SLO SLMCPR for LGS, Unit 1 Cycle 12 is contained in the Enclosure 4 letter.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration Exelon has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

- License Amendment Request Docket No. 50-352 Evaluation of Proposed Change License No. NPF-39 December 14,2005 Page 3 of 5

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The derivation of the cycle specific Single Loop Operation Safety Limit Minimum Critical Power Ratio (SLO SLMCPR) for incorporation into the Technical Specifications (TS), and its use to determine cycle specific thermal limits, has been performed using the methodology discussed in General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-15 (GESTAR-II), and U. S.

Supplement, NEDE-24011-P-A-15-US, September, 2005, which includes Amendment 25. Amendment 25 was approved by the NRC in a March 11, 1999 safety evaluation report.

The basis of the SLO SLMCPR calculation is to ensure that greater than 99.9% of all fuel rods in the core avoid transition boiling if the limit is not violated. The new SLO SLMCPR preserves the existing margin to transition boiling. The GE-14 fuel is in compliance with Amendment 22 to General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-15 (GESTAR-II), and U. S. Supplement, NEDE-24011-P-A-15-US, September 2005, which provides the fuel licensing acceptance criteria. The probability of fuel damage will not be increased as a result of this change. Therefore, the proposed TS change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The SLO SLMCPR is a TS numerical value, calculated to ensure that transition boiling does not occur in 99.9% of all fuel rods in the core if the limit is not violated.

The new SLO SLMCPR is calculated using NRC approved methodology discussed in General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-15 (GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-15-US, September 2005, which includes Amendment 25. Additionally, the GE-14 fuel is in compliance with Amendment 22 to General Electric Standard Application for Reactor Fuel, NEDE-2401 1-P-A-15 (GESTAR-II), and U. S. Supplement, NEDE-24011-P-A-15-US, September, 2005, which provides the fuel licensing acceptance criteria. The SLO SLMCPR is not an accident initiator, and its revision will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

There is no significant reduction in the margin of safety previously approved by the NRC as a result of the proposed change to the SLO SLMCPR, which includes the use of GE-14 fuel. The new SLO SLMCPR is calculated using methodology

- License Amendment Request Docket No. 50-352 Evaluation of Proposed Change License No. NPF-39 December 14,2005 Page 4 of 5 discussed in "General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-15 (GESTAR-11), and U. S. Supplement, NEDE-24011-P-A-15-US, September, 2005, which includes Amendment 25. The SLO SLMCPR ensures that greater than 99.9% of all fuel rods in the core will avoid transition boiling if the limit is not violated when all uncertainties are considered, thereby preserving the fuel cladding integrity.

Therefore, the proposed TS change will not involve a significant reduction in the margin of safety previously approved by the NRC.

Based on the above, Exelon Generation Company, LLC, concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirementdcriteria Safety limits are required to be included in the Technical Specifications by 10 CFR 50.36. The SLO SLMCPR ensures sufficient conservatism in the operating SLO MCPR limit that during normal operation and during abnormal operational transients, at least 99.9% of all fuel rods in the core do not experience transition boiling considering the power distribution within the core and all uncertainties.

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation," or would change an inspection or surveillance requirement. However, the proposed amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," Paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, Paragraph (b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

a) NEDE-24011-P-A-15 (GESTAR-II), "General Electric Standard Application for Reactor Fuel", and U.S. Supplement, NEDE-2401l-P-A-15-US, September 2005, which includes Amendment 25.

b) NRC Safety Evaluation Report dated March 11, 1999 (F. Akstulewicz (NRC) to G. A.

Watford (GE), "Acceptance for Referencing of Licensing Topical Reports NEDC-32601P, Methodology and Uncertainties for Safety Limit MCPR Evaluations; NEDC-32694P, Power Distribution Uncertainties for Safety Limit MCPR Evaluation; and Amendment 25 to NEDE-24011-P-A on Cycle-Specific Safety Limit MCPR," {TAC Nos. M97490, M99069, and M97491)).

- License Amendment Request Docket No. 50-352 Evaluation of Proposed Change License No. NPF-39 December 14,2005 Page 5 of 5 c) NEDC-32601P-A, Methodology and Uncertainties for Safety Limit MCPR Evaluations.

d) NEDC-32694P-A, Power Distribution Uncertainties for Safety Limit MCPR Evaluation.

e) Letter from M. J. Mneimneh (Global Nuclear Fuel) to J. Tusar (Exelon Generation Company, LLC), dated November 9,2005.

ENCLOSURE 2 LIMERICK GENERATING STATION UNIT 1 DOCKET NO. 50-352 LICENSE NO. NPF-39 LICENSE AMENDMENT REQUEST MARKUP TECHNICAL SPECIFICATION PAGES Page 2-1 B 2-1

2 . 1 SAFETY L I M I T S THFRMAL POWER. Low Pressure or Low Flow 2 . 1 . 1 THERMAL POWER shall n o t exceed 25% o f RATED THERMAL POWER with the reactor vessel steam dome pressure l e s s t h a n 785 p s i g o r core flow l e s s t h a n 10% o f rated f l o w .

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% o f RATED THERMAL POWER and the reactor vessel steam dome pressure l e s s t h a n 785 p s i g or core f l o w l e s s t h a n 10% of rated flow, be i n a t l e a s t HOT SHUTDOWN w i t h i n 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements o f Specification 6.7.1.

THERMAL POWER. H i s h Pressure and Hiah Flow 2.1.2 The MINIMUM C R I T I C A L POWER R A T I O (MCPR) shall n o t be 1 two recirculation loop operation and shall n o t be l e s s t h a n recirculation loop operation with the reactor vessel steam ssure greater t h a n 785 psig and core flow greater t h a n 10% o f rated f l o w .

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION 9 PR l e s s t h a n 1.07'for two recirculation loop operation or l e s s t h a n r single recirculati,pn loop operation and the reactor vessel steam dome ure greater t h a n 785 psig and core flow greater t h a n 10% of rated flow, be I

l e a s t HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply w i t h the requirements o f fication 6 . 7 . 1 .

ANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, a s measured in the reactor vessel steam dome, shall n o t exceed 1325 p s i g .

APPLICABIU: OPERATIONAL CONDJTIONS 1, 2, 3, and 4.

ACT IOh. :

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be i n a t l e a s t HOT SHUTDOWN with the reactor coolant system pressure l e s s t h a n or equal t o 1325 p s i g within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply w i t h the requirements o f Specification 6.7.1.

LIMERICK - UNIT 1 2-1 Amendment No. .Z., 38, a, 34.7, 15Ej, 170

2 . 1 SAFETY LIMITS i 2.0 INTRODUCTIOtl The fuel c l a d d i n g , reactor pressure and primary system p i p i n g are the principal barriers t o the release i o a c t i v e materials t o the envi r o w . Safety Limi t s are establ ish t-otect the integrity of these barriers during normal p l a n t operation nticipated transients. The fuel cladding integrity Safety Limit i s set h a t no fuel damage i s calculated t o occur i f the l i m i t i s not violat cause fuel damage i s not dire observable, a step-back approach i s o establish a Safety Limit su the MCPR i s n o t less t h a n 1.07 for irculation l o o p operation and f o r single recirculation loop op MCPR greater t h a n 1.07 for t w recirculation l o o p operation and single recirculation loop ope represents a conservative margin t o the conditions required t o m a i n t a i n fuel cladding integrity. The fuel cladding i s one of the physical barriers which separate the radioactive materials from the environs. The integrity of t h i s cladding barrier i s related t o i t s relative freedom from perforations or cracking. A l t h o u g h some corrosion or use related cracking may occur during the l i f e of the cladding, fission product migration from t h i s source i s incre-mentally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions a n d the L i m i t i n g Safety System Settings.

While fission product migration from c l a d d i n g perforation i s just as measurable as t h a t from use related cracking, the thermally caused cladding perforations signal a threshold beyond which s t i l l greater thermal stresses may cause gross rather t h a n incremental cladding deterioration. Therefore, the fuel cladding Safety Limit i s defined w i t h a margin t o the conditions which would produce onset of transition boiling, MCPR o f 1.0. These conditions represent a signi-ficant departure from the condition intended by design for planned operation.

2 . 1 . 1 THERMAL POWER, Low Pressure or Low Flow The use of the (GEXL) correlation i s n o t v a l i d for a l l c r i t i c a l power calculations a t pressures below 785 psig or core flows less t h a n 10% o f rated f l o w . Therefore, the fuel cladding integrity Safety Limit i s established by other means. This i s done by establishing a limiting condition on core THERMAL POWER w i t h the f o l l o w i n g basis. Since the pressure drop i n the.bypass region i s essentially a l l elevation head, the core pressure drop a t low power and flows w i l l always be greater t h a n 4.5 psi. Analyses show t h a t w i t h a bundle flow o f 28 x lo3 7 b / h , bundle pressure drop i s nearly independent of bundle power a n d has a value of 3.5 psi. Thus, the bundle flow w i t h a 4.5 psi driving head will be greater t h a n 28 x 10: l b / h . F u l l scale ATLAS t e s t d a t a taken a t pressures from 1 4 . 7 psia t o 800 psia indicate t h a t the fuel assembly c r i t i -

cal power a t t h i s flow i s approximately 3.35 M W t . With the design pedking factors, t h i s corresponds t o a ThERMAL POWE9 o f more t h a n 50% of RATED THERMAL POWER. Thus, a THERMAL POWER l i m i t o f 25% of RATED THERMAL POWER for reactor pressure below 785 psig i s conservative.

LIMERICK - UNIT 1 e 2-1 Amendment No. J , 38, 444,4.27, 4-56

,- ,- 170

ENCLOSURE 3 LIMERICK GENERATING STATION UNIT 1 DOCKET NO. 50-352 LICENSE NO. NPF-39 LICENSE AMENDMENT REQUEST CAMERA READY TECHNICAL SPECIFICATION PAGES Page 2-1 B 2-1

2 . 0 SAFETY LIMITS AND-LIMITING

-~

.~ SAFETY SYSTEM SETTILGS -_

2 . 1 SAFETY LIMITS THERMAL POWER, Low P r e s s u r e o r Low Flow 2 . 1 . 1 THERMAL POWER s h a l l n o t exceed 25% o f RATED THERMAL POWER w i t h t h e r e a c t o r v e s s e l steam dome p r e s s u r e l e s s t h a n 785 p s i g o r c o r e f l o w l e s s t h a n 10% o f r a t e d f l o w .

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

W i t h THERMAL POWER e x c e e d i n g 25% o f RATED THERMAL POWER and t h e r e a c t o r v e s s e l steam dome p r e s s u r e l e s s t h a n 785 p s i g o r c o r e f l o w l e s s t h a n 10% o f r a t e d f l o w ,

be i n a t l e a s t HOT SHUTDOWN w i t h i n 2 h o u r s and comply w i t h t h e r e q u i r e m e n t s o f S p e c i f i c a t i o n 6.7.1.

THERMAL POWER, H i q h P r e s s u r e and H i s h Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) s h a l l n o t be l e s s t h a n 1.07 f o r two r e c i r c u l a t i o n l o o p o p e r a t i o n and s h a l l n o t be l e s s t h a n 1.09 f o r s i n g l e I r e c i r c u l a t i o n l o o p o p e r a t i on w i t h t h e r e a c t o r v e s s e l steam dome p r e s s u r e g r e a t e r t h a n 785 p s i g and c o r e f l o w g r e a t e r t h a n 10% o f r a t e d f l o w .

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACT I ON :

W i t h MCPR l e s s t h a n 1.07 f o r two r e c i r c u l a t i o n l o o p o p e r a t i o n o r l e s s t h a n 1.09 f o r s i n g l e r e c i r c u l a t i o n l o o p o p e r a t i o n and t h e r e a c t o r v e s s e l steam dome I p r e s s u r e g r e a t e r t h a n 785 p s i g and c o r e f l o w g r e a t e r t h a n 10% o f r a t e d f l o w , be i n a t l e a s t HOT SHUTDOWN w i t h i n 2 h o u r s and comply w i t h t h e r e q u i r e m e n t s o f S p e c i f i c a t i o n 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The r e a c t o r c o o l a n t s y s t e m p r e s s u r e , as measured i n t h e r e a c t o r v e s s e l steam dome, s h a l l n o t exceed 1325 p s i g .

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

W i t h t h e r e a c t o r c o o l a n t s y s t e m p r e s s u r e , as measured i n t h e r e a c t o r v e s s e l steam dome, above 1325 p s i g , b e i n a t l e a s t HOT SHUTDOWN w i t h t h e r e a c t o r c o o l a n t system p r e s s u r e l e s s t h a n o r e q u a l t o 1325 p s i g w i t h i n 2 h o u r s and comply w i t h t h e r e q u i rements o f S p e c i f i c a t i o n 6.7.1.

LIMERICK - UNIT 1 2-1 Amendment No. 4 , 3 , u;Z-, 4-2.7, 256,

- 9

2.1 SAFETY LIMITS BASES

2.0 INTRODUCTION

The f u e l c l a d d i n g , r e a c t o r pressure vessel and p r i m a r y system p i p i n g are the p r i n c i p a l b a r r i e r s t o t h e release o f r a d i o a c t i v e m a t e r i a l s t o the e n v i r o n s . S a f e t y L i m i t s are e s t a b l i s h e d t o p r o t e c t t h e i n t e g r i t y o f these b a r r i e r s d u r i n g normal p l a n t o p e r a t i o n s and a n t i c i p a t e d t r a n s i e n t s . The f u e l c l a d d i n g i n t e g r i t y Safety L i m i t i s s e t such t h a t no f u e l damage i s c a l c u l a t e d t o occur i f t h e l i m i t i s n o t v i o l a t e d . Because f u e l damage i s n o t d i r e c t l y observable, a step-back approach i s used t o e s t a b l i s h a S a f e t y L i m i t such t h a t t h e MCPR i s n o t l e s s than 1.07 f o r two r e c i r c u l a t i o n l o o p o p e r a t i o n and 1.09 f o r s i n g l e r e c i r c u l a t i o n l o o p o p e r a t i o n . MCPR g r e a t e r t h a n 1.07 f o r two r e c i r c u l a t i o n l o o p o p e r a t i o n and 1.09 f o r s i n g l e r e c i r c u l a t i o n l o o p o p e r a t i o n represents a c o n s e r v a t i v e margin r e l a t i v e t o t h e c o n d i t i o n s r e q u i r e d t o m a i n t a i n f u e l c l a d d i n g i n t e g r i t y . The f u e l c l a d d i n g i s one o f t h e p h y s i c a l b a r r i e r s which separate t h e r a d i o a c t i v e m a t e r i a l s from t h e e n v i r o n s . The i n t e g r i t y of t h i s c l a d d i n g b a r r i e r i s r e l a t e d t o i t s r e l a t i v e freedom f r o m p e r f o r a t i o n s o r cracking. Although some c o r r o s i o n o r use r e l a t e d c r a c k i n g may occur d u r i n g t h e l i f e o f t h e c l a d d i n g , f i s s i o n p r o d u c t m i g r a t i o n from t h i s source i s i n c r e -

m e n t a l l y cumulative and c o n t i n u o u s l y measurable. Fuel c l a d d i n g p e r f o r a t i o n s ,

however, can r e s u l t from thermal s t r e s s e s which occur f r o m r e a c t o r o p e r a t i o n s i g n i f i c a n t l y above design c o n d i t i o n s and t h e L i m i t i n g S a f e t y System S e t t i n g s .

While f i s s i o n p r o d u c t m i g r a t i o n from c l a d d i n g p e r f o r a t i o n i s j u s t a s measurable a s t h a t from use r e l a t e d c r a c k i n g , t h e t h e r m a l l y caused c l a d d i n g p e r f o r a t i o n s s i g n a l a t h r e s h o l d beyond which s t i l l g r e a t e r thermal s t r e s s e s may cause gross r a t h e r t h a n incremental c l a d d i n g d e t e r i o r a t i o n . Therefore, t h e f u e l c l a d d i n g S a f e t y L i m i t i s d e f i n e d w i t h a margin t o t h e c o n d i t i o n s which would produce onset o f t r a n s i t i o n b o i l i n g , MCPR of 1.0. These c o n d i t i o n s r e p r e s e n t a s i g n i -

f i c a n t d e p a r t u r e f r o m t h e c o n d i t i o n i n t e n d e d by d e s i g n f o r planned o p e r a t i o n .

2.1.1 THERMAL POWER. Low Pressure o r Low Flow The use o f t h e (GEXL) c o r r e l a t i o n i s n o t v a l i d f o r a l l c r i t i c a l power c a l c u l a t i o n s a t pressures below 785 p s i g o r c o r e f l o w s l e s s t h a n 10% o f r a t e d f l o w . Therefore, t h e f u e l c l a d d i n g i n t e g r i t y S a f e t y L i m i t i s e s t a b l i s h e d by o t h e r means. T h i s i s done by e s t a b l i s h i n g a l i m i t i n g c o n d i t i o n on c o r e THERMAL POWER w i t h t h e f o l l o w i n g b a s i s . Since t h e pressure drop i n t h e bypass r e g i o n i s e s s e n t i a l l y a l l e l e v a t i o n head, t h e c o r e pressure drop a t low power and f l o w s w i l l a l w a y s be g r e a t e r than 4.5 p s i . Analyses show t h a t w i t h a bundle f l o w o f 28 x l o 3 l b / h , bundle pressure drop i s n e a r l y independent o f bundle power and has a v a l u e o f 3 . 5 p s i . Thus, t h e bundle f l o w w i t h a 4.5 p s i d r i v i n g head w i l l be g r e a t e r than 28 x l o 3 l b / h . F u l l s c a l e ATLAS t e s t d a t a taken a t pressures from 14.7 p s i a t o 800 p s i a i n d i c a t e t h a t t h e f u e l assembly c r i t i -

c a l power a t t h i s f l o w i s approximately 3.35 M W t . With t h e design peaking f a c t o r s , t h i s corresponds t o a THERMAL POWER o f more than 50% o f RATED THERMAL POWER. Thus, a THERMAL POWER l i m i t o f 25% of RATED THERMAL POWER f o r r e a c t o r pressure below 785 p s i g i s c o n s e r v a t i v e .

LIMERICK - U N I T 1 B 2-1 Amendment No. 2, 38, u.l,W, 446

, 440,

ENCLOSURE 5 LIMERICK GENERATING STATION UNIT 1 Docket No. 50-352 License No. NPF-39 LICENSE AMENDMENT REQUEST NON-PROPRIETARY VERSION OF GLOBAL NUCLEAR FUEL LElTER

Attachment Additional information Regarding the November 4,2005 Cycle Specific SLMCPR for Limerick 1 Cycle I2 Proprietary Information Notice This document is the GNF non-propnetaq version of the GNF proprietary report. From the GNF pmprietary version, tho information noted as GNF proprietary (enclosed in double brackets) was deleted to generate this version.

page 1 of 9 0000-0042-2435

Attachment Additional Information Regarding the November 4,2005 Cycle Specific SLMCPR for Limerick 1 Cycie 12 References Letter, Frank Akstulewicz (NRC) to Glen A WatfoFord (GE), Acceptance for Referencing of Licensing Topical Reparts NEDC-32601P, Methodology and Uncertainties for Safety Limit MCPR Evaluations; NEDC-32694P. Power Distribution Uncertainties for Safety Limit MCPR Evaluation: and Amendment 25 to NEDE-2401 l-P-A on Cycle Spectfic Safety Limit MCPR, fTAC Nos. M9745H). M99069 and M97491), AMarch11, 1999.

Letter, Thomas H. Essig (NRC) to Glen A. W d o r d (GE), Acceptance for Referencing of Licensing Topical Report NEDC-32SOjP, Revision I, R-Factor Calculation Method for GE 11, GE 12 and GE 13 Fuel, (TAC Nos.M99070 and M95081). January t 1,1999.

General Electric BWR Tbermal Analysis Basis (GETAB): Data, Conetation and Design Application, N E m - 10958-A, J ~ 1977.

w ~

Letter, Glen A. Watford (GNF-A) to U. S.Nucfear Regulatory Commission Document Control Desk with attention to R. Pulsifkr (NRC), Confirmation of 10x10 Fuel Design Applicability to Improved SLMCPR Power Distribution and R-Factor Methodologies, FLN-2001-016, September 24,2001, Letter, Glen A. Watford (GNF-A) to U. S. Nuclear Regulatory Commission Document Control Desk with attention to J. Donoghue (NRC),Confirmation of the Applicability of the GEXL14 Correlation and Associated R-Factor Methodology for Calculating SLMCPR Values in Cores Containing GEI4 Fuel, FLN-2001-0 17, October I, 200 1 Letter, Jason S. Post (GE Energy) to U.S. Nuclear Regulatory Commission Document Control Desk, Part 2 1 Repottable Condition and 60-Day Interim Rcpon Notification: Non-conservative SLMCPR, MFN-04-08 1, August 24,2004.

Letter, Glen A. Watford (GNF-A) to U. S. Nuclear Regulatory Commission Document Control Desk with attention to J. Donoghue WRC), Final Presentation Material for GEXL Presentation - February 11.2002, FLN-2002-004, Februap 12,2002 page 2 of9 0000-0042-243 5

Attachment Additional Information Regarding the November 4,2005 Cycle Specific SLMCPR for Limerick 1 Cycle 12 Discussion The Safev Limit Minimum Critical Power Ratio (SLMCPR) evaluations for the Limerick I Cycle I2 were performed using NRC approved methodo1ogy and uncertamties Ill. Table 1 summarizes the relevant input parameters and results of Cycle 12 and Cycle 1 1 cores. Additional information is provided in response to NRC questions related to similar submittals regarding changes in Technical Specification values of SLMCPR. NRC questions pertaining to how GE14 applications satisfy the condtions of the NRC SER"'

have been addressed in Reference 141. Other generically applicable questions related to appiication of the GEXL14 cornlation. and to the applicable range for the R-factor methodology, are addressed in Fkference

[j]. Items that require a plantkycie specific response are presented below.

Previously, the SLMCPR was calculated on the upper boundary of the powedflow operating map only at 100% flow I 100% power (rated flow'rated power), which had been shown in NEDC-32601P-A to result in conservdve SLMCPR evaluation values using the same control rod pattern used for rated flow/rstted power evaluations. Recent cvaluations for BWR plants fuclcd by GNF fuci bundle designs determined that limiting control blade patterns developed for less than rated flow at rated power condition sometimes yield morc limiting bundle-by-bundle MCPR distributions and/or more limiting bundle a d power shapes than the limiting contfol blade patterns developed for a rated Bowlrated power SLMCPR evaluation, as reported in Reference 16). Thcrefore, to conservativety account for operation at lower flow/rated power conditions, SLMCPR evaluations were also performed at h e lowest core flow rate (81% rated flow) at med power condition for the same exposure points that were previously calculated for the rated flodrated power evaluations. The results for Limerick 1 Cycle 12 at the lower flow condition bounded those at the rated core flow condition.

In general, the cdcuiated safety limtt is dominated by two key parameters. (1) flatness of the core bundle-by-bundle MCPR distributions, and (2) flafness of the bundle pin-by-pin powerili-factor distributions.

Greater flatness in either parauneter yields more mods susceptible to boiling transition and thus a higher cdcufated SLMCPR. The impact of these parameters on the Limerick I Cycle 12 and Cycle f I SLMCPR valnes is summarized in Table 1.

The core loading information for Limerick 1 Cycle I 1 is provided in Figure 1. For comparison the care loading information for Limerick I Cycle 12 is provided in Figure 2. The impact of the fuel load1118pattern differences on the calculated SLMCPR is correIated to the d u e s of [ [

The uncontrolled bundle pin-by-pin power distributions were conipared between the Limerick I Cycte 12 bundles and the Cycle 1 I bundles. Pin-by-pin power distributions are charactenzed in terms of R - h t o r s using the NRC approved methodology ['I. For the Limerick 1 Cycle 12 liminng case analyzed at EOC-MELLLA? ff 13'ljthe Limerick 1 Cycte 12 bundles have a more peaked power distribution than the bundles used for the Cycle I 1 SLMCPR analysis.

page 3 of 9 0000-0042-2435

Attachment Additional Information Regarding the November 4,2005 Cycle Specific SLIIilCPR for Limerick 1 Cycle 12 The SLMCPR was calculatcd for Limerick I Cycle 12 using uncertaintics that haw been previously reviewcd and approved by the NRC. These uncertaintics are shown in Table 2a and described in Reference

[ 11. Where wananted, higher plant-cycle-specific uncertainties were used. as listed tn Table 2b.

Tablc 1 summarizes the rcicvant input parameters and results of Cyclc 1 1 at mcd flow and powcr and for Cycle 12 evaluated at the limiting condition of 81% rated flowhated power. The SLMCPR values were calculated for Limerick 1 using uncertamties that have becn previously reviewed and approved b> the NRC as listed in Table 2a and described in Reference [ I ] and where warranted, higher plant-cycle-specific uncertainties as listed in Table 2b. In addition to using a ((

' 3 t ] ] consistent with current GNF fuel operation, for the tower flow evaluations, the Core Flow Rate and Random effective TIP reading unccrtainties were ((

g3})) for the Limerick 1 low flow Cycle 12 evaluation.

These calculations use the GEXL14 correiation for GE.14 fuel. if Pt))

For single loop operations (SLO} the calculated d e t y limit MCPR for the limiting case is 1.09 as determined by spcific calculations for Limerick 1 Cycle 12 at EOC-MELLLA. The DLO and SLO SLMCPR vdues calculated for Limerick 1 Cycle i2 are shown in Table 1 Summary The calculated 1.07 SLMCPR and 1.09 SLO SLiMCPR f i x Limerick I Cycle 12 are consistent with expectations "

{?a

+'jjthese vafues are appropriate when the approved methodology and the reduced uncertainties given in NEDC-32601P-A and NEDC-32694P-A are uscd.

Based on the mhrmation and discussion presented above, it is concluded that the calculated SLMCPR of 1.07 and 1.09 for SLO are appropriate for the Limerick 1 Cycle 12 core.

page 4 of 9 0000-0042-243 5

Attachment Additional Information Regarding the November 4,2005 Cycle Specific SLMCPR for Limerick 1 Cycle 12 Prepared by : Verified by:

Technical Program Wager GtoM Nudear Fuel Americas Global Nuclear Fuel - Americas page 5 of 9 0000-0042-2435

Attachment Additional Information Regarding the November 4,2005 Cycle Specific SLMCPR for Limerick 1 Cycle 12 Table 1 Comparison of the Limerick 1 Cycle 12 and Cycle 11 SLMCPR QUANTITY, DESCRIPTION Limerick 1 Limerick 1 Cycle 11 Cycle 12 Nurnbcr of Bundfes in Corr: 764 764 Limiting Cycle Exposure Point 13600 EOC-I'viELLLA Cycle Exposure at Limiting Point 13600 I1400 (MWd/STU)

Rcload Fuet Type GE14 GE14 Latest Reload Batch Fraction, YO 34.6 36. I Latest ReIoad Avemge Batch Weight % 4.16 4.03 Enrichment Core Fuel Fraction for GE14 (%) 71.2 100.0 Core Fuel Fraction for GE13 (%) 28.8 0.0 Core Average Weight % Enrichment 4.17 4.12 Core MCPR (for limizing rod pattern) 1.34 I .34 rt '3'11 SlJ]

' [I U'lf Power distribution methodology Revised NEDC- Revised NEDC-3260 1P-A 32601P-A Power distribution uncertainty Reduced NEDC- Reduced NEDC-32694P-A 32694P-A Non-power distribution uncertainty Revised NEDC- Revised NEDC-3260 1P-A 3260 IP-A Calculated Safety Limit MCPR (DLO) 1.07 1.O?

Calculated Safety Limit MCPR (SLO) 1.08 1.09 page 6 of 9 0000-0042-243 5

Attachment Additional information Regarding the November 4,2005 Cycle Specific SLIMCPR for Limerick 1 Cycle 12 Table 2a Standard Uncertainties I i Limerick I Cvcle 11 I Limerick 1 Cvcie 12 1 DESCRIPTION I Non-power Distribution Uncertainties f Revised NEDC-32601P-A Revised NEDC-326OlP-A Core flow rate (derived from pressure drop) 1 2.5 DLO 2.5 DLO Iadividual channel flow area It 131 11 " (3) 11 1

Individual channel friction factor f 50 5.0 Fnction factor multiplier fr (31 11 [I 131 11 Rexctor pressure 1 " m f] rr I]

Feedwater temperature ri (31 If I]

if D: 11 Feedwater flow rate J I (3~)

rr i3 Power Distribution Uncertainties Reduced NEDC-32694P-A Reduced NEDC-32694P-A GEXL R-fkctor rr i3) 11 II 131 Systematic cffictive TIP reading " f3)

[I f3)

Integmcd effective TIP reading Bundle power rr 13) 11 I]

rf (.1) 11 II f3l

[r {3f 11 Effectivc total bundle powcr uncertainty I[ (3,I r[ (3)

Table 2b Exceptions to the Standard Uncertainties Used in Limerick 1 Cycle 12 Core flow rate (DLO analysis only) If 11 GEXL R-fktor I[ I1 Random effective TIP reading " (31 11 page 7 of 9 0000-0042-243.5

Attachment Additional Information Regarding the November 4,2005 Cycle Specific SLMCPR for Limerick 1 Cycle 12 Figure 1 Reference Loading Pattern - Limerick 1 Cycle 11

- m 58 56 3.5 52 50 48 46 44 42 Number Cycle Code Bundk, Name Loaded Loaded A GEI 4-PlOCNAB417-15GZ-100T-1fjo-T6-2594 160 fl 6 (;El 4-PlOCNAB414-l4GZ-100T-t 50-T6-2690 la6 11 c GE13-P9CfB417-13GZ-lOoT-l4&T8-3833 100 9 0 GE13PQCTB417-11GZ-100T-14&T53834 56 9 E G El 3-PgCTB417-13GZ-Z OOT- 146-T&3833 48 9 F GET 3-P9CTB417-11CIZ4OOT-t 46-T6-3834 16 9 G GEt 4 4 1OCNA0417-7GB.018G70-1OOT-t SO-Tt?-2529 126 10 H GE14-P10CNA3417-13GZ-1O#T-1sO-fs-2530 80 10 I GE14-P 1OCNAB417-7G8.0/8G7,0-1OOT-15046-2523 24 10 J GEI 4-P?OCNA~17-13GZ-IMIT-150-T6-2530 48 10 K GEI d-P10CNAW17-7GB.W8G7 0-80U45R-150-T52532 2 10 page 8 of 9 0000-0042-2435

Attachment Additional Information Regarding the November 4,2005 Cycle Specific SLMCPR for Limerick t Cycle 12 Figure 2 Reference Loading Pattern - Limerick 1 Cycle 12 58 56 54 52 50 48 46 44 42 40 38 36 34 32 w

28 26 24 22 M

If3 16 14 12 I0 8

6 4

2 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 46 47 49 5t 53 55 57 5!

Number Cycfe Code Bundle Name Loaded Loaded A GEI 4-PIOCNAB417-15cZ-IWT-150-T6.2594 160 11 B GEI4-PlOCNAB414-14GZ-IWT-I50-T~2690 104 11 C GEldPIOCNAB403-14GZ-12oT-150-f$-2882 84 12 a GEI 4-P1 oCNA8403-15c42-120f-1!%T6-2883 192 12 E CE14-Pt OCNAB417-7080@070-10073150-TB-2529 93 10 F GE14-PI (X;NAB417-13CZ-lW)T-1SPT62530 65 10 0 GE14-P10CNAi3417-7G8 OBG7 O-TOOT-lSI-T&2S29 15 10 H GE14-PlOCNAB417-1302-1OOT-i5O*T6-2530 48 10 I GE14-PlNNAB417-7138OI8G7 0.80U4sR-1SPT6-2532 2 10 page 9 of 9 0000-0042-2435