ML20207L659

Proposed Tech Specs Section 2.1, Safety Limits, Revising MCPR Safety Limit

ML20207L659
Person / Time
Site:	Limerick
Issue date:	03/11/1999
From:	PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
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ML20137A820	List: ML20207L656 ML20207L659
References
NUDOCS 9903180214
Download: ML20207L659 (14)

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ATTACHMENT 2 LIMERICK GENERATING STATION UNIT 2 DOCKET NO. 50-353 LICENSE NO. NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST NO. 98-07-2 LIST OF AFFECTED PAGES UNIT 2 2-1 B2-1 P

P a '"88 M 3 88sssa PDR ,

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER. Low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

i ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least 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 with the requirements of Specification 6.7.1.

THERMAL POWER. Hiah Pressure and Hiah Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.12 for two recirculation loop operation and shall not be less than 1.14 for single

-recirculation loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With MCPR less than 1.12 for two recirculation loop operation or less than 1.14 j for single recirculation loop operation and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow, be in at least 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 with the requirements of Specification 6.7.1..

REACTOR COOLANT SYSTEM PRESSURE .

2.1.3.The reactor coolant system pressure, as measured in the reactor vessel I l

steam dome, shall not exceed 1325 psig. 1 APPLICABILITY OPERATION CONDITIONS 1, 2, 3, and 4.

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 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 of Specification 6.7.1.

LIMERICK - UNIT 2 2-1 1

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2.1 SAFETY LIMITS BASES 2.0 INTRODUCTIGd i

The fuel cladding, reactor pressure vessel and primary system piping are the principle barriers to the release of radioactive materials to the environs.

Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.12 for two recirculation loop operation and 1.14 for single recirculation loop operation.. MCPR greater than 1.12 for two recirculation loop operation and 1.14 for single recirculation loop operation represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the enviroris. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thennal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.

Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation.

listed above, are valid only Thefor MCPR Cycle values for both dual-loop and single loop operation 6 operation.

2.1.1 THERMAL POWER. Low Pressure or L e Flgg The use of the (GEXL) correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is ,

essentially all elevation head, the core pressure drop at low power and flows will i always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 10' lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be i greater than 28 x 10' lb/hr. Full scale ATLAS test data taken at pressures  !

from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this -)

flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

LIMERICK - UNIT 2 B 2-1 I

2.0 SAFETY LIMITS AND LIMITIfgG SAFFTY SYSTEM 5FTTINGS 2.1 SAFETY LIMITS THERNAL poWEB Law Pressure or I a= Flow 2.1.1 THERMAL POWER shall not saceed 25K of RATED THERMAL POWER with the reactor vessel flow. steam done pressure less than 785 psig or core flow less than 10E of rated APPLICARILITY; OPERATIONAL ColelTIONS I and 2.

ACIlmh With THEmiAL POWER exceeding 255 of RATED THDgmL POWER and the reactor vessel steam dome pressure less'than 785 psig or core flow less than 105 of rated flow, he in at least HOT SNUTDOWN within 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 requiremente of Specification 6.7.1. #

/ l'/

THERMAL POWER. Hiah Pressure and Hioh Flow /./e2 2.1.? The MINIMlm CRITICAL POWER RATIO (MCPR) shall not be oss than or two recirculation loop operation and shall not be less than 4-for single recirculation loop operation with the reactor vessel steam done pressure greater than 785 psig and core flow areater than IDE of rated flow.

APPLICABILITYt DPERATIOliAL CONo1TIONS I and 2.

ACTION? //.2 ' jd t!1th MCPR less than .

-for two recirculation loop operation or less than . f- l for single recirculation loop operation and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 105 of rated flow, be in at least 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 with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PREstuar 2.1.3 The reactor coolant systes pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

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

ACTION-With the reactor coolant system pressure, as measured in the reactor vessel steam dime, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 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 requitements of Specification 6.7.1.

LIMERICK - UNIT 2 2-1 h h t No. 18,53, 87 FEB 2 01997 o

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2.1 SAFETY LIMITS BASES i

Le INTEDDUCTItNI

' principle barriers to the release of radteactive anterials to the envi safety normal Limits are established plant operations to protect and anticipated the integrity of these barriers during transients.

The fusi cladding integrity  !

Safety is not violated. Limit is set such that ne fuel damage is calculated to occur if the limit a /,1Q approach is used to establish a safety Limit such that ss the /, M EpR is est leB than -

i operation. for tus recirculaties loop operation and h4pror stasie racirenlaj en loop l i

EpR greater than h4Ffor tus recFrculation loop operation and ,

/* /d  !

to the conditions rugstred te maintain feel classiing integrity.for single r is one of the physical barriers which separate the radioactive materials from theThe fuel clad environs.

freedom from perforattens or cracking.The integrity of this cladding barrier is related Although same corrosion er use related D to i i

. cracking may occur during the life of the cladding, fission product migration free j this source is incrementally cumulative and continuously measurable. Fuel i

cladding perforations, however, can result from thermal stresses which occur fres reactorSettings.

system operation significantly above design conditions and the Limiting Safety

While fission product migration from cladding perforation is

• just as measurable as that from use related cracking, the thermally cacsed j clading perforations signal a thresh ld beyond which still greater thermal stresses may cause gross rather thac ircramental cladding deterioration.

i Therefore, the fuel cladding Safety Limit is defined with a margin to the

.onditions which would produce onset of transition boiling, E PR of 1.0. These conditions for planned represent operation. a significant departure from the condition intended by design listed above, are valid only for CycleThe MCpR values for both dual-loop and single loop op operation.

l j L1.1 THERMAL POWER. Low pressure or f ow Flow I The use of the (EEXL) correlation is not valid for all critical power calculations flow. at pressures below 785 psig or core flows less than 101 of rated means. Therefore, the fuel cladding integrity Safety Limit is established by other  !

This with the following basis. is done by establishing a Itaiting condition on core THutMAL POWER Since the pressure drop in the bypass region is assentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flew of 28 x 10' a value lb/hr,ofbundle 3.5 ps1.pressure drop is nearly independent of bundle power and has greater than 28 x 10*Thus, lb/hr. theFull bundle flew with a 4.5 psi driving head will be scale ATLAS test data taken at pressures from 14.7 flow is approximately 3.35 MWt.psia to 800 psia indicate that the fuel assembly critical power at this to a THERMAL POWER With the design peaking factors, this corresponds of more than 50% of RATED THERMAL POWER. Thus, a 111LRMAL POWER liett of 251 of RATED THERMAL POWER for reactor pressure below 735 psig is censervative.

LIMERICK - UNIT 2 8 2-1 !_- - c No. 18, 33, 87 FEB 2 0 G7 a

AlTACHMENT 3 LIMERICK GENERATING STATION UNIT 2 DOCKET NO. 50-353 LICENSE NO. NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST NO. 98-07-2 Letter: R.M. Butrovich (GENE) to K.W. Hunt (PECO Energy)

" Limerick Unit 2 Cycle 6 Safety Limit MCPR" dated February 4,1999 (PROPRIETARY VERSION) o

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l GE Nuclear Energy i

i February 4,1999 cc: J. M. Carmody LIM RMB:99-030 J. A. Baumgartner S. B. Shelton l G. A. Watford i Mr. K. W. Hunt, Director Fuel & Services Division PECO NUCLEAR 965 Chesterbrook Boulevard Wayne, PA 19087-5691

SUBJECT:

Limerick l' nit 2 Cycle 6 Safety Limit MCPR

REFERENCE:

1. Attachment dated February 4,1999,

Subject:

AdditionalInformation Regarding the 1.12 Cycle Specific SIMCPRfor Limerick Unit 2 Cycle 6.

2. " Contract between Philadelphia Electric Company and General Electric I Company for Reload Fuel Supply for Limerick Generating Station Units 1 and 2", June 1,1994, as amended.

Dear Ken:

Attached for your information and use is reference 1 regarding the Limerick Unit 2 Cycle 6 l

cycle specific SLMCPR. The dualloop Cycle 6 SLMCPR is 1.12. A separate cycle specific  !

calculation was performed for Single Loop Operation. The single loop SLMCPR value obtained is 1.14.

Please note that reference I contains GE Company Proprietary Information contained within the double brackets as indicated in the attachment and should be handled in accordance with the proprietary information provisions contained in the reference 2 Contract between PECO Energy Company and GE.

RECEIVED Very truly yours, FEB 0 51999 N/Alag Kenneth W. Hunt Vendor: A' E l , / ,G,l Fueland Services Division ,/ '

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R 1. Butrovich Attachment

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['T Forward To:

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GE Nuclear Energy GeneralEJoctric Company P. o. Box 780. Wiltrington. Ne 28402 Affidavit I, Glen A. Watford, being duly sworn, depose and state as follows:

(1) I am Manager, Nuclear Fuel Engineering, General Electric Company ("GE") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in the letter, G. D. Edwards (PECO Energy i Company) to the U. S. Nuclear Regulatory Commission Document Control Desk, Limerick Generating Station, Unit 2 License Change Application ECR 99-00275, Docket No. 50-353, License No. NPF-85.

(3) In making this application for withholding of proprietary information of which it is the owner, GE ,

relies upon the exemption from disclosure set forth in the Freedom ofInformation Act ("FOIA"),  !

5 USC Sec. 552(b)(4), and the Trade Secrets Act,18 USC Sec.1905, and NRC regulations 10 CFR 9.17(a)(4) and 2.790(a)(4) for " trade secrets and commercial or financial information ,

obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all " confidential commercial information," and some ,

portions also qualify under the narrower definition of" trade secret," within the meanings assigned 1 to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Eneruv Proiect1 Nuclear Reculatorv Commission. 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA. 704F2dl280 (DC Cir.1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of General Electric, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential conunercial value to General Electric;

e. Information which discloses patentable subject matter for which it may be desirable to obtaiu patent protection.

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Affidavit l

De information sought to be withheld is considered to be proprietary for the reasons set l

forth in both paragraphs (4)a. and (4)b., above.

(5) He information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GE, and is in fact so held. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in (6) and (7) following. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GE, no public disclosure has been nude, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.

(7) He procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for 1 technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements. ,

l (8) De information identified in paragraph (2) is classified as proprietary because it contains details of GE's Safety Limit MCPR analysis and the corresponding results which GE has applied to this specific plant and cycle's actual core design with GE's fuel.

The developmer i se metiMs used in these analysis, along with the testing, development and approval of the supporting critical power correlation was achieved at a significant cost, on the order of several million dollars, to GE.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities.

He stability analysis is part of GE's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

He research, development, engineering, analytical, and NRC review costs comprise a substantial imestment of time and money by GE.

The precise value of the expertise to devise an evaluation process and apply the correct analpical i

methodology is difficult to quantify, but it clearly is substantial.

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GE's competitive advantage will be lost ifits competitors are able to use the results of the GE i experience to normalize or verify their own process or if they are able to claim an equivalent  :

understanding by demonstrating that they can arrive at the same or similar conclusions. t The value of this information to GE would be lost if the information were disclosed to the public.

, Making such information available to competitors without their having been required to undertake l l a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive  !

GE of the opportunity to exercise its competitive advantage to seek an adequate return on its large '

investment in developing these very valuable analytical tools.

l State of North Carolina ' ) gg, County of New Hanover )

Glen A. Watford, being duly sworn, deposes and says:

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That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his l

knowledge, information, and belief. i l

g l Executed at Wilmington, North Carolina, this 4 day of Nb/t49 ,19 ff  !

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Glen A. \ atford General Electric Company Subscribed and sworn before me this 4/M day of flt4m2a .1999 '

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Notary Public, State of North Carolina My Commission Expires S- 3 M -- Rod /

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ATTACHMENT 4 LIMERICK GENERATING STATION UNIT 2 DOCKET NO. 50-353 3 LICENSE NO. NPF-85 1

TECHNICAL SPECIFICATIONS CHANGE REQUEST NO. 98-07-2 i

i Letter: R.M. Butrovich (GENE) to K.W. Hunt (PECO Energy)

" Limerick Unit 2 Cycle 6 Safety Limit MCPR" dated February 4,1999 (NON PROPRIETARY VERSION)

O

NON-PROPRIETARY Attachment Additional Information Regarding the 1.12 February 4,1999 Cycle Specific SLMCPR for Limerick Unit 2 Cycle 6 References

( l} General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, NEDO-10958-A, January 1977.

[ 2} General Electric Standard Applicationfor Reactor Fuel (GESTAR 11), N EDE-240 l 1 -P-A- 11, November 1995.

( 3} General Electric Standard Applicationfor Reactor Fuel (GESTAR 11), NEDE-24011 P-A-13, August 1996.

[ 4] GeneralElectric FuelBundle Designs, NEDE-31152-P, Revision 6, April 1997.

[ 5} Sfethodology and Uncertaintiesfor Safety Limit AfCPR Evaluations, NEDC-32601 P, Class 1ll, December 1996.

[ 6} R-Factor Calculation Afethodfor gel 1, GE12 and GE13 Fuel, NEDC-32505P Revision I, June 1997.

Control Rod Pattern Development for the Limerick Unit 2 Cycle 6 SLMCPR Analysis Projected control blade patterns for the rodded burn through the cycle were used to deplete the core to the cycle exposures to be analyzed. At the desired cycle exposures the bundle exposure distributions and their associated R-factors were utilized for the SLMCPR cases to be analyzed. The use of different rod patterns to achieve the desired cycle exposure has been shown to have a negligible impact on the actual calculated SLMCPR. An estimated SLMCPR was obtained for an exposure point near beginning of cycle (BOC), middle of cycle (MOC), and the end of cycle (EOC) in order to establish which exposure points would produce the highest (most conservative) calculated SLMCPR.

The Safety Limit MCPR is analyzed with radial power distributions that maximize the number of bundles at or near the Operating Limit MCPR during rated power operation. This approach satisfies )

the stipulation in Reference I that the number of rods susceptible to boiling transition be maximized. l GENE has established criteria to determine if the control rod patterns and resulting radial power distributions are acceptable based on importance parameters described later. Different rod patterns l were analyzed until the criteria on the above parameter was satisfied. The rod pattern search was narrowed by starting from a defined set of patterns known from prior experience to yield the flattest possible MCPR distributions. This was done f3r the two most limiting exposure points in the cycle since the BOC point was excluded by criteria as non-limiting based on the value from the estimation procedure. A Monte Carlo analysis was then performed for the MOC peak hot excess point and the EOC-1.0 GWd/STU exposure point to establish the maximum SLMCPR for the cycle.

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NON-PROPRIETARY l 4

Attachment AdditionalInformation Regarding the 1.12 February 4,1999 Cycle Specific SLMCPR for Limerick Unit 2 Cycle 6  !

Comparison of the Limerick Unit 2 Cycle 6 SLMCPR to the Generic GE13 SLMCPR Value Table I summarizes the relevant input parameters and results of the SLMCPR evaluation for both the generic GE13 and the Limerick Unit 2 Cycle 6 core. The generic evaluation and the plant / cycle specific evaluations all were performed using the methods described in GETABl ll. The evaluations yield different calculated SLMCPR values because the inputs that are used are different. The quantities that have been shown to have some impact on the determination of the safety limit MCPR (SLMCPR) are provided. Much of this information is redundant but is provided in this case because it has been provided previously to the NRC to assist them in understanding the differences between

, plant / cycle specific SLMCPR evaluations and the generic values calculated previously for each fuel product line. (())

Prior to 1996, GESTAR 11[ 2] stipulated that the SLMCPR analysis for a new fuel design be performed for a large high power density plant assuming a bounding equilibrium core. The gel 3 product line generic SLMCPR value was determined according to this specification and found to be 1.09. Later revisions to GESTAR 11[ 3] that have been submitted to the NRC describe how I plant / cycle specific SLMCPR analyses are used to confirm the calculated SLMCPR value on a plant / cycle specific basis using the uncertainties defined in Reference [ 4]. l l

The Limerick Unit 2 Cycle 6 core is a mixed core with Gell, GE13 and ex-Shoreham GE6 fuel.

The latest reload consists of GE13 fuel. gel 3 fuel also makes up (()) of the total bundles in the core. l The fresh GE13 fuel has an average bundle enrichment of (()), as compared to a core average enrichment of(()). By way of comparison, the generic gel 3 equilibrium core has batch and core average enrichments of(()). liigher enrichment in the fresh gel 3 fuel for the Limerick Unit 2 Cycle 6 core (compared to the average of the core) produces slightly higher power in the fresh bundles relative to the rest of the core. ((.))

(())

((1]

[0]

i The core MCPR distribution for the Limerick Unit 2 Cycle 6 analysis is by all measures much flatter than the MCPR distribution i. 'umed for the generic GE13 evaluation. (())

((])

s

(()) From this comparison (()) it can be concluded that the core MCPR distribution for Limerick Unit 2 Cycle 6 is flatter overall than the MCPR distribution evaluated generically for GE13 and that based on this reason alone the calculated SLMCPR for Limerick Unit 2 Cycle 6 should be higher than the 1.09 generic GE13 SLMCPR.

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NON-PROPRIETARY Attachment AdditionalInformation Regarding the 1.12 February 4,1999 Cycle Specific SLMCPR for Limerick Unit 2 Cycle 6 l The uncontrolled bundle pia-by-pin power distributions were compared between the Limerick Unit 2 Cycle 6 bundles and the bundles used for the generic GE13 evaluation. Pin-by-pin power distributions are characterized in terms of R-factors using the methodology defined in Reference [ 6].

((1]

)

((1]

]

(())

Table 1 Comparison of Generic GE13 and Limerick Unit 2 Cycle 6 Core and Bundle Quantities that Impact the SLMCPR

(())

l Summary The calculated nominal 1.12 Monte Carlo SLMCPR for Limerick Unit 2 Cycle 6 is consistent with ,

what one would expect (()) the 1.12 SLMCPR value is appropriate. I Various quantities (()) have been used over the last year to compare quanti. ties that impact the calcuated SLMCPR value. These other quantities have been provided to the NRC previously for other plant / cycle specific analyses using a format such as that given in Table 1. These other quantities have also been compared for this core / cycle (()) The key parameters in Table I support the conclusion that the Limerick Unit 2 Cycle 6 core / cycle has a much flatter radial power distribution than was used to perform the GE13 generic SLMCPR evaluation. This fact is significant enough to more than compensate for the fact that the Limerick Unit 2 Cycle 6 bundles are less flat than the bundles used for the generic GE13 SLMCPR evaluation.

Based on all of the facts, observations and arguments presented above, it is concluded that the I calculated SLMCPR value of L12 for the Limerick Unit 2 Cycle 6 core is appropriate. Itis I reasonable that this value is higher than the generic GE13 SLMCPR evaluation.

i For single loop operations (SLO) the safety limit MCPR is 0.02 greater than the two-loop value. (())

Prepared by: Verified by:

NM l S. B. Shelton G. M. Baka l Technical Program Manager Technical Program Manager

Nuclear Fuel Engineering Nuclear Fuel Engineering I

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