ML20135F096

Proposed Tech Specs 2.1 Re Safety Limits

ML20135F096
Person / Time
Site:	Limerick
Issue date:	12/06/1996
From:	PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
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ML20135F094	List: ML20135F091 ML20135F096
References
NUDOCS 9612120350
Download: ML20135F096 (7)

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ATTACHMENT 2 LIMERICK GENERATING STATION UNIT 2 -

1 DOCKET NO. 50-353 LICENSE NO. NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST NO. 96-11-2 LIST OF AFFECTED PAGES UNIT 2 j 2-1 i B 2-1 i-I 9612120350 961206 PDR ADOCK 05000353 P PDR

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," 2,0 SAFETY LIMITS AND LXMfTfNG 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 re vessel flow. steam dome pressure less than 785 psig or core flow less than 10% of rated APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

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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, High Pressure and Hiah Flow I*!\ '

2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not.b ess than . for two recirculation loop operation and shall not be less than . 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. i APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION: l, l l l.12- - ;

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With MCPR less than Yor two recirculation loop operation or less than @

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 measure"d in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL 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 w.ith the requirements of Specification 6.7.1.

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LIMERICK - UNIT 2 2-1 Amendment No.14 MAY 0 6 !!91

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• 2.1. SAFETY LIMITS i

j BASES j ,

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2.0 INTRODUCTION

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The fuel cladding, reactor pressure vessel /

I principle barriers to the release of radioacti ndprimarysystempipingar(the Safety Limits are established materials to the environs.

] to protect the i tegrity of these barriers during j normal plant operations and a iticipated transi'nts. The fuel cladding inte lrity 1 Safety Limit is set such that no fuel damage i s calculated to occur if the limit 1

' is not violated. Because fue damage is not d approach is used to establish lia Safety Limit t the MCPR is not lessirectly ch $ hat than ob j I. O operation and . or single recirculation I

Mfor twoMCPR operation. recirculation greater thanlopp(gior two recirculation loop operation ane '

1 for single recirculation loop operation represents a conservative margin relative j

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 '

environs. 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 thermal stresses which occur from i

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 stresses may perforations'csse gross signal rather a than threshold beyond inciemental which deterioration?

clTdding still greater theragl",

Therefore, the fuel cladding Safety Limit is defined with a margin to the i conditions which would produce onset of transition boiling, MCPR of 1.0. These  ;

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' conditions represent a significant departure from theJooditiottintended by design

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$\ (c $ttf operation. W CW 7bs, MCPR&vtuhe5 fr kri durf~6re l

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i lh(Cd mt , m u f& ptQ /,p M WC cle 5 M

QT;1.1 RMAL POWER. Low Pressure or low Flow

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f 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 P0WER with the following basis. Since the pressure drop in'the bypass region is

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• essentially all elevation head, the core pressure drop at low power and flows will alyays 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. 3Thus, the bundle flow with a 4.5 psi driving head will be 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.

4 LIMERICK - UNIT 2 B 2-1 Amendment No. 14 M4Y 0 81991 4

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ATTACHMENT 3 LIMERICK GENERATING STATION UNIT 2  ;

i DOCKET NO. 50-353  :

i LICENSE NO. NPF-85  !

TECHNICAL SPECIFICATIONS CHANGE REQUEST NO. 96-11-2  !

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Letter, R. M. Butrovich (GE) to H. J. Diamond (PECO Energy),

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" Limerick Unit 2 Cycle 5 Safety Umit  :

MCPR ," dated November 11,1996 i

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, GE NuciscrEnergy

%chste M Butrovoch -- - -

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z.:e w - t _

November 11,1996 '#~

RMB:96-243 l

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l Mr. H. J. Diamond, Director '

Fuel & Services Drvision PECO NUCLEAR 965 Chesterbrook Boulevard Wayne, PA 19087-5691

SUBJECT:

Ilmerick Unit 2 Cycle 5 Safety Limit MCPR ,

REFERENCE:

1. Letter, R. M. Butrovich to H. J. Diamond, " Generic GE11 Safety Lima MCPR Calculation", April 2,1996

2. NEDC-32505P, R-Factor Calculation Methodfor GEI1, GE12 and GE13 Fuel. November 1995.

3. Licensing Topical Report, GeneralElectnc BWR Thermal Analysis Basis (GETAB): Da:a, Correlation andDesign Application, NEDO-10958-A, January 1977.

4. GeneralElectric Standard Applicationfor Reactor Fuel, NEDE-24011-P-A-13, and US Supplement, NEDE-24011-P-A-13-US, August 1996.

5. Letter, A. M. Olson to R. M. Butrovich, " LGS Unit 2 Reload 4 SLMCPR Determination", October 30,1996

Dear Hugh:

Reference 1 advised PECO Nuclear ofdiscoveries related to the methodology used by GE to calculate SLMCPR that indicated the generic SLMCPR may not ahvays yield the most conservative resuk. GE has performed a plant unique evaluation for Limerick Unit 2 Cycle 5.

A separate cycle specific calculation was performed for Single Loop Operation. The t

SLMCPR value obtained was 1.12. These calculations were based upon (" General Electric Standard Application for Reactor Fuel," NEDE-24011-P A-13, and US Supplement, NEDE-

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Mr. H. J. Diamsnd 2 11/11/96

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l 24011-P-A-13-US, August 1996) and the Technical Design Procedure ('TiETAB Safety l Limit", TDP-0049, Revision 0, July 1996). Revision 13 of the aforementioned document, I "GESTAR IT', requires that the SLMCPR be reconfirmed each cycle. His reconfirmation was performed using technical design procedure which replaced the interim implementing procedures which the NRC staffdiscussed with GENE during their meetings on April 17,1996 and May 6,1996 through May 10,1996. He technical design procedure includes 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. '

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he following is clarification of " conservative variations ofprojected control blade patterns" i I

used to calculate the SLMCPR for Limerick Unit 2 Cycle 5.

Projected control blade patterns for the rodded burn through the cycle are used to deplete the '

I core to the cycle exposures to be analyzed. At the desired cycle exposures, the bundle exposure distributions and their associated R-factors, determined in accordance with Reference 2, are then utihzed for the SLMCPR cases to be analyzed. Using different rod pattems, to achieve the desired cycle exposure, has been shown to have a negligible impact on the actual SLMCPR calculated. An estimated SLMCPR is obtained for an exposure point near beginning of cycle (BOC) and near end of cycle (EOC), as well as the calculated cycle exposure for peak hot excess reactivity (PHE), in order to establish which exposure point (s) will produce the highest (most conservative) calculated SLMCPR. At these exposure points, the goal of page IV-7 of Reference 3 is applied:

"%e objective in establishing the initial condition power distribution is to satisfy  ;

total power and local limits and to maximize the calculated number of rods  !

expected to experience boiling transition."

His is achieved by following the rule on page IV-6 in Reference 3.

"For a given reactor, at a particular exposure, there is a variety of rou patterns which produce k, = 1.0 [wnhin the established tolerances] and satisfy local power and MCPR constraints. For conservatism, the statistical analyses of the core are performed for only those operating states yielding MCPR equal to the limit, unless this involves an unreasonable power distribution or gross violation ofkW/ft limits."

To maximize the calculated SLMCPR value at the egosure point ofinterest, different control rod patterns are evaluated to find the patterns that yield the most bundles on or near the operating limit MCPR at this cycle exposure point. His is what is meant by " conservative variations ofprojectedcontrol blade patterns". %e variations are conservative since they are selected to maximize the number ofrods susceptible to boiling transition by mamminng the

- number of contributing bundles. The highest SLMCPR, from these patterns that met the

{ criteria above, was chosen as the cycle-specific SLMCPR value. His occurred at an exposure

point near end ofcycle.

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F Mr. H. J. Diamond 3 1 til1/96 Limerick Unit 2 Cycle 5 SLMCPR is 1.11 versus the generic GE13 value of1.09. He Single Loop Operation SLMCPR is 1.12 as determined by a separate cycle specific calculation.

GESTAR 11 requires that the SLMCPR analysis for a new fuel design shall be performed for a bounding equilibrium core. This is how the 1.09 value was obtained for the GE13 product line. However, Limerick Unit 2 Cycle 5 is not an equilibrium core, it is a mixed core of GE13/ gel 1/GE6 fuel. Over the last several cycles, the average enrichment for the fresh fuel has progressively increased. Higher enrichment in the fresh fuel (compared to the rest of the core) produces higher power in the fresh bundles which causes more rods to be suscepulle to boiling transition in the statistical analysis than for an equihirium core. He fresh Cycle 5 GE13 bundles have a flatter RFACTR distribution than the GE13 bundles analyzed in the generic case. In cores that operate for two years it is typical that almost all ofthe bundles close to the limiting MCPR ofthe core are fresh fuel. Also, Limerick Unit 2 Cycle 5 is loaded aggressively with a high batch fraction to achieve a two year cycle. With this loading it is easier !

to put more bundles in this core on MCPR limits than is generally possible for assumed  !

equihtrium core designs. As requested by PECO Nuclear (Reference 5), the SLMCPR calculations for Limerick Unit 2 Cycle 5 used a feedwater flow uncertainty of 2.24% rather than the generic value of 1.76% (Reference 4). For the above reasons the Limerick Unit 2 '

l Cycle 5 SLMCPR is higher than the generic GE13 1.09 value.

Ifyou have any questions, please give me a call.

Very truly yours, l

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/? ;pW R. . Butrovich F el Project Manager j l

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