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Proposed TS Section 2.1, Safety Limits, Revising MCPR Safety Limit.Nonproprietary Supporting Info Encl
	ML20202G787
Person / Time
Site:	Limerick
Issue date:	02/09/1998
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
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ML19317C910	List: ML20202G759; ML20202G787;
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ATTACHMENT 2 LIMERICK GENERATING STATION

UNIT 1 i

lxD:ET NO. 50-352 4 LICENSE NO. NPF 39 l

l TECHNICAL SPECIFICATIONS CHANGE REQUEST

! NO. 97-03-1 t

i LIST OF AFFECTED PAGES UNIT 1 P

2-1 B 2-1 i

5 9802200221 980209 PDR ADOCK 05000352 p PDR ._

. 2.0 SAFETY l!MITS AND t!MITING S3FETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER'. Low Pressere or Bw Flow 2.1.1 THERMAL POWER shall .;ot exceed 25% of RATED THERMAL POWER with the esactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 Rd 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less thac. 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the rcquirements of !

Specification 6.7.1.

THERMAL POWER. Hioh Pressure and Hich Flow l.12 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than for two recirculation loop operation and shall not be less than '.E for single recirculation loop operation with the reactor vessel steam domeiressuregreater than 785 psig and core flow greater than 10% of rated flow. '

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. I*N ACTION:

1. I2.

G '

With MCPR less than .

for two recirculation loop operation or less than '

i

. 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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 steam dome, shall not exceed 1325 psig.

APPLICABILITV 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 the reactor coolant i

system pressure less than or eoual to 1325 psig within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.

FEB 2 01996 LIMERICK - UNIT 1 2-1 Amenenent No. 7,30,111

2.1 SAFETY LIMITS BASFS 2.0 NTRODUCT[0N

The fuel cladding, reactor pressure vesse and primary system piping are the principal barriers to t a release of r ioactive materials to the environs. Safety Limits are e tablished to tact the integrity of these barriers during normal plant perati .is l*g cladding integrity Safety Li t is set su that no fuel damage anticipated transients. The fuel is calculated to occur if the limit is not violated. ause fuel damage is not directly observable, a step-back app ach is us the MCPR is not less than for t to establish a Safety Limit such that for single recirculation loop operat' n. recirculation loop oparation and MCPR greater than +:49 der two recirculation loop operation and bd 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 ph which separate the radioactive materials from the environs. ysical barriers ;*q this claddinq cracking. A barrier is related to its relative freedom from perforations orThe integrity of the life of the thouqh some corrosion or use related cracking may occur during c' adding, fission product migration from this source is incro-mentally cumulative and continuously measurable.

however, can result from thermal stresses which occur from reactor operationFuel cla <

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 themally caused cladding perforations signalthan rather a threshold incrementalbeyond cladding which still greater thermal stresses any cause gross deterioration.

Safety Limit is defined with a margin to the ;onditions which would produceTherefore, t onset of transition boiling, MCPR of 1.0.

ficant departure from the condition intended by design for planned operation.These 2.1.1 THERMAL POWER. Low Pressure or low Flow The use of the (GEXL) correlation is not valid for all critical power calculations flow. at pressures below 785 psig or core flows less than 101 of rated Therefore, other means. the fuel cladding integrity Safety Limit is established by POWER with the followtog basis.This is done by establishing a limiting condition on core TH is essentially all elevation head, the core pressure drop at low power andSince flows will always be greater than 4.5 psi. Analyses snow that with a bundle flow of 28 x 10' lb/h, bundle presstire drop is nearly independent of bundle power and has a value of 3.5 psi.

head will be grater than 28 x 10' lb/h.Thus, the bundle flow with a 4.5 psi driving Full scale ATLAS test data taken at pressures from 14.7 psia to_800 psia indicate that the fuel assembly criti-cal power at this flow is approximately 3.35 MWt.

With the design peaking factors, POWER.

this corresponds to a THERMAL POWER of more than 505 of RATED THERMAL Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

w The MCPR values for both dual-loop and single loop operation listed above, are valid only for Cycle 8 operation.

-m ~

FB20196 LIMERICK - UNIT 1 B 2-1 Amendment No. 7,39.111

B

.- ATTACHMENT 4 LIMERICK GENERATING STATION

. UNIT 1 DOCKET NO. 50-352 LICENSE NO. NPF 39 -

TECHNICAL SPECIFICATIONS CHANGE REQUEST NO. 97-03-1 i

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

l.!merick Unit 1 Cycle 8 Safety Limit MCPR" dated Jar.uary 20,1998 1

(NON-PROPRIETARY VERSION) i i

NON PROPRIETARY Attachment Additional Information Regarding the 1.12 January 20,1998 Cycle Specific SLMCPR for Limerick Unit 1 Cycle 8 References I

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

( 2) GeneralElectric Standard Applicationfor Reactor Fuel (GESTAR11), NEDE 240l1 P A ll, November 1995.

( 3} GeneralElectric StandardApplicationfor Reactor Fuel (GESTAR11), NEDE 240l1 P.A 13, August 1996.

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

( 5) Methodology and Uncertaintiesfor Safety Limit MCPR Evaluations, NEDC 3260lP, Class 1Il, December 1996.

( 6) R Factor Calculation Methodfor GEi1. GE13 andGE13 Fuel, NEDC-32505P Revision 1, June 1997.

Control Rod Pattern Development for the Limerick Unit 1 Cycle 8 SLMCPR Analysis i

Projec.:d 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 arder 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 transitio, be maximized.

GENE has established criteria to determine if the control rod patterna and resulting radial power distributions are acceptable based on importance parameter.= described later Different rod patterns 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 i possible MCPR distributions. This was done for 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 l.1 GWd/STU exposure point to establish the maximum SLMCPR for the cycle.

a

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t NON PROPRIETARY Attachment Additional Information Regarding the 1.12 la~n~uary 20,1998 Cycle Specific SLMCPR for Limerick Unit 1 Cycle 8 Comparison of the Limerick Unit 1 Cycle 8 SLMCPR to the Generic GE13 SLMCPR Value Table I sumrciarizes the relevant input parameters and results of the SLMCPR evaluation for both the generic GE13 and the Limerick Unit 1 Cycle 8 core. The generic evaluation and the plant / cycle specific evaluations all were performed using the methods described in GETAB I '3.

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 ir. 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 111I 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 GE13 product line generic SLMCPR value was determined according to this specification and found to be 1.09. Later revisions to GESTAR lii 3I that have been submitted to the NRC describe how 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].

The Limerick Unit 1 Cycle 8 core is a mixed core with Gell, GE13 and ex-Shoreham GE6 fuel.

The latest reload consists of GE!3 fuel making up ([)) cf the total bundles in the ore. 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 oad core average enrichments of(()). Higher enrichment in the fresh gel 3 fuel for the Limesick Unit I Cycle 8 core (compared to the average of the core) produces slightly higher power in the fresh bundles relative to the rest of the core. (())

(())

The core MCPR distribution for the Limerick Unit I Cycle 8 analysis is by all measures much flatter than the MCPR distribution assumed for the generic GE13 evaluation. (())

([))

(()) From this comparison (())it can be concluded that the core MCPR distribution for Limerick Unit 1 Cycle 8 is flatter overall than the MCPR distribution evaluated generically for GE13 and that based on this reason alone the calculated SLMCPR for Limerick Unit 1 Cycle 8 should be higher than the 1.09 generic GE13 SLMCPR.

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NON PROPRIETARY l .

~ Attachment AdditionalInformation Regarding the 1.12 January 20,1998 a

- Cycle Specific SLMCPR for Limerick Unit 1 Cycle 8

The uncontrolled bundle pin by pin power distributions were compared between the Limerick Unit I Cycle 8 bundles and the bundles used for the generic gel 3 evaluation. The pin by pin power distributions for the bundles used in the the Limerick Umt I Cycle 8 SLMCPR evaluation were also compared. Pin by pin power distributions are characterized in terms of R factors using the methodology defined in Reference [ 6). (())

The flatness of the pin R factor distribution within a particular bundle is characterized (()]

l ((l]

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NON PROPRIETARY

Attachment Additionallnformation Regarding the 1.12 January 20,1998 Cycle Specific SLMCPR for Limerick Unit 1 Cycle 8 Table 1 Comparison of Generic GE13 and Limerick Unit 1 Cycle 8 Core and Bundle Quantities that impact the SLMCPR ((j]

Summary The calculated nominal 1.12 Monte Carlo SLMCPR for Limerick Unit I Cycle 8 is consistent with what one would expect ((]) the 1.12 SLMCPR value is appropriate.

Various quantities ((]) have been used over the last year to compare quantities that impact the calculated 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 1 support the conclusion that the Limerick Unit 1 Cycle 8 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 1 Cycle 8 bundles are less flat than the hundles used for the generic GE13 SLMCPR evaluation.

Based on all of the facts, observations and arguments presented aoove, it is concluded that the calculated SLMCPR value of 1.12 for the Lirnerick Unit i Cycle 8 core is appropriate. Itis reasonable that this value is higher than the generic GE13 SLMCPR evaluation.

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

Prepared by: )

Verified by:

A. V. Austin B.R. Fischer Technical Program Manager Technical Program Manager Nuclear Fuel Engineering Nuclear Fuel Engineering

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ML20202G787

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