ML20070L689

Proposed Tech Spec Sections 2.1.2,3.2.1 & 3.2.3 Re MCPR Safety Limit/Revised Calculational Methods

ML20070L689
Person / Time
Site:	Limerick
Issue date:	03/12/1991
From:	PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
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ML20070L688	List: ML20070L685 ML20070L689
References
NUDOCS 9103200161
Download: ML20070L689 (10)

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LIMERICK GENERATING STATION Docket No. 50-353 License No. NPF-85 PBOFOSED TECHNICAL SPECIFICATIONS QWKIES List of Attached Pages IJnit._2 xviii 2-1 B 2-1 3/4 2-1 3/4 2-8 B 3/4 2-1 B 3/4 2-2

, B 3/4 2-3 B 3/4 2-5 l

i 9103200161 F'DR 910312 ADOCK 05000333 PDR

, , INDEX BASES SECTION PAGE 3/4.0 APPLICABILITY................................................. B-3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN............................................. B 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES......................................... B 3/4 1-1 3/4.1.3 CONTROL R00S................................................ B 3/4 1-2 3/4.1.4 CONTROL ROD. PROGRAM CONTR0LS................................- B 3/4 1-3.

3/4.1.5 STAND 8Y LIQUID CONTROL. SYSTEM...............................

B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE........................................................ -

B 3/4 2-1 LEFT INTENTIONALLY BLANK............................................ B 3/4 2-3 l 3/4.2.2 APRM SETP0lNTS.............................................. B 3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATI0................................ 8 3/4 2-4 3/4.2.4 LINEAR HEAT GENERATION RATE................................. B 3/4 2-5 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION...................- B 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION......................... B-3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION' INSTRUMENTATION............................................. B 3/4 3-2 3/4.3.4 RECIRCULATION PUMP TRIP' ACTUATION INSTRUMENTATION........... B 3/4 3-3 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION-INSTRUMENTATION............................................. B 3/4 3-4 3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION........................... B 3/4 3-4 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation........................ b 3/4 3-4 LIMERICK - UNIT 2 xviii

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l 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.

APPLICASILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of P.ATED 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 H0T 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 High Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.07 for two recirculation loop operation and shall not be less than 1.08 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 1.07 for two recirculation locp operation or less than 1.08 l for single recirculation loop operation and the resctor 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 reacter coolant system pressure, as measured 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 with the requirements of Specification 6.7.1.

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-LIMERICK - UNIT 2 2-1

l l J 2.1 SAFETY LIMITS B'ASES i

l

2.0 INTRODUCTION

l 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 1.07 for two recirculation loop operation and 1.08 for. single recirculation loop operation. MCPR greater than 1.07 for two recirculation loop operation and 1.08 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 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 m this source is incrementally cumulative and continuously' measurable. Fuel cladding perforations, however, can result from thermal 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 l conditions which would produce onset of transition boiling,-MCPR of 1.0. These l conditions represent a significant departure from the condition Intended by design l for planned operation.

2.1.1 THERMAL POWER. Low Pressure or low Flow 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 f-lows will algaysbegreaterthan4.5 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 criticai power at this flow is approximately 3.35 MWt. With the design peaking factors. this corresponds i

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

3 4.2 POWER-DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE-LIMITING CONDITION TOR OPERATION 3.2.1 All AVERAGE P.LANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of axial location and AVERAGE PLANAR EXPOSURE shall be eithin limits based on applicable APLHGR limit values which have been determined by approved methodology for the respective fuel and lattice types for two recirculation loop operation. When hand calculations are required, the APLHGR for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not excet:1 thelimitingvalueforthemostlimitinglattice(excludingnaturaluranium)as shown in the CORE OPERATING LIMITS REPORT (COLR). 10uring single loop operation, the APLHGR for each fuel type shall not exceed the:above-values multiplied by- '

the reduction factors shown in the COLR.

APPLICABILITY: OPERATIONAL C@ DITION 1, when. THERMAL POWER-is greater than or equal to 25% of RATED THERMAL ?OWER.

ACTION:

With an APLHGR exceeding the limiting value, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce hours, THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4-l' SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or:less than the limiting-value:

a. At least once per.24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after-completion of a THERMAL POWER increase of at least' 15% of RATED THERMAL-POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL R00 PATTERN for.APLHGR.

d. The provisions of Specification 4.0.4 are not applicable.

LIMERICK - UNIT 2 3/4 2-1

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', POWER DISTRIBUTION-LIMITS 3J4.2.3 MINIMUM CRITICAL POWER RATIO LlHITINGCONDITIONFOROPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit times the Kr, both values shown in the CORE OPERATING l LIMITS REPORT, provided that the end-of-cycle recirculation pump trip (EOC-RPT) system is OPERABLE per Specification 3.3.4.2, with:

T = (Tave TB)

TA - TB where:

TA = 0.86 seconds, control rod average scram insertion time limit to notch 39 per Specification 3.1.3.3, TB

= N i )ij(0.016), l 0.672 + 1.65 ( n I

N j i=1 n

I tave i*l i i i n

I N j i=1 n =

number of surveillance tests performed to date in cycle, Hj =

number of active control rods measured in the'ith surveillance test, tj = average scram time to notch 39 of all rods measured in the ith surveillance test,-and N1

=

total number of active rods measured in Specification i

4.1.3.2.a.

l< APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater:than or equal to l 25% of RATED THi.RMAL POWER.

L LIMERICK - UNIT 2 3/4 2-8

3/4.2 POWER' DISTRIBUTION LIMITS B'ASES 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature (PCT) following-the postulated design basis Loss-of-Coolant Accident (LOCA) will not-exceed the limits specified in 10CFR50.46 and that.the fuel design analysis-limits specified in NEDE-24011-P-A (Reference 2) will not be exceeded.

Mechanical Design Analysis: HRCapprovedmcthods(specifiedin Reference 2) are used to demonstrate that all fuel rods in a lattice operating at the bounding power history, meet the fuel design limits specified in Reference 2. No single fuel rod follows, or is capable of following, this-bounding power history. This Lounding power history is used as the basis for the fuel design analysis MAPLHGR limit.

LOCA Analysis: A LOCA analysis is performed in accordance with 10CFR50 Appendix K to demonstrate that the permissible planar power (MAPLHGR) limits comply witn the ECCS limits specified in 10CFR50.46. The analysis is performed for the most limiting break size, break location, and single failure combination-for the plant.

The MAPLHGR limit es shown in the CORE OPERATlHG LIMITS REPORT is the most limiting composite of the fuel mechanical design analysis MAPLHGR and the ECCS MAPLHGR limit.

Only the most limiting MAPLHGR values are shown in the-CORE OPERATING LIMITS REPORT for multiple lattice fuel. Compliance with the-specific lattice MAPIHGR operating limits, which are available in Reference 3, is ensured by use of the process computer.

The MAPLHGR limits shall be reduced to the value shown-in the CORE OPERATING LIMITS REPORT times the two recirculation loop operation limit when in single recirculation loop operation. The constant-. factor shown in the CORE OPERATING LIMITS REPORT is derived from LOCA. analyses initiated from single loop operation to account for earlier boiling transition at the limiting fuel node compared to the standard LOCA evaluation.

LIMERICK - UNIT 2 B 3/4 2-1

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POWER DISTRIBUTION LIMITS-BASES '

3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which vould yield the design LHGR at RATED THERMAL POWER.

Th2 flow biased neutron flux-upscale scram trip _setpoint and flow biased neutron

,:ux-upscale control rod block functions of the APRM instruments must be i adjusted to ensure that the MCPR does not become-less than the Safety Limit-NCPR or that > 1% plastic strain does not occur in the' degraded situation.- The scram and rod Eleck setpoints are adjusted in accordance with the formula in this specification when the_ combination of THERMAL POWER and_CMFLPD indicates a higher peaked power distribution to ensure that an LHGR transient would not be- -!

ircreased in the degraded condition.

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LIMERICK - UNIT 2 8 3/4 2-2 t yg- e 9 1-W9- M.'q-"? ===N#

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LEFT INTENTIONALLY BLANK LIMERICK - UNIT 2 B 3/4 2-3

'; POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued) 1 For operation in the automatic flow control mode, the same procedure was employed except the initial power distribution was established such that the MCPR was equal to the operating limit MCPR at RATED THERMAL POWER _and rated thermal flow.

The Kg factors shown in CORE OPERATING LIMITS REPORT are conservative for the General Electric Boiling Water Reactor plant operation because the operating limit MCPRs of Specification 3.2.3 are greater than the original 1.20 operating limit MCPR used for the generic derivation of Kr.

At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWtR, the reactor will be operating at minimum recirculation pump' speed and the moderator void content will be very small. For all designated control rod patterns which may; e be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin. During initial startup testing of the plant, a MCPR evaluation will be made at_25% of RATED THERMAL g POWER level with minimum recirculatioti pump speed. The MCPR margin will thus be  !

demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power '

distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when a limiting. control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.

3/4.2.4 LINEAR HEAT GENERATION RATE' This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.

References:

1. General Electric Company Analytical Model for Loss-of-Coolant Analysis j in Accordance with.10 CFR 50, Appendix K, NEDE-205L5, November 1975'.

2. " General Electric Standard Application for Reactor Fuel," NE0E-24011-P-A (latest approved revision).

3. " Basis of MAPLHGR Technical Specifications for Limerick Unit 2," HEOC-31930P (as amended).

4. Deleted.

5. Increased Core Flow and Partial Feedwater Heating Analysis for Limerick Generating Station Unit 2 Cycle 1, NEDC-31578P, March 1980 including Errata and Addenda Sheet No. I dated May 31, 1989.-

6. General-Electric Boiling Water Reactor Extended Load Line Limit Analysis for Limerick Generating Station Unit 2, Cycle 1, NE0C-31677P, March 1989.

LIMERICK - UNIT 2 8 3/4 2-5

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