ML20101C957

Amend 99 to License NPF-29,revising Safety Limit MCPR Values for two-loop Operation & Slo,Slo MAPLHGR multiplier,flow- Dependent MCPR Operating limits,power-dependent Operating Limits & exposure-dependent MCPR Operating Limits

ML20101C957
Person / Time
Site:	Grand Gulf
Issue date:	05/28/1992
From:	Larkins J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20101C961	List: ML20101C957 ML20101C963
References
NUDOCS 9206100219
Download: ML20101C957 (15)

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ENTERGY OPERATIONS. INC.

SYSTEH ENERGY RESDV8CES. INC.

SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION MISSISSIPPI POWER AND LIGHT COMPANY QQ ZET NO. 50-416 GRAND GULF NUCLEAR STATION. UNIT 1 i

l AMEN 0 MENT TO FACIllTY OPERATING LICENSE Amendment'No. 99 License No. NPF-29 1.

The Nuclear Regulatory Commission (the Commission)_ has found that:

y A.

The application-for amendment by Entergy Operations,_Inc. (the licensee) dated December 5, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended _(the Act),

and the Commission's rules and regulations set forth in 10 CFR Chapter I; l_

B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules _ and regulattons of the-Commission; i

C.

There is reasonable assurance (i) that the activities ~ authorized by-this amendment can be conducted without endangering _the health and safety of the public, and (ii) that-such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common-defense and security or to-the health and safety of the public; and E.

The. issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and~ all applicable requirements have.

been satisfied; 9206100219 920528 PDR ADOCK 05000416

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Accordingly, the license _is amended by-changes to the Technical Specifications, as-indicated in the attachment to this license amendment; and paragraph 2.C.(2)'of Facility Operating L.icense No.-NPF-29 is heraby amended to read as follows:

-(2) Technical Soecifications The Technical' Specifications contained in Appendix _ A'and the-Environmental Protection Plan contained in Appendix B,. as reviod through Amendment No. 99, are hereby incorporated into this. in license.

Entergy Operations, Inc. shall-operate the facility-accordance with the Technical Specifications and the Environmenta'-

Protection Plan.

3.

This-license amendment is-effective as of'its date of-issuance.

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FOR THE NUCLEAR REGULATORY: COMMISSION N

John T. Larkins, Director Project Directorate:IV-1 Division of Reactor Projects III/IV/V.

Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: May 28 -1992

t ATTACHMENT TO LICENSE AMENDMENT NO. 99 FACIllTY OPERATING LICENSE NO. NPF-29 DOCKET NO. 50-416 Replace the follo',ving pages of the Appendix A Technical Specifications with the attached pages.

The revised pages are identified by amendment number and contain vertical' lines indicating the area of change.

The corresponding:

overleaf pages are also provided to naintain' document completeness.

REMOVE PAGES INSERT PAGES 2-1 2-1 B 2-1 B 2-1 B :-la B 2-la B 2-2 B.2-2 3/4 1-1 3/4 2-1 3/4:2-2 3/4 2-2 3/4 2 3/4-2-5 3/4 2-6 3/4 2-6 3/4 2-6a 3/4'2-6a 3/4 2-7a 3/4-2-7a 3/4-2-7b 3/4 2-7b 3/4 2-7c 3/4 2-7c B 3/4 2-1 B 3/4 2-1 I

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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.

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 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s-and comply with the requirements oi Speci fication 6.7.1.

THERMAL POWER, High Pressure and High Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.06 during both two loop operation and 1.07 during single 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 the above limits 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-Specifi-cation 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.

APPLICA8ILITY:

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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GRAND GULF-UNIT 1 2-1 Amendment No 7), 99 nna y

s 2.1 SAFETY LIMITS l

BASES

2. 0 INTRODUCTION The fuel cladding, reactor pressure vessel and primary system piping are the principal 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 inteority 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.for the MCPR.- MCPR greater than the applicable Safety Limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of tne 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 reactor operation significantly above design condi-tions 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.

l 2.1.1 THERMAL POWER, Low Pressure or Low Flow l

The Siemens Nuclear Power Corporation (SNP) ANFB critical power correla-tion is applicable to the SNP core.

The applicable-range of the-ANFB correla-tion is for pressures above 585 psig and bundle mass flux greater than 0.25M1bs/

hr-f t2 For low pressure and low flow' conditions, a THERMAL _ POWER = safety limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig and below 10%

l RATED CORE FLOW was justified for Grand Gulf cycle 1 operation based on ATLAS test data and the GEXL correlation.

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 fue1 ~ cladding integrity Safety Limit was established by other means.

This was done by. establishing.a limiting condition on core THERMAL POWER with the following basis.

Since the L

pressure drop in the bypass region is essentially all elevation head, the core i

pressure drop at low power and flows will always be greater than 4.5 psi.

GRAND GULF-UNIT 1 B.2-1 Amendment No. 77, 99 m

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i 2.1 SAFETY LIMITS BASES-THERMAL POWER, Low Pressure or Low Flow (Continued)

Analyses show that with a bundle flow of 28 x 103 lbs/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 greater than 28 x-103 lbs/hr.

Full scale ATLAS test data taken at pressures from 14.7-psia to 800 psia indi-cate 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 THERRAL POWER for reactor pressure below-785 psig is conservative.

Overall, because of the design thermal-hydraulic compatibility of the.SNP fuel designs-with the cycle 1 fuel this iustification and the associated low pres--

sureandicwflowlimitsremalnapplcableforfuturecyclesofcorescontain-ing these fuel designs.

With regard to the low flow range, the core bypass region will be flooded at any flow rate greater than 10% RATED CORE FLOW.

With the bypass region flooded, the associated elevation head is sufficient to assure a bundle mass.

flux of greater than 0.25 Mlbs/hr-ft2 for all fuel assemblies which can approach critical heat flux.

Therefore, the ANFB critical power correlation is appro-priate for flows greater than 10% RATED CORE FLOW.

The low pressure range 'or cycle I was defined at 785 psig.

Since the ANFB correlation is applicable at any pressure greater than 585 psig, the cycle 1 low pressure boundary of 785 psig remains valid for-the ANFB correlation.

GflAND CULF-UNIT 1 B 2-la Amendment No. 71, 99 I

5 SAFETY LIMITS BASES 2_._1. 2 THERMAL POWER, High Pressure and Hiah Flow The onset of transition boiling-res'ults in a decrease in heat: transfer from the clad, elevated clad temperature, and the possibility-of clad failure.

However, the existence of critical power, or. boiling transition, is not a di-rectly observable parameter in an operating reacter.

Therefore, the margin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature and core power distribution.

The mar-ginforeachfuelassemblyischaracterIzedbythecriticalpowerratio'(CPR),

which is the ratio of the bundle power.which-would produce onset of transition.

boiling divided by the actual bundle power.

The minimum value of this ratio for any bundle in the core is the minimum critical-power ratio (MCPR). -

The Safety Limit MCPR assures sufficient conservatism such that, in the event of a sustained steady state operation _at the MCPR safety limit, at least 99.9% of'the fuel rods in the core would be expected to avoid boiling transi-tion.

The margin betweer calculated boiling transition (MCPR = 1.00) and the Safety Limit MCPR is based on a detailed stttistical procedure-which. considers the uncertainties in monitoring the core operating state and includes the -

effects associated with channel bow.

One specific uncertaintyLincluded in the-safety limit is the unct.rtainty inherent in the ANFB critical _ power correlation.

SNP report ANF-524 (P)(A), Rev. 2, " Advanced-Nuclear Fuels Corporation Critical Power Methodology _ for Boiling Water Reactors " April 1989, including supplements, describes the methodology'used in determining the Safety Limit MCPR.

The ANFB critical power correlation is based on a significant body of-practical test data, providing a high degree of assurance that the critical power as evaluated by the correlation is within a-small-percentage-of-the actual critical power being estimated.

The assumed reactor condMions used in defining the safety limit introduce conservatism-into the limit-scause bound-ing radial power factors and bounding flat local peaking distriLtions are used to estimate the number of rods in boiling transition.

Still'further con--

servatism is induced by the tendency.of the ANFB correlation.to overpredict the number of rods in boiling: transition.

These conservatisms and.the inherent-accuracy of the ANFB correlation provide assurance that during: sustained opera-tion at the Safety Limit MCPR there would be essentially no transition boiling in the core.

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GRAND GULF-UNIT 1 B 2-2

-Amen sent No. 77, 99-

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3/4.2-POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 During two loop operation, all AVERAGE PLANAR LINEAR-HEAT GENERATION-RATES (APLHGRs) for each type of f'.el as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figure _3.2.1-1.

During single loop operation, the APLHGR for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limit shown in Figure 3.2.1-1 multiplied by 0.86, APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than 1

or equal to 25% of RATED THERMAL POWER.

ACTION:

During two icop operation or single loop operation, with an APLHGR exceeding.

l t*.s limits of Figure 3.2.1-1 as corrected by the appropriate multiplication factor, initiate corrective acticn 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 THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the required limits 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 Initially and at least once-per 12: hours when the reactor _is-c.

operating with a LIMITING. CONTROL-R00 PATTERN for APLHGR.

d.

The provisions-of Specification 4.0.4 are not applicable.

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3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temper-ature following the postulated design basis loss-of-coolant accident will not exceed the 2200'F limit specified in 10 CFR 50.46, 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46, The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the-average beat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly.

The Maximus Average Planar Linear Heat Generation Rate (MAPLHGR) limits of Figure 3.2.1-1 are applicable to two loop operation.

For single-loop operation, a MAPLHGR limit corresponding to the product of the MAPLHGR, Figure 3.2.1-1, and 0.86 can be conservatively used to ensure that the PCT for single loop operation is bounded by the PCT for two loop operation.

l The daily requirement for calculating APLHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribu-tion shifts are very slow when there have not been significant power or control GRAND GULF-UNIT 1 B 3/4 2-1 Amendment No.

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