ML17286A676

Proposed Tech Specs Re Safety Limits,Thermal Power & High Pressure/High Flow

ML17286A676
Person / Time
Site:	Columbia
Issue date:	03/21/1991
From:	WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
Shared Package
ML17286A675	List: ML17286A674 ML17286A676
References
NUDOCS 9103280263
Download: ML17286A676 (6)

Text

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2.0 SAFET IMITS and, LIMITIHG SAFETY SYSTEM SETTINGS gp T() 4sgo lyvJ 9 /~T'~ ~/ ion intended by design for planned operation. The MCPR fuel cl ntegnty safety limit assures that during normal operation and d 'nticipated operational occur (P) rences, at least 99.9 percent of rods in the core do not ex eri transition boiling (

Reference:

H-H 52 A), ev. ; B om eport UK90-126; GE11 Lead Fuel Assembly Report or Mashington Public Power Supply System Nuclear Project No. 2, Reload 5, Cycle 6). The latter two references support application of the above establishe fety limit to GE11 and SVEA-9 LFA fuel in MNP-2 2.1 SAFETY IMITS

2. 1.1. THERMAL POMER Low r

~sQ Ubg TO 4 80 ~VJ>/NiN CYCLE, EXPOSNPE h.>K FOR. Cgcm SXPOSqQ.e, Cqem~i re or Low

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o)0 p ~~ pE For cer tain conditions of'ressure and flow, the correlation is not valid for all critical power calculations The correlation is not valid for bundle mass velocities less than . x 10 lbs/hr-ft or pressures less than . Therefore, the fuel cladding integrity Safety Limit is estab-

>she by other means. This is done by establishing a limiting condition on core THERMAL POMER 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 always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 10'bs/h (approximately a mass velocity of XS x lgs Ibs/hr-fts), bundie pressure drop is nearly independent offbundle power MASHINGTOH HUCLEAR - UNIT 2 B 2"1 Amendment Ho. 84 5'103230263 PDR- ADOCK 910321 P 05000397 PDR

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SAFETY LIMITS

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BASES ut THERMAL POWER. Low Pressure or Low Fl ow (Continued) suress and has a value of 3. 5 psi. Thus, the bundle flow with a 4. 5 psi driving head vill be greater than 28 x 10'bs/h. Ful 1 scale ATLAS test data taken at pres-from 14. 7 psi a to 800 psi a indicate that the fuel assembly critical power at this f ow i s 1 approximately 3. 35 Nt. With the design peaking factors, thi s corresponds to a THERMAL POWER of more than 50K of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25" of RATED THERMAL POWER for reactor pressure below

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SSS-y5iy is. conservative.

'2. 1.2 THERMAL POWER Hi h Pressure and Hi h flow The fuel cl adding i ntegri ty Safety Limit i s set such that no fuel damage is calculated to occur i f the limit i s not violated. Since the parameters which resul t in fuel damage are not directly observabl e during reactor opera-tion, the thermal and hydrau1 i c condi tions res ulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not neces sari ly result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient 1 imit. However, the uncertainti es in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertai nty in the val ue of the critical power. Therefore, the fuel cl adding integrity Safety Limit i s def i ned as the CPR in the limiting fuel assembly for whi ch more than 99. 9X of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

Methodo 1 ogy for boi 1 ing water reactors 'hi The Safety Limit MCPR is determiny$ ~using the AHF Critical Power combi nes al 1 of the uncertainties in operating parameters and the procedures ch is a stati sti cal model that p,~t 9 ~ used to calculate critical power. The probabi 1 ity of the occurrence of boi 1-ing transiti on is determined using the AHF nuclear critical heat fluxenthalpy correl ation. The correlation is valid over the range of conditions used in the tests of the daga used to deveIop the correlation.

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The requi red i nput to the stati sti cal model are the uncertainti es listed in Bases Tabl e B2. 1. 2-1.

a typical 764 assembly core in which the rod pattern was arbitrarily chosen to produce a skewed power distribution having the greatest number of assemblies at the highest power 1 evel s.. The worst distribution during any fuel cycle would not be as severe as the distribution used in the analysis.

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a. Huc 1 eadCri tical Power Methodology for Boiling Mater Reactors, N-H 524(A), Rev. ~

WASHINGTON HUCLEAR " UNIT 2 B 2" 2 Amendment No. 84 i~~~ii~ ~p."'.re're'r r . 'rn  %~iv>> i ~i. ~~.pre s',cc s>cj'o'F~.e.j. sf p~<g%< ~

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~ COWmOCLen COVA BASES TABLE B2.1.2-1 UNCERTAINTIES CONSIOEREO IN THE MCPR SAFETY LIMIT STANDARD Parameter DEVIATION" Feedwater Flow Rate . 0176 Feedwater Temperature .0076 Core Pressure .0050 Total Core Flow Rate .0250

~ Critical Assembly Flow Rate Power Correlation

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.0280 Power Distribution: p q~g py~y cia//gal go~@ g L

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" Fraction of'ominal Value.

g e g.~ ~I v C l.o c a c R o 0 Po ~L R, WASHINGTON NUCLEAR - UNIT 2 B 2"3 Amendment. No. 28

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