ML20062E361

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Proposed Tech Specs Re Upgrades to Safety Limit Min Critical Power Ratio & Rev to Operating Limit Min Critical Power Ratio
ML20062E361
Person / Time
Site: Pilgrim
Issue date: 11/08/1990
From:
BOSTON EDISON CO.
To:
Shared Package
ML20062E357 List:
References
NUDOCS 9011200135
Download: ML20062E361 (7)
v • d • e


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..t ATTAC4 MENT;C to'BECo 90'136

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1 1.1 SAFETY LIMIT

- 2 '.1 L TLIMITING SAFETY. SYSTEM SETTING ~

. { 2 '.1FUELCLADDINGINTEGRIE.

lf l'.1 FUEL CLADDING INTEGRITY

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' Applicability:

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' Applicability:T r

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Applies to the interrelated

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Applies' to. trip settings. of the variables associated with fuel instruments and devices'which are-thermal behavior.
provided to prevent:the reactor-

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' system'_ safety limits'from being-0,

exceeded..

Objectivet Objective:-

l To establish limits-below'which To define thellevel of thel pro cess variables at:which automatic; the integrity.of the fue1~ clad H s c ding-is preserved.

i protective'_ action. is initiated to

. prevent theifuel cladding:inte--

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=grity safety 111mits from being:

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exceeded.:

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-Specification:

. Specification;-

R A.

Reactor' Pressure >B00 psia and -

'A.-

Neutron FluxLSeram:

Cor Flow >100 of Rated The existente of a minimum,

F Thellimiting safety' system trip 1-

. settings shall be as'specified i

critical pbwer-ratioL(MCPR) h o g lest, than shall consti,

-below:

.tute viola, tion of:the fuel e

cladding. integrity ety?

t-ol.

? Neutron Flux Trip Settings

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I 1imiti A MCPR of is here -

V a.A LAPRM Flux" Scram Trip;

.s inafter referred to as the 1, f N Safety Limit'HCPR.

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Setting (Run Mode):

l' B.

Core Thermal Power Limit (Reac-I Whe'n'the)ModebSwitch"isi 1

tor Pressure 1800 psia and/or

.!L sin (the RUN1 position,>

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~ Core Flov 110%)

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. 3 the'APRM fluxLscram-y

~ trip; setting.shall be:

When'the' reactor pressure is-T S5.58W?+6$p2 loop' y

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.$ 800 psia or' core flow'is less l

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than or equal to'10% of' rated,

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2

. Where:

U the steady ' state core' thermal, l power shall not. exceed l25% of 3 'e " Setting; in percent 1 design thermal power..

of rated!thermali C.

Power Transtent poweri(1998,MWt)-

p EW =' :PercentL ofudrive -

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The safetyLlimit.sball'be as sumed to;be exceeded whan scram

' flow.to produce aH is known to.have been aecomplished' rated coreJflov of'

-69.M Ib/hr.

by a means other than;the~ expected s

Escram signal unless analyses demont

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strate.that the fuel cladding J

integrity safety limits defined'.in.

Specifications 1.1A1and 1.1B were,'

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'not exceeded during the actual l

transient.

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.;; niin;;t-ly-ehees; ;; produ;; ; eks;d van dinr'L.;in. '. ;isp

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Cere Ther a1 Pover' L1=.1: (Reaeter Pressure < BOO psig or Core Tiow

< 100 o f Ra t ed.

The use of the GEC correlation is not valid for the c.riticC i

power calculations at pressures belov 800_psis or core flows less than 10% of rated. Therefore,3the fuel cladding integrity saf ety

'd ' t is established by other.means.

This.is done by establishing a W ring condition of core thermal power operation l

with the following basis.

.l since the pres'sure drop.in the bypass region is essentially all elevation head which is 4.56 psi the core pressure drop at low f

power. and all flows vill always be graaAnalyses show that with a flov 'of 28x10g lbs/hr bundle flow, bundle pressure drop is nearly 1.ndependent of Lbundle power. and has a value of' 3.5 psi. -Thus,'the bund driving head vill be grester ' than.28x10ge now with a,4.56 psi lbs/hr 1rrespective of total core flow and independent of-bundle power for the range j

.of bundle powers of concern.- Full scale Alus test. data' takan at pressures from'14.7 psia' to.800 psia indicate that the' fuel assembly critical power at. this flov f.s approximately 3.35 MWt.

With the design peaking f actors,the - 3.35 MWt bundle power cor-responds, to a core thernal power of-more than 50%. Therefore a core thermal power lir.it of 25% for reactor pressures below 800

- psia', or core flov less than 10% is : cons trvative.

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Amendment No.

12 I

  1. 4 Insert A The statistical analysis used to determine the MCPR safety limit is based on a The model of the BWR core which simulates the process computer function.

reactor core selected for these analyses was a large 764 assembly, 251 inch Results from the large reload core analysis apply for all reload core.

operating reactors for all reload cycles, including equilibrium cycles.

f Random Monte Carlo selections of all operating parameters based on the uncertainty ranges of manufacturing tolerances, uncertainties in measurement l

of core operating parameters, calculational uncertairties, and statistical uncertainty associated with the critical power correlations kre imposed upon i

the analytical representation of the core and the resulting bundle critical Details of this statistical analysis are presented in power ratios.

Reference 2.

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Pever Transient _

Plant safety analyses have shown that the scrams; caused by'ex-ceeding:any.safsty:*etting' vill; assure that the-Safety Limit of Scrau times Specification 1.1A wr 1.13 vill not be' exceeded.

are checked periodically. to assure' the insertion times are adequate. The ther:nal pover transient resulting.when's scramD J

'is accomplished ;otharl than by the expected scram l signal' (a g.,

m seram from neutron fluz following closuresLof 1 the main turbinef y

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Bovaver,-

stop valves) does not necessarily cause fuel 1 damage.

for this' specification a Safery LimitLviolation will be' assumed

.vben a scram is only; accomplished by maans' of f a backup.- featura -

't of the plant design. ; ne coneapt of not approaching: a Safsty.

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Limit-provided'sers= -signals are operable is supported by tha

-e extensive p. tant safety analysisi f

ne computeriprcE.ded with Filgrim t1 cit 1 has a:: sequence-annunciation programfvhich v111Jindicate ths sequence in which 1

events such as. scra=, APPM trip' initiation, pressure : scram. Bis programL also > i.I in! tiation, -etei occur.

seram setpoint is cleared. 5 This vill prov$ desinfornation on hov long a scram' condition; exists and thus provida soma _ measure of the anergy = added during. a transient.

SI

' Raactor Water Level (Shutdove Con $1tiotN' L

D.

During periods when the reactor =1s shutdown,?cc didaration naist' also be given to water level requirements due. to J.ha zeffect If rametor water-level:should : drop below the :

of decay heat.

top' of.the' active: fual'during this ; time, thei ability to cool This : reduetion ' in' core cooling. capability.

(

the' core is reducad.

could lead to elevaeed cladding;temperaturas and' clad perforation.

The core can be cooled sufficiently.should the varar level-be-i; reduced to two-thirds the core height.: 1Estab11shment of the safety -

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l limit at 12 inches above the top of. the - fuel provides: adequate r

l This larel will;be continuously. monitored. -

3 margin.

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t References Genaral Ilectric The: mal Analy's'is 3 asis ;(GEIAB): ; ! Data,

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' Corralation:and Design Application, tGeneral' Electric Co.

1.

NovemberJ1973 /(NEDO-10958), f 3WR Systems Department, W-

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General' Ilectric Beiling Water Reactor Gener1c Reload,

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d Tue1~ Application, NEDE-24011-P.

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TABLE 3.11'l-OPERATING LIMIT.MCPR VALUES-

.A.

MCPR Operating Limit from'Beginning of; Cycle.(BOC)Lto BOC +

MHD/ST.:

P8x8R/BP8x8R L

for all values of t-7,ggy gj,43 f

B.

MCPR Operating Limit from BOC +

MHD/ST to End if'C,scle.

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m ETTACRIEN,T.D'to BECo 90-136.

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Supplemental-Reload" Licensing Submitted' forLPilgrim NuclearLPower Station-reload 7 Cycle 8,~23A4800 Rev.'-l1 1

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