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 u

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Marked Un Technica1 Soecifications' ,

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

1.1 SAFETY LIMIT FUEL CLADDING INTEGRITY ,. . { 2 '.1 FUELCLADDINGINTEGRIE. lf l' .1 'U

' Applicability: -[r ' Applicability:T

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Applies' to . trip settings . of the

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

' system'_ safety limits'from being- >

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exceeded..

Objective:-

Objectivet '

l To define thellevel of thel pro To establish limits-below'which cess variables at:which automatic; the integrity.of the fue1~ clad H sc ding-is preserved. i protective'_ action. is initiated to

. prevent theifuel cladding
inte-- a
  • J =grity safety 111mits from being: ~

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exceeded.: '1
. Specification;- R

-Specification:

A. Reactor' Pressure >B00 psia and - 'A.- Neutron FluxLSeram: 1 Cor Flow >100 of Rated Thellimiting safety' system trip 1-The existente of a minimum , F critical pbwer-ratioL(MCPR) ,

. . settings shall be as'specified i shall consti , -below: .

h o g lest, than

.tute viola, tion of:the fuel e

? Neutron Flux Trip Settings

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cladding. integrity ety? t- ol .

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.s 1imiti A MCPR of . is here -

a.A LAPRM Flux" Scram Trip; inafter referred to as the V 1, f N Safety Limit'HCPR. 1  : Setting (Run Mode):

l' Core Thermal Power Limit (Reac- I Whe'n'the)ModebSwitch"isi 1 B. sin (the RUN1 position,> J tor Pressure 1800 psia and/or .!L ,

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~ Core Flov 110%) h n . 3 the'APRM fluxLscram- "

~ trip; setting.shall be:

When'the' reactor pressure is- T S5.58W?+6$p2 loop' i y 1

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

than or equal to'10% of' rated, [ '

U L_

the steady ' state core' thermal, 2

. Where:

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 - '

j The safetyLlimit.sball'be as e ' flow.to produce aH  :

sumed to;be exceeded whan scram  : '

is known to.have been aecomplished' rated coreJflov of' by a means other than;the~ expected -69.M Ib/hr.

Escram signal unless analyses demont , s strate .that the fuel cladding ,

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integrity safety limits defined'.in. J Specifications 1.1A1and 1.1B were ,' [

t' .

'not exceeded during the actual l

- transient. , , ,

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

< 100 o f Ra t ed. .

i The use of the GEC correlation is not valid for the c.riticC 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 lbs/hr bundle flow, show that with a flov 'of 28x10g I bundle pressure drop is nearly 1.ndependent of Lbundle power. and has a value of' 3.5 psi. -Thus,'the bund l driving head vill be grester ' than.28x10ge lbs/hr 1rrespective now with a,4.56 psi l

' of total core flow and independent of- bundle power for the range j

.of bundle powers of concern.- Full scale Alus test. data' takan l 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. l 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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1 Amendment No. 12 -  ;

. . . . . . . . . . - . I

Insert #4 A

The statistical analysis used to determine the MCPR safety limit is based The on a  !

model of the BWR core which simulates the process computer function. .

reactor core selected for these analyses was a large 764 assembly, 251 inch  !

reload core.

Results from the large reload core analysis apply for all I operating reactors for all reload cycles, including equilibrium cycles.

Random Monte Carlo selections of all operating parameters based on the f uncertainty ranges of manufacturing tolerances, uncertainties in measurement l of core operating parameters, calculational uncertairties, and statistical i uncertainty associated with the critical power correlations kre imposed upon ('

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

Reference 2.

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y C. Pever Transient _

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

are checked periodically. to assure' the insertion times are  ;

adequate. The ther:nal pover transient resulting.when's scramD m J

'is accomplished ;otharl than by the expected scram l signal' (a g., , y seram from neutron fluz following closuresLof 1 the mainBovaver,- turbinef J stop valves) does not necessarily cause fuel 1 damage. -!

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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.

(

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

-e l extensive p. tant safety analysisi . ,1

.. , f ne computeriprcE.ded with Filgrim t1 cit 1 has a:: sequence- 1 annunciation programfvhich v111Jindicate ths sequence in which events in! tiation, such -eteiasoccur.

. scra=, APPM trip' initiation, pressure : scram. Bis programL also > .Iin seram setpoint is cleared. 5 This vill prov$ desinfornation on ,

l hov long a scram' condition; exists and thus provida soma _ measure .

of the anergy = added during. a transient.

SI L

D. ' Raactor Water Level (Shutdove Con $1tiotN' , ,,

During periods when the reactor =1s shutdown,?cc didaration naist'  ;

also be given to water level requirements due. to J.ha zeffect of decay heat. If rametor water- level:should : drop below the :

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

the' core is reducad.  :

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

i

- The core can be cooled sufficiently.should the varar level-be- ~

reduced to two-thirds the core height.: 1Estab11shment of the safety -

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

l margin. This larel will;be continuously. monitored. -

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References  ;

1. Genaral Ilectric The: mal Analy's'is 3 asis ;(GEIAB): ; ! Data, '!

'I

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

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

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tie.:=i: -0:np :; r? S7:tr 0:7 r9 t, Jre,1?M -

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General' Ilectric Beiling Water Reactor Gener1c Reload , -

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i 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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ETTACRIEN,T.D'to BECo,,90-136 =

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

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