ML20128D916

Forwards TS Bases for Monticello Nuclear Generating Plant Which Set Forth Bases for Established Thermal Limits of Specifications 2.1,3.11C,4.11C & Bases for Specification 3.11A Entitled, Aplhgr

ML20128D916
Person / Time
Site:	Monticello
Issue date:	12/05/1975
From:	Stello V Office of Nuclear Reactor Regulation
To:	Goller K Office of Nuclear Reactor Regulation
References
NUDOCS 9212070421
Download: ML20128D916 (19)

Text

.

.m - -. .

t .

l .

UNITED STAf ts /

NUCLEAR REGULATORY COMMISSION 0 W ASHINGTON. D. C. 20555 Docket No. 50-263 DEC 51975 Karl R. Goller, Assistant Director for Operating Reacto's, RL MONTICELLO TECHNICAL SPECIFICATION BASES FOR THE THERMAL LIMITS Plant Name: Monticello Docket No.: 50-263' Responsible Branch ORB-2 and Project Leader: B. Buckley Technical Review Branch Involved: Reactor Systems Branch Review Status: Complete Enclosed are the Technical Specification bases for the Monticello Nuclear Generating a. ant which set forth the bases for the established thermal-limits of Specifications 2.1, 3.11C, and a

4.11C. Also, the bases for Specification 3.11A, entitled " Average

• Planar Linear Heat Generation Rate" (APLEGR)lare provided.

Note that the Technical Specification bases for the thermal'

, limits were applied to the Monticello Cycle 4 safety analysis.

As stated by the licensee, the Cycle 5 analysis was-found to be bounded by the Cycle 4 analysis. Therefore, the enclosed bases j justify the Technical Specification. thermal limits for Cycle 5 plant operation.

l

/ ~~

Victor Stello( .Jr.gssistant Director for Reactor- Safety Division of Technical Review

• Office of Nuclear Reactor Regulation.

Enclosure:

l Tech. Spec. Bases .. ..._

I l cc: S.-Hanauer l

R. Heineman

!- R. Boyd

, V. Stello I i .D. Ziemann B. Buckley j.

P. Check T. Novak <

W. Minners i R. Woods

'A. Ignatonis-l l

9212070421 751205 PDR ADOCK 05000263-

. P_.

PDR

- a + 4. -- n 4....... .. . . . . . . .

,LJ.

l i'...

O *

• ENCLOSURE TO MONTICELLO TECHNIC, /.L SPECIFICATION BASES Bases: ,.

2.1 Fuel Cladding Integrity A. Fuel Cladding Integrity Limit at Reactor Pressure >800 psia and Core Flow >10% of Rated s

The fuel cladding integrity safety limit is set such that no

~

fue1 damage is calculated to occur if the limit is not violated.

Since the parameters which result in fuel damage are not directly observable during reactor operation the thereal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region wher,e fuci damage could occur. Although it is recognized that a departure from nucleate, boiling would not necessarily

. result in damage to Ek'R fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However,the uncertainties in monitoring the core operating state and in the procedure used to calculate

. the critical power result in an uncertainty in the value of the critical power. Therefore the. fuel cladding integrity safety limit is defined as the critical power ratio in the limit'ing 3

fuel assembly for.which more than 99.9% of the fuel rods in the core are expected'to avoid boiling transition considering the power distribution within the core and all uncertaintfes.

The Safety Limit MCPR is, determined using the General Electric Thermal Analysis Basis, CETAB .

, which is a statistieni model that combines n11 of the 'neertainties in oper' ting a parameters

- . - _ , _ . ._, -u ,, ,, _ _ _ . - _ _ . _ _ . . . _ . _ . . _ . _ . . _ _ _ _

o

.L2 e,, .

-- 2'

) and the procedures used to. calculate critical power.-

l ,

.: The. probability of'the occurrence of boilin's transition i

is determined using the- General' Electric- Critical Quality - (X) -

1 l' .

Boiling Length (L), CEXL, correlation.-

i i

.The CEXL correlation is valid over the range of conditions f,N 4 ,

used in the tests of the _dat,a used to develop the correlation.

I . f i

l These conditions'are:- .

4 .

4 l Pressure:- 800 to 1400 psia Mass flux: 0.1 to 1.25 10 61b/hr-f t2 i

0 to 100 Btu /lb' Inlet Subcooling:

ll Local Peaking: 1.61 at-a corner rod'to l 1.47 at an-interior rod 2'

-Axial Peaking:

j . S'hape Max / Avg.-

'. Uniform 1.0' I

I " Outlet. Peaked 1.60-i .

' ' Inlet Peaked 1.60 l- .

! Double Peak 1.46 and 1.38 L ,

i Cosine -1.39- .

t Rod Array- 16, 64 Rbds-in an 8 x B' array- y

' 49 Rods'in a.7.x 7 array c _ _ _

The required input to the statistical model are the uncertaintied i

• listed on Table 2.1-1, the nominal' values of 'the core paramete[rs t

listed in Tabic 2.1-2, andzthc relative assembly power'distrii' ,

. .- . i 1

bution shown:in- Table . 2.1-3. Tabic 2.1-4 shows.the R-factor ,

e distributions that arc input to,)he statistical model wnich is usel to establish =the safety limit MCPR. ,'he T R-factor distributions

~

shown areltaken near the beginning of the, fuel. cycle.:

- . ._ - . _. _ . ._ _ _. _ _ .u . _ . .. a. . _ . ~ _ . _ _ . . - . . - . _ _ _ . .

,.~, .u.,... , . . . . . . . - _ . .

-3 .

e. . ,

'a The b:ses or the uneartcintics in tha co. .

.j parameters are given in NEDD-20340.(2) and the basis for the uncertainty in the CEXL correlation is given in NEDO-10958(1) . The'p,o'wer distribution is based on a typical 764 assembly core in which th*e L rod pattern was arbitrarily chosen to produce a skewed power distribution

~

having the greatest number of-assemblies at the highest-power. levels . The worst distribution in Millstone Unit 1 during any fuel cycle would not be as severe as the distri-bution used in the analysis, i

B. Core Thermal Power Limit (Reactor Pressure 4 800' psia on

. Core F3mi 5 10% of Rated)

~

~

The use of the GEXL correlation is not valid for the critical power calculations at pressures below 800 psia or core flows less than 10% of rated. Therefore, the fuel' eladdingL integrity safety limit-is established by.other means. This is done by-establishing a limiting condition of core thermal power _ operation i with the following basis.

Since the pressure drop in the bypass region is essentially all i

elevation head which-is 4.56_ psi .the core _ pressure drop at low power-and all flows will always be greater ;than 4.56 psi.

3 LAnalyses- show that with a flow of 28 x 10 lbs/hr bundle flow, t

i bundle pressure drop is nearly_ . independent of bundle power and ias a .value of 3.5_ psi. Thus, the bundle flow with a. 4.56 psi ,

l 8

3 driving head will be greater than'28 x 10 lbs/hrl irrespective i of' total core flow and independent of bundle power for the range

l. '

of bundle powers o f concern. Full scale ATLAS _ test. data taken-l L

i k

. . - , , . . . . ,m _ _ . . _ . . . , ~ . , . _ , . _ , , _ . . . ~ - . _ . , , - - . -

, u . , t. . s . . m , -

. . , - m -- - - - - ~ - - - - - - - -- -

' 4 .4. - .

e

'4 i --

~

,(

j . .

, 7,

. at pressures from 14.7. psic to 800 psia indicato that ths -

fuel assembly critical power at this flow is approximately-4 3 35 MWe, with the design peaking factors the 3.35 MWe

" bundle power corresponds to a core thermal' power of more than s i

' 50%. Therefore a core thermal power limit of 25% for reactor-i pressures below 800 psia or core flow less.than 10% is conservative.

C. Power Transient  ;

" Plant safety analyses have shown that the scrams caused by -

(

exceeding any safety setting will assure that the Safety Limit

]. Scram I of Specification 2.1.A or . 2.1.B will not _ be exceeded.

J i times are checked periodically to assure the insertion-times i

i

- are adequate. The thermal povar transient resulting when a scram is accomplished other than by the expected scram signal 4

(e.g., scram from neutron flux,following closure of the main-I turbine stop valved) does not necessarily'cause fuel damage, j i However, for this specification a Safety Limit violation will t '

be assumed when a scram is.only accomplished by means of-a

~

a i -

i

I backup. feature of the plant design.- TL concept of not-4 approaching a Safety' Limit provided scram signals are operabic i

is supported by the extensive plant saf ety analysis, -

The computer provided Oith LMonticello has a sequence- ,

i .

annunciation program which-will indicate the sequence'in which

! z events'such as scram..APRM trip-initiation, pr' essure' scram.

initiation, etc. occur. -This program also indicates when . .

the scram setpoint is cleared. This wi11 provide information r

l on how long a-scram condition e,xists and thus' provide some

' me'asure of 'the -energy added during la transient.

1 . - .

[ ,

t .

. o

. . , _ _ _., _ .. _ _ _ . _ , , - , _ _u,_ _ . _ . , _ - . _ _ ._ ._,. _ . _ . _ . . . . . _

. . .w . ,.

w.- en..".:.

. -, s

+

... .. ,.a. .. ..,.-,n - - -

=

..~ ,

~ - ~

.. c t

2 .

1 l.

a .

. t

~';

I D. Reactor-Water Level (Shutdown Condition) 3 5

During periods when the reactor-.-is shutdown, consideration ,

. 'must also.be given to. water-level requirements.due to the effect t.

E of decay heat. If reactor water' level should drop.below the l l .

top of the active fuel during this time the' ability:to cool-

~

i.

!' c the core is reduced.. ' This reduction in core? cooling- capability n

[

i could lend to elevated cladding-.-temperatures and clad perforation.

The core can be cooled sufficiently;should the water level be l .

reduced to'two-thirds the core height. -. Establishment of'the-safety limit at 12 inches above the' top.of the fu'el provides-adequate margin. - This level kill.be' continuously-monitored. '

References l -.

l 1.: General Electric Thermal' Analysis Basis (GETAB):: Data - ,

[ . Correlation and Design Application, Cencral-Electric Co.

=.

. .BWR Systoms Department, November 1973 . (NEDO-10958) ~. '.-

• t- . . . .

l

'2. Process Computer Performa'nce Evaluation Accuracy', Cener'al:

[. Electric Company BWR Systems Department, June,1974 '

(NEDO-20340). ,

e .

d

. 'I_

o n,

4 l

j. . .

i o

n. . . -

l ,

, .a- --. g i . *

- . . - , ,, c , , , - , n._,, -m,, , , , , , a,.,,. .n..--, .-,e -,-i,...r.~.6,,.. m..-..--..

. w a y +y ,e g . .s.n._w.._.. . _;m . =;_ ;,, __;;, ~.

. . .. _ = = - .-. . .-.

.] ' *

6. '

r j . . . . .

2 l  ;-

• i Table 2.1 i

• UNCERTAINTIES USED IN THE DETERMINATION ,

t i

0F THE FUEL CLADDING SAFETY LIMIT t -

4-

. - 4!

t-Standard Deviation Quantity . . (% of Point) ,

i l Feedwater Flow 1.76:

Feedwater Temperature 0.76 i .

Reactor Pressure 0.5 f

i,

. Core Inlet Temperature .

b.2 .

r .

Core Total Flow 2.5 .

Channel Flow Area - -

3.0 .

i Friction Factor Multip, lier 10.0 Channel Friction Factor F

Hultiplier * *

'5.0 -

1 . .

b

. TIP Readings 8.7 .

i R Factor . . 1.6-j Critical Power -

3.6 - -

?

e -- . .

4 .

- 4 0 4 4

e

.i j ,

i. .

h

, e .

. g 6 Y 4

. 4 .

, =

] . ,- .-,- . .

i . . . .

M 4 ,

4 4

e au., g y -- , # m,. , s h , - g- +4, e--g wi a - .-s a-s rt v w +s-5% <r'fe-+'wrt- e r'e

• e- w se rte +'-f' t- * ** V- r -=. *W- er

La

. -1

. Table 2.1-2 NOMINAL VALUES OF FARAMETERS USED IN_

THE STATISTICAL ANALYSIS OF FUEL CLADDING INTEGRITY SAFETY LIMIT Core Thermal Power 3293 W .

, . Core Flow ,

102.5 M1b/hr Dome Pressure 1010.4 psig Channel Flow. . ~Area 0.1078 ft R-Factor 1.098 (7x7) 1.100 (8x8) e t

G

• 9

1. e e

6 y

e G .

. e O

Maesteoef e

e

4 . -- c. . 6 m .. f -v . . ... - c . - , ,; . ,_ , .., ,- . . _ . . _ _ _ _ .__ _._ _ _ _ _ _

. . . a .ma F 4

e.e. .

. g.

t- ..

1 . .

I- . Tabic 2.1-3 ,

c .

1 *

• RELATIVE BUNDLE POWER DISTRIBUTION- ,

I USED IN THE CETAB STATISTICAL ANALYSIS f.

4 1

i ..-

e

!- Range of Relative Bundle Power Percent of- Fuel Bund 16s Wit.hin Power' Interval-l .

1.375 to 1.425 6.6

? ,

1

'1.325 to 1.375 3.2 l

i

'., 1.275 to 1.325 .

15.6-1.225 to 1.275 10.8 w

+

1.175 to 1.225 6.6 3

1.125 to 1.175 ,

4.9 1.075 to l'.125 ,

9.0 t

i 1.025 to 1.075'. ~

4.0 .

i.

i 0.175 to 1.025 .3 9,3 '

j -

i- .

Sum = 100-e t

• 4.

F j . .

i i

p .

i-

• I .- * .

i . I i

t t<

o .

6 . I s

' s', --

c. .

i e e e g

r . .

I e

, , ,,v, ..m.... , , , . ,_..--,,,-.,-,I-,

. m. .._ , , . ~ , . . . . . .. - , . _ , . r -_,,.._.--m.,..,,-n-..., . . ~ , . . , . . . ,

, - .  :. u , = .: w . a ;. . ;.. ..;.. _ , ,... . .

i .. . ..

i

-:9- , , .

6 .

j. J. .

, Tab.le 2.1-4 .

. R-FACTOR DISTRIBUTION USED IN CETAB. STATISTICAL ANALYSIS 7x7 Rod Array 8x8 Rod Array R-Factor Rod Sequence No.- R-Factor Rod Sequence No. '6 1.098 1 1.10'O 1 4

• t.

i .

.1.083 , 2 1.100 2 i

l '. 1.075 3 - 1.095- 3s

! 1.062 4 1.095 4 1.052- 5 1.093 5 i -

^

1.042 6- 1.093 . 6--

i '

1.042 7 1.092-- 7-

[! . .

i 1 1.027 8 thru 49 1.1.077 8 thru 63 j

i . .

4 4

t

' S 3

i* ~ .

. e , e

+-

a j .

i . ,

1. 1: . ,

• r

[; *

s. .-

j_ . .

8 e

• e.

e n

• - *wm > , , ,e-w-s m- ,-e, t- rr r ft w- +- w*wrrr--gre l , +--*-<q==+e-- '

• ht r e w >,e w- +

:.- .w . .

'~

/ , BASES .

3.11A Average Planar _ Linear Heat Generation Rate (APLHGR).

This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR SO, Appendix K.

The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent, secondarily on the rod to rod power distribution within an assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification. This LHGR times 1.02 is used in the heat-up code along with the exposure dependent steady state 4

gap conductance and rod-to-rod local peshing factors. The Technical Specification APLHGR is this LHGR of the highest 4 powered rod divided by its local peaking f actor. The limiting value for APLEGR is shown in Figure 3.11.1.

The calculational procedure used to establish the APLHGR shown on Figure 3.11.1 is based on a loss-of-coolant accident analysis.

j The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in Reference 1. Differences in this analysis as compared to previous analyses performed with Reference 1 are:

(1) The analyses assumes a fuel assembly planar power consistent with 102% of the MAPLHGR shown in Figure 3.11.1, (2) Fission product decay is computed assuming an energy release rate of 200 FEN / Fission; (3) Pool boiling is assumed after nucleate boiling is lost during the flow stagnation period; (4) The effects of core spray entrainment and counter-current flow limiting as described in Reference 2, are included in the reflooding calculations.

A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Table 1. .

t

.. . . . . . . . - ~ . . . .

4

/ ,

111 ,

TABLE 1 SIGN 1f! Calli INPUTS PARAMETERS TO THE LOSS-OF-COOLAtti ACCIDElli ANALYSIS ,

'FORHONTIC{LLO .

PLANT PAPAMETERS:

1703 MWt which corresponds to Core Thermal Power...............'.....~._

102% of licensed core newer

• 6 lbm/h which corresponds to

,i Vessel Steam Output............. f.o13 x 10 i 102 % of rated steam flow '

1040 psia ,

Vessel Steam Dome Pressure............ _

Design Basis Recirculation Line 2 1.0 ft' l

Break Area-for Larce Breaks

" 3. 4 ft -

Recirculation Line Break Area

^ _1. 0 _f t2 - 0.07 ft2 for Small Breaks ,

FUEL PARAMETERS: ,

PEAK-TECHNICAL .

INITIAL

I '

DESIGN MINIMUM

' SPECIFICATION AXIAL CRITICAL LINEAR. HEAT PEAKING POWER FUEL BUNDLE GENERATION RATE RATIO FUEL TYPE GE0 METRY (kw/ft) FACTOR t i Anittai u re 1.18 7x7 17.5 1.57 i ( 7D225 __

i .

Reload 1 1.18

! 1.57 7x7 17.5 70230

' ~ Reload-2 1.18 l

8x8 13.4 1.57 80262 .

/-

! , Reload 3 13.4 1.57- l'.18 8D250 8x8 - ~

Reload 4 13.4 -1.' 57 1.18-

+ 80219 8x8 i

~

A more detailed list of. input to each model and its source .is presented in Section II-of Reference 1. The core heatup calcu-

• This power level equals- the Appendix K requirement of'102%

at.102% of its maximum (technical specification). linear heat generation rate. * -

,. p A e

~

e t

e d

1. "- -n4-- ,y,-i

i:

c .,

, r -'... , .

[. ,

i.

• 1

.. +

i .

I 1  ;

t w i

t .

3 a

4 4

• REFERENCES .

4

{

• 4:

j 1. General Elcetric Company. Analytical Model for Loss-of-Coolant _ Analysis-y in Accordance with 10 CFR 50, Appendix: K,:' NEDE-20566 (Draf t) .. submitted August 1974. -

+

3 i 2. General Electric Refill Reflood-Calculation (Supplement to SAFE Code -

' Description) transmitted to USAEC by~ 1etter, G. L; Gyorey to V. Stello, 4

Jr., dated December 20, 1974.

'1 1

1 ,.

i .. .

i 4 .

i.

a

• +

1

=

t t- ,

i .*. . . . ,

4 1 It i

• e l .

i' .

e f:

.- , .9 .,.

+

t

• J g' .

'g

• e

- 9

3 , . . ..-

jl 4 . ,

^

, '._T v

s

,_ 6

'T .

.p

, ~.,

j *

+

-l' ,

a o,

4

- e ,

i. ' .!. . ,

g- , ,

• 8._

9__

0

, ,- -, . ., , , . . - - _ - , , c. . - - . ,u-,-.,

_. .~ .. > a . .- .m. . . . . , ,,

,  ?

~ .

l

.j 9

Bases:

j . .

4 1

1 l 3.11C Minimum Critical Power Ratio (MCPR) - ,

f Operating Limit MCPR *'

i

. The required operating _ limit MCPR's.at steady state operating i

conditio.ns as specified in Specification.3.11C are derived 1

from the established fuel cladding integrity Safety Limit MCPR-of 1.06, and- an analysis of abnormal operational- transients (1) . ,

For any abnormal operating transient analysis evaluation-with l

~

the initial condition-of the reactor being at the steady state

~

l operating limit,it is required that the_resulting MCPR does not

decrease below the Safety Limit MCPR _at any = time during the j -- .

! . transient assuming instrument trip' setting giv.en in Specifica .

O tion 2.3, .

1- . .

l 4 To assure that the fuel cladding integrity Safety-Limit 'is not

. exceeded .during any anticipated abnormal operational: transient,

{ the most limiting transients;have bee'n' ahalyzed { to determine 4 . .

which ' result in the lar' gest reduction in; critical p'over ratio ,

(CPR). -The type of-transients-evaluated were loss of flow, 1- . .

increase in pressure and: power, positive reactivity insartion, 4 i and' coolant temperature decrease. .

i . .

d l'

• s $

t -

4

. e .

1 i ,

.- . u. . . . . . . :. . - .. ;.... .,a.-....,,

- , - w. g e s. , <. , , , .,- , - , - - - - - - - - ~~ -- -.-- - -

j. '. j * -

. e , -.- . ,e ,

i i . . .

{

The limiting transient which determines the required steady 4.

i 3 state MCPR limit is tha turbina trip htth, failure of the i turbine bypass, This transient yields.tha laraest AMCPR{,

(

When added to the Safety Limit MCPR of 1.06 the reonired j minimum operating limit MCPR . of specificatton 3.11C. is obtainedi .

3 h '

i l'

i i

^

Prior to the analysis of abnormal operational transients an-

• 4 i * * .

l

- initial fuel bundle MCPR was determined. This parameter _is ,

l based on the bundio flow calculated by a GE multi-channel i .

steady state flow distribution model.as described in 1

Section 4.4 of NEDO-20360(3) and on core parameters shown in t

j

• Reference 2 (response to Items 2 and 9), ,

l The evaluation of' a- given transient begins with the system

i. . . .

! initial parameters shown ,in Table 6-l'(page 6-10) of Refere ence 1

• that are* input - to a CE -core dynamic behavior transient a

computer program describ'ediin NEDO-10802(3)--.- Also, the to the= transient j void reactivity coefficients that were. input calcu1Etional proce' dure are based on=a new method of calcula-

{ ,

tion termed NEV which provides'a better agreement between the.* <

j - . .

O

i. .

L

.- , , , _ . ,, , . , _ . , a ..a _,. - . . _. . . ,, ..-...__,.-......;.. . . ~ . - - _ - _ . _ . . _

. _ . - . , _._, _ _ . . _. ,;g ,,_ , , ,._ , 3 ; .. . - ,

m,,; m m .. ,. , , ,

J

-b l calculhted and plant instrument power distributions. The 1

4 3

a outputs of this' program .along with ' the t'nitial MCPR form '

the input for further analyses of the thermally limiting-bundle with the single channel transient thermal hydraulic

-!' . SCAT code described in NEDE-20566.(4) The principal re-1 a -

, sult of this evaluation is the reduction in MCPR caused by-2 2 .

the transient.

T j- .

J j ,,

i

.i l-f

• j- s

• g .

i i

9 e

4 * '

e a .

4 i-

.E.

• +

__j j

+

, , . .. -I

-1 t

. *l .

j, -i-

, J '.

• f'  :. . .

. e g

s . .

l. . . ,

?,.

4 .

-. ~ . ..2..~,... . _ . . . . . . .. ~.. ......,- _ -_ ,, ._. ...;.._.. . . . . . ., . _ . . _ . - . -

. . , . _ ,, .,w.p ..w, -

3,w , _.. .a. _. m=- - - - -

- a

's . * ,

- - I l

s 1 .

j MCPR Limits for Core Flows Other than Rated ,

4 .

i '

The purpose of the gK factor is to define ' operating limits 4

at other than rated-flow conditions. At less than 100% flow <.:

) ,

'the required MCPR is the product of ,the operating limit MCPR and the K factor.

f Specifically,'the K ffactor provides I the required thermal margin to protect against a flow in-crease transient. The most limiting transient initiated from less than rated flow conditions is the recirculation pump i

! speed up caused by a motor-generator speed control failure.-

For operation in the automatic flow control mode, the K g factors

[ .

assure that the operating limit MCPR of Specification 3.11C will i not be v'iolated should the most limiting transient occur at 1 -

less than rated flow. In the ' manual flow control mode, the Kg j ,

factors assure that the Safety Limit MCPR will not be violated j for the same postulated transient event.

i The Kg factor curves shown in Figure .}.11.2 were developed _

generically which ate applicable,to all BWR/2, BWR/3, and BWR/4 f -reactors. The K gfactors'were derived unine the flotr conerni 4

line corresponding to rated thermal power at rated core flow.

..i For the-manualiflow control mode, the K factora were calculated f

such that at the maximum flow state' (as limited by. the pump scoop i'

tube set point) an'd the corresponding core power (along the rated a

l flow control line), the limiting bundle's' relative power was -

e

~S 9 j . . .

a 4 e

+

--mp -

• rw y---v- e me im.-,ea u ,-< my,- + ways,w -etv- t 5 *g" Fu 9' F- S r -5'F"% -*-*"91*-tr e-p we-.,g-ate.as-e e

o

, -17 . ,

adjusted until the,MCPR was siightly above the Safety Limit.

Using this relative-bundle power, the MCPR's were calculated

(

i ,

at different points along the rated flow control line i

corresponding to-different-~ core flows. 'The ratio of the MCPR calculated at a given point of core flow, divided by the -

operating

• limit MCPR determines the Kg ..

, For operation in .the automatic flow control mode, the same 4

procedure was employed except the initial-power distribution

~

was established such that the MCPR was equal-to the operating limit MCPR at rated power and flow.

i The K g f actors shown in Figure 3.11.2, are conservative for i

the Monticello Nuclear Generating Plant operation because the .

operating limit MCPR's of Specification 3, llc are greater than the

{ original 1.20 operating limit MCPR used for the generic-derivation of Kg. .

i ,

4 g

e *

, . e s .-

Y e a

d i . .

s .. .

t

, 'b e

9 q

) ,

, , , , , s. * --

[

e. . . . . -

, e .

e i . ,

4

- a .. . ,..;.-. ,  :- . -- . . . . . - . -

, y .. _ ._, _ ,.._. . _ .

7_

.-7.

. A

. .. I'  :

i-j.

p ,

I J .

4.11C Minimum Critical Power Ratio (HCPR)

- Surveillance Reouirement -

l At core thurmal power levels less than or equal to 25%, the

. _ reactor vill be operating'at minimum recirculation pump speed and the moderator void content v111'be very small. For all f- .

j ,

designated control rod patterns which may be employed at this l

-point, operating plant experience indicated that the resulting MCPR value is in excess of--requirements by a considerabic j margin. Uith this low void' content, any inadvertent core flow V

increase would only place operation in a more-conservative mode i ,

relative to MCPR.. During initial start-up testing of the plant,

! a MCPR evaluation vill be made at' 25% thermal power level with minimum recirculation pump spedd. The MCPR margin' will thus

be demonstrated such that future.

tCPR evaluation below this 1

l power 1, sel vill be shown to be unnecessary. The daily rc~

i -

l ,

quirement for calculating MCPR above 25%. rated thermal power is 1

sufficient since power distribution shif ts are very slow when -

there have not been significant- gewer or control rod changes.

i The requirement for calculating MCPR when-a limiting control:

e' rod' pattern is app ~ roached ensures that MCPR will.be known i following a change in power.or power shape (regardless of 5 --magnitude) that could place' operation at a. thermal limit.

• - . I l'

~

}

j

. -t

. 1 "

'~ '

. . L

. i.

. . e

'g

~

- - e v , y ,,<---w , .~r4 + e +~y,. .e- ---c.e,v,,ce--- .-*+ww-newH++* -*,. wee ~