ML19344D214

Revised Pages to Tech Specs Incorporating New Max Average Planar Linear Heat Generation Rate to Sections 2.1.a.1,3.10 & 4.10 of App a

ML19344D214
Person / Time
Site:	Oyster Creek
Issue date:	03/04/1980
From:	JERSEY CENTRAL POWER & LIGHT CO.
To:
Shared Package
ML19344D209	List: ML19344D208 ML19344D212 ML19344D214 ML19344D216
References
NUDOCS 8003110548
Download: ML19344D214 (11)

Text

. .

2.1.1 SECTION 2-

_ SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMIT - FUEL CLADDING INTEGRITY Applicability: Applies co the interrelated variables associated with fuel thermal behavior.

Objective: To establish limits on the important thermal hydraulic variables.to assure the integrity of the fuel cladding.

Specifications: A. When the reactor pressure is greater than 600 psia, the combination of reactor core flow and reactor thermal power to water shall not exceed the limit shown on Figure 2.1.1 for any fuel type.

A.1 Figure 2.1.1 applies directly when the total peaking factor is less than or equal to the following:

Fuel Type IIIF

a. Axial peak at core nidplane or below of 2.74

b. Axial peak above core nidplane of 2.50 For 8 x 8 Fuel

a. Axial peak at core midplane or below of 2.78

b. Axial peak above midplane of 2.61 A.2 For total peaking factors greater than those specified in Specification 2.1.A.1, the safety limit is reduced by the following:

PFo

, SL = slo x PF where: SL = reduced safety limit SLo = safety limit from figure 2.1.1 PFo = peaking factor specified l B. When the reactor pressure is less than 600 psia or reactor flow is less than 10 percent of design, the reactor thermal l - power shall not exceed 354 Mwt.

l C. The neutron flux shall not exceed its scram setting for longer than 1.75 seconds.

Anendment No.jd',' 24 800311g f g

_ _ - , - _ =

3.10-1 3.10 CORE LIMITS Applicability: Applies to core conditions required to meet the Final Acceptance Criteria for Emergency Core Cooling Performance.

Objective: To assure conformance to the peak clad temperature limita-tions during a postulated loss-of-coolant accident as speci-fled in 10 CFR 50.46 (January 4, 1974) and to esaure confor-mance to the 17.2 KW/ft. (for 7 x 7 fuel) and 14.5 KW/ft.

(for 8 x 8 fuel) operating limits for local linear heat generation rate.

Specification: A. Average Planar LHGR During power operation, the average linear heat generation rate (LHCR) of all the rods in any fuel assembly, as a function of average planar exposure, at any axial location shall not exceed the product of the maximum average planar LHGR (MAPLHGR) limit shown in Figures 3.10-1 (for 5-loop operation) and 3.10-2 (for 4-loop operation) and the axial MAPLHGR multiplier in Figure 3.10-3. If at any time during

. power operation it is aetermined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated to restore operation to within the prescribed limits. If the APLHGR is not returned to within the prescribed limits within two (2) hours, action shall be initiated to bring the reactor to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. During this period surveillance and corresponding action shall continue until reactor operation is within the prescribed limits at which time power operation may be continued.

B. Local LHGR During power operation, the linear heat generation rate (LHGR) of any rod in any fuel assembly, at any axial location shall not exceed the maximum allowable LHGR as calculated by the following equation:

LHCR < LHGRd[l- max ( )]

Where: LHGRd = Limiting LHGR oP = Maximum Power Spiking Penalty P

LT = Total Core Length - 144 inches L = Axial position above bottom of core Amendment [, )'[, 39

~

3.10-2 Fuel Type LHGRd aP/P IIIF 17.2 .033 V 14.5 .033 VB 14.5 .039 If at any time during operation it is determined by normal surveillance rhat the limiting value for LHGR is being exceeded, action shall be initiated to restore operation to within the prescribed limits. If the LHGR is not returned to within the prescribed limits within two (2) hours, action shall be initiated to bring the reactor to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. During this period, surveillance and corresponding action shall continue until reactor operation is within the prescribed limits at which time power operation may be continued.

C. Assembly Averaged Power Void Relationship ,

(Applicable to Fuel Type IIIF for 4-loop Operation Only) l During power operation, the assembly average void fraction and assenbly power shall be such that the following relation-ship is satisfied:

I-VF

(

PR X FCP

) --> B Where: VF = Bundle average boid fraction PR = Assembly radial power f actor FCP = Fractional ccre power (relative to 1930 MWt)

B = Power-Void limit The limiting values of "B" for fuel type IIIF is .377.

D. Mininum Critical Power Ratio (MCPR)

During steady state power operation, MCPR shall be greater tha; or equal to the following:

ARPM Status MCPR Limit

1. If (ny two (2) LPRM assemblies which 1.64 are ic.put to the APRM systen and are separated in distance by less than Amendment No. , 7 39

• 3.10-4 4

affect the calculated p~eak clad temperature by less than

+ 20*F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are below the limits specified in 10 CFR 50.46 (January 4, 1974).

The maximum average planar LHGR limits shown in Figure 3.10-1 -

for Type IIIF, V and VB fuel for five loop opera *on and in Figure 3.10-2 for Type V and VB fuel for four loop operation are the result of LOCA analyses performed by Exxon Nuclear Company utilizing an evaluation model developed by Exxon Nuclear Company in compliance with Appendix K to 10 CFR 50 (2). Operation is permitted with the four-loop limite of Figure 3.10-2 provided the fif th loop has its discharge valve closed and its bypass and suction valves open. In addition, the maximum average planar LHGR limits shown in Figures 3.10-1 and 3.10-2 for Type V and VB fuel were sna-lyzed with 100% of the spray cooling coefficients spec.fied in Appendix K to 10 CFR Part 50 for 7x7 fuel. These spray heat transfer coefficients were justified in the ENC Spray Cooling Heat Transfer Test Program (3).

The maximum average planar LHGR limits shown in Figure 3.10-2 for Type IIIF fuel for four loop operation is the result of LOCA analyses performed by Exxon Nuclear Company utilizing blowdown results obtained f rom a General Electric Company evaluation model in compliance with 10 CFR 50, Appendix K(1).

Single failure considerations were based on the revised Oyster Creek Single Failure Analysis submitted to the Staff on July 15, 1975.

The effect of axial power profile peak location is evaluated for the worst break size by performing a series of fuel heat-up calculations. A set of multipliere is devised to reduce the allowable bottom skewed axial power peaks relative to center or above center peaked profiles. The major factors which lead to the lower MAPLHGR limits with bottom skewed axial power profiles are the c.hange in canister quench time at the axial peak location and a deterioration in heat trans-fer during the extended downward flow period during blowdown.

The MAPLHGR multiplier in Figure 3.10-3 shall only be applied to MAPLIIGR determined by the evaluation model described in reference 2.

The-possible. effects of fuel pellet densification are : 1) creep collapse of the cladding due to axial gap formation;

2) increase in the LHGR because of pellet column shortening;

3) power. spikes due to axial gap formation; and 4) changes in stored energy due to increased radial gap size.

. Calculations show that clad collapse is conservatively pre-dicted not to occur during the exposure lifetime of the fuel.

Therefore, clad' collapse is not considered in the analyses.

Amendment No. )[, f, 33

3.10-5 Since axial thermal expansion of the fuel pellets is greater than axial shrinkage due to densification, the analyses of peak clad temperature do not conrider any change in LHGR

. due to pellet column shortening. Although the formation of axial gaps might produce a local power spike at one location on any one rod in a fuel assembly, the increase in local power density would be on the order of only 2% at the axial midplane. Since small local variations in power distribution have a small effect on peak clad temperature, power spikes were not considered in the analysis of loss-of-coolant ,

accidents.

Changes in gap size affect the peak clad temperatures by their effect on pellet clad thermal conductance and fuel pellet stored energy. Treatment of this effect combined with the effects of pellet cracking, relocation and subse-quent gap closure are discussed in XN-174. Pellet-clad '

thermal conductance for Type IIIF, V and vb fuel was calcu-lated using the GAPEX model (XN-174).

The specification for local LHGR assures that the linear heat generation rate in any rod is less than the limiting

. linear heat generation even if fuel pellet densification. I is postulated. The power spike penalty for Type IIIF, V and

• VB fuel is based on analyses presented in Facility Change ,

Request No. 5, Facility Change Request No. 6 and Amendment No. 76, respectively. The analysis assumes a linearly increasing variation in axial gaps between core bottom and top, and assures with 95% confidence that no more than one fuel rods exceeds the design linear heat generation rate due to power spiking.

The General Electric non-jet pump BWR BCCS model (1) utilizes an empirical correlation to determine the duration of nucleate boiling heat transfer in the early period following the pos-tulated pipe break. This correlation for time to dryout is found to be proportional to the ratio of assembly water volume

' to power. Dryout time is a significant parameter in determin-ing the extent of nucleate and transition boiling heat transfer, and consequently the peak cladding temperature.

By maintaining reactor power and void fraction as specified in 3.10.C, dryout times at least as long as that used in the LOCA 4

Amendment-No. )fI, 33 i p

3.10-6 analysis will be assured. The limiting value of B in Specifi-cation 3.10.C was developed for core conditions of 100%

power and 70% flow, the minimum flow that could be achieved

.without automatic plant trip (flow biased high neutron flux scram). Such a condition is never achieved during actual operation due to the neutron flux rod block and the inherent j reactor powerflow relationship. The MAPLHGR results for fuel type IIIF shown in Figure 3.10-2 were evaluated for 102% power l and 70% flow, thus the 2% conservatism for instrument uncer-tainty is retained in the limiting value of B. Additional s i

conservatism is provided by the following assumptions used in determining the B limit.

1 1. All heat was assumed to be removed by 1 octive channel flow. No credit was taken for heat removal by leakage j flow (10% of total flow).

I

2. Each fuel type was assumed to be operating at full ther-mal power rather than the reduced power resulting from the more limiting conditions imposed by Figure 3.10-2.

The loss of coolant accident (LOCA) analyses are performed using an initial core flow that is 70% of the rated value.

The rationale f or use of this value of flow is based on the possibility of achieving full power (100% rated power) at a reduced flow condition. The cagnitude of the reduced flow is limited by the flow relationship for overpower scram.

.The low flow condition for the LOCA analysis ensures a con-l servative analysis because this initial condition is assoc-iated with a higher initial quality in the core relative to higher flow-lower quality conditions at full power. The high quality-low flow condition for the steady-state core operation results in rapid voiding of the core during the blowdown period of the LOCA. The rapid degradation of the coolant conditions due to voiding results in a decrease in the time to boiling transition and thus degradation of heat

^

transfer with consequent higher peak cladding temperatures.

Thus, analys.s of the LOCA using 70% flow and 102% power provides a conservative basis for evaluation of the peak cladding temperature and the maximum average planar linear heat generation rate (MAPLHGR) for the reactor.

C The minimum critical power ratio (MCPR) calculated for the initial conditions of the LOCA represents the thermal margin

'of the hot assembly to the boiling transition point. An increase in core flow from 70% would result in additional

- thermal margin (higher MCPR value). The conservative ECCS analysis bounds the range of permitted reactor operating conditions so long as operating MCPR's are above the values Amendment No. 33 l

, . m e,. +- .v. . -- .a-e

• c==

I 3.10-9 ,

I .

o .

O .

. ... . . . . . . . // og i .

. . . i. . . . . . // . .

e i, . .  ;

. .. i . i

..i .. ..

i

. . . - ,-i . .

. . . 4J,

. -i . . i .

. . , . . , i

. _ . . i iff.... i e

.4 i .//.. , . ,

. . . . . i i i i i .i.. . i e i

. , _ . i,i.. , . . . ... ... _ ff.....! . . i . ,

t . i1; . . . 4 i.,,, .1 . . i /f. ! . . i e l . i . 6

. . I!!, j i i e i i . . . t 4 i ./ f i e i . , 1 i i ii . 6 > >

. // . . . . i W4 i . . . , . , g . . i . , g //i , . i . M i ' i i . 6

. . . , ,i . . . . g . e i . . . i > // 4 i e t . ! ie . . , e i H + . , i . e . t i .J/ .. . t i i i i i i ,i. i i , , . i , . 0 i

,- i 4 r

U . . , 4

, i i i . . i6  !,,

6

. 6 . .- y I ff

, i i n -

gg e i a  ! i e i .  ! e i6 i . ,

x . if , , i i p . ... W  ! . , .

g

. . . . i e . . i . p f.. .  ! F_ ii 1-* i ! , , . i a

. . , e i '

ti . . i  ; . , . . . .,  ! 4 , i 4 t . ,

i . i . . 4 1,if- . . . . . . i i i . i e i I

. .f i .

o

• .  ! I , , n

. . . . , .fr ,. . . . . r. . ' . . 6

. . i . .

6i _

f , i e i. , i .

...m.

. . i

. i . . . . ,

.f i . .

. . r . .f. i :1 ,

i m

. . . , . t i i . . , i i gg fi

.g . . i e t i j . . . . .

ZW . i i f . I e u)  %

< F* f . t,. . N 3

• i !l # i * . . .

6!

% ^

f . i, i g , , e i _f. . t f .

' . i I . V f.d = 0 . i. - 1 6 oOM . .. , . ,f f. 1 .

M b . .  !, ./ .. . .# j . . t . .

g H . . . . . ./. . . . . f. f . . . , . . . . ! 3 g

O > 42

' ' . ' ' I' i ' . 'l' t t . m

- <wm . . i . . e f i . .' . i f1 '. i . i , , , o M

W*;

=0 . . f . i # I. . O N

g 4 . f g j W

g A0%

g . / . #6 i . i . .e g 4g . . . . . i i , ... -# #-- .. . . . , ,

p 4 . . . 4 . . . .. . . . i . f ... . , i . . i e i 3< . ! ./ l

. . . . i C C h3 . . * '

.I i . . f- . . .

>* 4 ;"': M . . . / . . .l. j e ... . . . . . . i

% 4 > .

.! , >p . . ... . . . # # 1 .i.., . . , . . g 4gg . .. . . . 6 i i . . . . p , f. . . . . . . . i . . . ,

• =

4 W r.g V i . . . . . I i ii . .I t

./

, .,i . . f !1 , , .. , . . . . i . . . i W Z . pi . r . . . v1 0 hM

. . . .f. i

.n . ./ , i f. / . . . . . . . . ie 4 M3 . . . .. .

./ i . .... . i . . . / - I. . ... . . i . . 1 . 4 .

i e i . ,. . . / . # . I i. . . i . p / , i i., . , i . , , , , 4 4

i . . . if; . . i .i& . j i

../ sf. a j i i . . i e i i , I i

. . i .

. i ,

. 4 4 .

,f

• /. .

i e s

. . i !

. i i .

'. . i .

4

. . .ff ..,

f.

i !I..i

f. .. .

i i

.,ig

. i . .

j

, i , .

. Ii i  ;

. fl i . . . . 4 ;f . i . . . i . . .f. ,p. i . . . : i . ,  ; , .

~~~

, . . . i t i f. t i t i ! . . i , i Ia 3 . .  ! + a i i ("'

. e i 6 . i .. I i ;.f. '. . . . ,

i . . i o r . . . i 1 . i e i . . . i i !I .. .=

• i . w 6 . . i e i . )i , i i i i. . i ,

i .I t/i , ,a . . i . . e i 4 i t . . I ! . . i , I i.iii i i , i :.  ! 6 f f , .f . i i' !ie i e i i i

. . 4 . . . . i .i,.i . i ; . . , p.;e i i, . . . i . i i i i I . t , i . , ,. t. 1.e , .  ! J;fie e . , , i . . e i i i i . , i . . i e 6 i ...,, . . 4 .. .  ! if. . . . . . . i . . . . e i

. . . i . i . . . . '. 4 i . i i.1 . : i , . . ! . e . , i...ie 6 i i . 1 i . I e . 6 , i . , . i . . , . . . . i . }4 . . i , i

• , . . . . I i . ii i t i ei.  ! i , . ! . . . , , !i..i, i

. . . . . i . .i , i . 6 . . . .. . . t , , Wi

. .. . . . . . , , , e , ,e

,i . , . . . . . , , . . . . , . , i . . ,

. . . . i e i i , . . i i i i i , . . . . .i.. i . i i . . . . i . ,

,i . ! I i . i e tiii .

1 i . i . . .,ii...  ! i . . i . , 4 i

. . 8 e i , t e , f i ! I I ' I

! i i I . , *.,! i . t i , i I . I i i . i . . . . . , . _ . . . , .

i i i i . , . i i i i . i . i i .  ! . 4 r . 1 i i . , . . i . , . I i i i 6 . I e

. i . , i i i , .

... . . t + . ie i i i . # . e i l . .  ! . i i . . i i i i i.. . i+ i

• 4 .  !. i f , i -i -i

... . . i ' . . . . . 3 . i . . . ,. . t . , , , , ., . , i

! . . I i 6 Y,ili . i i i i i i i i i ! i i . 4 1 i , I i i ii . . 4 i i , !

~ O i

k-.i

. . 6 . . i N . . . , . . , 6 i , , . .

w w .-e .-e e e

(.II/tDO IDH'I WV!WId 30VH3AV HnNIXVH

' C FIGURE 3.10-2 . .

MAXIMUM ALLOWABLE AVERAGE PLANAR LINEAR llEAT GENERATION RATE (FOUR LOOP OPERATION) 14 _ ., . _ _ _ . _ ._ , . _ _ _ _

,g _ _ _ _ _

.g .;

}

13

^ -_ .. _ . . . . . .. . . . . . . _ . . . . _ _ _ _

gs . . . ... .. . . . . .. .. . _ . . .

12 - - - - - - ---

12

• 3 . _ . . . . _ _ _ . _ _ _ . .

Q .

j.. - . . . . . _ . - 2. . . .

, -s s . .

y y ..

\ g . - .-- . . . .. _.. _ _ . _ .. .

/ ..\ % _ . . . _ . _ . .

Q .- . . ._ . . . . . . . . . . . . .. .

o g

11 _ . .. _ . .. _ ... _ .. _._. .

, / .. . . ._ . . . . .

-N ._----- -

s Type HIF '-~ . .

r_ .

.4

( .-. . . . . . ~_~. .

s . .

' ' ' - ~

~ ~ ~ ~

N Type VB --  :

x 1_ ~_T .~ .~~. .. . s .____ _ - .

zw a

p

_ . 5

_ s 10

_-----m;;;r m _

---

. s _ .

_m;_N Type V .

Ne _ . . .

... 4 N

g '

r

:  : .

T . . . . T.. _ _. .

9 . _ _ _ . . _ __. _ . . .. _ . . _ _ .

. . _ . . . . . _ . _._.. . . . _ _ .. _ .i

.. ..d a

_ . . e . .e

_ e e e .

. _ .g e

_ . . _ . . .. ..,_ e g . ._ e

. e.

.. _ . pg.e_

p Q

8 r 8 0 5 10 15 20 25 30 35 40 .

g AVERAGE PLANAR EXPOSURE (GWD/MTM) . .

t 9

FIGURE 3.10-3 AXIAL MAPL 11GR MlILTIPLIER l :_- :~ . .  : _ .: :

- ~

~ - ~ ~

~

~ ' -

^

1.0 e-

... - . .. _. . .- - . _ _ . . . . . . _ . . , s ,_ . . . _ . . . . . _.. . . .

7 _ .

M _.. . . .-

/

p g

/ .

- .7 m . _ .. _ . -

0.9 _/ - .... .

s '

E-4 .

e...

7 _ _-.

g _ . _.._ _ _ _ _. .._. . . . . . . .. . . .

o 0.8 ...__.- . _ __ -. . . . _ _ _ _. . . .. _ _.. . _ .. . . . . . . . . . . . _.. ._ _. . .._ .._ ._,

. 0.7 ~

0 2 4 6 8 lo 12 e

V CORE HEIGHT (FT)

~

L O

I

. ~

W e

8 i

1

4 10-1 4.10 ECCS RELATED CORE LIMITS Applicability: Applies to the periodic measurement during power operation of core parameters related to ECCS performance.

Objective: To assure that the limits of Section 3.10 are not being violated.

Specification: A. Average Planar LHGR.

The APLHGR for each type of fuel as a function of average planar exposure shall be checked daily during reactor operation at > 25% rated thermal power.

B. Local LHCR The LHGR as a function of core height shall be checked daily during reactor operation at > 25% rated thermal power.

C. Assembly Averaged Power-Void Relationship (Applicable to Fuel Type IIIF for 4-Loop Operation Only)

Canpliance with the Power-Void Relationship in Section 3.10. will be verified at least once during a startup between 50% and 70% power, when steady state power operation is attained and at least every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> thereafter during power operation.

D. Minimum Critical Power Ratio (MCPR).

MCPR and APRM status shall be checked daily during reactor operation and > 25% rated thermal power.

Basis: The LHGR shall be checked daily to determine whether fuel burnup or control rod movement has caused changes in power distribution. Since changes due to burnup are slow, and ,

only a few control rods are moved daily, a daily check of power distribution is adequate. l The Power-Void Relationship is verified between 50% and .

70% power during a startup. This ' single verification l during startup is acceptable since operating experience has I shown that even under the most extreme void conditions encountered at lower power levels, the relationship 'is not violated. Additionally reduced power operation involves less stored heat in the core and lower decay heat rates which would add further margin to limiting peak clad tem-peratures in the event of a LOCA.

Amendment No. , 33

4.10-2 Verification when steady state power operation is attained and every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> thereafter is appropriate since once steady state conditions are achieved, the void fraction, radial peaking factor, and power level that combine to form the relationship are unlikely to change so rapidly to result in a significant change during.that period.

The minimum critical power ratio (MCPR) is unlikely to change

-significantly during steady state power operation so that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is an acceptable frequency for surveillance. In' the event of a single pump trip, 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> surveillance interval remains acceptable because the accompanying power reduction is much larger than the change in MAPLHGR limits for four loop operation at the corresponding lower steady state power level as compared to five loop operation. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> frequency is also acceptable for the APRM status check since neutron monitoring system failures are infrequent and a downscale failure of either an APRM or LPRM initiates a control rod withdrawal block thus precluding the possibil-ity of a control rod withdrawal error.

At core power levels less than or equal to 25% rated thermal power the reactor will be operating at or above the minimum recirculation pump speed. For all designated control rod patterns which may be employed at this point , operating plant experience and thermal hydraulic analysis indicate that the resulting APLHGR, LHGR and MCPR values all have considerable margin to the limits of section 3.10. Conse-quently, monitoring of these quantities below 25% of rated thermal power is not required.

l i

l Amendment' No.J16 l