ML20236R863

Reload Safety Evaluation,Beaver Valley Nuclear Plant Unit 1 Cycle 7

ML20236R863
Person / Time
Site:	Beaver Valley
Issue date:	10/31/1987
From:	Dzenis E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20236R843	List: ML20236R840 ML20236R863
References
NUDOCS 8711240070
Download: ML20236R863 (18)

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' RELOAD SAFETY EVALUATION m

, BEAVER VALLEY NUCLEAR PLANT

'i >.- UNIT'1 CYCLE 7 J , '

" ' October, 1987 4

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' Edited.by: B. W. Gergos 1

e g-l Approved: 80 72 x

M' E. A. Dzenis,' Manager 44 ' Core Operations Nuclear Fuel Division 1

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TABLE'0F CONTENTS v , n. , '

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1.04 ; INTRODUCTION;AND.

SUMMARY

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Introduction:

1 1;2'GeneralLDescription-1-

'T L1;3 Conclusions' 2

~'2.0 . REACTOR DESIGNj 3 o

L 2.14 iMechanical' Design , , 3 2.2; Nuclear Design

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3.0 POWERz CAPABILITY.AND ACCIDENT EVALUATION 5'

~ 3.1L-Power Capability 6

~3.2 -Accident Evaluation 6

--' Kinetic, Parameters 6

' : 3. 2.1. -

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3.2.2 -Control Rod Worths 7 3.2.3 Core Peaking Factors 7

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!4.0: TECHNICAL SPECIFICAT: ION CHANGES 8

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511 Fuel: Assembly Design Parameters 10 s

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'2. '. f - ' Kinetic- Characteristics 11

- c: : 3 ' . Shutdown Requirements and Margins ~ 12 ty s , c;

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a 1.0^ INTRODUCTION AND

SUMMARY

hp w 1.1- ' INTRODUCTION This report documents the results of an evaluation for Beaver Valley Unit 1,

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' Cycle 7, which demonstrates that insertion of reload fuel _ into the core will not adversely affectLthe safety of the plant. This evaluation was accomplished utilizing the NRC approved methodology described in WCAP-9273-A,

" Westinghouse Reload Safety Evaluation Methodology" (Reference 1).

Based upon the above referenced methodology, only those incidents analyzed and reported in the FSAR (Reference 2) as well as safety evaluations for 10% Steam Generator Tube Plugging (3) and Upflow Conversion / Thimble Plug Removal (4) ,

which could potentially be affected by this fuel reload, have been reviewed for the Cycle 7 design described herein. The applicability of previous results has been evaluated.

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. 1.2 GENERAL DESCRIPTION

. The Beaver Valley Unit i reactor core is composed of 157 fuel assemblies arranged in the core loading pattern configuration shown in Figure 1. The Cycle 7 core configuration features a low leakage pattern. During the Cycle 6/7 refueling, all twelve (12) of the Region 6 assemblies and sixty (60) of the Region 7 assemblies will be replaced with seventy-two (72) Region 9 assemblies (Region 9A, 98). One Region 1 assembly will be replaced with another~ Region 1 assembly. A summary of the Cycle 7 fuel invento-v is given in Table 1. j l'

. Nominal core design parameters utilized for Cycle 7 are as follows:

Core Power (MWt) 2652 3 4 System Pressure (psia) 2250 x Core. Inlet Temperature (*F) 542.5 l 1

Core Average Temperature ('F) 580.2 j'

... Thermal Design Flow (gpm) 265,500 Average Linear Power Density (kw/ft) 5.19 Vessel Average Temperature 576.2 s7a m e-s m 2 1

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UV T 1>.3. CON'CLUSIONS3 ze ok -From the evaluation, presented'in this report, it-is concluded that the Cycle 7 design;does.not:cause the previously' acceptable' safety limits.for:any incident a" .

to!beaxceeded.LTheseconclusionsarebasedonthefollowingassumptions:

, f:"w i , ?1.-lCyb1'e'6burnupisbetween12900and14900 MWD /MTU.-

2. ' Cycle'7 bur'nup isilimited to the end-of-life full power' capability

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( ,, J1000 MWD /MTUpower/coastdown.. i

3. There _is adherence to plant operating limitations given-in the Technical Specifications.  ;

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2.0' REACTOR DESIGN y,

Y 0 2.17 MECHANICAL DESIGN l

F The mechanical:designiof the 72 Region-9 fuel assemblies is the same as the

Region 8 assemblies with the~ exception of the features detailed in Table 1 g,

O hich' compares-pertinent design parameters of the various-fuel regions. The L  : fuel pellet' stack holddown springs in Region 9 fuel rods exert a minimum 4G

- axial force :instead of a minimum 6G axial force in the other regions. The ,

j pellets _in the.. Region 9. fuel have chamfered edges. A total of 8448 Integrated Fuel Burnable Absorber;(IFBA) (Reference 5) rods are being' introduced into the Cycle' 7Leore; ; These rods are the same as the other Region 9 fuel rods except ifor the addition of a thin boride coating .on the cylindrical surface of the 1 enriched fuel.' pellets along the central portion (120 inches) of the fuel stack l l length. :In order to offset the effects of the He gas release from the IFBA 7 coating'during irradiation, a lower initial He backfill pressure is used in b the_IFBA rods compared to the non-IFBA fuel rods. The Region 9 fuel will also 1 . have six inch axial blankets of natural uranium (Reference 5). The natural

' pellets are.of<the'same design as the enriched and IFBA pellets except for the

-absence of a dish on the natural pellets. The Region 9 fuel has been designed according tofthe fuel performance model in Reference 6. The fuel is designed

' to operate so that clad flattening .is not predicted to occur, when the fuel is

• .analymd with the Westinghouse model (Reference 7). The fuel rod internal L ' pressure design basis, Reference 8, is satisfied for all fuel regions.

Westinghouse's experience with Zircaloy clad fuel is described in WCAP-8183,

'" Operational' Experience with Westinghouse Cores," ' brence 9, which is L updated annually.

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Jconditionsfisi+ 7 0 Table'2 summarizes'the current, limits for kinetics gf " . ' 2

, characteristics which'arelbasedLon previously submitted accident analyses.

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[ [ , }The Aoderato temperature' coefficient (MTC) at'all rods out (AR0), hot zero:

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• power]HZP)(no xenon conditionsLcould be measured positive near beginning of l .

(life !(BOL)'.t. The e MTCLatiHZP may.be kept negative by. imposing. administrative-.

M vy" contrditon D bank insertionL(i.e'. rei withdrawal limits)'.- This-assures g6 ,

consistency wittnthe" existing safety analysis-and operat* ion in'accordance with i

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,IFBAfuelwasTevaluandtoLdeterminethapotentialimpact'ontheFSAR'Large.

I' 'Br$akLLOCA analysis.,Thef IFBA fuel rods have been shown, by analysis,

\.g . (ReferencesD10 and 11) to'be less limiting than,the FSAR analysis of-record due to anl inherent powerTdensity reduction caused by the neutron poisoning and fluxidepression,offtheabsorber. . IFBA' power density reduction input to the

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Emergency Core Cooling System'(ECCS) evaluation model has;been verified to be

[M/^ fcons'ervativefrelative to'theiCycle 71 core.. The limiting fuel type assumption E jfor~ FSAR.ljarge[ Break LOCA continues to. apply:and the IFBA fuel rod is bounded

', by th's FSARLanaljsis of record. Thus, operation of the Cycle 7 with IFBA fuel JK >

E, meets the: requirements of'10CFR50.46 and Appendix K to 10CFR50.

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\ Cycle 7 control rod worths and requirements'are compared in Table 3 with those M [3for' Cycle 6 at the most limiting condition (end-of-life).. The available l 3 t

shutdown l margin exceeds.the minimum required margin.

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'h .4 lTheElo^ading pattern for Cycle 7.is shown in Figure 1. It contains 8448 IFBA k< ' rods located in.64 fuel assemblies. Two secondary' sources, retained from the L

LCycle 6l core,'are located in positions H3 and H13.

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3, p r2.3.LTHERMAL'AND HYDRAULIC.DESIGNL

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d theconditEnsgiven;i5Section1.0.

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.TheNhermal and hydrauliclevaluation"of the~ Cycle 7 reload. assumes- the. removal 1Sflallithimble'pluggingdedcas'from.thecore. The major = impact of thimble f . '[

l plug [remosliis?a' reduction in the core effective flow rate due to an increase

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' 'linjcore-bypassflow,(fromL4.5%;to~6.5%oftotalreactor' flow). This change

_ ,has'besn evaluated ^and found to be acceptable as' documented.in Reference 4.

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Gi _ m, Cinthe;FSARfsothatitheDNBidesigncriteria.(References 2and10)aremet

'4 without? changing;thscorelDNB. limits.

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c 3.0 POWER CAPABILITY AND ACCIDENT EVALUATION

• 3.1 POWER CAPABILITY The plant powor capability is evaluated considering the consequences of those  ;

incidents examined in the FSAR using the previously accepted design basis. It i is concluded that the core reload will not adversely affect the ability to safely operat,e at design power levels (Section 1) during Cycle 7. For the overpower transient, the fuel centerline temperature limit of 4700*F can be accommodated'with margin in the Cycle 7 core. The time dependent densification model (Reference 13) was used for fuel temperature evaluations.

The LOCA. limit at rated power can be met by maintaining gF at or below 2.32 according to the normalized Fg envelope (Figure 2 ).

3.2 ACCIDENT EVALUATION The. effects of the reload, including-mechanical design changes described in Section 2.1, on the design basis and postulated incidents analyzed in the FSAR (Refe'rence 2) were examined. In all cases, it was found that the effects were accommodated within the conservatism of the initial assumptions used in (1) the previous applicable' safety analysis, (2) the safety evaluation performed in support of the 10% steam generator' tube plugging level (3)and (3) the safety evaluation performed in support of thimble plug removal and upflow

. ' conversion (4) .

A core reload can typically affect accident analysis input parameters in the following' areas: core kinetic characteristics, control rod wortns, and core l peaking factors. Cycle 7 parameters in each of these three ereas were l examined as discussed in the following subsections to ascertain whether new l

> accident analyses were required. l K

- 3.2.1 KINETICS PARAMETERS

• Table 2 is a summary of the current limits for kinetics parameters. All the  ;

l Cycle 7 kinetic values fall within the bounds of the current limits.

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13.2.2? .. CONTROL ROD WORTHS:

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W JF ' Changes infcontrol rod worths may1 affect differential rod worths, shutdown y- marhin,Lejected rod ' worths, and. trip reactivity. . Table 2 showsLthat the A, smaximum differentialtrod worth'of two.RCCA control; banks moving together in .

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9,, u their; highest woftiregion'far:Cscle 7 meets the current limit. Table3 shows

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...;  :; Peaking factors 4 following control fod ejection are within the bounds'of the r a Lcurrent: limits. jThe peaking. factors for:steamline' break have been evaluated

-n ..and?arewithinitheboun'dsoftheprevioussafetyanalysislimits.

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1 - 5.0 = REFERENCES (

1. k q.

w,

> [1. (Dayidson,lS. L.,L(Ed.), Let; dal.,L"Westingh'ouse; Reload Safety Evaluation M' C,... i Methodology,.".WCAP-9273.A,: July 1985..

~

" _ ;;g; y '

2 2 Beaver Valley UnitLNoUliFinal Safety Analysis! Report," Docket Number '

a "

50-334;-

m's 4

o- , . . . . .

n .

s(%  % , 30Jan ..YCA.,;(etE al.). -" Beaver. Valley' Unit 1,101 Steam Generator Tube-

?, ,

! Plugging Licensing Report," WCAP 11591, September 1987.

y .

', r - L4..-(TolbeLissue'd)i"UpflowlConversionSafetyEvaluationReportforBeaver

%;R -.  ; Valley:Unitil,"gWCAP 11639,fOctober 1987.;

, ,f. , ,

/' ~ l5.LDavidson,iS;LJand.Kramer,-W.R.-(Ed.),"ReferenceCoreReportVANTAGE5

~

l

Fuel Assembly,"~WCAP ,10445-NP-A, September 1985.

9 f.g~ .

, M' 6J Miller,"J.iV.,{(Ed.),-.l." Improved Analytical Model used in. Westinghouse

, Fuel-Rod Design Computations,"?WCAP-8785, October 1976.

?7. George,:RE A.,i(et."al.-),;" Revised Clad Flattening Model," WCAP-8381, July

, -1974.H i

X ' '

K8.MRi M r(D. H., (et. al'.), " Safety Analysis for the Revised Fuel Rod

-Internal' Pressure' Design Basis," WCAP-8964, Jun'e.1977.

n ', '

9.: Skditka, J. ,clorii, 'J. A. ,;" Operational. Experience with Westinghouse ol 1

' Cores,"LWCAP-8183,: Revision 15,; July,.1987.

+

4

_;10..Davidson,;S. L., (Ed.)', Led.-al.,~" Extended Burnup Evaluation of

.Wwestinghouse: Fuel," WCAP-10126-A, December 1985. ,

. c V,,

.. .;1 i

Davidson, S. ;L., (Ed.)',' ed. ali,. " Reference Core Report. VANTAGE-5 Fuel

=11'.;LAssembly;"~ WCAP-10445-A', September 1985.

- ~ .

c12.iletter from EL P. Rahe,-Jr. (Westinghouse) to H. Berkow (NRC),

NS-NRC-86-3116,. dated March 25, 1986, Westinghouse Response to Additional

-Requestion WCAP-9226-P/WCAP-9227-NP,' " Reactor Core Response to Excessive

. LSecondary. Steam Release," (Non-Proprietary).

' 15M Hd11 man, J. M., (Ed.)', " Fuel Densification. Experimental Results and Model

for Reactor Operation,"'WCAP-8219-A, March 1975. ,

I L14.-.Morita, T. Osborne, M. P., et. al., " Dropped Rod Methodology for Negative

,j;( ,

Flux Rate Trip Plants, WCAP-10298-A (Non-Proprietary), June 1983.

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, - TABLE 2

(., ,

i '

N 4 ,, , LKINETICS' CHARACTERISTICS g%3,

?,ly ; ,a

w. -

. BEAVER' VALLEY UNIT 1 CYCLE:7 t

&E ,,

, m. ,

y <

hR?

y,

'a' ;T +

, Cycle-7 Changesz r..

. Current Limits to-Current Limits m- ..'

T Jbc

~

iMcdritor Density **: -

, 2n :s foefficientf(ap/gm/cc):- . 0cto 0.43. --

m m o

e. g"R , DopplerL Tempera ture

> H Coefficient ((pem/ k)*

~

] 1

-2.9lto .

~

Lsast;NegativefDopplerf-0nly' Power Coefficient, ZeroL.to

- -6.68'-

,xiA

+ Ful_1 PowerL(pcm/%; power)*

y

"~

JMEstLNegative Doppler-: 'Only;

? Power. Coefficient: Zeror te t w;, s l Fullf Power".'(pcm/% power)*r -~-19 4- .

LDelayedLNeutronFraction:

16fff,(%). 0.44 to 0.75 --

~ ,

+ iMinimum1 Delayed Neutron-Fraction l  : Rod. Ejection 80C;Bgff,(%) . 0.52 --

g(  ; ~

g; Rod Ejection E001Bfff,(%) '0.47- --

JMaxbm Prompt; Neutron Lifetime. '

- :(u: sec).1 26- --

l 2, 1 Maximum Differential Rod Worth lof Two Banks ~ Moving Together-

- (p'em/_i n. )* : -100

. a-

> t

• pcm V 10;53 ,

S 4 N -**The moderator density coefficient for the hot zero power, all rods out W , iphysics test condition mayLbe negative near B0C 7. The coefficient will W. ' ?."' ~ be kept positive at .that zero power by administrative controls (with

' appropriate D bank position and boron concentration).

$,g.- 1--Ind.icates 'no change. .

M) <>

.1 c

J. . '.

5701L:P-871027 <

, j a

n.u E !$6',j. ~ du  :

, 3;(; 1F '

'j- , , t  : !.

f: W'

p;; w~ ,

4

, TABLE 3:

s: , . . .

Sin .  ; . END-0F-CYCLE SHUTDOWN REQUIREMENTS AND MARGINS i

,s .

I, .< . BEAVER VALLEY UNIT 1 - CYCLE 7

. di i:

~

' ~

. Control Rod Worth'-(%Ao)' Cycle 6'. Cycle 7 a:

I t Allblods11nserted' 8.63= 8.69 l All; Rods' Inserted Less Worst. Stuck' Rod

~

um -7.66 7.61

. c.

' (1)?Less 10% 6.89: 6.85

., a 4,  : Control Rod Requirements 1

' Reactivity Defects:(Combined Doppler, T,yg, LVoid and Redistribution' Effects) 3.05 2.97  ;

- Rod'.Inse~rtion Allowance - 0.50 0.50

- '(2)[TotalRequirements 3.55 3.47 ,

l y .- l Shutdown' Margin-[(1) .-(2)] (%ap) 3.34. 3.38 l I

, LRequired. Shutdown Margin (%ap)' 1.77 1.77 d

.y ._"

ty -

s

. i2  ;.

s s

. . 12 570106-871027

= --

MTv

- ~ ~ ' - ~~ ^ ~ - -~- - - - - ~ ~ - - -

,&l.c

, ;;o 8- i ' ^', t , ,v; 4 ,j (

$d. ~fi;y;.s{ f ' ,.

. .1

+

7>

j 'f lgT p .. .. N - M -- -L K- J' H G 'F 'E D C B A

g if i

'h; '

i

'1' -8 8 8 m u, . . ,

9

- 7:

b b 7

' 2 M- 9' Bb= ,

3

'7. g9 g 8: -

8-b 6' 9 7 c 4: 7' f g 8 8- 8 8 8 b 6 9 7 1

7 5 '7 g8 ; g .8- -

.8 1g0 .1kb 1k0 1$0 6k i

.6- -9 '

ElI0 I

1k0 1k0 1k0

~

A1

[i I n

~ .

.W[ :7. g' 8h 21k0 - 1ko 1k0 Bk fp[ -8 90 =. 8 -;

,g jg0 '8

'b 8 b 1 lk0 8. g 8 g g 8

9' 38 ' '

.1k0 8k

~

of ,

8 1$0 1k0 10 k 19 '

1g0 1kb 1k0 1$0

- 11f '7 '

, 6'4 1$0 -

1k0 1$0 1k0 1$0 6k

<, q

^ 12-7 '

6'4 1$0' .1$0 6'4

' l SS ,

[. '

113 .7 6k 1k0 1k0 1b0 d 9 7 f 14 '.- 7 92 gg b 8k 15 8 8 8 0 REGION NUMBER

.g44 gF g IN-L- NUMBER OF ABSORBERS

[.: SS - SECONDARY SOURCE y

4. LI ..

FIGURE 1 BEAVER VALLEY UNIT 1 CYCLE 7 .

1 CORE LOADING PATTERN

'I 13 f[

u _ _ _ _ _ - - _ _ -

_J

.. y J 'G , .', ./

\ ', -.. .,,

? j l

)

v .

bll l

'1.2-.

)

g - (6.0. 1.0) l

- 1. 0 --===--------umaungg """'~

1

---

mnummina (to.s. o e4)

. ~

- 4

e. 3 o- '

u.;

O.8 e- \

. . . 'w. \ \

N. . '.

d-j

<- . g E .: 0.6

e- CE , l 1 1

-E. 1 1

1 4 ..

I m.

s' : o.4 )

J l(12 o. .

431) F -

) ,

L. ; p O.2

/ I O,0

- O. 2 4 6 8 1o 12 14 p

CORE-HEIGHT (feet) o l- '

e 9 .

.,'g,l FIGURE 2 h- - * '

K(Z) - NORMALIZED FQ(Z) AS A ,

.- FUNCTION OF CORE HEIGHT l BEAVER VALLEY UNIT 1  !

i l

I

)

14 l- . . ---~. a.. - . . . . . - - - _ - - _ _ . - - -