ML20073A234

Supplemental Reload Licensing Submittal for Pilgrim Nuclear Power Station Unit 1,Reload 5

ML20073A234
Person / Time
Site:	Pilgrim
Issue date:	09/30/1982
From:	Charnley J, Hilf C, Leaser J GENERAL ELECTRIC CO.
To:
Shared Package
ML20073A090	List: ML20073A087 ML20073A103 ML20073A149 ML20073A234
References
Y1003J01A28, Y1003J01A28-R01, Y1003J1A28, Y1003J1A28-R1, NUDOCS 8304110698
Download: ML20073A234 (19)

Text

- - - - - - - - - -

%?ES?W*

SEPTEMBE 1982 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR PILGRIM NUCLEAR POWER STATION UNIT 1, RELOAD 5 l

l l

\\

8304110690 830405

" ^"""

"Es!

Y1003J01A28 Revision 1 Class I September 1982 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR PILGRIM NUCLEAR POWER STATION UNIT 1, RELOAD 5 Prepared:

08 b

/J. D. Leaser A

/

Verified; C

E

1. /

f

/'

C. L. Hilf

/,/

Approved:

J. f Charnley, Acting Manager Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GENER AL h ELECTRIC i

Rav. 1 Y1003J01A28 IMPORTANT NOTICE REGARDING CONTENTS OF ThIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Boston Edison Company (BECo) for BECo's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending BECO's operating license of the Pilgrim Nuclear Power Station.

The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or pro-vided to General Electric at the time this report was prepared.

The only undertakings of the General Electric Company respecting informa-tion in this document are contained in the contract between Boston Edison Company and General Electric Company for reload fuel fabrication for the nuclear system for Pilgrim Nuclear Power Station, dated July 14, 1972, and nothing contained in this document shall be construed as changing said con-tract. The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any representation or warranty (express or implied) as to the completeness, accuracy or useful-ness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any

' responsibility for liability or damage of any kind which may result from such use of such information.

ii

Rsv. I Y1003J01A28 1.

PLANT UNIQUE ITEMS -(1.9)*

New Control Rod Withdrawal Error Analysis Procedure: Appendix A Revised End-of-Cycle Target Exposure Distribution: Appendix B Transient Analysis Initial Conditions: Appendix C 2.

RELOAD FUEL BUNDLES (1.0, 2.7, 3.3.1 AND 4.0) l Fuel Cycle Designation

_. Loaded Number Number Drilled Irradiated 8DB219H 4

68 68 8DB219L 4

156 156 P8DRB265L 5

120 120 P8DRB282 5

64 64 New P8DRB265H 6

60 60 P8DRB282 6

112 112 Total 580 580 3.

REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at 14.0 GWd/T end of cycle:

Minimum previous cycle core average exposure at 14.0 GWd/T end of cycle from cold shutdown considerations:

Assumed reload cycle core average exposure at 15.7 GWd/T l

end of cycle:

Core loading pattern:

Figure 1

• ( ) refers to areas of discussion in " General Electric Boiling Water Reactor Generic Reload Fuel Application," NEDE-240ll-P-A-1 and NEDO-24011-A, July 1979.

1

}

Y1003J01A28 RLv. 1 4.

CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2)

C k,ff Uncontrolled 1.113 Fully Controlled 0.952 Strongest Control Rod Out 0.985 R, Maximum Increase in Cold Core Reactivity 0.001 with-Exposure Into Cycle, Ak 5.

STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (Ak) pg (20*C, Xenon Free) 700 0.05 6.

RELOAD UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 AND 5.2)(1)

(REDY Events Only)

EOC6 Void Coefficient N/A* (C/% Rg)

-6.1/-7.6 Void Fraction (%)

36.9 Doppler Coefficient N/A (C/*F)

-0.22/-0.21 Average Fuel Temperature (*F) 1205 Scram Worth N/A ($)(

Sc. ram Reactivity vs Time ( }

+

• N = Nuclear Input Data A = Used in Transient Analysis Applies to Loss of Feedwater Heating Event only.

- 2 Generic, exposure independent values are used as given in " General Electric Boiling Water Reactor Generic Reload Fuel Application," NEDE-240ll-P-A-1, Amendment.10, April 1981.

2

Rsv. !

Y1003J01A28 7.

RELOAD binQUE CETAB TRANSIENT ANALYSIS INITIAL JONDITION PARAMETERS (5.2)

Peaking Factors un e Flow Fuel Exposure (Local, Radial, Bundle Power Initial 3

Design (GWd/T)

Axial)

R-Factor

[MWt)

(10 lb/hr)

MCPR 8x8 E0C6 1.22, 1.51 1.10 5.10 100 1.40 1.40 P8x8R EOC6 1.20, 1.63 1.05 5.48 101 1.43 1.40 8.

SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)

Transient Recategorization: No Recirculati6n Pump Trip:

No Rod Withdrawal Limiter:

No Thermal Power Monitor:

No Measured Scram Time:

No Exposure Dependent Limits:

No 9.

CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)

ACPR Exposure Range Q/A (GWd/T)

(% NBR)

_{%)

8x8 P8x8R Figure l

Load Rejection EOC6 597 123 0.33 0.36 3

without Bypass Loss of 100*F E0C6 117 115 0.14 0.15 4

Feedwater Heater Feedwater EOC6 385 123 0.28 0.30 5

Controller Failure

[.

10.

LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)

TRANSIENT

SUMMARY

(5.2.1)

See Appendix A.

3

Y1003J01A28 Rsv. 1 11.

CYCLE MCPR VALUES (5.2)

Exposure Range (GWd/t)

Pressurization Events Option A Option B BOC to EOC 8x8

_P8x8R 8x8 P8x8R Load Rejection w/o 0.39 0.42 0.34 0.37 Bypass Feedwater Controller 0.34 0.36 0.26 0.28 Failure Nonpressurization Events 8x8 P8x8R Loss of Feedwater Heating 0.14 0.15 Rotated Bundle Error 0.17 Rod Withdrawal Error 0.22 0.22 12.

OVERPRESSURIZATION ANALYSIS

SUMMARY

(5.3) i sl v

Plant Transient (psig)

(psig)

Response

MSIV Closure 1346 1360 Figure 7 (Flux Scram) 13.

STABILITY ANALYSIS RESULTS (5.4) i Rod Line Analyzed:

Extrapolated Rod Block Figure 8 Decay Ratio:

Reactor Core Stability Decay Ratio, x /*0 0.65 2

Channel Hydrodynamic Perfcrmance Decay Ratio, x /*0 2

8x8 Channel:

0.22 P8x8R Channel:

0.18 l

4 t

k Y1003J01A28 Rev. 1 14.

ROTATED BUNDLE ERROR RESULTS (5.5.4)

Variabic Water Gap Misoriented Bundle Analysis: Yes Includes 2.2% Power Spiking Penalty:

Yes Initial MCPR Resulting MCPR Resulting LHCR (kW/ft) 1.22 1.07 17.67 15.

CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)

Maxtcum incremental control rod worth: 0.70% Ak 16.

LOSS-OF-COOLANT ACCIDENT RESULTS. NEW FUEL (5.5.2)

See " Loss-of-Coolant Accident Analysis Report for Pilgrim Nuclear Power Station," August 1977, NEDO-21696, as amended.

l 5/6

Y1003J01A28 Rev. 1 BBBEBEMBiBsM M M BEM BBBEM M M BEMBEBEMBEMMBBMM

BE M M Bs BE M Bi BE M M M Bi H

M M Bi BE M M M M Bi Bi M M M

BEMMBEMMMMMMBiBBBE

M BE M M BEBEBE M M M M BE M

M M BEM BEBBBEBEM BEM BEM

BsBEMMMBEBEBEMBEBsBEM

BBBBBsMMBEMMBEMBEBias

i iBEMMMBiBiBEMMBiBi l

l MMMMMMMMM I

' Ti3 2 G T O T @ TiG tie E 8 4

53IS E @E 8ETEE3 @YS YYY 2

l l I IIlIIIIIIIII 1 3 5 7 9111315171921232527293133353733414345474931 y

FUEL TYPE 8DB219 L, C4 A=

8DB219L C4 E

=

8DB219 H, C4 8 = 8DB219H, C4 F

=

C=

P8DRB265 L, C5 G=

P8DR B265 H, C6 0=

P80 RB282, C5 H=

P8DRB282, C6 Figure 1.

Reference Core Loading Pattern 7

Y1003J01A28 Rev. 1 DELETED See Section 6 1

Figure 2.

Scram Reactivity and Control Rod Drive Specifications 8

l 1

ii 4!

l

$4. -

dseS&o s

)

I S

- #L P

f'Mll

~

o.

Y (1

T 6

e IT Y

TVTT tIUiL TI!T IFR Jt ITvV E

VCiI S V 4;L

!ATT 4

V o 4 TECC ELi L 6

1 e t 2

6 CIFT A FE

'f W v_V ERfR l

1 3

1E ffiW LML 9

LY f

l Sltl OPAR t

Sl H

lPHT ii t 0CCO E J i

V Y (2t

~

VDST 12's lit. ~U 0

!23C 8

s 1

s 4) 1 a

u C

9-E p

1 S

C y

e_

E

' [

(

E B

3 S(

M t

I 2.E u

2.T L\\

M 3

)

3I o

T hti u

W

• /

a n

-'f 6

o 6

I

[

i

}

. 1

/

Y

- 1 tce h

je L

R 2

d I

..0

'.0 h

a g

0 0

g 0

0 1

2 oL 0

0 2

1

@6gu0Eg ro tar T

e R

n X

K e

U S

i G

d L

%I d

P F

o E W SWO t

T OL nW 3

FLFW FO e

EFMO HL RMAL s

X F

AEF 4

n UE 4

HET LCI 6

o M

6 C1EF 8SI g

l'lt NS E

p fL 1

I ET s

NtNf 0t (LNA e

LEIH ESB U

R 1

E UEh VSR E

EEUE EV0 t

NAC LVTF 3

n 8

h 0

1214S 123g a

4) l m

4)

N%

C C

E P

E 4 ~,

S S

I

~

M I

E EM M

3

" 2.T 4

2.T e

I I

3 3

3 r

^

ug V

iF 4

g A

3 pl 6

6 Y

- 1 I t N

I

- 1 3 W 7

D h

~

0 0

0 0.

0.

0 0*

0 D

0 5

0 5

0 0

0 1

1 2

1 1 -

Ei $

e ll l

1l!fl1 li

1l ga 8owN$>N" l

i S

P(WW T

=-

Y EOO IYT SLL YVT T

IFF TIII R

ITVV EE VCII

!RT I

q 0

SVV 0

TECC 3

6 ELL 3

Fff 6

CfFF 1

1 PVv 1

A EE 1

ERR R

LFS RE EES LM L

SIA DPAA SLP IPRT T

O0CG EEB V0SI V

0 123456 0

1234 2

21)C y

1 a

E

)

S 1

1 C

g

(

E E

S n

M I

i t

.I

.E T

0 0M a

8 8I e

T H

y r

y 1a e

e t

1 aw 0

0

- 4

- 4 dee F

F m

0

~.O.

3

.O I

0 1

~

5 5

5 g

O 1

2 n

2 2

f 2

1 o

g $ o.6 u E 2 Q we sso L

TR o

I K

t X

U S-

=-

L

=-

P e

F E W s

SWO T

A OL n

Eg B FLFH o

HiU EFM O

p X pS RMA L

s AEF u

0 UE 0

HET 1

C15 t

6 e

s LC7i 2

6 R

1 t

3 1

NS I

Ff t

%g LN E H t

1 I

Nf OU (L

RS;I LEIW n

E ESB U

T n R VSR L

a UE O

EEU L

l EV NAcC LVT t

P 0

12345 0

I2345 2

2 1) 1)

n C

C a

E 2

E 4

S g

S i

u I

w.

i I

e E

E r

M M

u

.I

.I g

T T

0 0

i 8

8 F

u a

m i

0 0

- 4

- 4 us

[. '

0"O 0

a, m

O 0

0 0

0 S

5 0

5 1

1 g

I awh b ~5 HC

(

ilLlll l

)!-

fl[

d8to$NC g0.

O

)

I S

P Y

(

T SLM W

IYY EO q

TVTT 3

TI'-

V IFH 4

R r

T F.

Et;U j

h 3

EFFf 2

3 TE 2

2 Fi F

EE PVH I

ERR B RE LY LML ETI DPA A SE IPH T EAL'F SF OOCO VDST VS 1234 4

12 5G 4

2 2l 1

C 3

E

?

l 2

S 1

C

(

E e

S E

r

(

M u

.I

.E l

T 6M i

6 1I 1

T a

F u_

r z

e t

l l

o

- 6

-8 r

t C

no

=

re

-1

.O

.D t

i 5

2 a

s 3

O 1

5 z

w 2

2 2-d 2

z6 OE~l e

e F

o T

t R

I K

e X

S s

U n

L

'4 P

o F

E W T

SWO p

OL EC FLF W s

A" EFMO e

Hb t

RMA L R

X S

l J-AEF r-HET 2

I 2

3 t

lk FB,

3 C15 F

I M

fI NS E

n NR n-I ET u

a OUH (LNA 2-LEIW l

RS 1,

ES8 D

P T

VSR E

UE EEU E

EV8 2

NAH5 LVT F

12345 4

5 12 4

21 E

f 2)

C C

E e

y S

r S

u

(

I u

g E

E i M

F I

I

.T

.T 6

5 X

1 1

n w

u 1

> 8

-8

=

4 V'

4,.

_ : ~.O

.O A

o-w 0

D 0

0 0

0 5

5 0

5 s

1 1

i 1

_g E b rgh-HH l

liIl

.ltl1

Y1003J01A28 Rev. 1 DELETED See Appendix A J

Figure 6.

Limiting RWE Rod Pattern 12

ll l

U8w4SDC oo4, -

o

t S

P(L u.

Y l

T WW EdOO ITY SLLL TVT T TIII IEFF R

ITV V E E,E VCII IAT T S

u TFCC EWV,V

! L 8

PVVI 8

2 EFFEE F

/

Fif V

ERRR 3

RE LM($

F LML E

b DPA A

D Sfl

/

ih IPRT O0CO E

V Kb V0ST 1

1234 1

45S 6

6)

/

u CE

)

2 S

C

(

3 E

E S

(

M

)

I

.T

.E m

7

(

M a

4 4

I r

T c

S u

xu W

l

/

F

. 2 3

-2

(

e r

i n

u m

s o

l

.t

-.O

.D C

0*

0 0

D g

0 2

1 V

0 0

0 3

2 1

I S

8 0'5=g M

o t

TR e

I X

K s

S n

U

~

o

%3 P

3 p

L m_

F E W SNO s

T OL e

AH FLFN R

EC EFM HL 1

RMR(L X F AEF t

UE HET 8

n LCT 8

C15 t

% N I

Eit 2

a FfE NS i

fL l

NRN (LN R

OUI LEIN P

RS ES8 U

3 T

E VSR L

UER EEU L

EVO LVT t

NRC 7

12345 12345 6

6) e M

1 C

C r

E E

?

S u

S g

1 I

(

i EF EM 3

I I

.T

.T 4

4

/

d N

t A

. 2

. U' 4

,'i

-2 t

t 1

0

,0 0

0 0

g 0

0 0

0 5

0 5

0 0

0 1

2 1

1 -

g.E25g-C

l lIll ll

Rev. 1 Y1003J01A28 ULTIMATE PERFORMANCE LIMIT i.oo

'o 0.75 21 x"

A Op E>

NAT;JRAL CIRCULATION E 0.50

^

s 100% ROD LINE 0.25 8

l I

I l

o 0

20 40 60 80 100 120 POWER (PERCENT)

Figure 8.

Reactor Core Decay Ratio versus Power 14

Y1003J01A28 Rev. 1 l

l APPENDIX A LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT

SUMMARY

(NEW PROCEDURE)

The Local Rod Withdrawal Error results are reported below in accordance with Letter, R. E. Engel (GE) to T. A. Ippolito (NRC), " Change in General Electric Methods for Analysis of Control Rod Withdrawal Error," May 18, 1981.

ACPR Rod Block Reading

• 8x8/P8x8R 104 0.13 105 0.16 106 0.19 107*

0.22 108 0.28 109 0.32 110 0.36 1

• Indicates set point selected.

15/16

Y1003J01A28 Rev. 1 APPENDIX B REVISED END-OF-CYCLE TARGET EXPOSUEE DISTRIBUTION The original reload licensing analyses for Pilgrim Unit 1 cycle 6 were per-formed using a standard end-of-cycle exposure distribution. Subsequently, a revised end-of-cycle target exposure distribution was established to pro-vide for increased uranium utilization.

To account for this change, a revised set of end-of-cycle nuclear parameters were developed and the affected events were reanalyzed. The revised ODYN transient results are presented in Revision 1.

?

l k

9 17/18

Y1003J01A28 Rav. I

~

1 l

APPENDIX C TRANSIENT ANALYSIS INITIAL CONDITIONS i

Lowest S/RV Setpoint 1125 psig +1%

l S/RV Capacity 41.1%

Dome Pressure 1025 psig 1.

Y i

l 19/20 (FINAL)

~

a e

GEN ER AL $ ELECTRIC