ML20033A009

Supplemental Reload Licensing Submittal for Ja Fitzpatrick Nuclear Power Plant Reload 4

ML20033A009
Person / Time
Site:	FitzPatrick
Issue date:	08/31/1981
From:	Charnley J, Engel R, Leaser J GENERAL ELECTRIC CO.
To:
Shared Package
ML20032F017	List: ML20032F016 ML20032F022 ML20032F028 ML20033A002 ML20033A009
References
Y1003J01A25, Y1003J01A25-R00, Y1003J1A25, Y1003J1A25-R, NUDOCS 8111230633
Download: ML20033A009 (25)

Text

-- -

"EUsl AUGUST 1981 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR JAMES A.

FITZPATRICK NUCLEAR POWER PLANT RELOAD 4 lAA'ES8efoS$$$h!!

GENER AL h ELECTRIC

Y1003J01A25 Revision 0 Class I August 1981 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT R.rLOAD 4 Prepared: 0 8. b r--

I J'.

D. Leaser Verified:

o-

/S. Charnley' J

c/

Approved:

R. E. Engef, Manager Reload Fuel Licensing NUCLEAR POWER SYSTEMS OlVISION e GENERAL ELECTRIC COMPANY SAN JOSE, CAU FORNI A 95125 GEN ER AL h ELECTRIC i

i 1

Y1003J01A25 Rev. O IMPORTANT NOTI' E REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for The Power Authority of the State of New York (The Authority) for The Authority's use with the U.S.

h Nuclear Regulatory Commission (USNRC) for amending The Authority's operating license of the James A. FitzPatrick Nuclear Power Plant.

The information con-tained in this report is believed by General Electric to be an accurate and truc representation of the facts known, obtained or provided to General Electric the time this report was prepared.

at The only undertakings of the General Electric Company respecting information in this document are contained in the contract between The Authority and General Electric Company for nuclear fuel and related services for the nuclear system for The James A. FitzPatrick Nuclear Power Flant, dated June 12, 1970, and nothing contained in this document shall be construed as changing said The use of this information exc'ept as defined by said contract, contract.

or for any purpose other than that for which it ais 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 usefulness 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.

11

~

Y1003J01A25 Rev. 0 1.

PLANT UNIQUE ITEMS (1.0)*

Transient Analysis Initial Conditions: Appendix A 2.

RELOAD FUEL BUNDLES (1.0, 2.7, 3.3.1 and 4.0)

Fuel Cycle Designation Loaded Number Number Drilled Irradiated 8DB274L 2

20 20 8DB274H 2

56 56 8DRB265L 3

36 36 8DRB283 3

100 100 P8DRB265L 4

24 24 P8DRB283 4

136 136 New P8DRB284H 5

128 128 P8DRB299 5

60 60 Total 560 560 3.

REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at end of cycle:

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

16.4 GWd/T l

Assumed reload cycle core average exposure at end of cycle:

17.2 GWd/T Core loading pattern:

Figure 1 h

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

July 1979, as revised by amendments 2-10.

L 1

Y1003J01A25 Rev. O s

4.

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

BOC k,ff Uncontrolled 1.116 Fully Controlled 0.957 Strongest Control Rod out 0.987 R, Maximum Increase in Cold Core Reactivity with Exposure Into Cycle, Ak 0.000 5.

STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (ak) pga (20*C, Xenon Free) 600 0.03 6.

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

EOC-1 EOC GWd/T Void Coefficient N/A* (c/% Rg)

-8.8/-11.0

-9.6/-12.0 Void Fraction (%)

41.7 41.7 Dopplci Coef ficient N/A (c/*F)

-0.23/-0.22

-0.23/-0.22 Average Fuel Temperature (*F) 1343 1343 Scram Worth N/A (%)( }

Scram Reactivity vs Time (

CN = Nuclear Input Data A = Used in Transient Analysis (1) Applies to Inadvertent Startup of HPCI Pump Event Only (2) Generic, exposure independent values are used as given in " General Electric Boiling Water Reactor Generic Reload Fuel Application," NEDE-24011-P-A-1, Amendment 10, April 1981.

2

(

Y1003J01A25 Rev. 0 7.

RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2)

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

Design (GWd/T)

Axial)

R-Factor (MWt)

(10 lb/hr)

MCPR 8x8 EOC

-1.22, 1.35, 1.40 1.10 5.75 115 1.29 EOC-1 1.22, 1.39, 1.40 1.10 5.92 114 1.25 8x8R EOC 1.20, 1.50, 1.40 1.05 6.40 115 1.29 EOC-1 1.20, 1.54, 1.40 1.05 6.58 114 1.25 l

1 f

P8x8R EOC 1.20, 1.48, 1.40 1.05 6.29 116 1.31 EOC-1 1.20, 1.52, 1.40 1.05 6.46 115 1.28 8.

SELECTED MARGIN IMPROV1NENT OPTIONS (5.2.2)

Transient Recategorization: No Recirculation Pump Trip:

No Rod Withdrawal Limiter:

No Thermal Power Monitor:

Yes Measured Scram Time:

No Exposure Dependent L!mits:

Yes Exposures Analyzed (GWd/T): EOC EOC-1

~

9.

CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)

Exposure Range Q/A ACPR Transient (GWd/T)

(% NBR)

(%)

8x8/8x8R P8x8R Figure k

Load Rejection EOC 653 125 0.22 0.24 3a without Bypass EOC-1 609 122 3.18 0.21 3b Inadvertent Start BOC to 128 120 0.14 0.15 4

OC 1

of HPCI Pump

{

Feedwater EOC 362 122 0.17 0.19 Sa l

Controller Failure EOC-1 311 120 0.15 0.16 Sb i

Y1003J01A25 Rev. G 10.

LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)

TRANSIENT

SUMMARY

(5.2.1)

Limiting Rod Pattern:

Figure 6 Includes 2.2% Power Spiking Penalty: Yea Rod Position Rod Block (Feet ACPR MLHGR (kW/ft)

Readin g

• Withdrawn) 8x8**8x8R/P8x8R 8x8**8x8R/P8x8R 104 3.0 0.12

'16.1 105 3.5

0. 15 16.5 106 4.0 0.17 16.8 107 5.0 0.20 16.9 108*

6.5 0 24 16.9 109 7.0 0.24 16.9 110 8.0 0.25 16.9 3

11.

CYCLE MCPR VALUES (5.2)

Exposure Range (CWd/t)

Pressurization Events Option A Option B BOC to EOC-1 8x8/8x8R/P8x8R 8x8/8x8R/P8x8R Load Rejection w/o Bypass 0.23/0.23/0.27 0.04/0.04/0.06 Feedwater Controller Failure 0.19/0.20/0.21 0.13/0.14/0.15 EOC-1 to EOC Load Rejection w/o Bypass 0.28/0.28/0.30 0.16/0.16/0.18 Feedvater Controller Failure 0.22/0.22/0.25 0.16/0.16/0.18 BOC to EOC Non-Pressurization Events 8x8/8x8R/PJx8R Inadvertent PCI Pump SC'Et 0.14/0.14/0.15 Rotated Bundle Error

-/-/0.13 Rod Withdrawal Error

-- 0.24/0.24 jg

l 0 Indicates set point selected.

ooNot Limiting 4

Y1003J01A25 Rev. 0 12.

OVERPRESSURIZATIOy ANALYSIS

SUMMARY

(5.3) si y

Plant Transient (psig)

(psig)

Response

l MSIV Closure 1236 1275 Figure 7 (Flux Scram) 13.

STABILITY ANALYSIS RESULTS (5.4)

J Rod Line Analyzed:

Extrapolated Rod Block Decay Ratio:

Figure 8 Reactor Core Stability Decay Ratio, x f*0:

0.87 2

Channel Hydrodynamic Performance Decay Ratio, x /*0 2

8x8 Channel:

0.37 8x8R/P8x8R Channel:

0.30 14.

ROTATED BUNDLE ERROR RESULTS

(.~

4,4)

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

Yes f

Initial Resulting Resulting

}

MCPR MCPR LEGR (kW/ft) 1.18 1.07 15.32 1

15.

CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)

Bounding Analysis Results:

1 Doppler Reactivity Coefficient:

Figure 9 Accident Reactivity Shape Functions: Figures 10 and 11 Scram Reactivity Functions:

Figures 12 and 13 J

f 16.

LOSS-OF-COOLANT ACCIDENT RESULTS, NEW FUEL (5.5.2) l l

See " Loss-of-Coolant Accident Analysis Report for James A. FitzPatrick Nuclear Power Plant (Lead Plant)," July 1977, NED0-21662, as amended.

l l

5

Y1003J01A25 Rev. 0

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• A =

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8 P90RB283

=

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80R8283 H = P80R8299 Figure 1.

Reference Core Loading Pattern 6

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Y1003J01A25 Rev. 0 ULTIMATE PERFORMANCE UMlT 1.00 - - - - emme

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Figure 8.

Reactor Core Decay Ratio versus Power 14

Y1003J01A25 Rev. 0 m

i 0.O A CAL CULATED VA UE-COLD B CAL CULATED VAI.UE-HS8 C BO', W VAL 280 C C G COLO

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Doppler Reactivity C afficient Comparison for RDA 15

Y1003J01A25 Rev. O 0.020 A ACCIDENT FUNCTICN 8 BOUNCING VALUE 280 CAL /0 d

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RDA Reactivity Shape Function, Cold I

16

.--i.-

Y1003J01A25 Rev. 0 9

0.020 A ACCIDENT FUNCTION 8 BOUNDING VALUE 290 CAL /0 0.015

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Figure 11.

RDA Reactivity Shape Function, Hot Standby 17

Y1003J01A25 Rev. 0

}

0.04 A SCF AM FUNCTIC J B B0t NOING VALU : 290 C /0 5

5

'~

0.03 2

<3 l

p' h 0.02

/

^

p l

=

o

.s f

~

s-1 0.01

/

?

r i

W 0

0. 0

1. 0

2. 0 3.0 4.0

5. 0

6. 0 E LAPSED TIME (sec)

?

i Figure 12.

RDA Scram Reac'ivity Function, Cold 18

l Y1003J01A25 Rev. 0 0.06 A SCF AM FUNCTIO.4 8 BOL NJING VALUl: 280 CAL /G 0.05

/

\\

0.04 x

<3 l

I D

b 0.03

$e 0.02

{

0.01

_M.

O. 0

0. 0 1.0

2. 0 3.0 4.0

5. 0 6.0 ELAPSED TIME (sec) t i

Figure 13.

RDA Scram Reactivity Function, Hot Standby 19 g

Y1003J01A25 Rev. O APPENDLT A TRANSIENT ANALYSIS INITIAL COcQITIONS S/RV Capacity 84.2%

This valve more accurately represents the capacity of the S/RVs in the "as-installed" conditions.

e 20 (FINAL)

'b.

GEN ER AL h ELECTRIC n

e i-i

__