ML20212E256

Supplemental Reload Licensing Submittal for Ja Fitzpatrick Nuclear Power Plant Reload 7

ML20212E256
Person / Time
Site:	FitzPatrick
Issue date:	12/23/1986
From:	Charnley J, Lambert P, Plotycia G GENERAL ELECTRIC CO.
To:
Shared Package
ML19292G521	List: ML20212E152 ML20212E185 ML20212E221 ML20212E256
References
23A4825, NUDOCS 8701050282
Download: ML20212E256 (21)

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_ _ _ _

O 23A4825 Class I

g November 1986 SUPPLEMENTAL RELOAD LICENSING SU3MITTAL l

FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT U

RELOAD 7 O

Prepared P. A. Lambe'rt Fuel Licensing Q

Verified:

G. D. Plotycia /

Fuel Licensing O,

Approv J

, Chan61ef, Manage d,.

i Licensing O

NUCLEAR ENERGY BUSINESS OPERATIONS + GENERAL ELECTRIC COMPANY SAN JOSE. C ALIFOANIA 95125 GENERAL h ELECTRIC g?S?,$$SNSh!i

./2 g

23A4825 Rev. O IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for The Power Author-ity of the State of New York (The Authority) for The Authority's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending The Authority's oper-ating license of the James A. FitzPatrick Nuclear Power Plant. The informa-tion contained in this report is believed by General Electric to be an accu-9" rate and true representation of the facts known, obtained or provided to General Electric at the time this report was prepared.

The only undertakings of the General Electric Company respectir.g informa-tion 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 Plant, dated August 1, 1981, and nothing contained in this document shall be construed as changing said O

contract. 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 nor any of the contributors to this document makes any representation or warranty 3

(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.

3

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O 3/4 O

'O 23A4825 Rev. O O

ACKNOWLEDGMENT The engineering and reload licensing analyses which form the technical basis of this Supplemental Reload Licensing Submittal, were performed by P. V. Vo and J. L. Casillas of the Nuclear Fuel Engineering Department.

O O

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,o i

O iO l

5/6 O

J 23A4825 Rev. 0 1.

PLANT UNIQUE ITEM (1.0)*

GEMINI System of Methods Used 3

Appendix A: Transient Operating Parameters 2.

RELOAD FUEL F"NDLES (1.0, 2.0, 3.3.1 AND 4.0)

Fuel Type Cycle Loaded Number

)

Irradiated P8DRB299 6

176 BP8DRB299 7

196 D

New BP8DRB299 8

4 BD319A 8

184 O

Total 560 3.

REFERENCE CORE LOADING PATTERN (3.3.1)

O Nominal previous cycle core average exposure at end of cycle:

21609 mwd /MT Minimum previous cycle core average exposure at end O

of cycle from cold shutdown considerations:

21278 mwd /MT Assumed reload cycle core average exposure at end of cycle:

21042 mwd /MT O

Core loading pattern:

Figure 1 9

• ( ) Refers to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-8, dated May 1986. A letter "S" preceding the number refers to the appropriate country-specific supplement.

7 6

O 23A4825 Rev. 0 4.

_ CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS, 20'C (3.3.2.1.1 and 3.3.2.1.2) gp Beginning of Cycle, K,f f Uncontrolled 1.113 gp Fully Controlled 0.963 Strongest Control Rod Out 0.989 (p

R, Maximum Increase in Cold Core Reactivity with 0.000 Exposure into Cycle, A K O

5.

STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin G K) jyyn (20*C, Xenon Free)

O 660 0.041 1

6.

RELOAD UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2) l (REDY EVENTS ONLY)

O Void Fraction (%)

42.2 Average Fuel Temperature ('F) 1085 Void Coefficient N/A* (f/% Rg)

-7.63/-9.54 Doppler Coefficient N/A* (4/'F)

-0.207/-0.197 Scram Worth N/A* ($)

GD l

t

• N = Nuclear Input Data A = Used in Transient Analysis II

• • Generic exposure independent values are used as given in " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-8, May 1986.

8 O

23A4825 Rev. O 7.

RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS

)

(S.2.2)

Fuel Peaking Factors Bundle Power Bundle Flow Initial Design Local Radial Axial R-Factor (MWt)

(1000 lb/hr)

MCPR Exposure: EOC BP/P8x8R 1.20 1.49 1.40 1.051 6.323 113.7 1.28 GE8x8IB 1.20 1.49 1.40 1.051 6.319 115.4 1.28 l

Exposure: E0C-1102 mwd /MT BP/P8x8R 1.20 1.52 1.40 1.051 6.449 113.0 1.25 GE8x8EB 1.20 1.52 1.40 1.051 6.440 114.7 1.26 Exposure: E0C-2205 mwd /MT j

BP/P8x8R 1.20 1.56 1.40 1.051 6.620 112.0 1.22 GE8x8EB 1.20 1.57 1.40 1.051 6.659 113.5 1.21 8.

SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)

)

Transient Recategorization:

No Recirculation Pump Trip:

No Rod Withdrawal Limiter:

No Thermal Power Monitor:

Yes

)

Improved Scram Time:

Yes (ODYN Option B)

Exposure Dependent Limits:

Yes Exposure Points Analyzed:

3

)

9.

OPERATING FLEXIBILITY OPTIONS (S.2.2.3)

Single Loop Operation:

Yes Load Line Limit:

Yes

)

Extended Load Line Limit:

No Increased Core Flow:

No Flow Point Analyzed:

N/A Feedwater Temperature Reduction:

No

)

j

)

ARTS Program:

No Maximum Extended Operating Domain:

No

)

9

O 23A4825 Rev. 0 10.

CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1) g Flux Q/A aCPR Transient

(% NBR)

(% NBR)

BP/P8x8R GE8x8EB Figure Exposure: Independent g

Inadvertent HPCI Activation 120 115 0.15 0.14 2

Exposure: EOC8 Load Rejection w/o Bypass 538 124 0.21 0.21 3a S

Feedwater Controller Failure 355 119 0.16 0.16 4a Exposure: E0C8-1102 mwd /MT Load Rejection w/o Bypass 463 121 0.19 0.19 3b O

Feedwater Controller Failure 306 116 U.14 0.14 4b Exposure: EOC-2205 mwd /MT Load Rejection w/o Bypass 334 116 0.13 0.13 3c 9

Feedwater Contrc11er Failure 217 111 0.08 0.08 4c l

11.

LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT

(

SUMMARY

S.2.2.1)

O Limiting Rod Pattern: Figure 5 Rod Block Rod Position aCPR MLHGR (kW/ft)

Reading (ft withdrawn)

BP/P8x8R GE8x8EB BP/P8x8R GE8x8EB g

l 104 3.5 0.09 0.09 15.12 16.12 l

105 4.0 0.11 0.11 15.84 16.84 1

106 4.0 0.11 0.11 15.84 16.84 i

107 4.5 0.13 0.13 16.36 17.36 9

108 6.5 0.20 0.20 17.63 18.63 109 12.0 0.26 0.26 17.93 18.93 110 12.0 0.26 0.26 17.93 18.93 O

Setpoint Selected: 108 i

9

O.

23A4825 Rev. O g

12.

CYCLE MCPR VALUES (S.2.2)

BP/P8x8R GE8x8EB Non-Pressurization Events O

Exposure Range: BOC to EOC Inadvertent HPCI Activation 1.22 1.21 Fuel 1,oading Error 1.19 Rod Withdrawal Error 1.27 1.27 O

Option A Option B BP/P8x8R GE8x8EB BP/P8x8R GE8x8EB Pressurization Events

• 1 O

Exposure Range: EOC Load Rejection Without Bypass 1.36 1.36 1.30 1.30 Feedwater Controller Failure 1.27 1.27 1.24 1.24 0

Exposure Range: EOC-1102 mwd /MT O

L ad Rejecti n Without Bypass 1.36 1.36 1.29 1.29 Feedwater Controller Failure 1.27 1.27 1.25 1.25 Exposure Range: EOC-2205 mwd /MT O

Load Rejection Without Bypass 1.30 1.30 1.23 1.23 Feedwater Controller Failure 1.21 1.21 1.19 1.19 O

• 0DYN Adjustment Factors are documented in a letter from J.S. Charnley (GE) to O

H. N. Berkow (NRC), " Supplementary Information Regarding Amendment 11 to GE Licensing Topical Report NEDE-24011-P-A," January 16, 1986.

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O 23A4825 Rev. 0

13. OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.3) g s1 y

Transient (psig)

(psig)

Plant Response O

MSIV Closure 1217 1256 Figure 6 (Flux Scram) i

14. LOADING ERROR RESULTS (S.2.5.4)

GD l

Variable Water Gap Misoriented Bundle Analysis: Yes Event Initial CPR Resulting CPR O

l Misoriented 1.17 1.07 t

l 15.

CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1) l ED i

l Plant Specific Analysis Results:

Parameter (s) not Bounded, Cold:

None Resultant Peak Enthalpy, Cold:

N/A Parameter (s) not Bounded, HSB:

Accident Reactivity II Resultant Peak Enthalpy, HSB:

252.2 cal /gm l

16.

STABILITY ANALYSIS RESULT (S.2.4)

[

9 l

l GE SIL-380 recommendations have been included in the James A. FitzPatrick Nuclear Power Plant operating procedures and/or Technical Specifications l

and, therefore, no stability analysis is required.

l db i

kh 12 O

l

!I__

D l

23A4825 Rev. 0 l

17.

LOSS-OF-COOLANT ACCIDENT RESULT (S.2.5.2) l LOCA Method Used: SAFE /REFLOOD/ CHASTE See " Loss-of-Coclant Accident Analysis for James A. FitzPatrick Nuclear Power Plant (Lead Plant)," July 1977. NEDO-21662 (as amended).

LOCA Method Used: SAFER /GESTR-LOCA D

See " James A. Fitzpatrick Nuclear Power Plant SAFER /GESTR-LOCA Loss of Coolant Accident Analysis," October 1986, NEDC-31317P (as amended).

O

3 3

3 3

e 13 O

O 23A4825 Rzv. 0

.MMMMM.

.MMMMMMMMM.

MMMMMMMMMMM

.MMMMMMMMMMM.

M M M M M M M M M M M M M

': M M M M M M M M M M M M M

:M E M M M M M M M M M M M

';: M M M M M M M M M M M M M

':: M M M M M M M M M M M M M

"MMMMMMMMMMM" l'

MMMMMMMMMMM "MMMMMMMMM"

-l "MMMMM" 1IIIIIIIII 1 35 7 9111315171921232527293133353739414345474951 FUEL TYPE A = BD319A (Cycle 8)

C BP8DRB299 (Cycle 7)

=

BP8DRB299 (Cycle 8)

B = P8DRB299 (Cycle 6)

D

=

O Figure 1.

Reference Core Loading Pattern 14 9

l D'

23A4825 Rev. 0 t

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Figure 2.

Plant Response to Inadvertent HPCI Activation 15 0

i o

23A4825 Rsv. O t

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Figure 3a.

Plant Response to Generator Load Rejection w/o Bypass (EOC) l 16 9

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v 23A4825 Rev. O

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.O 17

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' 23A4825 Rev. 0

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O 1 NEUTRON FLUt I n'ESSEL PRESE R!sttPSI) 2##E $URFACE HEAT FLUX 2 SAFEff VALVE FLOW f 5 6..

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O Figure St.-

Plant Response to Generator Load Rejection w/o Bypass (E0C8-2205 mwd /MT) 18 g

n v

23A4825 Rev. O O

iS...

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I VESEEL PRESS R!SE(PSI) 2 AVE SURFA E >( AT LUX 2SM lTV VAL LCe 3 E05: 1%1 F Cs 3 REL IEF W AL E Cu L

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O Figure 4a.

Plant Response to Feedwater Controller Failure (EOC)

O

O 23A4825 Rsv. O l

9 158..

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$UEF9'E HEAT UX 2 $ AF :7T VAL F *e h

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Figure 4b. Plant Response to Feedwater Controller Failure (E0C-1102 hvd/MT) 20 0

3 23A4825 Rev. O J

i s...

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Figure 4c.

Plant Response to Feedwater Controller Failure (E0C8-2205 mwd /MT) 21 0

4 23A4825 Rev. 0 02 06 10 14 18 22 26 30-34 38 42 46 50 l

51 8

8 47 40 40 10 40 40 9:

43 16 8

30 30 8

16 39 40 4

8 4

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30 40 40 30 8

31 16 10 8

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8 NOTES:

1.

NUMBER INDICATES NUMBER OF NOTCHES WITHDRAWN OUT OF 48.

g BLANK IS A WITHDRAWN ROD.

2.

ERROR ROD IS (26,31) 4:

Figure 5.

Limiting RWE Rod Pattern 22 O

V 23A4825 Rsv. 0

.O-v I hEUTRON F.UX 1 VESSEL PREt$ RISECP5I) q 2 AVE SURFA:E HEAT FLUX 2 SAFETY VA.VT FLOW

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-]

Figure 6.

Plant Response to MSIV Closure (Flux Scram) 23/24 O

7 23A4825 Rev. O APPENDIX A O

TRANSIENT OPERATING PARAMETERS All of the transients and overpressure protection analyses were run considerin6 a power uncertainty of 2%. According to Reference A-1, the O

uncertainty in the ODYN transient analysis is included in the statistical adjustment factors and the transient is initiated at 100% rated power. The other initial conditions reflect this initial power level.

O

REFERENCE:

A-1. Letter, J. S. Charnley (GE) to H. N. Berkow (NRC), " Revised Supplementary Information Regarding Amendment 11 to GE Licensing Topical Report NEDE-24011-P-A," dated January 16, 1986.

O O

O O

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'O 25/26 CFI"*'}

O

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