ML20214L686

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Rev 0 to Supplemental Reload Licensing Submittal
ML20214L686
Person / Time
Site: Millstone Dominion icon.png
Issue date: 05/21/1987
From: Charnley J, Lambert P, Plotycia G
GENERAL ELECTRIC CO.
To:
Shared Package
ML19292H313 List:
References
23A4845, 23A4845-R, 23A4845-R00, NUDOCS 8706010052
Download: ML20214L686 (25)
v • d • e


Text

O 23A4845 Revision 0 Class I April 1987 O

i O

SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR O

MILLSTONE POINT NUCLEAR POWER STATION UNIT 1, RELOAD 11 O

O Prepared:

P. A. Lambert Fuel Licensing A

O verified:

~G. D. Plotycia Mf Fuel Licensing I l

)

, /

O 3pproy

( arnley, Manager el Licensing h

O l =

NUCLEAR ENERGY BUSINESS OPERATIONS

  • GENERAL ELECTRIC COMPANY O SAN JOSE. CALIFORNIA 95125 GENER AL $ ELECTRIC

~'

.0 e706010052 e7o521 1/2 PDR ADOCK 05000245' p PDR,

O 23A4845 Rsv. 0 IMPORTANT NOTICE REGARDING

!O CONTENTS OF THIS REPORT PLEASE READ CAREFULLY i

l

O This report was prepared by General Electric solely for Northeast Utilities Service Company (NUSCo) for NUSCo's use with the U. S. Nuclear Regulatory Commission (USNRC) for amending NUSCo's operating license of the Millstone Point Nuclear Power Station. The information contained in this
O report is believed by General Electric to be an accurate and true representa-tion of the facts known, obtained or provided to General Electric at the time this report was prepared.

lO The only undertakings of the General Electric Compacy respecting informa-tion in this document are contained in the contract between Northeast l Utilities Service Company and General Electric Company for nuclear fuel and related services for the nuclear system for Millstone Point Nuclear Power '

O Station, dated April 14, 1967 and March 13, 1980 and nothing contained in this

- document shall be construed as changing said contracts. The use of this information except as defined by said contracts, for any purpose other than that for which it is intended, is not authorized; and with respect to any such 40 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 d cument r that such use of such information any not infringe privately iO I owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.

O O

d

!O 3/4

O 23A4845 Rev. O O ACKNOWLEDGMENT The engineering and reload licensing analyses which form the technical basis of this Supplemental Reload Licensing Submittal, were performed by O R. D. Nourse and T. P. Shannon of the Nuclear Fuel and Engineering Services Department.

O O

5 1

O O

1 10 O

O O

5/6

O 23A4845 Rw. 0

1. PLANT-UNIQUE ITEMS (1.0)*

O GETAB Analysis Initial Conditions: Appendix A Feedwater Temperature Reduction Analysis: Appendix B lO 2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)

Fuel Type Cycle Loaded Number

,0 Irradiated BP8DRB300 10 184 BP8DRB300 11 200  !

O 3,,

BD338A** 12 196 i

Total 580

'O

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure 19,947 mwd /MT O at end of cyeie:

Minimum previous cycle core average exposure at 19,947 mwd /MT end of cycle from cold shutdown considerations:

Assumed reload cycle core average exposure at 21,592 mwd /MT O end of cycle:

Core loading pattern: Figure 1 O *( ) Refers to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-8, dated May 1986. A letter "S" preced-

! ing the number refers to the U.S. Supplement, NEDE-24011-P-A-8-US, May 1986.

    • The BD338A fuel bundle description was provided in a letter from J. S. Charnley (GE) to G. C. Lainas (NRC), " Additional Information

!g Pertaining to Proposed Amendment 18 to NEDE-24011-P-A," January 29, 1987.

The BD338A fuel bundle description is also contained in " Loss-of-Coolant Accident Analysis for Millstone Unit 1 Nuclear Power Station, Supplement 1,"

NEDE-24085-1-P, April 1987.

o 7

0 23A4845 Rsv. 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) lO Beginning of Cycle, Keff Uncontrolled 1.106
O Fully Controlled 0.958 i

Strongest Control Rod Out 0.976 i

R, Maximum Increase in Cold Core Reactivity with 0.011 Exposure into Cycle,

  • K Q,
5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (AK)

O m (20*C, Xenon Free) 660 0.055 g 6. RELOAD UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2)

(Cold Water Injection Events Only) g Void Fraction (%) 37.1 Average Fuel Temperature (*F) 987 Void Coefficient N/A* (d/% Rg) -5.85/-7.31 Doppler Coefficient N/A* (d/*F) -0.273/-0.259 Scram Worth N/A* ($) **

O l D

D *N = Nuclear Input Data, A = Used in Transient Analysis

    • Generic exposure independent values are used as given in " General Electric  ;

Standard Application for Reactor Fuel," NEDE-24011-P-A-8-US, dated May 1986. 1

, 8

'J

'O 23A4845 Rsv. 0

7. RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2)

Fuel Peaking Factors __ Bundle Power Bundle Flow Initial Design Local Radial Arial R-Factor (MWt) (1000 lb/hr) MCPR Exposure: BOC12 to EOC12 lO BP8x8R 1.20 1.72 1.40 1.051 5.807 97.3 1.32 GE8x8EB 1.20 1.73 1.40 1.051 5.852 99.4 1.32 40

! 8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)

.i i

j Transient Recategorization: No Recirculation Pump Trip: No lO Rod Withdrawal Limiter: No Thermal Power Monitor: No j Improved Scram Time: Yes (ODYN Option B)

Exposure Dependent Limits: No Exposure Points Analyzed: 1

9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3) i
O Single-Loop Operation
Yes
Load Line Limit
Yes a Extended Load Line Limit: Yes
g Increased Core Flow
No Flow Point Analyzed: N/A j Feedwater Temperature Reduction: Yes ARTS Program: No Maximum Extended Operating Domain: No O

)

I i

i iO 2

i b

!O 9 i

v, ~ , - - - , - ----,.. .-, ,-,,,n - - , , ,,--w ., .-, , . , -

O 23A4345 Rev. 0
10. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)

(O Methods Used: GEMINI Exposure Range: BOC12 to EOC12 Flux Q/A AN "O

Transient (% NBR) (% NBR) BP8x8R GE8x8EB Figure Load Rejection Without Bypass 452 126 0.26 0.25 2 Loss of Feedwater Heating 117 117 0.15 0.15 3

!O Feedwater Controller Failure 108 107 0.06 0.06 4 O

11. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT

SUMMARY

(S.2.2.1)

Limiting Rod Pattern: Figure 5

O -

Rod Block Rod Position ACPR MLHGR (kW/ft) l Reading (ft withdrawn) BP8x8R GE8x8EB BP8x8R GE8x8EB 104 3.5 0.12 0.12 15.92 16.92

O 105 3.5 0.12 0.12 15.92 16.92 106 4.0 0.15 0.15 17.02 18.02 i 107 4.0 0.15 0.15 17.02 18.02  !

108- 6.0 0.24 0.24 18.02 19.02'  !

O 109 7.0 0.26 0.26 18.02 19.02 i 110 8.0 0.26 0.26 18.02 19.02 l Setpoint Selected: 108 l \

1 ib g l

'O 10

12. CYCLE MCPR VALUES (S.2.2 and S.2.5.4) lO Non-Pressurization Events Exposure Range: BOC12 to EOC12 BP8x8R GE8x8EB O

Loss of Feedwater Heater 1.22 1.22 Fuel Loading Error 1.07 1.07 Rod Withdrawal Error 1.33 1.33 O

Pressurization Events

  • Exposure Range: BOC12 to EOC12 O

Option A Option B BP8x8R GE8x8EB BP8x8R GE8x8EB O Load Rejection Without Bypass 1.39 1.39 1.34 1.34 Feedwater Controller Failure 1.19 1.19 1.14 1.14

13. OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.3)

O s1 Y Transient (psig) (psig) Plant Response MSIV Closure (Flux Scram) 1249 1268 Figure 6 O

O

  • 0DYN Adjustment Factors are documented in a letter from J.S. Charnley (GE) to G.C. Lainas (NRC), " GEMINI /0DYN Statistical Adders for BWR/2,3 Plants (without g RPT) - MOC/EOC," March 13, 1987.

O 11

lC 23A4845 Rsv. 0

14. STABILITY ANALYSIS RESULTS (S.2.4) 1 O

Millstone Point Nuclear Power Station Unit 1 is exempt from the current requirement to submit a cycle-specific stability analysis as documented in the letter from C.O. Thomas (NRC) to H.C. Pfefferlen (GE), " Acceptance

.O for Referencins of Licensins Topical Report NEDE-24011, Rev. 6, Amendment 8, ' Thermal Hydraulic Stability Amendment to GESTAR II'," April 24, 1985.

15. LOADING ERROR RESULTS (S.2.5'.4)

O Variable Water Gap Misoriented Bundle Analysis: Yes Event aCPR O

Misoriented 0.0

16. CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1)

O Millstone Point Nuclear Power Station Unit 1 is a Banked Position With-drawal Sequence plant, so the control Rod Drop Accident Analysis is not required. NRC approval is documented in NEDE-24011-P-A-8-US, May 1986.

O I

l, O

O t

O O 12 L - -

O 23A4845 Rev. 0
17. LOSS-OF-COOLANT ACCIDENT RESULTS (S.2.5.2) l iO LOCA Method Used: SAFE /REFLOOD/ CHASTE i See " Loss-of-Coolant Accident Analysis Repcrt for Millstone Nuclear Power lO Station Unit 1 , ceneral Electric Company, July 1980 (NEDO-24085-1, as amended).

Technical Specification MAPLHGR Limits for.BD338A Fuel Bundle iO MAPLacR (kw/Ft)

Exposure Most Least Oxidation (GWd/ST) Limiting

  • Limiting
  • PCT (*F)** Fraction **

0.2 10.38 10.70 2076 0.065

.O 1.0 10.55 10.80 2107 0.070 5.0 n.40 n.40 2195 0.085 10.0 n.50 n.50 2199 0.083 15.0 n.50 n.50 2198 0.083 20.0 n.40 n.40 2194 0.081 25.0 n.04 11.06 2197 0.082

O 35.0 9.24 9.27 2058 0.053 45.0 7.44 7.47 1822 0.020 0.005 50.0 6.54 6.58 1672
O  ;

I l

i l

l

!O

!O

  • Enriched Lattices only. Natural lattices are non-limiting for LOCA events.
    • Maximum lattice values.

!O 13

~O

O' 23A4845 Rw. O IO 1

MMMMMMM o  :: MMMMMMMHH
MEMMMMMMMMM
MMMiGMMMMMMMMiG
M M M M M M M M M M M M M
MMMMMMMMMMMMM o : 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

'::MMMMMMMMMMMMM

M M M M M M M M M M M M M
MEMMMMMMMMM l o  : MMMMMMMMM
MMMMMMM IIIIIIIIIIIIIi 1 1 357 9111315171921232527293133353739414345474951 O

O FUEL TYPE A = BP8DRB300 (Cycle 10) C = BD338A O B = BP8DRB300 (Cycle 11)

Figure 1. Reference Core Loading Pattern O ii~

23A4845 R:;v. O D

1 hfu1RON FLth  ! VESSEL Pats ' RISttPSI) 2 Avt SLFF AC(' p( AT FLUI 2 $AF(Tf VA(W , FLQg Os g e g, ,

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D D Figure 2. Plant Response to Generator Load Rejection Without Bypass '

(EOC12) 15'

)

23A4845 Rsv. 0

.O I 3 vts itt r I esturam FLuz 2 REL ltf .atS5 ALyt alSEtP513 Ft0s .

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O Figure 3. Plant Response to Loss of 100*F Feedwater Heating 16

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23A4845 Rsv. 0 l

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I geturatps Flux i VE5LEL PRESS RISC(P51)

! 2 $AF TT VALv[ FLOW 2 AVE $URfACE p(AT FLUE 3 REL l[F WALvt Flow 3= COR:

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O Figure 4. Plant Response to Feedwater Controller Failure (EOC12)

J 17

() - 23A4845 Rsv. O l

' C) i C) 2 6 10 14 18 22 26 30 34 38 42 46 50 51 12 12 12 47 0 10 10 0 0

43 12 32 12 39 0 10 6 6 10 0 35 12 32 24 32 24 32 12

.O 31 10 14 0 0 14 10 27 12 32 16 32 12 23 10 14 0 0 14 10

)

19 12 32 24 32 24 32 12 4

15 0 10 6 6 10 0 11 12 32 12

C) 7 0 10 10 0 3 12 12 12 C)

NOTES:

lc)

1. No. indicates number of notches withdrawn out of 48. Blank is a Withdrawa Rod. .
2. Error Rod is (22, 31).

O Figure 5. Limiting Rod Pattern

-O 18

n Rsv. O V 23A4845 .

l 1

O 8 VESSEL PRESS RISEEPSI)

I Neuf RW F.ur 2 $4F( f 7 V A,vt FLOW 2 Avf $W A:( W AT FLUE 8 c""E 8*E' ' ' " .....

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2 OtPPLER AC7iv!TY M

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nar int =u riac <u. = u lO 40 Figure 6. Plant Response to MSIV CLOSURE (Flux Scram) (EOC12) 19/20

O

7_

O 23A4845 Rt.v. O APPENDIX A O GETAB AND TRANSIENT ANALYSIS INITIAL CONDITIONS l

The values used in the GETAB and fransient Analysis for non-fuel power l fraction and Omfaty/ Relief valve capacities and setpoints are given in Table O A-1. The followins values differ from the values reported in NEDE-24011-P-A-8-US, May 1986.

O Table A-1 i PLANT PARAMETER Parameter Analysis Value NEDE-24011 Value O

Non-Fuel Power Fraction 0.038 0.035 Safety / Relief Valve (SRV)

Number of SRVs at:

O Lowest Setpoint Capacity I (psig) (1b/hr) 1095 791,000 0 4 O 1095 829,000 3 2 1125 791,000 3 0 O

O O

o 21/22

23A4845 Rev. 0

.O APPENDIX B jQ FEEDWATER TEMPERATURE REDUCTION AT EOC12 Analyses were performed for end-of-cycle (EOC) 12 operation with the last-stage feedwater heaters valved out-of-service, in order to justify opera-lO tion with feedwater temperature reduced by 75'F. The pressurization events of Section 12 were reanalyzed for operation at the reduced feedwater tempera-ture. This appendix presents the results of these transient analyses.

O The balance of the safety analysis required to justify operation at a reduced feedwater temperature (as defined in Reference B-1) will be provided by NUSCo.

O

REFERENCES:

B-1. " General Electric Standard Application for Reactor Fuel",

NEDE-24011-P-A-8-US, dated May 1986.

.O 4

B.1 CORE AVERAGE EXPOSURE

Assumed reload core average exposure 22,371 mwd /MT iO for Feedwater Temperature Reduction (FWTR) analysis (Extended EOC12)

B.2 RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS

O Peaking Factors Fuel Bundle Power Bundle Flow Initial Design Local Radial Arial R-Factor (MWt) (1000 lb/hr) MCPR

!O Exposure: EOC12 to Extended EOC12 BP8x8R 1.20 1.76 1.40 1.051 5.951 96.4 1.31 GE8x8EB 1.20 1.75 1.40 1.051 5.926 99.0 1.32 10 23 10 4

!O 23A4845 Rsv. 0 B.3 CORE-WIDE TRANSIENT ANALYSIS RESULTS

o-Exposure: EOC12 to Extended EOC12 Flux Q/A ACPR Transient (% NBR) (% NBR) BP/P8x8R GE8x8EB Figure O Load Rejection Without Bypass 431 124 0.24 0.25 B-1 Feedwater Controller Failure 130 112 0.12 0.11 B-2 B.4 CYCLE MCPR VALUES *
O Exposure Range: EOC12 to Extended EOC12 Option A Option B BP8x8R GE8x8EB BP8x8R GE8x8EB

!O i

i Load Rejection Without Bypass 1.38 1.39 1.33 ' 1.34 j Feedwater Controller Failure 1.24 1.24 1.19 1.19 20 l

! l 3

lO l

IO ,

1 l i

O i
  • 0DYN Adjustment Factors are documented in a letter from J.S. Charnley (GE) to J G.C. Lainas (NRC), " GEMINI /0DYN Statistical Adders for BWR/2,3 Plants (without

!O " " ) ~ ""#E "'~ "***h 13' 1987*

24 O

O 23A4845 Rev. O O

1 NEUTRON Flug i VESSEL PREd5 RISECP51) 2 AVE SURFACE PEAT Flux 2 SAFETY VAOE Flow 3 CORE INLET rLov g ,, , 3 g E{ VALVE QOW ggg,,

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O l LEVEL ([NCH-hEF.5EP.SMRT) [ t h!D REACT!ktTY M 2 VESSEL $ TEAR 0v 2 DCP8tY.M ne u aviiI

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O Figure B-1. Plant Response to Generator Load Rejection Without Bypass, FWTR O 25

0 23A4845 Rev. O o O

=

134.s .

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o Figure B-2. Plant Response to Feedwater Controller Failure, FWTR 26 rJ FINAL

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