Text

Rev 0 to Supplemental Reload Licensing Rept for Brunswick Steam Electric Plant Unit 2,Reload 7,Cycle 8
	ML20238D777
Person / Time
Site:	Brunswick
Issue date:	08/31/1987
From:	Lambert P; GENERAL ELECTRIC CO.
To:	;
Shared Package
ML20238D723	List: ML20238D721; ML20238D749; ML20238D777;
References
	23A5855, 23A5855-R, 23A5855-R00, NUDOCS 8709110387
	Download: ML20238D777 (42)
	v • d • e

_ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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3 23A5855 Revision 0 i

Class I j

August 1987 j

)

1 (23A5855, Rev. 0)

SUPPLEMENTAL RELOAD LICENSING REPORT i

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FOR BRUNSWICK ST"AM ELECTRIC PLANT UNIT 2, RELOAD 7, CYCLE 8

-l Prepared:

P. A. Lambett Fuel Licensing l

Verified 8

P. E. F.111ott i

Fuel Licensing l

\\

Approved:

J. S.

ha ey, Manager Fuel ic sing i

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NUCLEAR ENERGY BU$iNESS OPERATIONS

GENERAL ELECTRIC CCMPANY SAN JOSE, CALIFORNIA 95125 87R91 P

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IMPORTANT NOTICE RIGARMNG 1

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CONTENTS OF. T!IIS REPORT e

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PLEASE READ CAREFULLY

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N-p This report was prepared by General Electric sniely for Carolina Power

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and Light Company (CP&L)'for CP&L's use'with the U.S.LNuclear Regulatory Com-mission.(USNRC) to amend CP&L's operating license of the Brunswick Steam

/

Electric Plant Unit l2.' The information' contained in,this report isthelieved f.t 4

by.Ganaral' Electric to be an accurate and true representation of'the. facts.'

i known,obtained'orprovidedtoGeneralElectric.atthetimethisreportNas-

~

prepared.

i The only undet:akings of the General Electric Compcny respecting informs-tion in this document are contained ~in th'a Supplemental Agreement.' to the' Con-

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1

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tract between Carolina Power and Light Company and General Electric Cespany for Raioad Fuel Supply and Related Services for Erunswick Steam Electric Plant i

Unit 2, and nothing contained in this' document shall be construed as changing-said contract. The use of this'information'except as defined by sa'id: con-tract, 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 5

Electric Company nor any of the contributors to this document makes any repre-sentation 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 liabilif f or damage of any kind which may result from such use of such information.

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l r-ACKNOWLEDGMENT s

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The. engineering.and reload ~' licensing analyses, which form the technical basis. of this Supplemental. Reload Licensing Report, were performed. by T. P.

f Lung and R. E. Polonik of the' Nuclear Fuel and' Engineering Services Department.

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t 23A5835

.Rev. O y

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PLANT-UhXQUE ITEMS (1.0)*

Appe ; dix '. A:- Limiting Conditions for Operation Appendix B: Bases for Limiting Conditions for Operation Appendix C: Safety Reifef Valve Out-of-Service g-Appendix D: : Transient Operating Parameters.

Appendix E: Turbine Control Valve Configuration Appendix F: Use of GEMINI Methods for Cycle 8 ir

-2.

RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)

Cycle Loaded.

Number

?.

Irradiated

-P8DRB265H 5

44 BP8DRB299' 6

184 BP8DRB299 7

148 y

New BD317A.

8 92 BD323A 8

92 Total 560

,o 3.

REFERENCE CORE LOADING PATTERN (3.3.1) i Nominal previous cycle core average exposure at and of cycle:

20,_449. M)!d/MT _

~

Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations:

20,006 mwd /MT Assumed reload cycle core average exposure'a't end of-

)

cycle 20,814 mwd /MT Core loading pattern Figure 1 3

( ) 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 section in the United Statec Supplement, NEDE-24011-P-A-8-US, May 1986.

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23A5855 Rev. 0 4.

CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WOR.TH - NO VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2) r

" ; tuning of Cycle, K,ff Uncontrolled 1.114

~

e Fully Controlled 0.968 Strongest Control Rod Out 0.988 R, Maximus Increase in Cold Core Reactivity with 0.0 Exposure into Cycle, AK 5.

STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)'

Shutdown Margin..(AK)

(20*C, Xenon Free) ppm 600 0.031

- 6..

RELOAD-UNIQUE 11tANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2)

(Cold Water' Injection Events Only)

Void Fraction (%)

41.71 Average Fuel Temperature (*F) 1104 Void Coefficient N/A* (4/% Rg)

-7.34/-9.18 Doppler Coefficient N/A*-(d/*F)

-0.205/-0.195 Scram Worth N/A* ($)

l

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, dated May 1986.

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RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2) g Peakins Factors yg Bundle Power Bundle Flow Initial-Design Ldcal Radial Axial R-Factor (MWe)

(1000 lb/hr)

MCPR Exposures BOC8 to EOC8-2000 mwd /ST

].

^

,1.22 BP/P8x8R 1.20 11.55 1.40 1.051 6.574 111.5 GE8x8EB 1.20 1.56-1.40 1.051 6.602-113.6 1.23 Exposure:.EOC8-2000 mwd /ST to EOC8

.j

'BP/P8x8R 1.20 1.49 1.40 1.051 6.320 113.0 1.28 GE8x8EB 1.20 1.50-1.40 1.051 6.382 114.8 1.27 t

8.

SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)

~

Transient Recategorization:

No Recirculation Pump Trip:

.No-

1) '

Rod Withdrawal Limiter:

No Thermal Power Monitor Yes Improved Scram Times No Exposure Dependent Limits:

Yes 3'.

Exposure Points Analyzed:

EOC8 and EOC8-2000 mwd /ST 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 l'

Flow Point Analyzed:

N/A Feedwater Temperature Reduction:

No ARTS Program No Maximuu Extended Operating Domain:

No J

i 9

23A5855 Rev. O s

~10.

CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)

Methods Used: GEMINI Flux Q/A ACPR Transient

(% NBR)

BP/P8x8R GE8x8EB. Figure Exposure Range: BOCS'to EOC8 Inadvertent'HPCI-123 119 0.15

'O.15 2

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Exposure Range:

BOC8 to EOC8-2000 mwd /ST e

i Load. Rejection Without 379 118 0.15 0.15

.3 Bypass Feedwater Controller 107 105 0.04 0.04 4

Failure Exposure Range: EOC8-2000 mwd /ST to EOC8 l

' Load Rejection Without 381 122 0.20 0.20 5

Bypass l

Feedwater Controller 153 107 0.05

'0.05 6,

Failure I*

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

SUMMARY

(S.2.2.1).

Limiting Rod Pattern: Figure 7 Rod Block

' Rod Position aCPR MLHCR'(kW/ft)

Reading -

(feet withdrawn)

BP/P8x8R

'GE8x8EB

.BP/P8x8R GE8x8EB 104 4.5 0.15-0.15 14.08 15.08

~

l 105 5.0 0.16 0.'16 14.52 15.52 l

L-106 5.5 0.18 0.18 14.90 15.90 107 5.5 0.18 0.18 14.90 15.90 108 6.0 0.19 0.19 15.24 16.24 L

109 8.5 0.24 0.24 16.28 17.28 110 9.5 0.24 0.24 16.28 17.28 Setpoint Selected:

107 10 L____---_-____

D 23A5855-Rsv. 0 '

S 12.

CYCLE'MCPR VALUES'(S.2.2) 7 Non-Pressurized Events BP/P8x8R GE8x8EB' Exposure Ranges' BOC8 to EOC8 4

Inadvertent HPCI 1.22

'1.22 1.20

' Fuel Loading Error Rod Withdrawal Error 1.25 l'25 b

Pressurization Events Option A Option B BP/P8x8R-GE8x8EB BP/P8x8R GE8x8EB Exposure Rang'es BOC8 to EOC8-2000 mwd /ST

)-

Lose Rejection Without Bypass 1.32 1.32 1.25 1.25 Feedwater Controller Failure 1.16 1.16 1.14 1.14 Exposure Range: 'EOC8-2000 mwd /ST to EOC8 y-Load Rejaction Without Bypass 1.33 1.33 1.29

~ 1.29 Feedwater Controller Failure 1.17 1.17 1.14 1.14 f

13. OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.3) 8 si' y

(psig)

- Plant Response Transient-P l-MSIV Closure 1213 1251 Figure 8

('

(Fluz Scram) 11 k,C

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23A5855 Rev. 0

14. LOADING ERROR RESULTS (S.2.5.4)

Variable 1 dater Gap Disoriented Bundle Analysis

.Yes*

Event ACPR e

Disoriented 0.11 15.

CONTROL ROD DROP ANALYSIS RESULTS (5.2.5.1)

Bounding Analysis Results Doppler Reactivity Coefficients Figure 9 Accident Rasctivity' Shape Functions:

Figures 10 sed 11-Scram Reactivity Functions:

Figuree 12 and 13 Plant-Specifit Analysis Results:

Resultant Peak Enthalpy, Colds 139.4 Resultant Peak Enthalpy, HSB:

192.7 s

16.

STABILITY ANALYSIS RESULTS (S.2.4)

Rod Line Analyzed:'

Extrapolated Decay Ratio:

Figure 14 Rosetor Core Stability Decay Ratio, x /x :

0.80 3 0 Channel' Hydrodynamic Performance Decay Ratio, x /*0 2

i i Channel Type 0.31 5P/P8x8R- -

GE8x8EB 0.28 l

ACPP. penalty of 0.02 for the tilted miseriented bundle is applied to the i

cycle MCPR value reported in Section 12.

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17. LOSS-OF-COOLANT ACCIDENT RESULT (S.2.5.2)

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l See " Loss-of-Coolant Analysis Report for Brunswick Steam Electric Plant Unit No. 2," NEDO-24053, September 1977 (as amended), and NEDE-24053-P,

.j August 1987.

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'23A5855-Rev. 0 n

eMMBEMM.

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.BEMMMMMMEL i

i MME8BEMMMMMMM

.NE818MHHBEMBEMM.

MMMMBfMHHBEMMBEM

':MMMMMMBEMMMMBEM

'::BEBRE8888E8258BEBEE88888M L

': M M BER M BsBEM M MBEM M

'::MMBEMMBsBfMMMMMM

"MMMMBsMBEMMMM"

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l' MBEMMMME8BEMMM "BuiGM M M MBEM M" "MMMMM" l l l l 11IIlI 1 357 91113151719212325272S3133353739414345474951 FUEL TYPE B = P8 65H E = BP8DRB299 (Cycle C = BD317A Figure 1.

Reference Core Loading Pattern j

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23A5855 Rsv. O t-k l7 v

Y 1 NEUTRCN FLUX

1. VESSEL PRESS RISE frssi) 4, 4~ 2 AVE SURFACE HEAT FLUX 4~4<

~ 150 3 BYPASS VALVE FLC'N

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' 2 RELIEF VALVE FLOW 3 CORE INLET FLOW 150 4 CORE INLET SU8 4 HPCI FLOW (% of tw)

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3--3-3 3

3 33 100 8 100 3

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50 50

^4-4-4 4e 4 4-4 4 -4 1-1%- 1.

1 1

0 04~3 2-3-2 3 32-3-2 3-2-3 2 0

50 100 0'

50 100 TIME teocondel TIME (seconds) t e.

1 LEV'EL (inch-REF SEP SKRT) 1 VOIO REACTIVITY 2 VESSL1,STEAMFLOW 2 DOPPLER REACTIVITY 1

_3 SCRAM REACTIVITY 150 3 TURaitdE STEAMFLOW 4 FEEDWATER FLOW 4 TOTAL REAC11YlTY g

1-

.- r g _

-1 1

T 2 3 32-3 2 - 32

~

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d h3 34-3---4 434 100 3

0 4-4-4-4 4 "4 - 4 ' 4 ---4 -4 2

2--2 2-2 -- -2 u

E.

D 5

.0 "I

1-1-

_j 1-

~

s 0

-2 J

0 50 100 0

50 100 TIME (seconds)

TIME (secends) j J

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Figure 2.

Plant Response to Inadvertent Startup of HPCI 15 l

I

23A5855 Rev. 0 1 NEUTRON FLUX 1 VESSEL PRESS RISE (psil 2 AVE SURFACE HEAT FLUX 2 SAFETY VALV2 FLOW 3 REUEF VALVE FLOW 3 CORE INLET FLOW 3M 150 4 BYPASS VALVE FLOW Su 200

100, g

2' M N 1

(

g

~2

~3 3-3 1

i 0

01 2 4 2-d2 4-2 1

1 e

2-4-4 0

2 4

6 0

2' 4

6 TIME (seconds)

TIME (secondel i

I LEVEL (hREF SEFSKRT)

/ V 1 VOtD REACTIVITY 2 VESSEL STEAMFLOW 1

2. DOPPLER REACTIVITY 3 SCRAM REACTIVITY j

3 TURSINE STEAMFLOW j

200 4 FEEDWATER FLOW 4 TOTAL REACTIVITY 1

4

/

O

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/2 rn 4

b 2*

2 N

2 i

2 2-

3..

1 2

3 1 1

0 3-3 3

3 3-3

-1 l

. ~..

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~ 0 2-4 6

0 2

4 6

TIME (seconds)

TIME (secondel l

Figure 3.

Plant Response to Generator Load Rejection Without Bypass (EOC8-2000 mwd /ST) i l

16 l

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23A5855 R v. 0 S

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1 NEUTRON FLUX.

1 VESSEL PRESS RISE (pe0 2 AVE SURFACE HEAT FLUX

- 2. SAFETY VALVE FLOW 150 3 CORE INLET FLOW 4 CORE INLET SUS 3 REUEF VALVE FLOW 4 BYPASS VALVE FLOW A.

4 100 4

_4- -s 3-3-3 4

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2 4

tt

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4 50

~

2 50 1

0,6 342-34 3--4 32-3-23--2 0

1 1

-\\

1

'. 0 50 100 0

50 100

[

TIME (seconds)

I q

e 1 LEVEL (inch-REF SEP SKRT) 1 VOID REACTIVITY 2 VESSEL STEAMPLOW 2 DOPPLER REACTIVITY

.I 150 3 TUMNE STEAMFLOW 3 SCM REAGW

);

4 4-4 FEEDWATER FLOW

._ 9 4 TOTAL REACTIVITY I

/

2 2

2 t

1 100 O'

22W '.2 j

2 h

E 4

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50

-t i

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't 4

3 0

3-3-3M

-2 0

50 100 0

50 100 TIME (seconds)

TIME (seconds) l 1

w Figure 4.

Plant Response to Feedwater Controller Failure (EOC8-2000 mwd /ST) g, 17

w E

'2.h5855 Rev. 0 8 NEuta0N FLUE-1 MESSEL PRE 5B RISE (P$1) 2 AVE SURFACE ISAT FLUM 2 SAFETY WALW tPLOW 3 CORE Ile.ET PLOW 3 MLIEF VALW ! FLOW 180 7

4-Y 2-

,.s NN, x

i=

..e s

.o o--

-.ss o

2-4 0

0 2

4 6

L~

TIM (secondal TIME (seconde) i v010 AEACTIkITY 1 LEY t ItsCH.IEF.SEP.$suti) 2 VES L STEAE'Lov 2 DOPPLER Rt fivtTY j

3 SCRAM #EAC vity 3 g ag yu_grLow 200 gj j

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1 100 lo

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v C

o ' fp :

-1 a

-100

-2 0

2 4

S o

2 4

6 TIME (secondel TIME (seconds) i Figure 5.

Plant Response to Generator Load Rejection Without Bypass (EOC8) 18

F 23A5855-

'Rev. 0 t

+ f k

s VEMEL PRES 5 RISE (PSI) f

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t NEurRON Flux 2 SA' [TY WALVE Flow 2 AVE SURFACE HEAT FLUX 3 RELlFF VALVE FLOW 3 CORE IPLET Flow -

4 SYPLS$ VALVE FLOW 2 raar

- =

+ - =

150

~o D.

100 1

M ut i :

B 300,

N 50, t

t 2

60 3

,j O&c ;;,

O-10 20 0

10

20 TIME (seconds)

TIME (seconds)

I, 1 v01 ) REActICITV i lev :L(INcM*REF.SEP.SMRT3 2 00P'LER REAl'T!v!TY~

2 VESLEL STEAMFLOW 3 SCRLM RLACT, vtTY j

3,,f ue, l l.N.E.. STE,,ATL0v 3

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100-

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W

-1

,1 50/

g

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f 0

-2 0

10 20 0

10 20 TIME (seconds)

TIME (seconds) l I

lb Figure 6.

Plant Response to feedwater Controller Failure (EOC8) i b-19 4

l

- - _L_- _ _-____-___:__

4frf

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23A3855 R2v. 0-I 2

.6 10 14 18 -

22 26 30 34 38 42 46-50 52-18 18

~48 36 44 44 44 36

. 44 14 2

10 10 2

14 44 40 44 44 44 44 36 2.

10 2

2 10 2

32 44-44 44 44 10 2

0 0

2 10 28 24 44 44 44 44 20 2

10 2

2 10 2

s 16 44 44 44 44 44 12 14 2

10 10 2

14 8

36 44 44 44 36 l

4 18 18 l

NOTES:

1.

NUPEER INDICATES NUMBER OF NOTCHES WITHDRAWN OUT OF 48. BLANK IS A WITHORAWN ROD.

2.

ERROR ROD IS (22,28).

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Figure 7.

Limi. ting Rod Pattern 20

( :-

l

)

23A5855 Rsv. 0 L

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1. VESSEL PRESS RISEtPSI) l 1 NEUTRON F.UX 2 SAFETY YA.YE FLOW 2 AVE SURFAZ W AT FLUX 3 RELIEF VA.YE FLOW 3 CORE IPLET FLOW

^ ~ ~ ' " " ' ' '

" " " ' ^ "

300 150 y

- 8 5

^

200

100 i

1 100 y

50 -

p 0

0--

5 0

TIME (seconds)

TIME (seconds) e

! LEVELtINC4REF.SEP.SKAT3 O REACTIVITY PLER REACTIVITY 2 VESSEL STEAMFLOW U& !!!

?

1

a 200 V

-2 en O

' 'y ~

-w__

=

100 U

D g -t 0

E

- 100

.2-4 O

5 0

5 TIME (seconds)

TIME (seconds)

Figure 8.

Plant Response to MSIV Closure - Fluz Scram i

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g' as'f a y

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-30.0 T,

-35.0 A cAtct-LAltu vaLut -LULU B CALCl.'LATED VALUE -HSB C BOUNC VAL 280 C/ L/G COLD.

D BOUNC VAL 280 C/ L/G HSB

-40.0

0. 0 500.0 1000.0 1500.0 2000,0 2500.0 3000.0 FUEL TEMPERATURE DEG C.

F5 g' ure 9.

Fuel Doppler Coefficient in 1/A*C 22

23A5855 Rev. 0 6

20.0 17.5 i

15.0 m

A0 0 0

12.5

'4 W

_J.

/,

l.

g 10.0 r

7.5 a+

U to a:

5. 0 2.5 A ACCIDENT FUNCTION B BOUNDING '/ALUE 280 CAL /G 0, 0,,

0. 0 5.0 10.0 15.0 20.0 ROD POSITION, FEET OUT Figure 10.

Accident Reactivity Shape Function - Cold Startup 23

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20.0 E

s

'17.5 i

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10.0

-- O.

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A ACCIDENT FUNCTION l

B BOUNDING 'fALUE 280 CAL /G t

0. 0,

O. 0 5.0 10.0 15.0 20.0 L.

l-R00 POSITION, FEET GUT l

Figure 11. Accident Reactivity Shape Function - HSB i

1.

1'.

I-24

23A5855 R';v. 0 s

30.0 A SCRAM FLNCTION B BOUNDING VALUE 280 CAL /G 25.0 1

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4 3

q w

Q W

15.0 c.ow

-z v

i H

l 10.0 I

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1 5.0 1

. /"'

O. 0,;

- 0. 0

1. 0

2. 0 3.0

4. 0

5. 0 6.0 l

ELAPSED TIME, SECDNDS L

Figure 12.

Scram Reactivity Function - Cold Startup 25 9

~. _ _ _ _ _ _ _ _ _ _ _

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1 23A5855 Rsv. 0 i

50.0 l A SCRAM FLNCTION-B BOUNDING VALUE'280 CAL /G 40.0 i

m o

-1 w-x 30.0 l

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0. 0,,

0. 0

1. 0

2. 0 3.0 4.0

5. 0 6.0 ELAPSED TIME, SECONDS i

Figure 13.

Scram Reactivity Function - HSB 4

26

= - - __-_--- - - ___ -_ _

c.

u

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g.

23A5855 Rsv. 0 f

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1.25 W

A NATURAL CIRCULATION B 105 PEHCENT MOD UNE 1.00 4

/

0.73 AB' o

g x"

/

'k 0.50

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0.25 A

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POWER (%)

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i Figure 14. Reactor Core Decay Hatio Versus Power 3

27/28

~23A5855 Rev.'O APPENDIX A LIMITING CONDITIONS:FOR,0PERATION This appendix'provides the limiting condition.for operation (LCO)'for

'each of the power distribution limits identified belows (1) Average Planar Linear Heat Generation' Rate (APLHGR)

(2). Operating Limit MCPR (3) APEM Setpointa k-Surveillance requirements and required actions are specified.in the Tech-nical. Specifications. -The power distribution limit bases are given in Appen-

~

dix B..

.. A.1 APLEGR.

During steady-state power operation, the APLHCR for each type of fuel c.s g.

a functica of ar4al location'and average planar exposure shall not ezceed limits based on applicable'APLHGR limit values which have been approved for the resp.ctive fuel and lattice types determined by the approved methodology described in GESTAR-II (NEDE-24011-P-A). When hand calculations are required, the APLHGR for each type of fuel as s' function of average planar exposure

-shall not exceed the limiting value for the most limiting lattice (excluding natural uranius) shown in Figures A-1 through A-4,~during two recirculation loep operation.

A.2--OPERATING LIMIT MCPR n

-The fuel cladding integrity safety limit MCPR is 1.07.

During steady-P state power operatien, the MCPR for each type of fuel shall not be less than

'the limiting value (shown in Table A-1) times the Kf (shown in Figure A-5),

for two recirculation loop operation.

p-L

~

29 L

l 23A5855 Rev. 0

. In reference to Technical Specification 3.2.3.2,- the OLMCPR for t,y, less than or equal to T is1the greater of the non pressurization transient or' 3

the Option B OLMCPR (Table A-1), where T,y, and T3 are give:2 by:

I n[N T

g g ial ave n.[N y i=1 wheres I

'i = Surveillance test number.

n = Number of surveillance tests performed to date in the cycle (includ-ing BOC).

th N -= Number of rods tested in the i surveillance test.

g 5

Y = Average scras time to notch 36 for surveillance test 1.

g and-

\\1 Ny T = u + 1.65 e

fNg i=1 where N = Number of rods tested at BOC.

y p = 0.813 seconds (seen value for statistical scram time distribution

-from de,-energization of scram pilot valve solenoid to pickup on notch 36).

k'

.o = 0.018 seconds (standard deviatica of the above statistical distribution).

30

N 23A5855 R2v. 0 In reference to Technical Specification 3.2.3.2, the OLMCPR for T,y, I

greater than T3 shall be either:

a..

The greater.of the non pressurization transient (Table A-1) or the adjusted pressurization transient MCPR (MCPRadj) where:

T T

- - ~

MCPR

- MCPR MCPR

= MCPR adj Option B +

A ~'B Option A Option B T

D T = 1.05 seconds (control rod average scram insertion A

time limit to notch 36).

and MCIR as given in Table A-1 g

0ption A MCPR as given in Table A-1 0ption B or, p,

b.-

MCPR as g ven in able A-1.

0ption A y

A.3 APRM SETPOINTS The flow-biased APRM scram trip setpoint (S) and rod block trip setpoint (SRB) shall bes B

S 1 (0.66W + 54%)T, and S

1 (0.66W + 42%) T; RB where S and S are in percent of rated thermal power; g3 W n loop recirculation flow in percent of rated flow.

T is the ratio of Fraction of Rated Thermal Power (FRTP) divided by Core Maxi-g mum Average Planar Linear Heat Generation Rate Ratio (CMAPRAT):

T=

where T i 1.

I E

31

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ue s:

,i

l c'

l 23A5855 Rev. 0 8

4 8

8

.~

$d l

2 1

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6

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~

s m

$ N. Y N 0,

S 5

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2 v

it N

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5 m

I I

e o

9 N.

h Figure A-5.

K Factor g

36

_______m.___.

l l :,

23A5855 Rev. 0 Table A-1 0-MCPRs Fuel Type:.P8x8R, BP8x8R, and GE8x8EB

'O Non-Pressurized Transient MCPR = 1.25 Pressurization Transients Exposure Range MCPR0ption A Option B

~

BOC8 to EOC8-2000 mwd /ST.

1.32 1.25 EOC8-2000 mwd /ST to EOC8 1.33 1.29 9

4 9

ty E'

E4 37/38

)

= _ _

i b,,,,s

'I L

23&5855 Rev. O

,7 I

g' 4

APPENDIX B O

BASES FOR LIMITING CONDITIONS FOR OPERATION inis appendix provides the bases for each of the power distribution ~

limits identified in Appendix A.

B.1 APLHGR'

'This specification' assures that the peak cladding temperature (PCT) fol-3",

lowing the postulated. design basis loss-of-coolant accident (LOCA) will not-exceed the limits specified.in 10CFR50.46 and that the fuel mechaeical' design y,

analysis limits'specified in Reference B-1 will not be exceeded.

Thermal' Mechanical Design Analysis NRC' approved methods (specified in Reference B-1) are used to demonstrate. that all fuel. rods in a lattice oper-sting at the k - M as power history meet the fuel design limits specified in Reference B-1.

No single fuel rod follows,or is capable of following,.'this bounding power history. This bounding power history is used as the basis for the fuel design' analysis APLEC'a limit.

-LOCA Analysis A LOCA analysis is performed.in accordance with 10CFR50, Appendiz %.to demonstrate that the permissible planar power (maximus APLHGR)

. limits comply with the ECCS limits specified in 10CFR50.46. The analysis is performed for the most limiting break size, break location, and single failure combination for the plant. The methods used are discussed in Reference B-2.

The APLHGR limit is the most limiting composite of the fuel design analy-sis'APLHGR limit and the ECCS APLHGR limit.

i.,

B.2 OPERATING LIMIT MCPR

The required operating limit MCPRs at steady state operating conditions as specified in Appendix A are derived from the established fuel cladding integrity safety' limit MCPR specified in Appendix A and an analysis of s

Of 39

1

^b 23A5855 Rev. 0' abnorma1' operational transients. For any abnormal operating transient'ans 9-

' sis evaluation with tha initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the safety limit MCPR at any time during the transier..t, assus-ing instrument trip setting as:siven in Specification 2.2.1 ef the Technical Specifications.

To assure that the fuel cladding integrity safety limit is not exceeded during any anticipated abnormal operational. transient, the most limiting tran-sients have been analyzed to determine which ones result in the largest reduc-tion in Critical Power Ratio (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity ins'ert' ion, and coolant temperature decrease.

1 The codes used to perform the transient analyses that serve as the basis for the operating limit MCPR are described in Reference B-1.

Conditions at limu ing exposures are used for nuclear data to provide conservatism relative to core arposure aspects. Plant-unique initial conditions and system param-eters are used as inputs to the t*z,ssient codes. The ACPR' calculated by the transient codes is adjusted using NRC approved adjustment factors to account for code uncertainties and to provide a 95/95 licensing basis.

The limiting transient yields the largest ACPR. The ACPR for the limiting transient is added to the fuel cladding integrity safety limit to MCPR to establish the minimum operating limit MCPR.

c The purpose of the K factor is to define operating limits at other g

than rated flow conditions. At less than 100% flow, the required MCPR is the product of the operating limit MCPR and the K factor. Specifically, the f

K factor provides the required thermal margin to protect against a flow g

increase transient. The most limiting transient initiated from less than rated flow conditions is the recirculation pump speedup caused by a actor generator speed control failure.

o e

40

fc, 7

i 23A5855 Rsv. O L

t 4

for operation in the automatic flow control mode, the K " actors assure 1

F that the operating limit MCPR in Appendix A will not be violat- \\ should the L

most limiting transient occur at less than rated flow. In'the sanual flow I-fcontrol mode, the K factors assure that the safety limit MCPR will not be g

- violated should the most limiting transient occur at less than rated flow.

f-The K factor values are generically developed as described in f

- Reference B-1.

F The K factors are conservative for the General Electric plant opera-f tion because the operating' limit MCPRs in Appendix A are greater than the original 1.20 operating limit MCPR used for the generic derivation of K.

f At core thermal power levels less than or equal to 25%, the reactor will be operating at minimum recirculation pump speed and the moderator void con-

~

tant will be very small.

For all' designated control rod patterns which may be employed at this point, operating plant experience indicated that the result-ing MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place

-operation in a more conservative mode relative to MCPR.

During initial startup testing of the plant, a MCPR evaluation will be made at 25% initial power level with minimum recirculation pump speed.

The MCPR sargin will thus

'be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR above 25% rated thermal power is sufficient, since power distribution shifts are i

very slow when there have not been'significant power or control rod changes.

The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape, regardless of magnitude that could place operation at a thermal limit.

I I

41 9

i

C' 23A5855 Rev. O I

i B.3. APRM SETPOINTS The flow-biased thermal power upscale scram setting a,nd flow-biased neu-tron flux upscale control rod block functions of the APRM instruments are adjusted to ensure that fuel design and safety limits are not exceeded. The scram setting and rod block setting are adjusted in accordance with the for-mula in Appendix A when the combination of Thermal Power and CMAPRAT indicate a highly peaked power' distribution. This adjustuent may be accomplished by increasing the APRM' gain and thus reducing the slope and intercept point of the flow referenced APRM high flux scram curve by the reciprocal of the APRM gain change.

B.4 REFERENCES 1.

" General, Electric Standard Application for Reactor Fuel", NEDE-24011-P-A (latest approved revision).

e.

2.

" General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix K", NEDO-20566, January 1970.

h e

42

k 23A5855 Rsv. O i

f 4

L.'

APPENDIX C 1.

SAFETY-RELIEF VALVE OUT-OF-SERVICE The. analysis was performed for safety-relief valve out-of-service and there was no change in the ACPR. The change in pressure for MSIV with flux 4* '

scram is shown below.-

E P

s1 y

(psig)

Plant Response MSIV Closure 1225 1261 Figure C-1 (Flux Scram)

'/

k

/

4 9

i 4

43 J

l

23A5855 Rev. O o

l 1 NEUTRON FLUX 1 VESSEL P'RESS RtSE (psu i

2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLOW

{,

3 CORE INLET FLOW 3 REUEF VALVE FLOW 15 0.0 300.0 4 BYPASS VALVE FLOW 2/ \\\\

5 2#3~33 4

'1 m t

- 100.0 1

3 3-200.0 E

3%

2 8

3

.{

3

-g

(

/

2 iOO.0 3-3 3

3 --

-- 3 4

H--4'3 /

0. 0 1_

j 0.0 2 324-242 24--4-2 4

0. 0 5.0 C.9

5. O TIME (SECONDS)

TIME (SECONOS)

I i

c 1 LEVEL (Ir$:h-REF-SEP SKRT)

\\ 1 VOf0 Rf! ACTIVITY 2 VESSEL JTEAMFLOW 2 OOPPLER REACTIVITY 2H.1 3 TURBMF. STEAMFLOW 3.0 3 SCRAM F,EACTIVITY 4 FEEDWATER FLOW

/

4 TOTAL REACTIVITY

~

[\\

/*

'A b

/

E 0.4 w 4 2 4 4 4-4h

,4%,,,4

1. J 4

4 -7 -----

33 254 2"4

.; 2 3

2 2

\\

3

~~

3 0

0. 0 2

3s

-1.3 3

3-3 4

-l**.*

l

\\

l l

2.c O. 0 5.0

0. 0

5. 0 TIME (SELCh0A1 rgME (SECONOS) i 1

o Figure C-1.

Plant Response to MSIV Closure - Flux Scram (SRV005) 44

f.'

J 23&58S5

~Rsv.-0

..4 APPENDIX D p' '

PLANT P.UUV.ETER DIFFERENCES

.o m

-CETAB a.nd' Transient Analysis Initial Conditions

$2 The values used in the GETAB and Transient Analysis t.re given in Table D-1. lThe' following~ values ' differ from the values reported 1n

~

7

. Tables S.2-4.1 and S.2-6 in NEDE-24011-P-A-8-US, May 1986'.

Table D-1.

PLANT PARAMETER Paraineter Analysis Value NELE-24011. Vape

-Dome Pre'ssure 1005 1020 1 2 pai Rated Steam Flow 10.47 10.96-1 0.2%

P Turbine Pressure-.

950 960 1 2 psi Non-Fuel Power' Fraction 0.039 0.04 k

i l

l y

0 e

.s 4

L 45/45 g

I

I 4

h 23A3855 Rev. 0

)

4 APPENDIX E a

,)

TURAINE CONTROL VALVE CONFIGURATI0N

-13 Tha' traasient GEIAB analyses presented in the body of this report are based on turbina control valves in a full-arc. configuration and on the power D-supply to the recirculation Motor-Generator Sets from offsite power.

I l

1

.i

~.

r S

e 9

i

.e

l 3

L e

47/48

t k

23A5855 Rev. 0 j' '.

APPENDIX F USE OF GEMINI METHODS' FOR CYCLE 8 The analyses required for this cycle were performed with GE's' advanced reload licensing methods, known as GEMINI. Any differences between this

~

R reload and the previous one Are due not only to cycle differences, but also to the' difference in the methods. Therefore, making direct comparisons between the two cycles'will be inconclusive.

T,*1,

\\

g b

gy d

,o.

j l

49/50 9

(FINAL)

_ _____- __.. _, _ -

ML20238D777

Text

SUMMARY

SUMMARY

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