Text

Suppl Reload Licensing Rept for Bsep,Unit 1,Reload 9 Cycle 10
	ML20083R808
Person / Time
Site:	Brunswick
Issue date:	02/28/1995
From:	Hull G, Klapproth J; GENERAL ELECTRIC CO.
To:	;
Shared Package
ML19344C916	List: ML20083R747; ML20083R792; ML20083R808;
References
	24A5159, 24A5159-R, 24A5159-R00, NUDOCS 9505300160
	Download: ML20083R808 (27)
	v • d • e

_ _ _

l3 l

GENuclearEnergy 1

24A5159 Revision 0 s

Class I February 1995 24A5139, Rev. 0

)

Supplemental Reload Licensing Report for Brunswick Steam Electric Plant Unit 1 Reload 9 Cycle 10

.)

i 1

)

I i

s

)

Approved hh Appmved hk J.F. Klapproth, er G.R. Hull Fuel and FacilityI.icensing Fuel Project Manager 9505300160 950515 DR ADOCK 0500 5

BRUNSWICK 1 i

Reload 9

\\

24A515!

Rev. (

i Important Notice Regarding Contents of This Report Please Read Carefully

)

This report was prepared by General Electric Company and Light Company (CP&L) for CP8 L's use in defin o ey forCarolina Power Steam Electric Plant Unit 1. The information co to be an accurate and true representation of the facts k e runswick I

eportis believed by GE GE at the time this report was prepared.

o a ned orprovided to.

the contract between CP&L and GE for n ument are contained in shallbe construed as changing said con services forthe nuclear 3

fined by said contract, or for any purpose other than that f n sdocument authorized; and with respect to any such unauthorized exceptas de-contributors to this document makes any representa plied) as to the completeness, accuracy or usefulness of nor any of the document or that such use of such information may not infri ressed orim-s on containedin this from such use of such information.nor do they assume a w chmayresult

)

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i e

Page 2

BRUNSWICK 1 24A5159 Reload 9 Rev.0 1

Acknowledgement The engineering and n: load licensing analyses, which form the technical basis of this Supplemental Reload Licensing Report, were performed by M.R. Morris. The Supplemental Reload Licensing Report was pre-pared by M.R. Morris. This document has been verified by R.N. Anderson.

e i

b I

Page 3

1 BRUNSWICK 1 24A5159 Reload 9 Rev.0 The basis for this report is General Electric Standard Applicationfor Reactor Fuel, NEDE-24011-P-A-10, February 1991; and the U.S. Supplement, NEDE-24011-P-A-10-US, March 1991, i

1.

Plant-unique Items Appendix A: Analysis Conditions Appendix B: Main Steamline Isolation Valve Out of Service Appendix C: Reload Unique Anticipated Operational Occurrence (AOO) Analysis input Appendix D: Final Feedwater Temperature Reduction (FFWTR)

Appendix E: Maximum Extended Operating Domain 2.

Reload Fuel Bundles Cycle Fuel Type Loaded Number Irradiated:

GE8B-P8DQB339-11GL80M-4WR-150-T (GE8x8EB) 7 60 GE8B-P8DQB323-10GL80M-4WR-150-T (GE8x8EB) 7 60 GE10-P811X8322-11GL70M-150-T (GE8x8NB-3) 8 108 GE10-P811XB324-12GL70M-150-T (GE8x8NB-3) 8 52 GE10-P811XB320-11 GL100M-150-T (GE8x8NB-3) 9 68 GE10-P811XB322-11GL70M-150-T (GE8x8NB-3) 9 56 Ncm i

GE 10-P811XB346-12GL100M-150-T (GE8x8NB-3) 10 156 Total 560 3.

Reference Core Loading Pattern Nominal previous cycle core average exposure at end of cycle:

24864 mwd /MT (22556 mwd /ST)

Minimum previous cycle core average exposure at end of cycle 24547 mwd /MT from cold shutdown considerations:

(22269 mwd /ST)

Assumed reload cycle core average exposure at beginning of 15484 mwd /MT cycle:

( 14047 mwd /ST)

Assumed reload cycle core average exposure at end of cycle:

26734 mwd /MT (24253 mwd /ST)

Reference core loading pattem:

Figure 1 Page 4

s

- BRUNSWICK 1 I,.

Reload 9 24A5159 Rev.0 4.

Calculated Core Effective Multiplication and Control Syst

-)

Beginning of Cycle, kenecuve Uncontmlied 1.111 Fully controlled 0.%9

.)

Strongest control md out 0.989 R, Maximum increase in cold core reactivity with exposureinto cycle, Ak 0.000 5.

Standby Liquid Control System Shutdown Capability 1

Bomn Shutdown Margin (Ak)

(ppm)

(20*C, Xenon Free) 600 0.029

.)

6.

Initial Condition ParametersReload Unique GETAB Anticipa Exposure: BOC10 to EOC10-2000 mwd /MT (1814 mwd /ST) 7eaking Factors

')

Fuel Design Local Radial Axial R-Factor Power Flow MCPR Bundle Bundle Initial I

(MWt)

(1000lb/hr)

GE8x8NB-3 1.20 1.70 1.40 1.000 7.203 107.6 1.21 3

Exposure: EOC10-2000 mwd /MT (1814 mwd /ST) to EOC10 B Peaking Factors 1

Fuel 1

Design Local Radial Axial R-Factor Power Flow MCPR Bundle Bundle Initial

'i (MWt)

(1000lb/hr)

GE8x8NB-3 1.20 1,62 1.40 1.000 6.897 109.7 1.27 i

Exposure: BOC10 to EOC10 BRUNSWICK 1 CIO MSIVOOS Peaking Factors Fuel Design Local Radial Axial R-Factor Power Flow MCPR Bundle Bundle Initial (MWt)

(1000lb/hr)

GE8x8NB-3 1.20 1.64 1.40 1.000 6.944 109.4 1.26 i

Page5

m BRUNSWICK I i3 Reload 9 24A5159 Rev.0 Exposure: BRUNSWICK 1 CIO Extended EOCIO with FFWTR 3

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

(1000lb/hr)

GE8x8NB-3 1.20 1.67 1.40 1.000 7.070 108.1 1.26 7

7.

Selected Margin Improvement Options Recirculation pump trip:

No Thennal power monitor:

' 'es improved scram time:

Yes (ODYN Option B)

Measured scram time:

)

No Exposure dependentlimits:

Yes Exposure points analyzed:

2 (EOC10-2000 mwd /MT and EOC10) 1 8.

Operating Flexibility Options Single-loop operation:

Yes Load line limit:

Yes s

Extended load line limit:

Yes Maximum extended load linelimit:

Yes increased core flow throughout cycle:

Yes Flow point analyzed:

105.0 %

Increased core flow at EOC:

Yes Feedwater temperature reduction throughout cycle:

No

)

Final feedwater temperature reduction:

Yes Temperature reduction:

130.0'F ARTS Program:

Yes Maximum extended operating domain:

Yes Page 6

BRUNSWICK 1 Reload 9 24A5159 i

Rev.0 Moisture separator reheater OOS:

No Turbine bypass system OOS:

No Safety / relief valves OOS:

(credit taken for 9 of11 valves)

Yes

)

ADS OOS:

Yes (2 valves OOS)

EOCRIT OOS:

No Main steam isolation valve OOS:

Yes

')

9.

Core-wide AOO Analysis Results Methods used: GEMINI; GEXI -PLUS Exposure range: BOC10 to EOC10-2000 mwd /MT (1814 mwd /ST)

)

Uncorrected ACPR Event Flux Q/A GE8x8NB-3

(%NBR)

(%NBR) 5

)

Load Reject w/o Bypass 376 117 0,14 Turbine Trip w/o Bypass 2

354 116 0.14 3

Exposure range: EOC10-2000 mwd /MT (1814 mwd /ST) to EOCIO

)

Uncorrected ACPR Event Flux Q/A GE8x8NB-3

(%NBR)

Fig.

Load Reject w/o Bypass 508 122 0.20 i

Turbine 1Tip w/o Bypass 4

521 121 0.20 5

Exposure range: BOC10 to EOCIO BRUNSWICK 1 CIO MSIVOOS Uncorrected ACPR R

Event Flux Q/A GE8x8NB-3

(%NBR)

Fig.

)

l

\\

Load Reject w/o Bypass 435 120 0.19 Turbine Trip w/o Bypass 6

427 120 0.19 7

Page 7 i

i BRUNSWICK 1 Reload 9 24A5159

)

Rev.0 l

Exposure range: BRUNSWICK 1 C10 Extended EOC10 with FFWTR Uncorrected ACPR Event Flux Q/A GE8x8NB-3 Fig.

(%NBR)

FW Controller Failure 402 124 0.19 8

3 l

10. Local Rod Withdrawal Error (With Limiting Instrufnent Failure) AOO Summary The rei withdrawal error event in the maximum extended operating domain was originally analy GE BWR Licensing Report, Maximum Extended Operating Domain Analysisfor Brunswick Steam E Plant, NEDC-31654P, dated February 1989. The MCPR limit for rod withdrawal erroris bo erating limit MCPRs presented in Section 11 of this report for RBM setpoints shown in Tables 1 10-5(b) of NEDC-31654P.

)

11. Cycle MCPR Values,2 1

Safety limit:

1.07 Single loop operation safety limit: 1.08 Non-oressurization events:

Exposure Range: BOC10 to EOC10 GE8x8NB-3 f

Fuel Loading Error (misoriented) 1.25 j

Fuel Loading Error (mislocated) 1.24 LFWII

)

1.20

)

1. Brunswick Suom Flectric Plant liniu l and2 Single-Loop 0peranon. NEDC-31776P December 1989, Operating l ing limit MCPR need not be changed for St,0. loop operation (TLO) tounds the crerating limit MCPR f 3

2. The GE8x8NB-3 fuel type MCPR values bound the GE8x8 :!' MCPR values for all events.

Page 8 i

K1 1

24A5155 Rev._(

\\

Pressurization events l

Exposure point: EOC10-2000 mwd /MT (18 i

d/ST) BRUNSWICK 1 CIOICF

)

Option A GE8x8NB Option B i

3 GE8x8NB t

~

3 Load Reject w/o Bypass 1brbine Trip w/o Bypass 1.31 g

1.24

)

1.31 I

1.24 Exposure point: EOC10 Exposure range: EOC10-2000 mwd to EOC10 BRUNSWICK 1 C10IC l

Option A l

)

GE8x8NB Option B 3

GE8x8NB-3 Load Reject w/o Bypass Turbine Trip w/o Bypass 1.32 1.28

~

1.32 I

1.28 Exposure point: EOCIOExposure range: BOC10 to EOC10 BRU

]

VOOS 7

Option A I

y GE8x8NB Option B

~

3 GE8x8NB 3

Load Reject w/o Bypass Turbine Trip w/o Bypass 1.32 1_

1.28

~

1.31 1

I 1.27 Exposure point:EOC10 Exposure range: BR UNSWICK 1 C10 Exte

]

with FFWTR Option A 1

GE8x8NB-I Option B

~

FW ControllerFaiiu'n.

3 GE8x8NB 1

3 1.30 1.27 L

i 9

l 1

Page 9

. = _. - _

BRUNSWICK 1 24A5159 Reload 9 Rev 0

12. Overpressurization Analysis Summary Psi Py Plant Event (psig)

(psig)

Response

l MSIV Closure (Flux Scram) 1243 1272 Figure 9 t

1 1

i

'l j'

l Page 10 i

BRUNSWICK 1 l

Reload 9 24A5159 s

Rev.0

.k.

t

13. Loading Error Results t

I s

Variable water gap m'soriented bundle analysis: Yes 3 Misoriented Fuel Bundle ACPR r

G E 10-P8 HXB 346-12GZ-100M-150-T (GE8 x8 NB-3) 0.18 3

14. Control Rod Drop Analysis Results This is a banked position withdrawal sequence plant, therefore, the control rod drop accident analysis required. NRC approval is documented in NEDE-24011-P-A-US.

)

15. Stability Analysis Results GE SIL-380 recommendations and GE interim corrective actions have been included Electric Plant Unit 1 operating procedures. Regions of restricted operation defined in Attachment l' to NR E

Bulletin No. 88-07, Supplement 1, Power Oscillations in Boiling Water Reactors (BWRs), are applic Brunswick 1.

16. Loss-of-Coolant Accident Results LOCA method used: SAFER /GESTR-LOCA The GE8x8EB LOCA analysis results presented in Sections 5 and 6 of BrunswickSteam ElectricPlant Units 1 and2 SAFERIGESTAR-LOCA Loss-of-coolant Accident Analysis, NEDC-31624P. Revision 2, July conservatively bound the LOCA analysis of the GE8x8NB-3 fuel types. This analysis yieid a licensi a

basis peak clad temperature of 1533 *F, a peak local oxidation fraction of <0.30%, and a cort-wide n.?tal-wa ter reaction of 0.046%. The most and the least limiting MAPLHGRs for the new GE8x8NB-3 fuel de are as follows:

1 P

3. Includes a 0.02 penalty due to vanable water gap R4anor uncertainty.

Page 11 f

..I -)

BRUNSWICK 1 Reload 9 24A5159 Rev.0

16. Loss-of-Coolant Accident Results (cont) is Bundle Type: GE10-P81!XB346-12GZ-100M-150-T Average Planar Exposure MAPLHGR(kW/ft)

(GWd/ST)

(GWd/MT)

Most Limiting Least Limiting 0.00 0.00 10.92 11.44 0.20 0.22 10.95 11.52 1.00 1.10 11.04 11.65 2.00 2.20 11.21 11.83 3

3.00 3.31 11.41 12.00 f

4.00 4.41 11.65 12.19 l

5.00 5.51 11.90 12.37 6.00 6.61 12.08 12.56

)

7.00 7.72 12.26 12.75 8.00 8.82 12.47 12.%

9.00 9.92 12.71 13.05 10.00 11.02 12.94 13.09 3

12.50 13.78 12.86 13.18 15.00 16.53 12.73 12.95 20.00 22.05 12.05 12.43 25.00 27.56 11.33 11.67 35.00 38.58 10.06 10.28 45.00 49.60 8.76 8.93 51.31 56.56 5.86 6.16 51.37 56.62 6.13 51.92 57.24 i

5.89 52.06 57.39 5.83 Page 12

BRUNSWICK 1 Reload 9 24A5159 Rev.0 J s M M 8+s M s

1 eMMMMMEs+sMBis Es+8M M M Ms+8M M BsE

- s M M M M M B i s+s s+s M

MMMMMMM8+ss+sMMBEM

MB8BsMBsMBsBsBsBis+sH ll:M Ms+sM M MBsBEM M MB ll:MBas+sMs+sB8BsMs+sBis+

';: M M B i s+s M 8+s s+s M M s+s s+8

MMMBEMMMMMMMS i

M M 8+s M s+sBs M M M Bs E

MMMMMMMMss*'

+

eBEMMMMm IIIIIIIIII, 1

i>>

1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 4143 45 47 49 51 1

Fuel Type A =G E88-P8 DQB 339-i l GZ-80M-4 WR-150-T B =G E8 B-P8 DQB 323-10GZ-80M-4 WR-150-T (Cycle 7)

E=G E 10-P81DCB32G-i l GZ-100M-150-T C=GE10-P8HXB322-IICZ-70M-150-T (Cycle 7)

F=GElo-P8HXB322-11GZ-70M-150-T (Cycle 9)

D=GE10-P8HXB324-12GZ-70M-150-T (Cycle 8)

G=G E 10-P81DCB346-120Z-100M-150-T (Cycle 9)

(Cycle B)

(Cycle 10)

Figure 1 Reference Core Loading Pattern Page 13

BRUNSWICK 1 24A5159 Reload 9 Rev.0 e

Neutron Flux Vessel Press Rise (psi)

Ave Surface Heat Flux

- - - - Safety Vafve Flow

- Core inlet Flow 300.0 - --- Relief Valve Flow 150.0 - ---

--- Bypass Valve Flow

)

-/\\,s*. [ N%s J

5 g200.0 g 100.0 s

W

s'%s E

v CC CC y

~

~'

60.0 100.0 r-----

I I

I 0.0 O.0 O.0 3.0 6.0 0.0 3.0 6.0 Eme (Sec)

Time, (sec)

I

^ # oid Reactivity V

Levol(inch-REF-SEP-SKRT) h Doppler Reactivity

- - - - - Vessel Steam How 200.0.- --- _ Turbine Steam Flow 1.0

-- Scram Reactivity

--- Feodwater Flow

-- Total Reacevity W

0,

...... ~..

ug o 0.0 p 100.0 R

s *.... -

a g.

g s

i.

ct 0.0 34 '-

'1-------------.

N -1.0 I'

k

, s,

\\ '\\

l

-100.0

- 2.0 O.0 3.0 6.0 0.0 3.0 6.0 Time (sec)

Time (sec)

Figure 2 Plant Response to Load Reject w/o Bypass (BOC10 to EOC10-2000 mwd /MT (1814 mwd /ST) BRUNSWICK 1 CIO ICF)

Page 14

BRUNSWICK.1 24A5159 Reload 9 Rev.0 I

Neutron Flux Vessel Press Rise (psi)

- - Ave Surface Heat Flux

- - - - Safety Valve Flow 150.0 - ----- Core inlet Flow 300.0 - --- Relief Valve Flow

--- Bypass Valve Flow

)

/M. /\\

S'., \\

)

200.0

./

100.0 s

te co N

C c

~

'.,~.'

50.0 100.0 p---.--._,_.

I I

0.0 O.0 O.0 3.0 6.0 0.0 3.0 6.0 Eme(seC)

Eme (SOC)

oid Reactmty

^

V Level (inch-REF-SEP-SKRT)

Doppler Reactivity

- - - - Vessel Steam Flow 200.0 - --- Turbine Steam Flow 1.0

-- Scram Reactivity

--- Feedwater Flow

-- Total Reactivity G

O u

,'. ~. -

e 0.0 g 100.0 7-

, ~ ~~

\\ '... -

y l,

e 3

\\

s x

g i %..,

f

\\\\

W L - 9'- - - - - - - - - - - - - -

E -1.0

\\.

0.0

'.I E

\\\\

tQ l

I

-100.0

-2.0 0.0 3.0 6.0 0.0 3.0 6.0 Eme (sec) mme (sec)

Figure 3 Plant Response to'Ibrbine 11 rip w/o Bypass (BOC10 to EOC10-2000 mwd /MT (1814 mwd /ST) BRUNSWICK 1 C10 ICF)

Page 15

BRUNSWICK 1 Reload 9 24A5159 Rev. 0_ _

l I

- Neutron Flux

\\

-h < - Ave Surface Heat Flux

- VesselPress Rese(psi) 150.0 - --- - Coro inlet Row

- - - Safety Valve Flow 3MO 1

--- Relief Valve Flow

- - Bypass Valve Flow

\\,' 'A

~)

s-g 100.0

'. %'S E

g200.0 C

% N W

)

g

%s*

C g

50.0 -

100.0 3

/--- -

I o,o i

/

O.0 3.0 O.0 ' --

6.0 0.0 Time (SeC) 3.0 6.0 Eme (SOC)

)

I

- Lewl(inch-REF-SEP-SKRT)

- - - - Vessel Steam Flow

~

id Reactvity v

200.0 - - - Turt>ine Steam Flow

- - - Doppler Reactivity Feedwater Flow 1.0 -

- Scram Reactivity TotalReac6vity ij) mE

}

p 100.0 g

g 0.0 is

).

E g',

E-s.),....',

\\

34:'-d- --- -~~~~---

)

Q 0.0 N -1.0 -

\\

a>x

\\

I 1

- 100.0 -

I i

0.0 J

3.0

-2.0 -

A 1

6.0 0.0 Time (sec) 3.0 6.0 Eme (sec)

Figure 4 Plant Response to Load Re,iect w/o Bypass (EOC10 mwd /ST) to EOCIO BRUNSWICK 1 C10ICF)

(1814 1

Page 16

BRUNSWICK 1 Reload 9 24A5159 i

Rev.0

)

Neutron Flux Wssel Press Rise (psi)

- Ave Surface Heat Flux

- - - - Safety Valve Flow 150.0 - -----

Core Inlet Flow 300.0

--- Relief Valve Flow

--- Bypass Valve Flow

)

\\(' /s g 100.0

, e' g200.0 e

~s e

C N~'

C I

60.0 100.0 w

I I

/

0.0 O.0 O.0 3.0 6.0 0.0 3.0 6.0 mme (sec)

Time (sec) 8 Level (inch-REF-.SEP-SKRT) id Reac[

- - - - Vessel Steam Flow -

- - - - Doppler Reactivity 200.0

--- Turbine Steam Paw 1.0

- Scram Reactivity

--- Feedwater Flow Total Reactivity b

e o

g

\\

y

\\

g 100.0 7-g 0.0 e

1 N-E g.1.....',

C i, ' '.. **..

j

\\1 i,

I k.

0.0 t_ _ M - - - - - - - - - - - - -.

-1.0 g

C

}

\\

i l

\\

t

'I I

- 100.0

-2.0 O.0 3.0 6.0 0.0 3.0 6.0 Time (sec)

.Eme (sec)

Figure 5 Plant Response to 'Ibrbine Trip w/o Bypass (EOC10-2000 mwd /MT (1814 mwd /ST) to EOC10 BRUNSWICK 1 C10ICF)

Page 17

~

BR' NSWICK 1 U

i Reload 9 24A5159 Rev. 0_

)

--- Neutron Flux Ave Surface Heat Flux Vessel Press Rise (psi) 150.0 - ---

Core inlet Flow

- - - - Safety Valve Flow 300.0 -

-- Retsf Vale Flow I

Bypass Valve Flow

.,Y,}%

~,

100.0

[,

', s 200.0 m

C N., %

a

-)

%~~-

C

~

60.0 -

100.0 -

f-~~

\\

I

')

I

\\

0.0 --

0.0 O.0 -

3.0 6.0 0.0 Eme (SOC) 3.0 6.0 Eme (SeC)

)

l

- Lowl(inch-REF-SEP-SKRT)

Vessel Steam Flow oid Rehvity i

200.0 - -- Turbine Steam Flow

- - - - Doppler -..,

Feedwater Flow 1.0 N

Scram Roactivity TotalReacevity

.)

w

\\

~

63 E

u 100.0

- --La-g s

l',

0.0 to

\\<

,., - ~..

I'.

... ~'

s'L,'\\,,..

s *

~

r3

%l*.

n

~

\\)

\\.,

M. f-- - - - - - - - - - - - -.

N. -1.0 -

\\

)

0.0 2

C

\\

t.

-100.0-1 0.0

-2.0

'I 3.0 I

6.0 0.0 Eme (sec) 3.0 6.0 Eme (sec)

Figure 6 Plant Response to Load Reject w/o Bypass (BOC10 C10 MSIVOOS) 1 Page 18

l 1

. BRUNSWICK 1 Reload 9 24A5159 Rev.0 r

i Neutron Flux

- - Ave Surface Heat Flux Wssel Press Rise (psi)

- - - - Safety Valve Flow 150.0

-- Core inlet Flow 300.0

--- Relief Valve Flow

--- Bypass Wlve Flow

,)

)

/.N 100.0 200.0 m

N C

ss C

e 3

50.0 100.0 -

(~~~~~~~~_

I I

')

I

\\

O.0 "f

0.0 O.0 3.0 6.0 0.0 3.0 6.0 Ime (sec)

Time (sec)

)

Levet(inch-REF-SEP-SKRT) id Rehvity

- - - - Vessel Steam Flow

- - - - Doppler -g tj 200.0 - --- Turbine Steam Flow 1.0 s

--- Feedwater Row Scram Reactivity TotalReactwity n

'\\

y u

\\

y

\\

g 100.0 m-

-p.

p 0.0 w

i.

..._: :. ~

F t.t....

C 8

\\\\

Ih,I' x

g.

c I

i \\,

)

M.'

N, -1.0 I

0.0 e

\\

C l

\\

I'

\\'

I

-100.0

-2.0

'I I

O.0 3.0 6.0 0.0 3.0 6.0 Time (sec)

Eme (sec)

Figure 7 Plant Response to 'Ibrbine Trip w/o Bypass (BOC10 to EOC10 BRUNSWICK 1 CIO MSIVOOS)

I Page 19 i

f Reload 9 i

24A515(8 Rev._

t i

Neutron Fl p ',

y 150.0 - - - - - Avepce Heat Flum

-- Vessel Press Rise (psi)

,-,,,, e6e inlet Flow

- Core Inlet Subcooling 125.0

- Safety Valve Flow

-- Relief Valwe Flow

- Bypass Val.s Flow

s-100.0 E 7 Y---

~

i g

75.0

\\

'. K E

)

50.0 -

Y l~ ~I 25.0 -

--4 l-I l

e 3

s I

I i

0.0 '-

O.0 7.5 25.0 '-

l 15.0 0.0 Eme (SOC) 7.5 15.t Eme (Sec)

)

Level (inch-REF-SEP-SKRT)

- Void Reactivity

- - - Vessel Steam Flow 150.0

..--, Turbine Steam Flow

--- FeedwaterFlow 1.0 Doppler Reactivity

--- Scram Reactivity

--- TotalReactivity e

\\

')

E

{

g 100.0

+

w 0.0

- +,.

.,, fg '

~

e

-I g

l,'.s '

t..

o

)

O b

i a

l.'e

s 50.0 -.

g.

i

-1.0 -

j ll l .*

I '.',.' (

E e

g l.

4 N

I' L

I 0.0 t I

0.0 J

7.5

-2.0 ' -

I II 15.0 0.0 Eme (sec) 7.5 15.0 Eme (sec)

Figjure 8 Plant Response to FW Controller Failure (BRUNSW EOC10 with FFWTR)

Extended i

)

Page 20

[

l

)

i

,0 Y mwot Y EEg RB 2

1 1

2 eR 0

0 0

5 0

5 1

1 l

0 O.

0 0

oU.

0 0

a O.

0 0

dN O.

0 0

9

~

S

.a W

I C.

Eh.,..s

'A,.

K g

/

1 F

s FTVL i

s euee

~

CAk g

oWh er s v o

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%,-l..

dibs e r

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n Sc E

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r n s

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eh ef teSt m

aF t

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BRUNSWICK 1 3

Reload 9 24A5159 Rev.0

+

f, Appendix A Analysis Conditions 1-To reflect actual plant parameters accurately, the values shown in Table A-1 were used this c 7

Table A-1 Analysis Value Parameter ICF MSIVOOS FFWTR Thermal power, MWt 2436.0 2436.0 2436.0 Core flow, Mlb/hr 80.8 80.8 80.8 Reactor pressure, psia

~

1036.0 1036.0 1034.5 Inlet enthalpy. BTU /lb 528.0 528.0 515.8 Non-fuel power fraction 0.038 0.038 0.038 Steam flow analysis, Mlb/hr 10.48 10.48 8.94 Dome pressure,psig 1005.7 1020.3 992.1 Turbine pressure, psig 950.0 919.9 951.3 No. of Safety / Relief Valves 9

11 9

Relief mode lowest setpoint, psig I116.0 1116.0 1116.0 Recirculation pump power source On-sited On-site 4 Orssite4 Turbine control valve mode of operation Partial arc Partial arc Partial arc 3

)

4. Dmnds operatim wnh off-site power smrce for reload beensing events for CWie 10.

Page 22

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1 HRUNSWICK 1

_R_eload 9 24A5159 Rev.0 Appendix B Main Steamline Isolation Valve Out of Service Reference B-1 pmvided a basis for operation of Brunswick Steam Electric Plant (BSEP) w 1

Steamline Isolation Valve Out of Service (MSIVOOS)(three steamline operation) and all S For this mode of operation in BSEP Unit I throughout Cycle 10, the EOC10-2000 mwd /M MCPR limits presented in Section 1 I of the body of this report are bounding and ating in the MSIVOOS mode at any time during the cycle. The peak steamline and pea the limiting overpressurization event (MSIV closure with flux scram) were not calculated fo mode of operation. In this mode of operation it is required that all S/RVs be operational ve y

j 2 S/RVs OOS for the events evaluated during normal plant operation. Previous cycles aj that the MSIV closure with 11ux scram, evaluated in the MSIVOOS mode, has res l

sure being reduced by more than 25 psi, when compared to the same case evaluated w j

operational.

t

)

i Reference B-1. Main Steamline isolation Valve Out ofServicefor the Brun.swick Steam Electric GE Nuclear Energy, April 1988.

)

I i

Page 23

_--y+ar3

3 BRUNSWICK 1 Reload 9 24A5159 Rev.0 Appendix C Reload Unique Anticipated Operational Occurrence (AOO) Analysis input t

I The data previously recorded in this Appendix (input to be used in the analys transient)(Inadvenent startup of the liPCI system), has been eliminated. The basis contained in Reference C-1.

' )

Reference C-1. Determination ofLimiting Cold Water Event, March 14,1994 letter from J.E Kla Energy) to USNRC.

)

'b Page 24

1_

BRUNSWICK 1 Reload 9 24A5159 Rev.0 t

Appendix D Final Feedwater Temperature Reduction (FFWTR)

I j

An NEDC report will be issued that justifies the use of Final Feedwater Temperature Reduction (FFWTR) l7 for Brunswick 1 Cycle 10. The MCPR limits presented in section 11 are bounding and should be applied when operating with FFWTR.

)

f 4

Page 25

fchaNSWICK 1 T

24A5159 Rev.0 i

Appendix E Maximum Extended Operating Domain l

Reference E-1 provided a basis for operation of the Brunswick Steam Electric P tended Operating Domain (MEOD). The reload licensing analysis performed for

)

herein is consistent with and provide the cycle-specific update to the referen ne the GEXIePLUS correlation to the reload fuel has been confinned as required in refe Reference

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l GE Nuclear Energy (Proprietary), February 1989.Fel. Maximum Exte I

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9

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Page 26

^

ENCLOSURE 3 -

BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO.1

.NRC DOCKET NO. 50-325 OPERATING LICENSE NO. DPR-71 TRANSMITTAL OF CORE OPERATING LIMITS REPORT, SUPPLEMENTAL RELOAD LICENSING REPORT, AND LOSS-OF-ACCIDENT ANALYSIS REPORT l

i I

' ~

LOSS-OF-COOLANT ACCIDENT ANALYSIS REPORT FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 1

)

RELOAD 9, CYCLE 10 NEDC-31624P, SUPPLEMENT 1, REVISION 2 j

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ML20083R808

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