ML20235Z118

Rev 2 to Supplemental Reload Licensing Rept for Brunswick Steam Electric Plant Unit 1,Reload 5 (Cycle 6)(W/o Recirculation Pump Trip)

ML20235Z118
Person / Time
Site:	Brunswick
Issue date:	10/31/1987
From:	Charnley J, Elliott P, Lambert P GENERAL ELECTRIC CO.
To:
Shared Package
ML20235Z115	List: ML20235Z114 ML20235Z118
References
23A5814, 23A5814-R02, 23A5814-R2, NUDOCS 8710200610
Download: ML20235Z118 (40)

Text

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_23A3814

,_ , Revision 2 Class I October 1987 (23A5814, Rev. 2 )

SUPPLEMENTAL RELOAD LICENSING REPORT IOR ,

PRUNSWICK STEAM ELECTRIC PLANT UNIT 1, RELOAD 5 (CYCLE 6)

(WIlh00T RECIRCULATION PUMP TRIP) l Prepared: .

P. A. Lambert Fuel Licensing Verified:

O P. E. Elliott Fuel Licensin

/

[ ,.

l Approved * -

. . harnTey, Manager

.F el Licensing NUCLEAR ENERGY BUSINESS OPERATIONS

• GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GEN ER AL (h ELECTRIC B710200610 871014 PDR ADOCK 05000324 1/2 i- P PDR i

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23A5814 Reva 2.

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Carolina Power and Lf3ht Company (CP&L) for CP&L's use with the United States Nuclear Regula- j tory Cormission (USNRC) for amending CP6L's operating license of the Brunswick Steam Electric Plant Unit 1. The information contained in this report is believed by General Electric to be an accurate 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 respecting informa-tion in this document are contaired in the Contract between Carolina Power and Light Company and General Electric Company for Reload Fuel Supply and Related Services for Brunswick Steam Electric Plant Unit 1, effective recember 31, 1982, as amended, and nothing contained in this document shall be construed as changing said contract. The use of thic information except as defined by caid contract, for any purpose other than that for which it is intended, is not authoriced; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this docurent 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.

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23A5814 Rev. 2 ACKh0WLEDOMENT The engineering and reload licensing analyses which form the technical basis of this Supplemental Reload Licensing Report were performed by S. C. hoen, R. E. Polomik, and T. P. Lung of the Nuclear Fuel and Engineering Services'Eepartment.  !

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23A5814 Rev. 2 1

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' 1. ' PLANT-UN10UE ITEMS (1.0)*

Limiting Conditions for Operation Appendix A bases for Limiting Conditions for Operation Appendix B Information in Section 4.and Appendix C provided by Appendix C Carolina Power and Light Company Transient Operating Parameters: Appendix D Use of GEMINI Methods for Cycle 6 Appendix E

2. . RELCAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)

Fuel Type Cycle Loaded Number Irradiated l P8tRB285 3 20 P8LRB265H 4 72 F8LRL284H 4 72 l P8LRB299 4 36 BP8DRB299 5 184 New BP8DRL299 6 176 Total 560

3. REFEREhCE CORE LCA0 LNG PATTERN (3.3.1)

Nominal previous cycle core average exposure at end of cycle: 18,616 mwd /ST Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 18,616 mwd /ST Assumed reload cycle core average exposure at end of cycle: 17,982 mwd /ST Core loading pattern: Figure 1 l

• ( ') Ref ers 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 ref ers to the U.S. Gupplement , NEDE-24011-F-A-8-US, May 1986.

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23A5814 Rev. 2

)

4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO V01DS, 20*C (3.3.2.1.1 and 3.3.2.1.2)*

Beginning of Cycle, Keff  ;

Uncontrolled 1.125 Fully Controlled 0.968 Strongest Control Rod Cut 0.991 R, haximum Incresce in Cold Core Reactivity with 0.000 Exposure into Cycle, AK  ;

5. STAhDbi LICb1D CCNTRCL SiSTEM SHUIEUWN CAPALILITY (3.3.2.1.3)

Shutdown Margin (aK) ppm (20"C. Xenon Free) .1:

600 0.036

6. RELOAL UNICUE TRAASIENT ANAliSIS INPUI (3.3.2.1.5 ANE S.2.2)

(Cold Water Injection Events Only)

Void Fraction (%) 41.7 ,~

Average Fuel Temperature (*F) 1097 ,

I Void Coefficient N/A** (d/% Rg) -6.991/-8.739 Loppler Coefficient N/A** (d/*F) -0.204/-0.194 ,

Scram Worth h/A** ($) ***

I  !

,5 i r

• See Appendix C.

• • 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-240Il-P-A-8, dated May 19E6.

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7. RELO iD-UNIOUE GETAB TRAFSIENT ANALYSIS. INITIAL CONDITION PARAMETERG (S.2.2), , p 1 __

Fuel Peaking Factors -

Bundle Power Bundle Flow Initial Design Local Radial Axial / R-Factor (MWt) (1000 lb/hr) MCPR Exposure: BOC6 to E'JC6-2000 mwd /ST BP/P8x8R 1.20 1 [56 [. 40 1.051 6.651 109.1 1.20 1  ?

l Exposure: EOC6-2000 mwd /ST to E0C6 l

l P.P/P8x8R 1.20' 1.47 1.40 1.051 6.245 112.2 1.29 l

8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)

Transient Recategorization: No Recirculation Pump Trip: No Rod Withdrawal Limiter: No Thermal Power Monitor: Yes -

Measured Scram Time: No (r.gooure Dependent Limits: Yes f Exposure Points Analyzed: 2

9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3)

Single-Loop.0peration: Yes Load Line Limit:

• Yes Extended Load Line Limit: No increased Core Flow: No Flow Point Analyzed: N/A Feedwater Temperature Reduction: No ANTSPrvaram: No Maxfmum Extended Operating Domain: No

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23A5814 Rev. 2

10. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1) 1 Methods Used: GEMINI j Exposure Range: BOC6 to EOC6 Flux Q/A ACPR l

Transient (% NBR) (% NBR) BP/P8x8R Figure Inadvertent HPCI 123 118 0.16 2 Exposure Range: BOC6 to EOC6-2000 mwd /ST Flux 0/A aCPR Transient (~ NBR) (% NBR) BP/P8x8R Figure Load Rejection Without Bypass 333 115 0.13 3 Feedwater Controller Failure 237 112 0.10 4 Exposure Range: E0C6-2000 mwd /ST to EOC6 Flux O/A ACPR Transient (* NER) (% NBR) BP/P8x8R Figure Lead Rejection Without Bypass 504 123 0.22 5 Feedwater Controller Failure 342 118 0.16 6

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

SUMMARY

(S.2.2.1)

(Generic Bounding Analysis Results)

ACPR Rod Block Reading (%) BP/P8x8R 104 0.13 105 0.16 106 0.19 107 0.22 108 0.28 109 0.32 110 0.36 Setpoint Selected: 107 10

23A5814 Rev. 2

12. CYCLE MCPR VALUES (S.2.2)

Non-Pressurization Events Exposure Range: BOC to EOC BP/P8x8R Inadvertent HPCI 1.23 Fuel Loading Error 1.20 Rod Withdrawal Error 1.29 Pressurization Events Option A Option B BP/P8x8R BP/P8x8R Exposure Range:

BOC6 to E0C6-2000 mwd /ST Load Rejection Without Bypass 1.30 1.23 Feedwater Controller Failure 1.23 1.21 Exposure Range:

E0C6-2000 mwd /ST to E006 Load kejection Without Bypass 1.34 1.30 Feedwater Controller Failure 1.28 1.25

13. OVERP_ PRESSURIZATION ANALYSIS

SUMMARY

(S.2.3) s1 y Trancient (psig) (psig) Plant Response MSIV Closure 1214 1248 Figure 7 (Flux Scram) 11

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.g. r t 23A5814- Rev. 2.

3,

14. STABILITY ANALYSIS'RESULTS (S.2.4)

. Rod-Line Analyzed: Extrapolated' Decay Ratio: . Figure 8 mt '

Reactor Core Stability. Decay Ratio, r2 /*0: 0.85 ChanAel- Hydrodynamic Performance Decay Rati), x2/ *0 3

. Channel Type-BP/P8x8R 0.65 f

15. LOADING ERROR RESULTS (S'.2.5.4) ,

Variable Water Gap Disoriented Bundle Analysis: Yes 4

Event Initial'MCPR Resulting MCPR*

Disoriented 1.18 1.07

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

Bounding Analysis Results:

Doppler Reactivity Coefficient: Figure 9 Accident Reactivity Shape Functions: Figures 10 and 11 Scram Reactivity Functions: Figures 12 and 13 Plant Specific Analysis Results:

Parameter (s) not Bounded, Cold: None Resultant Peak Enthalpy, Cold: N/A Parameter (s) not Bounded, HSB: None Resultant Peak Enthalpy, HSB: N/A i'

• The ACPR for this event was calculated using the GENESIS nuclear method, consistent with the last reload.

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23A5814 Rev.2

17. LOSS-OF-COOLANT ACCIDENT RESULTS (S.2.5.2) te, i

LOCA Method Used: SAFE /REFLOOD/ CHASTE ,,

" Loss-of-Coolant Accident Analysis Report for Brunswick Steam Electric Plant Unit 1," General Electric Company, November 1978, (NED0-24165, as amended).

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23A5814 Rev. 2 b l l

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l IIIIIIIII 1 357 9,11 13151719 21 23 25 27 29 31 33 35 37 39 4143454749 51 FUEL TYPE A = P8LRB285 D = P8DRB299

= P8DRB284 F = BP8DRB2 Cycle 6 Figure 1. Reference Core Loading Pattern 14

23A5814 Rev. 2 1 NEUF RON FLUX 1VESEELPRESSRISE(PSI) 2 AVEI SURF ACE *# a? f t ov 2 RELBEF WALVE FLOW A- ' . **** WOW ' 3 BvPPS$ VALVE FLOW

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9. 0 50 0 100.0 0. 0 50.0 1 C 0. 0 fl*C ($t:cNCs) flat (SELON05)

Figure 2. Plant Response to Inadvertent Activation of HPCI 15

23A5814 Rev. 2 1 NEUTRON FLug l VESSEL PRESS RISE (PSI) 2 AVE $UM ACE HE AT F(Ux 2 SAFETY VALVE FLOW LOW ,,,,, {y(((( [g 15 0. 0 l

les.O k 200,0 l 7 '

N w

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\T ; 3 100,0 .g,3 6.0 0. 0 2. 0 4.0 8.0

0. 0 2.0 4.0 IIME (SECONOS) f!ME (SECONDS)

Figure 3. Plant Response to Generator Load Rejection k'ithout Bypass (EOC6-2000 mwd /ST) 16

23A5814 Rev. 2 4

9 150.0 i NEUTRON Fil/7 h I VIS EL PRESS #15ECPSI) s 2 AVE SURF 4 HE A FLVA 2 SAF[TY v4LVE FLCe Flow 3 RELg EF W A L v < . LC W 3 CDR: IPL 4 BrP'SS va;v LC=

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Figure 4. Plant Respor.ue to Feedwater Controller Failure (EOC6-2000 mwd /ST) 17 .-

23A5814 Rev. 2 1 NEUTRON PLUk I VEISEL PRISE RISIlP$l) l 2 Avf $Unr Act d at Flux 2 5AFETY VALVE FLou l 3C0Rt lHLE1 rL0w 3 REgl(( VALy [ Low 15 0. g 300. 0 g i00.0 ', -

s 200.0 l i

i

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100.0 N

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.s00,0 2.0 2.0 00 00 6.0 0. g 0.0 2.0 4. 0 Tiet ISECONOM flet ($ttCND51 Figure 5. Plant Response to Generator Load Rejection Without Bypass (EOC6) 18

23A5814 Rev. 2 r

156.9 jNEU'RONIINX 1 Vf$hEL PEES RISE (P$l) 2 AVE SL41F 't ME F(UX 2 SAFETY WALv \F OW L 3 COGi INLJ TLOW 3 REL)EF VALV LOW

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Figure 6. Plant Response to Feedwater Controller Failure (EOC6) 19

23A5814 Rev. 2 1

'lNEufRONFlur i VESSEL PRESS Pl$EtPSI) 2 $AFEf? VA.VE FLov 22coacAvf SVRFfsE rutt rtevHE AT FLt)X ,,,,, ggigyA.yjgray

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Figure 7. Plant Response to MSIV Closure (Flux Scram) 20

23A5814 Rev. 2 1.00 C

/A

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0.50 0.25 A = NATURAL CIRCULATION B = 105 PERCENT h0D LINE C = ULT PERFORMANCE LIMIT

-B l 06 0 20 40 60 80 100 120 ,

PERCENT POWER Figure 8. Reactor Core Decay Ratio 21

23A5814 Rev.2 l

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/  ! A = CALCULATED VALUE - COLD B = CALCULATED VALUE - HSB

-30 C = BOUND VAL 280 cal /g COLD D = BOUND VAL 280 cal /g HSB

- 35 500 1000 1500 2000 2500 3000 0

FUEL TEMPERATURE (*C)

Figure 9. Fuel Doppler Coefficient in 1/a*C 22

23A5814 Rey, 2 20.0 17.5 l

I l

15.0 g 12.5

+

w G

g 10.0 '

C E

O E 7.5 -'

5.0 A = ACCIDENT FUNCTION B = BOUNDING VALUE 280 cal /g 2.5 0

0 5 10 15 20 ROD POSITION (feet out)

Figure 10. Accident Reactivity Shape Function, Cold Startup 23

23A5814 Rev. 2 I

I 20.0 1 17.5 15.0

$ f h..

, 16.u A

G g 10.0

/

W 5

7.5 A = ACCIDENT FUNCTION B = BOUNDING VALUE 28C cal /g 5.0 2.5 .

- ~ - - -

0 v 5 10 15 20 ROD POSITION (feet ocd Figure 11. Accident Reactivity Shape Function, Hot Standby 24

23A5814 Rev. 2 l -.

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w A = SCRAM FUNCTION

<t B = BOUNDING VALUE 280 caug C ' ' >

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. 3 3 4 5 6 0 1 2 .,-_

ELAPSED TIME (sec)

Figure 12. Scram Activity Function, Cold Startup 25

s 23A5814 Rev. 2 50 A

/

40 R

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y 30

. w s A a SCRAM FUNCTION 0

o f B = BOUNDING VALUE 280 ca /g D

E -- B y 20 5

a 10 f s

0 4 5 6 0 1 2 3 ELAPSED TIME (sec)

Figure 13. Scram Reactivity Function, Hot Standby 26

23A5814 Rev. 2 APPENDIX A LIMITING CONEITIONS FOR OPERATION This appendix provides the limiting condition for operation (LCO) for each of the power distribution limits identified belew:

(1) Average Planar Linear Heat Generation Rate (APLHGR)

(2) Operuting Limit MCPR (3) APRM Setpoints 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 APLHGR During steady-state power operation, the APLHCR for each type of fuel as a function of axial location and average planar exposure shall not exceed limits based on applicable APLEGR limit values which have been approved for the respective 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 a function of average planar exposure shall not exceed the limiting value for the most limiting lattice (excluding natural uranium) shown in Figures A-1 through A-5, during two recirculation loop operation.

A.2 OPERATING LIMIT MCPR The fuel cladding integrity safety limit MCPR is 1.07. During steady-state power operation, the MCFR 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-6),

for two recirculation loop operation.

27

23A5814 Rev. 2 In reference to Technical Specification 3.2.3.2, the OLMCPR for T ave less than or equal to T is the greater of the non pressurization transient or B

the Option B OLMCPR (Table A-1), where T ave and TB are given by:

n

[N g i=1 T

g T ~ '

ave n

[N i=1 g

where:

i = Surveillance test number.

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

th Ng = Number of rods tested in the i surveillance test. ,

T g = Average scram time to notch 36 for surveillance test i.

and T

B

=p + 1.65 fN n 2

I')

. [N i=1 g

where:

Ny = Number of rods tested at BCC.

p = 0.813 seconds (mean value for statistical scram time distribution from de-energization of scram pilot valve solenoid to pickup on notch 36).

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

28

t t

23A5814 Rev. 2 In reference to Technical Specification 3.2.3.2, the OLMCPR for T ne " ' '

greater than T shall be either:

B l

a. The greater of the non pressurization transient (Table A-1) or the -

adjusted pressurization transient MCPR (MCPRadj) "h*#*

\

T T / )

~

M PR ~

' Option B adj Option B T -T Option A B \

T = 1.05 seconds (control rod average scram insertion A

time limit to notch 36),

and MCPR as given in Table M 0ption A MCPR as ghen in Table A-1 0ption B or,

b. MCPR as given in Table A-1.

0ption A A.3 APRM SETPOINTS The flow-blased APRM scram trip setpoint (S) and rod block trip setpoint (S shall be:

RB _

S 1 (0.66W + 54*.)T, and S

RB 5( .66W + 420 h where S and S RB are in percent of rated thermal powen W = loop recirculation flow in percent of rated flow.

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

• I

• T = CMAP T -

29 s

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23A5814 Rev. 2 .

P

l. Table A-1 l

l MCPRs Fuel Type: P8x8R and BP8x8R Non-Pressurized Transient MCPR = 1.29 Pressurization Transients Exposure Range MCPR 0ption A M PR 0ption B BOC6 to EOC6-2000 mwd /ST 1.30 1.23 i

E0C6-2000 mwd /ST to E0C6 1.34 1.30 .

4 e+

h 37/38

l 23A5814 Rev. 2 L

l APPENDIX B L

L BASES FOR LIMITING COND1TIONS FOR OPERATION This appendix provides the bases for each of the power distribution limits identified in Appendix A.

B.1 APLHCR This specification assures that the peak cladding temperature (PCI) fol-lowing the postulated design basis loss-of-coolant accident (LOCA) will not exceed the limits specified in 10CFR50.46 and that the fuel mechanical design analysis limits specified in Reference B-1 will not be exceeded.

Thermal hechanical Design Analysis: NRC approved methods (specified in Reference B-1) are used to demonstrate that all fuel rods in a lattice oper-ating at the bounding 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 AFLHGR limit.

LCCA Analysis: A LCCA analysis is performed in accordance .th 10CFR50, Appendix K to demonstrate that the permissible planar power (maximum AFLHGR) 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 APLHCR li ,it is the most limiting composite of the fuel design analy-sis APLHGR limit and the ECCS APLHCR limit.

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 39

23A5814 Rev. 2 abnormal operational transients. For any abnormal operating transient analy-sis evaluation with the 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 transient, assum-ing instrument trip setting as given in Specification 2.2.1 of 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 insertion, and coolant temperature decrease.

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 limiting exposures are used for nuclear data to provide conservatism relative to core exposure aspects. Plant-unique initial conditions and system param-eters are used as inputs to the transient codes. The 6CPR 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.

The purpose of the fK factc; is to define operating limits at other than rated flow conditions. At less than 100% flow, the recuired MCPR is the product of the operating limit MCPR and the fK factor. Spe_Afically, the Kg factor provides the required thermal margin to protect against a flow increase transient. The most limiting transient initiated from less than rated flow conditions is the recirculation pump speedup caused by a motor generator speed control failure.

40

23A5814 Rev. 2 For operation in the automatic flow control mode, the Kf factors assure that the operating limit MCPR in Appendix A will not be violated should the most limiting transient occur at less than rated flow. In the manual flow l

control mode, the Kf factors assure that the safety limit MCPR will not be violated should the most limiting transient occur at le,s than rated flow.

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

Reference B-1.

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 Kf.

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-tent will be very small. Fcr 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 margin will thus be demonstrated such that future MCPR evaluation below this power level vill be shown to be unnecessary. The daily requirement for calculating MCPR above 25% rated thermal power is sufficient, since power distribution shifts are 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.

41

23A5814 Rev. 2 B.3 APRM SETPOINTS The flow-biased thermal power upscale scram setting and 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 adjustment 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", NEDL 24011-P-A (latest approved revision).

2. " General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix K", NED0-20566, January 1970.

42

23A5814 Rev. 2 APPENDIX C This cold shutdown margin design evaluation was performed by Carolina Power and Light Company personnel using NRC approved methodology

• applied in conformance with Carolina Power and Light Company's Quality Assurance Program.

This evaluation provides a high degree of confidence that greater than the Technical Specification requirement of 0.38% AK/K cold shutdown margin will be maintained throughout the cycle, and that 0.38% aK/K plus "R" will be measured during the beginning of cycle shutdown margin demonstration.

• Letter f rom Domenic B. Vassallo to E. E. Utley, " Brunswick Reload Licensing Methodologies", Docket Nos. 50-325/324, May 18, 1984.

43/44

23A5814 Rev. 2 APPENDIX D TRANSIENT OPE!.ATING PARAMETERS All of the transients and overpressure protection analyses were run considering a power uncertainty of 2%. The uncertainty in the ODYN transient analysis is included in the statistical adjustment factors and the transient is initiated at 100% rated power, as approved in Reference D-1. The other initial conditions given in Table S.2-6 of NEDE-24011-P-A-8-US reflect this initial power level.

REFERENCE:

D-1. Letter, G. C. Lainas (NRC) to J. S. Charnley (GE), " Acceptance for Referencing of Licensing Topical Report NEDE-24011-P-A, ' General Electric Generic Licensing Reload Report,' Supplement to Amendment 11,"

Harch 22, 1986.

b 45/46

23A5814 Rev. 2 8

APPENDIX E USE OF_ GEMINI METHODS FOR CYCLE 6 .

The analyses required for this cycle were performed with GE's advanced s reload licensing methods, known as GEMINI. Any differences between this 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.

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(FINAL)

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