ML20207B908

Rev 0 to Brunswick Steam Electric Plant,Unit 1,Reload 6 (Cycle 7) (W/O Recirculation Pump Trip)

ML20207B908
Person / Time
Site:	Brunswick
Issue date:	06/30/1988
From:	Charnley J, Elliott P, Pentzien D GENERAL ELECTRIC CO.
To:
Shared Package
ML19292J182	List: ML20207B892 ML20207B899 ML20207B908
References
23A5896, 23A5896-R, 23A5896-R00, NUDOCS 8808040333
Download: ML20207B908 (39)

Text

___ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ . . _ _ _ _ . . _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

23A5896 REVISION O CLASS I  :

JUNE 1988 l l

I (23A5896, REY. 0)

SUPPLEMENTAL RELOAD LICENSING REPORT FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 1, RELOAD 6 (CYCLE 7)

(WITHOUT RECIRCULATION PUMP TRIP)

Prepared D. C. Pent c' l

~

Fuel Licens Verified: -

O P. E. Elliott Fuel Lice n n

Ap ro .

. Ctiarnley, Han Fuel Licensing 9

$$geo$$$$$$$o! 5 #ENwAearEnetyr P

175 Curw kenue SmJose. [A 95125

23A5896 REV. O IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by the General Electric Company (GE) solely for Cnrolina Power and Light Company (CP&L) for CP&L's use with the United States Nuclear Regulatory Commission (USNRC) for amending CP&L's operating license for the Brunswick Steam Electric Plant Unit 1.

The information contained in this report is believed by GE to be an accurate and true representation of the facts known, obtained or provided to GE at the time this report was prepared.

The only undertakings of GE respecting information in this document are contained 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 December 31, 1982, as amended, and nothing contained in this document shall be construed as changing said contract. The use of this information except as defined i

by said Contract, for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither GE 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 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 for such use of such information.

3/4

I l

23A5896 REV. O ACKNOWLEDGEMENT The engineering and reload licensing analyses which form the l technical basis of this Supplemental Reload Licensing Report were performed by R. E. Polomik and T. P. Lung of the Fuel Engineering Section.

5/6

I i

l 23A5896 REY. 0 l

l

1. PLANT UNIOUE ITEMS fl.0)* .

Limiting conditions for Operation Appendix A Bases for Limiting Conditions for Operation Appendix B Plant Parameter Differences Appendix C Use of GEXL-PLUS Methods f.or Cycle 7 Appendix D Use of New Safety Limit MCPR for Cycle 7 Appendix E

2. RELOAD FUEL BUNDLES (1.0. 2.0. 3.3.1 AND 4.0)

Fuel Tyne Cvele Loaded Number Irradiated P8DRB284H 4 8 P8DRB299 4 8 BP8DRB299 5 184 BP80RB299 6 176 Hilf BD339A 7 60 BD323B 7 111 Total 560

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at end of cycle: 21,072 mwd /MT Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 20,481 mwd /MT Assumed reload cycle core average exposure at end of cycle: 21,230 mwd /MT Core loading pattern: Figure 1

• ( ) Refers to area of discussion in "General Electric Standard Application for Reactor Fuel," NEDE 240ll-P A 8, dated May 1986. A letter "S" preceding the number refers to the U.S. Supplement, NEDE-240ll-P-A 8 US, May 1986.

7

sm a 23A5896 REV. 0

4. (AlfULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS. 20*C (3.3.2.1.1 and 3.3.2.1.21 Beginning of Cycle, K,ff Uncontrolled 1.111 Fully Controlled 0.966 Strongest Control Rod Out 0.988 R, Maximum Increase in Cold Core Reactivity 0.002 with Exposure into Cycle, AK

5. STANDBY LIOVID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

ShutdownMargin(AK) 225 (20*C. Xenon Free) 600 0.036

6. RELOAD-UNIOUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2)

(Cold Water Injection Events Only)

Void Fraction (%) 41.7 Average Fuel Temperature (*F) 1096 Void Coefficient N/A* (t/% Rg) -6.710/-8.387 Doppler Coefficient N/A* (t/*F) -0.201/ 0.191 Scram Worth L'/A* ($)

• 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-240ll-P-A 8, dated May 1986.

8

23A5896 REY. 0

7. RELOAD-UNIOUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION '

PARAMETERS (S.2.2)

Bundle Fuel 4eakina Factors Power Bundle Flow Initial Desian inn), itadial M R-Factor fMWt) (1000 lb/hr) ,(Epp Exposure: SOC 7 to EOC7-2000 mwd /ST SP/P8v8R 1.20 1.51 ' l.40 1.051 6.409 109.5 1.25 GE8X8EG 1.20 1.01 1.40 1.051 6.418 112.2 1.26 Exposure: E0C7-2000 mwd /ST to EOC7 8P/P8x8R 1.20 1.44 1.40 1.051 6.127 111.7 1.31 GE8x8E8 1.20 1.45 1.40 1.051 6.139 114.3 1.32  ;

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 Exposure Dependent Limits: Yes Exposure Points Analyzed: EOC7-2000 mwd /ST and EOC7

9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3)

Single-toop Operation: Yes Load Line Limit: Yes Extended Load Line Limit: No increased Core Flow: No Flow Point Analyzed: N/A Feedwater Temperature Reduction: No ARTS Program: No Maximum Extended Operation Domain: No 9

23A5896 REV. 0

10. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.11 Methods Used: GEMINI Exposure Range: BOC7 to EOC7 Flux Q/A ALEB Transient f% NBR) (% NBR) BP/P8x8R GE8x8EB Fiaure Inadvertent HPCI 122 119 0.15 0.15 2 Exposure Range: BOC7 to EOC7-2000 mwd /ST

~

Flux Q/A ALEB Transient (% NBR) (% NBR) BP/P8x8R GE8x8EB Fiaure Load Rejection Without Bypass 482 120 0.18 0.18 3 Feedwater Controller Failure 285 116 0.12 0.13 4 Exposure Range: EOC7-2000 mwd /ST to EOC7 Flux Q/A ALEB Transient (% NBR) (% NBR) BP/P8x8R GE8x8EB Fiaure 0 +

Load Rejection Without Bypass 592 125 0.24 0.25 5' Feedwater Controller Failure 418 121 0.18 0.19 6

11. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILUREi 1RANSIENT

SUMMARY

(S.2.2.1)

Limiting Rod Pattern: Figure 7 Rod Block Rod Position ALEB Readino f%) (Feet Withdrawn) BP/P8x8R GE8x8E8 104 3.5 0.11 0.11 105 3.5 0.11 0.11 106 4.0 0.13 0.13 107 4.5 0.14 0.14 108 5.5 0.18 0.18 109 12.0 0.21 0.21 110 12.0 0.21 0.21 l

l Setpoint Selected: 107 l 10 l

l l 23A5896 REY. 0 l j

12. CYCLE MCPR VALUES f S.2.2) ,

Non-Pressurization Events Exposure Range: BOC to EOC BP/P8x8R GE8x8EB Inadvertent HPCI .

1.19 1.19 Fuel Loading Error --

1.25 Rod Withdrawal Error 1.18 1.18

~

Pressurization Events Ootion A Dotion B BP/P8x8R GE8x8EB BP/P8x8R EBxB.EB Exposure Range:

SOC 7 to EOC7-2000 mwd /ST Load Rejection Without Bypass 1.32 1.32 1.25 1.25

, Fecdwater Controller Failure 1.22 1.23 1.20 1.21 Exposure Range:

EOC7-2000 mwd /ST to EOC7 Load Rejection Without Bypass 1.34 1.34 1.30 1.30 Feedwater Controller Failure 1.27 1.27 1.24 1.24 l 13. OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.3J P- P i

s1 y Transient 1211ql fosia) Plant Resconse MSIV Closure 1235 1266 Figure 8 (Flux Scram) 11

23A5896 REY. 0

14. LOADING ERROR RESULTS fS.2.5.4)

Variable Water Gap Misoriented Bundle Analysis: Yes*

LY.tn.t A2B Misoriented , 0.19

15. CONTROL ROD DROP ANALYSIS"RESULTS fS.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:

Resultant Peak Enthalpy, Cold: 149.6 Resultant Peak Enthalpy, HSB: 215.6

16. STABILITY ANALYSIS RESULIS fildl

~

GE SIL380 recomendations have been included in the Brunswick Steam l Electric Plant Unit 1 operating procedures and/or Technical

, Specifications and, therefore, the stability analysis is not required.

NRC approval for deletion of a cycle-specific stability analysis is documented in Amendment 8 to NEDE-240ll-P-A 8 US.

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

LOCA Method Used: SAFE /REFLOOD/ CHASTE "Loss-of-Coolant Accident Analysis Report for Brunswick Steam Electric Plant Unit No.1," General Electric Company (NEDO 24165, December 1978, as amended and NEDE 24165 P, April 1988.) i l

• ACPR penalty of 0.02 for the tilted misoriented bundle is applied to the cycle MCPR value reported in Section 12.

12

23A5896 RV. 0

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"MMMMMMMMM" "MMMMM" lIIIIIIIIi 1 3 5 7 9111315171921232527253133353739414345474951 YUEL TYPE A = P8DRB284H D = BP8DRB299 (Cycle 6)

B = P8DRB299 E = BD339A C = BP8DRB299 (Cycle 5) F = BD323B Figure 3. Reference Core Loading Pattern 13

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Figure 4. Plant Response'to Feedwater Controller Failure (EOC7-2000 mwd /ST) 16

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23A5896 REV. 0 2 6 10 14 18 22 26 30 34 38 42 46 50 51 36 36 47 5 6 6 6 6 43 36 36 35 36 36 36 39 6 6 14 6 6 35 36 36 36 36 36 36 31 6 6 14 0 14 6 6 27 36 36 44 44 36 36 23 6 6 14 0 14 6 6 19 36' 36 36 36 36 36 15 6 6 14 6 6 11 36 36 36 36 36 36 7 6 6 6 6 6 3 36 36 NOTES:

1. No. indicates nu-ber of notches withdrawn out of 48. Blank is a withdrawn rod.

2. Error rod is (26,31).

Figure 7. Limiting Rod Pattern 19 I

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1 A ACC*0ENT l' UNC T l '. H B 63UNDING ' ALUE 260 CAL /G

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25/26

23A5896 REY. 0 APPENDIX A LIMITING CONDITIONS FOR OPERATION This appendix provides the limiting condition for operation (LCO) for each of the power distribution limits identified below:

(1) Average Planar Linear Heat Generation Rate (APLHGR) l (2) Operating Limit MCPR (3) APRH Setpoints Surveillance requirements &nd required actions are spe:ified in the Technical Specifications. The power distribution limit bases are given in Appendix B.

A.1 APLHGR During steady state power operation, the APLHGR for each type of fuel as a function of axial location and average planar exposure shall not exceed

) limits based on applicable APLHGR limit values which have been approved for

! the respective fuel and lattice types determined by the approved methodology l ,, described in GESTAR II (NEDE-240ll-P A). When hand calculations are l 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 applicable fuel cladding integrity safety limit MCPR for this cycle is 1.04. This safety limit MCPR applies to Unit I during this cycle because it is a D-Lattice BWR with at least two successive reloads of P8X8F., BP8X8R, GE8X8E or GE8X8EB fwl types having nigh bundle R-Factors (>1.04), one of which is the fuel in its first cycle of operation. The use of this value has been approved in Amendment 14 of NEDE 240ll-P A 8. During steady state power operation, the MCPR for each type of fuel shall not be less than the limiting value (shown in Table A 1) times the K7 (shown in Figure A 6), for two recirculation loop operation.

27

23A589* REY. O In reference to Technical Specification 3.2.3.2, the OLHCPR for r ay, less than or equal to 78, is the greater of the non-pressurization transient or the Option B OLMCPR (Table A-1), where 7,y, and r8 are given by:

a l

h,'t't

' ave

• a l

[N g i=1 where:

i = "arveillance test number.

n = Numbar of surveillance tests performed to date in the cycle (including BOC).

th surveillance test, Ng - Number of rods tested in the i r g = Average scram time to notch 36 for curveillance test 1.

and t3 = u + 1.65 [N y

2 (c)

Ng y

where:

N3 - Number of rods tested at BOC.

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

23A5896 REV. O In reference to Technical Specification 3.2.3.2, the OLMCPR for 7,y, j greater that 7 shall be either:

8 1

a. The greater of the non-pressurization transient (Table A-1) or the adjusted pressurization transient MCPR (MCPR adj) d ere*

MCPRadj - MCPR0ption B + ve - B (MCPR0ptionA-MCPbptionB)

A7B 7A - 1.05 seconds (control rod average scram insertion time limit to notch 36).

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

b. MCP ption A as given in Table A-1.

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

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

RB 5 (0.66W + 42%) T; where S and S are in percent of rated thermal power; RB W = loop recirculation flow in percent of rated flow.

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

T= FRTP where T 5 1.

CMAPRAT 29

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23A5896 REV. 0 l

1.4 ,

1.3 -

1.2 -

AUTOMATIC FLOW CONTROL N

~

1.1 --

2 MANUAL ROW CONTROL SCOOP TUBE SETPOINT CAUBRATION POSITIONED SUCH THAT FLOWM Ax = 102.5%

107.0 %

i,o 112.0 %

117.0 %

0.9 I I I I I I I 30 40 50 60 70 80 00 100 CORE FLOW (%)

Figure A-6. Kg Factor for GEXL-PLUS 35/36 l

23A5896 REV. 0 Table A-1 MCPRs Fuel Type: P8X8R, BP8X8R, and GE8X8EB Non-Pressurized Transient MCPR = 1.25 a

Pressurization Transients MCPR HCPR Exposure Range 0ption A 0ption B BOC7 to EOC7-2000 mwd /ST 1.32 1.25 1

EOC7-2000 mwd /ST to EOC7 1.34 1.30 l

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23A5896 REY. O APPENDIX B BASES FOR LIMITING CONDITIONS FOR OPERATION This appendix provides the bases for each of the power distribution limits identified in Appendix A.

B.I APLHGR This specification assu es that the peak cladding temperature (PCT) following the postulated derign 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 Mechanical Desian 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 APLHGR limit.

LOCA Analysis: A LOCA analysis is performed in accordance with 10CFR50, Appendix X to demonstrate that the permissible planar power (maximum APLHGR) limits comply with the ECCS limits specified in 10CFR50.46. The analysis is perfomed 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 analysis APLHGR limit and the ECCS APLHGR 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

23A5896 REV. O abnormal operational transients. In the analysis of these abnormal opera-tional transients, the GEXL-PLUS thermal correlation has been used, wMrc applicable, to determine the appropriate initial conditions. For any.

abnormal operating transient analysis 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, assuming instrument trip setting as given in Specifica-l tion 2.2.1 of the Technical Specifications.

To assure that the fuel cladd'ag integrity safety limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which ones result in the largest reduction 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 para-meters are used as inputs to the transient 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 establish the minimum operating limit MCPR.

The purpose of the fK factor is to define operating limits at other than rated flow conditions. At less than 100% flow, the required MCPR is the product of the operating limit MCPR and the K f factor. Specifically, the K g 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 controller failure.

40 f

23A5896 REV. O 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 control mode, the Kf factors assure that the safety limit MCPR will not be violated should the most limiting transient occur at less than rated flow.

The K factor values are generically developed as described in f

References B-3 and B-4.

The K factors are conservative for the General Electric plant operation f

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 content will be very small. For all designated control rod patterns which may be 4,mployed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable marc S.

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 was made at 25% initial power level with minimum recirculation pump speed. The demonstrated MCPR margin was such, that future MCPR evaluations below this power level are unnecessary. The daily requirement for calculating MCPR above 25% rated thermal pcwer 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 the MCPR will be known following a change in power or power shape, regardless of magnitude that could place operation at a thermal limit.

41 l

1

_j

23A5896 REV. O B.3 APRM SETPOINTS The flow-biased thermal power upscale scram setting and flow-biased neutron 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 formula in Appendix A when the combination of Thermal Power and CHAPRAT in-dicate a highly peaked power distribution. This adjustment may be accom-plished 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-8.

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

3. Letter, J. S. Charnley (GE) to M. W. Hodges (NRC), "Application of GESTAR-II Amendment 15," March 22, 1988, MFN-027-88.

4. Letter, A. C. Thadani (NRC) to J. S. Charnley (GE), "Acceptance for Referencing of Application of Amendment 15 to General Electric Licensing Topical Report NEDE-24011-P-A, ' General Electric Standard Application for Reactor Fuel' (TAC No. 60903)," Hay 5, 1988.

42

~

l 23A5896 REY. 0 l

l APPENDIX C PLANT PARAMETER DIFFERENCES GETAB and Transient Analysis Initial Conditions The values used in the GETAB and Transient Analysis which differ from the values reported in Tables S.2 4.1 and S.2-6 in NEDE-240ll-P-A-8-US are  !

given in Table C-1.

. Table C-1 ,

PLANT PARAMETER DIFFERNCES Parameter Analysis Value NEDE-24011-P-A 8-US Value Rated Steam Flow l 10.47E+06 10.96E+06 0.2%

Dome Pressure l 1005 1020 i 2 psi Turbine Pressure l 950 960 2 psi Non-Fuel Power Fraction I 0.039 0.040 2

Number of S/RVs 10 11 1

The indicated changes are a result of the application of the pre-approved methods outlined in Amendment 11 to NEDE-24011-P-A 8.

2 The indicated change is a result of the simulation of a valve-out-of-service condition.

43/44

23A5896 REV. O APPENDIX D USE OF GEXL-PLUS METHODS FOR CVCLE 7 The analyses required for this cycle were performed with the GEXL-PLUS thermal correlation. In analyses prior to Cycle 7 (Reload 6), the GEXL thermal correlation was used. The incorporation of GEXL-PLUS into the fuel cycle analysis process is provided for in Arrendment 15 to GESTAR-II (NEDE-240ll-P-A-8). Any difference between this reload and the previous one are due not only to cycle differences, but also to the difference in the B methods. Therefore, making direct comparisons between the two cycles may be inconclusive.

I I

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23A5896 REV. O APPENDIX E USE OF NEW SAFETY LIMIT MCPR FOR CYCLE 7 The analyses required for this cycle were performet' with the upgraded safety limit MCPR of 1.04, instead of the previous safety limit MCPR of 1.07.

The implementation of this safety limit is a result of the utilization of fuel types with high bundle R-factors, as stipulated in Amendment 14 to GESTAR-II (NEDE-24011-P-A-8). Any difference 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 may be inconclusive.

Y l

47/48

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