ML20212M732

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Rev 3 to Grand Gulf Unit 1 Cycle 2 Reload Analysis
ML20212M732
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 08/09/1986
From: Chandler J, Morgan J, Ward G
SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML20212M724 List:
References
TAC-61930, XN-NF-86-35, XN-NF-86-35-R03, XN-NF-86-35-R3, NUDOCS 8608270103
Download: ML20212M732 (53)
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Text

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X \\-\\ :-86-35 REVIS O\\

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GRA\\ J GU_: U\\ ~~ 1 CYC_E 2 l

RE_OAJ A\\ A_YS S I

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AUGUS-~ 1986 I

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RIC

_A\\ J, WA 99352 I

EXXON NUCLEAR COM PANY, INC.

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Xh-NF-86-35 e

Revision 3 Issue Date: 8/9/86 I

GRAND GULF UNIT 1 CYCLE 2 RELOAD ANAL.YSIS

.r74 M f/

d Prepare:

N J. C. Chandler, Senior Engineer BWR Safety Analysis Concur:

,#MF

//t/c G. N. Ward, Manager Reload Licensing Concur:

di.92st by hhtff fffbg-v J. N. Morg Manag

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Customer S vices E 1 ering Approve:

hk L sirs) r)shc H.f.'Williamson, Manager I

Licensing and Safety Engineering Approve:

O M/2 W

</////4 T. W. Patten, Manager '

Neutronics and Fuel Management i

Approve:

N-II/7/fl.

I G. L. llitter, Manager Fuel Engineering and Technical Services I

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,g ERON h UC _ EAR CO V 3A \\ Y,

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I NUCLEAR REGULATORY CO415tISSION REPORT DISCLAlhtER I

1%1PORTANT NOTICE REG ARDING CONTENTS AND USE OF Tills DOCU5 TENT PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Exxon Nuclear Company,Inc. It is being submitted by Exxon Nuclear to the U1 Nuclear Regulatory Commission as part of a technical contribution to facilitate safety analyses by licensees of the U1 Nuclear Regulatory Commission which utilize Exxon Nuclear fabricated reload fuel or other technical services provided by Exxon Nuc! car for light water power reactors and it is true and correct to the best of Ennon Nuclear's knowledge,information, and belief. The information contained herein may be used by the U1 Nuclear Regulatory Commission in its review of this report, and under the terms of the respective agreements, by licensees or applicants before the U1 Nuclest Regulatory Commission which are customers of Exxon Nuclear in their demonstration of compliance with the U1 Nuclear Regulatory Commission's regulations.

Exxon Nuclear's warranties and representations concerning the subject matter of this document are those set forth in the agreement between Exxon Nuclear and the customer to which this document is issued. Accordingly, except as otherwise expressly provided in such agreement, neither Exxon Nuclear not any person acting on its behalf:

A-Makes any warranty, or representation express or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this document, of that the use of any information, apparatus, method, or process disclosed in this -

document will not infringe privately owned rights, or B.

Assumes any liabilities with respect to the use of, or for 5

damages resulting from the use of, any information, apparatus, method, or process disclosed in this documen t.

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4 XN-NF-86-35 Revision 3

SUMMARY

OF REVISIONS I

Revision 3 to XN-NF-86-35 is issued to incorporate the results of transient analyses included in Revision 3 of XN-NF-86-36, " Grand Gulf Unit 1 Cycle 2 I

Plant Transient Analysis," and summarized in Appendix G of this report.

Revisions 1 of XN-NF-86-36 and XN-NF-86-35 contained transient analysis results based on the use of the Updated COTRANSA Hot Channel model, which was suspected to have a nonconservatism.

Revisions 2 of these reports rescinded the suspect results.

Revisions 3

reinstate these results following confirmation of the Updated COTRANSA Hot Channel model results for Grand Gulf through analysis with XCOBRA-T.

For more information on the XCOBRA-T I

confirmation of the Updated COTRANSA Hot Channel model for Grand Gulf, refer to Appendix B of XN-NF-86-36, Revision 3.

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XN-NF-86-35 Revision 3 TABLE OF CONTENTS P

SECTION P_ASE L

1.0 INTRODUCTION

1 2.0 FUEL MECHANICAL DESIGN ANALYSIS.....................

2 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS...................

3 3.2 Hydraulic Characterization..........................

3 3.2.3 Fuel Centerline Temperature.........................

3 3.2.5 Bypass F10w.........................................

3 3.3 MCPR Fuel Cladding Integrity Safety Limit...........

3 3.3.1 Nominal Coolant Condition in Monte Carlo Analysis...

3 f

3.3.2 Design Basis Radial Power Distribution..............

3 3.3.3 Design Basis local Power Distribution...............

3 1

4.0 NUCLEAR DESIGN ANALYSIS.............................

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4.1 Fuel Bundle Nuclear Design Analysis.................

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4.2 Core Nuclear Design Analysis........................

4 4.2.1 Core Configuration..................................

4 1

4.2.2 Core Reactivity Characteristics.....................

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5.0 ANTICIPATED OPERATIONAL OCCURRENCES.................

5 5.1 Analysis Of Plant Transients At Rated Conditions....

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5.2 Analyses For Increased Flow Operation...............

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5.3 Analyses For Extended Power Operation...............

6 5.4 ASME Overpressurization Analysis....................

6 5.5 Control Rod Withdrawal Error (CRWE).................

6 5.6 Fuel Loading Error (FLE)............................

7 5.7 Determination Of Thermal Margins....................

7 6.0 POSTULATED ACCIDENTS...............................

8 6.1 Loss-of-Coolant Accident............................

8 I

6.1.1 Break Location Spoetrum.............................

8 6.1.2 Break Size Spectrum.................................

8 6.1.3 MAPLHGR Analyses for ENC Fue1.......................

8 6.2 Control Rod Drop Accident...........................

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I XN-NF-86-35 TABLE OF CONTENTS (Continued)

SECTION EaGE 7.0 SPECIFICATION OF OPERATING LIMITS...................

10 7.1 Limiting Safety System Settings.....................

10 7.1.1 MCPR Fuel Cladding Integrity Safety Limit...........

10 7.1.2 Steam Dome Pressure Safety Limit....................

10 7.2 Limiting Conditions For Operation...................

10 7.2.1 Average Planar Linear Heat Generation Rate..........

10 7.2.2 Minimum Critical Powar Ratio........................

11 7.3 Surveillance Requirements...........................

12 7.3.1 Scram Insertion Time Surveillance...................

12 7.3.2 Stability Surveillance..............................

12 3

7.3.3 Procedural Controls.................................

12 3

8.0 METHODOLOGY REFERENCES..............................

13 9.0 ADDITIONAL REFERENCES...............................

14 APPENDICES A.

SINGLE LOOP OPERATION WITH ENC 8x8 FUEL.............

A-1 8.

INCREASED CORE FLOW OPERATION.......................

B-1 C.

EXTENDED LOAD LINE 0PERATION........................

C-1 D.

COMBINATION OF LHGR AND MAPLHGR LIMITS..............

D-1 5

E.

CALCULATION OF REDUCED FLOW LHGR LIMITS.............

E-1 F.

COMBINED LOCA-SEISMIC EVALUATION....................

F-1 G.

XCOBRA-T CONFIRMATION OF COTRANSA E

HOT CHANNEL RESULTS.................................

G-1 W

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111 XN-NF-86-35 Revision 3 LIST OF TABLES TABLE TITLE PAGE g

4.1 Neutronic Design Values..............................

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LIST OF FIGURES FIGURE TITLE PEE 3.1 Design Basis Radial Power Distribution...............

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3.2 Design Basis local Power Distribution -

ENC XN-1 8x8 Fuel....................................

17 3.3 Design Basis Local Power Distribution -

G.E. 8x8 Fuel........................................

18 4.1 Enrichment Distribution for ENC Design 2.99 SGd3.0...

19 4.2 Grand Gul f Cycle 2 Reference Core Loading............

20 5.1 Operating Power-Flow Map.............................

21

~

5.2 Flow-Dependent MCPR Limit............................

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5.3 Power-Dependent MCPR Limit...........................

23 5.4 Flow-Dependent MAPLHGR Factor........................

24 5.5 Power-Dependent MAPLHGR Factor.......................

25 7.1 MAPLHGR Operating Limit for ENC Fuel.................

26 7.2 Grand Gul f Power / Flow Operating Hap..................

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XN-NF-86-35 Revision 3

1.0 INTRODUCTION

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This report provides the results of the analyses performed by Exxon Nuclear Company (ENC) in support of the Cycle 2 reload for Grand Gulf Unit 1, which is 1

I scheduled to commence operation in Fall 1986.

This report is intended to be used in conjunction with ENC topical report XN-NF-80-19(A), Volume 4, Revision 1, " Application of the ENC Methodology to BWR Reloads," which describes the analyses performed in support of this reload, identifies the methodology used I

for those analyses, and provides a generic reference list.

Section numbers in this report are the same as corresponding section numbers in XN-NF-80-19(A),

Volume 4, Revision 1.

The numbered sections which are missing in this report were not needed and the reader is referred to XN-NF-80-19(A), Vol. 4, Rev.1 for their content.

The Grand GulT Unit 1 Cycle 2 core will comprise a total of 800 fuel assemblies, including 264 unirradiated ENC XN-1 8x8 assemblies, and 536 lg previously irradiated assemblies of an 8x8 lattice configuration fabricated by u

General Electric.

The reference core configuration is described in Section 4.2.

I The design and safety analyses reported in this document were based on the design and operational assumptions submitted for Grand Gulf Unit I during tha previous operating cycle.

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XN-NF-86-35 Revision 3 i

2.0 FUEL MECHANICAL DESIGN ANALYSIS Applicable Fuel Design Report:

Reference 9.1 The expected power history for the fuel to be irradiated during Cycle 2 of Grand Gulf Unit 1 is bounded by the assumed power history in the fuel mechanical design analysis.

Combination of the design LHGR limit and the power distribution limits from the LOCA analysis is discussed in Appendix 0.

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3 XN-NF-86-35 Revision 3 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS I

3.2 Hydraulic Characterization 3.2.3 Fuel Centerline Temoerature Maximum Centerline Temperature at 120% Power 3896 F Minimum Melting Point of Fuel 4841 F Minimum Margin to Centerline Melting 945 F 3.2.5 B.yoass Flow Calculated Bypass Flow Fraction 10.6%

(Exclusive of water rod flow) 3.3 MCPR Fuel Claddina Intecrity Safety limit 3.3.1 Nominal Coolant Conditions in Monte Carlo Analysis Thermal Power 4128 MWt Feedwater Flowrate 17.76 Mlbm/hr Core Outlet Pressure 1050 psia Feedwater Temperature 420 F 3.3.2 Desian Basis Radial Power Distribution Figure 3.1 3.3.3 Desian Basis local Power Distribution I

ENC XN-1 8x8 Fuel Figure 3.2 GE Fuel Figure 3.3 I

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4 XN-NF-86-35 Revision 3 4.0 NUCLEAR DESIGN ANALYSIS 4.1 Fuel Bundle Nuclear Desian Analysis Assembly Average Enrichment 2.81 w/o Radial Enrichment Distribution Figure 4.1 Axial Enrichment Distribution Uniform Axial lattice arrangement is uniform 2.99 w/o with six-inch natural urania sections at top and bottom.

Burnable Poisons Figure 4.1 Non-Fueled Rods Figure 4.1 Neutronic Design Parameters Table 4.1 l

4.2 Core Nuclear 03sion Analysis 4.2.1 Core Confiauration Figure 4.2 Core Exposure at E0C1 (Nominal) 8,173 MWD /MTU Core Exposure at E0C1 (Maximum) 8,960 MWD /MTU l

Core Exposure at B0C2 (Nominal) 5,999 MWD /MTV Core Exposure at E0C2 (Nominal) 13,651 MWD /MTV Core Exposure at E0C2 (Maximum) 14,172 MWD /MTU 4.2.2 Core Reactivity Characteristics BOC Cold K-effective, All Rods Out 1.10632 BOC Cold K-effective, All Rods In 0.93014 B0C Cold K-effective, Strongest Rod Out 0.96092 Reactivity Defect (R-Value) 1.34% rho Standby Liquid Control System Reactivity, Cold Conditions, 660 ppm 0.9522 i

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XN-NF-86-35 Revision 3 I

l 5.0 ANTICIPATED OPERATIONAL OCCURRENCES ll Applicable Generic Transient Analysis Report Reference 9.2 5.1 Analysis Of Plant Transients At Rated Conditions Reference 9.3 Limiting Transient (s): Load Rejection Without Bypass (LRNB)

Feedwater Controller Failure (FWCF)

I INIT INIT PEAK PEAK PEAK DELTA I

EVENT POWER FLOW POWER HT FLX PRESSURE CPR MODEL

_7did 7ttd

_5ttd

%rtd osia LRNB 104.2 100 131.4 100.5 1229 0.02 COTRANSA FWCF 104.2 100 109.9 107.3 1194 0.04 COTRANSA I

5.2 Analyses For Increased Flow Operation Reference 9.3 (ICF Region in Figure 5.1)

Limiting Transient (s): Load Rejection Without Bypass (LRNB)

Feedwater Controller Failure (FWCF)

INIT INIT PEAK PEAK PEAK DELTA EVENT POWER FLOW POWER HT FLX PRESSURE CPR MODEL I

_Id1d 7tttd

.3r_td

%rtd osia LRNB 104.2 108 144.6 101.5 1232 0.02 COTRANSA LRNB 40 108 153.3 48.7 1137 0.28 COTRANSA FWCF 104.2 108 110.1 107.4 1198 0.04 COTRANSA FWCF 40 108 76.5 43.6 1012 0.09 COTRANSA I

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6 XN-NF-86-35 Revision 3 5.3 Analyses For Extended Power Ooeration Reference 9.3 (ELL Region in Figure 5.1)

Limiting Transient (s): Load Rejection Without Bypass (LRNB)

Feedwater Controller Failure (FWCF)

INIT INIT PEAK PEAK PEAK DELTA EVENT POWER FLOW POWER HT FLX PRESSURE CPR MODEL

_1r_td 3rld

_Er_td

%rtd osia LRNB 104.2 73.8 191.8 101.5 1231 0.05 COTRANSA LRNB 70 100 103.8 69.1 1192 0.04 COTRANSA LRNB 70 40 70.0 67.3 1185 0.16 COTRANSA LRNB 40 100 142.4 48.7 1143 0.27 COTRANSA LRNB 25 73.8 60.9 44.4 1180 1.05 COTRANSA E

LRNB 25 40 86.6 38.3 1174 0.80 COTRANSA E

FWCF 104.2 73.8 109.5 106.5 1189 0.06 COTRANSA FWCF 70 100 73.3 70.9 1071 0.03 COTRANSA FWCF 70 40 74.9 71.7 1064 0.15 COTRANSA FWCF 40 100 69.6 42.8 1009 0.07 COTRANSA FWCF 25 73.8 32.5 26.9 974 0.08 COTRANSA E

FWCF 25 40 27.8 25.5 966 0.07 COTRANSA E

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5.4 ASME Overoressurization Analysis Reference 9.3 Limiting Event MSIV Closure Worst Single Failure Direct Scram Maximum Pressure 1296 psia Maximum Steam Dome Pressure 1280 psia l

l 5.5 Control Rod Withdrawal Error (CRWE)

I See Reference 9.5 for generic statistical evaluation of CRWE.

Power dependent MCPR limits are established in the analysis.

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7 XN-NF-86-35 Revision 3 I

5.6 Fuel Loadina Error (FLE.1 Reference 8.1 I

Initial MCPR 1.31 Final MCPR 1.20 I

Delta-CPR 0.11 5.7 D.g. termination Of Thermal Marains

SUMMARY

OF THERMAL MARGIN RE0VIREMENTS AT RATED CONDITIONS Ey.fE EQWEg ELQM DELTA-CPR MCPR LIMIT LRNB 104.2 100 0.02 1.08 FWCF 104.2 100 0.04 1.10 I

LFWH Reference 9.4 0.08 1.14 CRWE Reference 9.5 0.10 1.16 FLE 0.11 1.17 MCPR OPERATING LIMIT AT RATED CONDITIONS 1.17 MCPR OPERATING LIMITS AT OFF-NOMINAL CONDITIONS Reference 9.3 Flow Dependent MCPR Limit Figure 5.2 Power Dependent MCPR Limit Figure 5.3 MAPLHGR LIMIT FACTORS FOR OFF-NOMINAL CONDITIONS MAPFAC(f) (APPENDIX E)

Figure 5.4 MAPFAC(p) (Reference 9.3)

Figure 5.5 I

  • 1.18 to be used in Cycle 2 for continued use of Cycle 1 limits.

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XN-NF-86-35

(

Revision 3 1

[

6.0 POSTULATED ACCIDENTS 6.1 Loss-of-Coolant Accident 6.1.1 Break Location Snectrum Reference 9.6 6.1.2 Break Size Spectrum Reference 9.6

[

6.1.3 MAPLHGR Analyses for ENC Fuel Reference 9.7 Limiting Break: Double-Ended Guillotine Pipe Break in Recirculation Pump Discharge Line with 1.00 Discharge Coefficient (1.0 DEG/RD)

Peak Local Average Planar Analyzed Peak Clad Metal-Water Exonsure MAPLHGR Temnerature Reaction 0 GWD/MTU 14.3 kW/ft l'738 F 0.4%

5 14.3 1685 0.3 10 14.3 1678 0.3 15 14.3 1687 0.3 20 14.3 1680 0.3

[

25 13.2 1642 0.3 30 12.1 1575 0.2 35 11.1 1496 0.1

(

40 10.0 1403 0.1 45 9.0 1321 0.1 C

[

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XN-NF-86-35 Revision 3 I

6.2 Control Rod Droo Accident Reference 8.1 Dropped Control Rod Worth 12.46 mk Doppler Coefficient

-9.5 x 10**(-6) g 6K/K/*F W

Effective Delayed Neutron Fraction 0.005 Four-Bundle Local Peaking Factor 1.300 Maximum Deposited Fuel Rod Enthalpy 218 cal /gm I

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L 10 XN-NF-86-35 Revision 3

[

7.0 SPECIFICATICN OF OPERATING LIMITS 7.1 Limitina Safety System Settinas 7.1.1 MCPR Fuel Claddina Intearity Safety limit l

MCPR Safety Limit 1.06 l

7.1.2 Steam Dome Pressure Safety Limit l

Pressure Safety Limit 1325 psig l

7.2 Limitina Conditions For Ooeration l

7.2.1 Averace Planar Linear Heat Generation Rate Figure 7.1 I

l The following values correspond to an average planar representation of the design LHGR limit.

The LOCA analysis was performed at higher MAPLHGR values for additional canservatism relative to 10 CFR 50.46 and Appendix K.

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I 11 XN-NF-86-35 Revision 3 I

Average Planar Exoosure MAPLHGR 0.00 GWD/MTU 13.20 kW/ft 3

0.25 13.20 E

1.00 13.38 2.00 13.54 4.00 13.89 6.00 14.26 8.00 14.26 10.00 14.12 g

15.00 13.78 g

20.00 13.30 24.00 13.03 25.00 12.96 25.40 12.94 30.00 11.77 35.00 10.48 E

l 40.00 9.15 E

l 42.00 8.61 I

MAPLHGR Limit Multipliers for Off-Nominal Conditions I

MAPFAC(f)

Figure 5.4 l

MAPFAC(p)

Figure 5.5 7.2.2 Minimum Critical Power Ratio Rated Conditions MCPR Limit 1.18 MCPR(f)

Figure 5.2 MCPR(p)

Figure 5.3 I

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Revision 3 I

E 7.3 Surveillance Reouirements 1

m 7.3.1 Scram Insertion Time Surveillance L

j Thermal margins are based on analyses in which scram perfornance was assumed L

consistent with the technical specification limits.

No additional surveillance for scram performance is required.

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7.3.2 Stability Surveillance Acceptable surveillance procedures for potentially unstable operation shall be instituted in the portion of the operating power-flow map bounded by lines L

defined by constant decay ratios of 0.75 and 0.90 as calculated by COTRAN.

Figure 7.2 presents a graphical description of this region.

I 7.3.3 Procedural Controls Procedural controls shall be instituted to assure that normal operation of the fuel remains within the steady state LHGR assumptions of the mechanical design analysis.

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I 13 XN-NF-86-35 Revision 3 8.0 METHODOLOGY REFERENCES

'I Section 8 references are contained in the following reference report:

" Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," XN-NF-80-19(A), Volume 4, Revision 1,

Exxon Nuclear Company, Richland, Washington (March 1985).

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14 XN-NF-86-35 Revision 3 9.0 ADDITIONAL REFERENCES I

9.1 " Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"

XN-NF-85-67(P), Revision 1, Exxon Nuclear Company, Richland, Washington (April 1986).

9.2 " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors,"

XN-NF-79-71(P), Revision 2, Exxon Nuclear Company, Richland, Washington I

(November 1981).

9.3 " Grand Gulf Unit 1 Cycle 2 Plant Transient Analysis," XN-NF-86-36, Revision 3, Exxon Nuclear Company, Richland, Washington (August 1986).

9.4 "A Generic Analysis of the Loss of Feedwater Heating Transient for Boiling Water Reactors," XN-NF-900(P), Exxon Nuclear Company, Richland, I

Washington (February 1986).

9.5 "BWR/6 Generic Rod Withdrawal Error Analysis, MCPRp," XN-NF-825(A), Exxon I

Nuclear Company, Richland, Washington (April 1985) and XN-NF-825(P),

Supplement 2 (January 1986).

9.6 " Generic LOCA Break Spectrum Analysis for BWR/6 Plants," XN-NF-86-37(P),

I Exxon Nuclear Company, Richland, Washington (April 1986).

9.7 " Grand Gulf Lnit 1 LOCA Analysis," XN-NF-86-38, Exxon Nuclear Company, Richland, Washington (June 1986).

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I 15 XN-NF-86-35 Revision 3 I

l Table 4.1 Neutronic Design Values l

Core Data I

l Number of fuel assemblies 800

!B Rated thermal power, MW 3833 lg Rated core flow, Mlbm/hr 112.5 Core inlet subcooling, BTU /lbm 22.9 Moderator temperature, F 551 I

Channel thickness, inch 0.120 Fuel assembly pitch, inch 6.0 Water gap thickness, inch 0.545 I

Control Rod Data Absorber material B4C Total blade span, inch 9.804 Total blade support span, inch 1.550 I

Blade thickness, inch 0.3280 Blade face-to-face internal dimension, inch 0.238 Absorber rods per blade 72 I

Absorber rod outside diameter,, inch 0.220 Absorber rod inside diameter,iinch 0.166 Absorberdensity,%oftheorQical 70

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I 17 XN.NF-86-35 Revision 3 I

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0.92 : 0.96 :

1.02 :

1.06 :

1.06 1.02 : 0.95 :

1.04 I

0.96 : 0.97
0.91 1.07 1.06
0.97 : 0.94 : 0.95 I

1.02 : 0.91 1.04 :

1.01

0.99 :

1.02 : 0.97 1.02 I

1.06 :

1.07 :

1.01

0.00 : 0.87
0.99 :

1.06 :

1.06 I

1.06 :

1.06 : 0.99 : 0.87 : 0.00 1.00 :

1.07 :

1.06 I

1.02 : 0.97 :

1.02 : 0.99 :

1.00 :

1.03 : 0.87 1.02 I

I

0.95 : 0.94 : 0.97 1.06 :

1.07

0.87 :

1.05

0.95 I

1.04 : 0.95 :

1.02 1.06 :

1.06 :

1.02 : 0.95 :

1.04 FIGURE 3.2 DESIGN BASIS LOCAL POWER DISTRIBUTION I

ENC XN-1 8X8 FUEL

  • Fuel rod adjacent to control blade position I

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18 XN-N.F.85-35 Revision 3 I

1.03 : 1.00 : 0.99 : 0.99 : 0.99 : 0.99 :

1.00 :

1.03 1.00 : 0.97

0.99 1.02 : 1.03 :

1.03 : 0.99 :

1.00

0.99 : 0.99 :

1.02 :

1.01

1.02 : 0.91 : 1.03 : 0.99 :
0.99 :

1.02 1.01

0.91
0.00 :

1.02 : 1.02 : 0.99 :

0.99 :

1.03 1.02 : 0.00 : 1.02 : 1.01

0.99 : 0.99 :
0.99 :

1.03

0.91 1.02 :

1.01

0.98 : 0.99 : 0.99 :

1.00 : 0.99 :

1.03 1.02 : 0.99 : 0.99 : 0.97 :

1.00 1.03 :

1.00 : 0.99 : 0.99 : 0.99 : 0.99 :

1.00 :

1.03 FIGURE 3.3 DESIGN BASIS LOCAL POWER DISTRIBUTION G.E. 8X8 FUEL

  • Fuel rod adjacent to control blade location I

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19 XN-NF-86-35 Revision 3 1

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LL L

ML :

M M

ML L

L I

L

ML*

ML :

H H

M

ML*

L I

ML :

ML H

H H

H M

ML I

M H

H W

M H

H M

I M

H H

M W

H H

M I

ML M

H H

H H

ML*
ML I

L

ML*

M H

H

ML*

M L

I I

L L

ML M

M ML :

L L

LL Rods ( 1) 1.50 w/o U235 L Rods (11)

--- 2.00 w/o U235 ML Rods (10) 2.64 w/o U235 M Rods (15)

--- 3.03 w/o U235 H Rods (20)

--- 3.84 w/o U235 ML* Rods ( 5)

--- 2.64 w/o U235 + 3.00 w/o Gd203 W Rods ( 2)

Inert Water Rod I

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ENC Design 2.99 5Gd3.0 I

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! B1 ! B1 ! B1 ! B1 ! B1 ! B1 ! C0 ! B1 ! B1 ! B1 ! C0 ! B1 ! C0 ! B1 ! B1 ! B1 !

! B1 ! CO ! B1 ! C0 ! B1 ! C0 ! B1 ! C0 ! B1 ! C0 ! B1 ! C0 ! B1 ! CO ! B1 ! Al !

! B1 ! B1 ! B1 ! B1 ! C0 ! B1 ! C0 ! B1 ! B1 ! B1 ! C0 ! B1 ! C0 ! B1 ! B1 !

! B1 ! C0 ! B1 ! C0 ! B1 ! C0 ! B1 ! B1 ! B1 ! C0 ! Bl ! CO ! B1 ! C0 ! Al !

! B1 ! B1 ! C0 ! B1 ! B1 ! B1 ! B1 ! B1 ! CO ! B1 ! C0 ! B1 ! C0 ! B1 ! Al !

! B1 ! C0 ! B1 ! C0 ! B1 ! C0 ! B1 ! C0 ! B1 ! C0 ! B1 ! C0 ! B1 ! C0 ! A1 !

i

! C0 ! B1 ! C0 ! B1 ! B1 ! B1 ! CO ! B1 ! C0 ! B1 ! C0 ! B1 ! C0 ! B1 ! Al !

! B1 ! C0 ! B1 ! B1 ! B1 ! C0 ! B1 ! Al ! B1 ! C0 ! B1 ! C0 ! B1 ! Al !

g 5

! B1 ! B1 ! B1 ! Bl ! C0 ! B1 ! C0 ! Bl ! C0 ! Bl ! C0 ! Bl ! Al !

! B1 ! CO ! B1 ! C0 ! B1 ! C0 ! B1 ! CO ! B1 ! C0 ! B1 ! B1 ! Al !

E

! C0 ? B1 ! C0 ! B1 ! C0 ! B1 ! C0 ! B1 ! C0 ? B1 ! B1 ! B1 ! Al !

E I

! B1 ! C0 ! B1 ! C0 ! B1 ! C0 ? B1 ! C0 ! B1 ! B1 ! B1 ! Al !

! C0 ! B1 ! C0 ! B1 ! C0 ! B1 ! C0 ! B1 ! Al ! Al ! Al !

j 5

! B1 ! C0 ! B1 ! C0 ! B1 ! C0 ! B1 ! Al !

! B1 ! B1 ! B1 ! Al ! Al ! Al ! Al !

l u

! B1 ! Al !

Fuel Type No. of Bundles Description A

80 GE 8x8 Type JI 1.60 w/o U-235 (enriched zone)

B 456 GE 8x8 Type III 2.10 w/o U-235(

)

C 264 ENC 8x8 XN-1 2.99 w/o U-235

(

)

Figure 4.2 Grand Gulf Cycle 2 Reference Core loading I

M M

m M

M mma m

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m 'm mm a

mmm

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FIG. 5.1 OPERATING POWER-FLOW MRP 120 -

(73.8,100)

(105,100) h80-ELL Region j-Region ICF

/

g 100% Rod Line g

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g 40 -

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I (40,25)

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FIG.

5.3 POWER-DEPENDENT MCPR LIMIT

2. 4 -

l

2. 2 -

Core Flow >50%

2-

- Core Flow 250%

f

1. 8 -

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10 20 30 40 50 60 70 80 90 100 110 120

-l CORE POWER, 7. OF RATED 1

1 i

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A l

FIG. 5.4 FLOW-DEPENDENT MRPLHGR FRCTOR 1.1 -

)

1-l

0. 9 -

1 Eb E

C

0. 8 -

ke E

i

0. 7 -

ENC 107% MAX FLOW GE 107% MAX FLOW ENC 102.5% MAX FLOW GE 102.5% MAX FLOW

0. 6 -

??

0.5

{y O

10 20 30 40 50 60 70 80 90 100 110 120 88 CORE FLOW, % OF RATED ws i

M U

U M

H W

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m m 'M ~ m M

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FIG.

5.5 POWER-DEPENDENT MRPLHGR FRCTOR 1.1 -

1-

0. 9 -

E

0. 8 -

b

{

Power >40% All Core Flows b

0. 7 -

Core Fiow $50%

Power $40%-

Core Flow >50%

0. 6 -

Power $40%

0. 5 -

_??

Gin 0

l0 2O 3O IO 50 6O 7O 8O 90 1$0 110 120 CORE POWER, 7. OF RATED

FIG. 7.1 MAPLHGR OPERATING LIMIT FOR ENC FUEL 18 -

16 -

14 -

g L

5 s

x g 12 -

cn I

_a 0E 10 -

8-a's

5. a 0

hh o

s la ls 2o 2s 3o 3s 4O 45 RVERAGE PLRNRR EXPOSURE, GWD/MTU w$

t i

m mm m

W W

W W

W h

M M

M M

M M

M W

W

I W W

W M

M M

M M

M M

M M

M M

M M

M M

M 110 i

i i

i i

i i

l 100 l

% Power = 197-8.59(% Flow)+.130(% Flow)2 l

Detect and Suppress Region go

% Power =72.99-2.59(% Flow)+.0547(% Flow)2 0

80 Eo 0-70 80% Rod Line o

G o

60 m

39% Flow - H j

g l

O oEe oL2 30 20 10 O

0 10 20 30 40 50 60 70 80 90 100 110 y5 j

~

Percent of Rated Flow M

88:

J Figure 7.2 Grand Gulf Power / Flow Operating Map us I

I A-1 XN-NF-86-35 Revision 3 I

APPENDIX A SINGLE-LOOP OPERATION WITH ENC 8x8 FUEL Analyses have been performed for Grand Gulf Unit I for normal two pump operation both by the NSSS vendor and Exxon Nuclear Company (ENC). Generally, both analyses showed similar results and yielded comparable allowed operating limits.

Since the ENC and vendor 8x8 fuel designs are very similar, this result is to be expected.

The ability to operate Grand Gulf Unit I with only one recirculation pump running is highly desirable in the event that a recirculation pump or other component maintenance renders one loop inoperative.

In order to justify single-loop operation, the NSSS vendor has performed additional accident and transient analyses for single-loop operating conditions (Reference A.1).

The single-loop operation analysis generally showed that operation within the I

full-power two pump operating MCPR limits will assure that the safety limit MCPR is not violated and that substantial margin to the safety limit exists for single-loop operation due to the reduced power.

For these cases, ENC fuel will likewise experience the benefit of the power reduction and application of two pump full-power MCPR limits for the ENC fuel designs is conservative and appropriate. This Appendix discusses appropriate limits for Grand Gulf Unit 1 Cycle 2 operation with ENC 8x8 fuel and their bases.

I A.1 R0D WITHDRAWAL ERROR The rod withdrawal error (RWE) transient analysis for BWR/6's was performed for both rated and off-rated con'ditions (Ref. XN-NF-825(A), as supplemented).

The analysis included a statistical evaluation of the minimum critical power ratio (MCPR) due to the withdrawal of ganged. control rods at rated and off-rated ccnditions. The analysis is valid for all rated and off-rated power I

I

1 I,

A-2 XN-NF-86-35 Revision 3 I

and flow state points (which include single-loop operation state points) shown in Figure 3.1 of the Reference.

A.2 TRANSIENT MCPR LIMITS Operating with one recirculation loop results in a maximum power output which is about 30% below that which is attainable for two pump operation.

Flow and Power dependent MCPR functions require a significantly increased operating MCPR over the allowed full-power full-flow limit.

Therefore, the NSSS vendor single-loop analysis showed that the consequences of abnormal operation transients will be considerably less severe than those analyzed from a two-loop operational mode.

These results are shown in Table 15.C.3-3 of Reference A.I.

The limiting transients from an allowed MCPR operating limit g

of 1.41 gave transient MCPRs of 1.24-1.41 which are well above the GE safety E

limit of 1.07 with a 0.17 or greater margin in delta CPR.

For pressurization, flow increase, flow decrease, and cold water injection transients, results for two-loop operation bourid both the thermal and overpressure consequences of one-loop operation.

It was concluded that the MCPR operating limits established for two-pump operation are also conservatively applicable to single-loop operation conditions. This is true even for the increased safety g

limit associated with single-loop operation (see A.4).

g i

The increased MCPR margin for single-loop operation at reduced flow and power is also applicable to ENC 8x8 fuel designs.

Therefore, the operating MCPR l

limits established for two-pump operation with ENC fuel will be conservative when applied to single-loop operation for the same reasons as for the vendor l

fuel. Applicability of two-pump limits for single pump operation is discussed g

phenomenologically in the following section.

E A.3 ABNORMAL OPERATING TRANSIENTS MCPR limits established for full flow two-loop operation are conservative for single-loop operation because of the physical phenomena related to part-power I

1 I

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A-3 XN-ilF-86-35 Revision 3 I

part-flow operation, not because of features in reactor analysis models or compatible fuel designs.

A review of the most limiting delta CPR transients I

for single-loop operation was conducted. Under single-loop conditions, steaay state operation cannot exceed approximately 70% power and 54% core flow because of the capability of the recirculation loop pump.

Thus, the MCPR limit at maximum single-loop operation flow and power is higher than the two-pump operating MCPR limit due to the flow and power dependent MCPR functions.

The MCPR flow dependence is based on a flow increase transient with two operating recirculation loops.

Flow increase transients with single I

recirculation loop operation would be much less severe but the conservative two loop limit is retained.

A.3.1 Load Re.iection Without Bypass I

A possible limiting system transient for the Grand Gulf is the Load Rejection Without Bypass (LRNB) pressurization transient.

In this transient, the I

primary phenomena is the pressurization caused by abruptly stopping the steam flow through rapid closure of the curbine control valve.

When the rapid pressurization reaches the core it causes a power excursion due to void collapse.

I At these reduced power and flow condition there is a corresponding reduction in steam flow.

With lower steam flow the maximum pressurization of the core I

is reduced in comparison to rated conditions when the control valve is closed.

The resulting power excursion and associated delta CPR show substantial margin to the two-loop operating limits.

Thus the MCPR limits based on LRNB analyses for two-loop operation are conservatively applicable to the lower powers associated with single-loop conditions based on the physics of the transient.

GE analyses (A.1) under I

single-loop conditions also confirm this trend.

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A-4 XN-NF-86-35 Revision 3 A.3.2 Feedwater Controller Failure Feedwater controller failure to maximum demand from the single-loop reduced power and flow initial conditions results in a larger increase :n feedwater flow, a faster vessel water level rise, and an earlier scram than the same transient from rated conditions. The relative increase in power and heat flux is less for single-loop operation than for rated two-loop conditions due to the earlier scram.

Application of two-loop operating MCPR limits is conservative for this event.

A.3.3 Pumo Seizure Accident I

Core flow drops rapidly during this event and power is shutdown by increased voiding.

The NSSS supplier demonstrated that the increased initial MCPR margin required for single-loop operation at reduced power and flow through the MCPR(f) and MCPR(p) limits more than offsets the increased transient delta CPR resulting from this event.

Because of similarity in fuel designs and the fact that the NSSS MCPR(f) and MCPR(p) Cycle 1 limits are being retained for g

Cycle 2, this event will be covered by Cycle 2 limits.

m A.3.4 Loss of Feedwater Heatina A generic statistical loss of feedwater heater analysis using a variety of conditions attainable while operating within the extended power flow maps (one-pump or two-pump operation) conservatively determined the MCPR for g

protection to LFWH limit.

This limit is independent of flow and power.

This W

j analysis applies for single-loop operating conditions.

I A.3.5 Summary It is very conservative to use the reduced flow and power dependent two-loop operating MCPR limit for single-loop operations.

Maintaining this two-loop g

limit assures that there is even more thermal margin under single-loop 5

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I A-5 XN-NF-86-35 Revision 3 I

conditions than under two-loop full power full flow conditions.

I A.4 SAFETY LIMIT MCPR For single-loop operation, the NSSS vendor found that an increase of 0.01 in the MCPR safety limit was needed to account for the increased flow measurement uncertainties and increased tip uncertainties associated with single pump operation.

ENC has evaluated the effects of the increased flow measurement uncertainties on the safety limit MCPR and found that the NSSS vendor determined increase in the allowed safety limit MCPR is also appropriate for application to ENC fuel during single-loop operation.

Thus, increasing the I

safety limit MCPR by 0.01 for single-loop operation (1.07) with ENC fuel is sufficiently conservative to also bound the increased flow measurement uncertainties for single-loop operation.

A.5 MAPLHGR LIMITS The NSSS vendor has also evaluated the changes in the two-loop MAPLHGR limits I

required to permit single-loop operation.

A multiplier of 0.86 is to be applied to the appropriate rated two-loop MAPLHGR limit to obtain the MAPLHGR limit for single-loop operation.

The need to reduce the allowed MAPLHGR arises because of the conservative assumption of early boiling transition (at 0.1 sec) in the LOCA-ECCS analysis applied for single-loop operation at reduced core flow.

I To support operation of Grand Gulf Unit I with Exxon Nuclear Company (ENC) 8x8 fuel with a single recirculating pump operating, the GE MAPLHGR limits for the highest enriched GE 8x8R fuel design (Type 8CR210) with a multiplier of 0.86 are to be applied to ENC 8x8 fuel for single-loop operation.

The basis for this is two-fold:

1)

The phenomena which require the reduction in MAPLHGR limits are a result of operation of the Grand Gulf Unit 1 system with single I

I

1 I

I A-6 XN-NF-86-35 Revision 3 I

active recirculation loop, and are therefore, equally applicable to both GE and ENC fuel designs, and l

2)

For the expected exposures during Cycle 2 operation the analysis methods used by GE have yielded conservative MAPLHGR limits relative to the MAPLHGR limits obtained using the ENC approved analysis g

models.

Therefore, applying the more conservative GE MAPLHGR limit 3

to ENC fuel provides a limit which assures conformance to NRC 10 CFR 50.46 criteria.

The major difference between operation with both recirculation pumps running and operating with o'ily one active recirculation pump are reduced operating core flow, reduced core power, and reverse flow through the inactive loop jet E

pumps.

Flow and power dependent MCPR limits assure reduced maximum assembly 3

power during single-loop operation. The primary system coolant inventory and LOCA break conditions are essentially unchanged from the two-loop operation.

Thus, the uncovery of the jet pump suction, recirculation suction line uncovery, and system depressurization rate would be expected to change little between one and two-loop operation.

The phenomena associated with these key parameters largely determine LOCA analysis results for both ENC and GE g

analyses.

The analyses performed by GE confirm this system behavior in that a

the limiting pipe break LOCA is essentially unchanged from the two-loor analysis, as are the break size and core uncovery and reflood times. Although ENC LOCA analysis methods differ from those of GE, similar results would be expected from an ENC analysis because the phenomena are governed by the system parameters.

The principal LOCA concern associated with sirgle-loop operation is the possibility of the LOCA break occurring in the operating loop, in which case there is no coastdown of an intact loop recirculation pump to sustain jet pump and core flow during the early portion of the system blowdown.

An early boiling transition (CHF) may result from this early loss of flow capability.

I I

I A-7 XN-NF-86-35 Revision 3 I

To account for this possibility, GE derived a single-loop operation MAPLHGR I

multiplier of 0.86 to be used with calculated two-loop MAPLHGR limits during single-loop operation.

The analyses which determined this multiplier assumed a near instantaneous boiling transition (0.1 sec) even though a longer boiling transition time may have been calculated using approved models.

This assumption is very conservative.

I The major difference between the ENC and GE methodologies that would effect I

analysis differences between single and two-loop operation is in the blowdown heat transfer.

ENC's more mechanistic model calculates boiling transition times that are equivalent to or later than those reported from the GE model, and the ENC model explicitly calculates the blowdown heat transfer throughout the blowdown period while the GE model assumes an adiabatic heatup period.

Thus, the conservative approach taken in the GE analysis of assuming an early boiling transition (0.1 sec) for single-loop operation would yield a greater I

penalty using ENC methodology than for the more conservative GE methods.

For this

reason, limits based on the more conservative GE analysis are recommended.

ENC's more mechanistic heat transfer during the GE adiabatic I

heatup period would partially offset this effect, thus, making the recommended limits conservative for ENC fuel.

I Application of GE calculated 8x8 type 8CR210 MAPLHGR limits times 0.86 to ENC 8x8 fuel for single-loop operation will conservatively assure that the NRC criteria of 10 CFR 50.46 will be met for the following reasons:

1)

MAPLHGR limits for ENC 8x8 fuel in Grand Gulf are higher than the equivalent GE 8x8 fuel limits in all cases for bundle exposures less than 19,000 MWD /MTV.

Even with the higher MAPLHGRs, the ENC analysis showed greater PCT margin than in the GE analysis.

An ENC analysis for the similar single-loop operating conditions would be expected to also yield MAPLHGR limits equal to or higher than those obtained by GE.

I I

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A-8 XN-NF-86-35 Revision 3 2)

The MAPLHGR reduction factor to protect against early boiling transition determined by GE is based on a conservative early boiling g

transition.

W 3)

ENC analysis for two-loop operation at expected exposures in Cycle 2 of Grand Gulf Unit I with 8x8 fuel justifies MAPLHGR limits equal to or greater than the GE 8x8 design.

The MAPFAC(f) and MAPFAC(p) multipliers on the MAPLHGR are not related to LOCA/ECCS conditions, but are to protect fuel design limits during transients initiated from off-rated conditions.

Since transient effects are substantially limited under single-loop operating conditions, the use of MAPFAC(p) and either GE or ENC MAPFAC(f) with the 0.86 MAPLHGR limit conservatively protects ENC transient LHGR limits.

For consistency, the GE fuel MAPFAC(f) factor should be used with the GE MAPLHGR limit for application to ENC fuel under single-loop conditions.

For Cycle 2 of Grand Gulf Unit I single-loop operation with ENC 8x8 fuel, a MAPLHGR limit corresponding to 0.86 times the MAPLHGR limits for the highest enriched Cycle 1 GE fuel type along with the GE MAPFAC(f) multiplier can be conservatively used.

A.6 STABILITY Grand Gulf Unit I has adopted a detect and suppress approach to avoid unstable reactor operation. This is consistent with single-loop operation requirements stated in NRC Generic Letter #86-09 (Reference A.2).

I A.7 REFERENCES A.1 General Electric Co.,

"GGNS Single-Loop Operation Analysis", General Electric Co., February 1986.

A.2 " Technical Resolution of Generic Issue No. B-59-(N-1) Loop Operation in BWRs and PWPs", (Generic Letter No. 86-09), March 31, 1986.

I I

W B-1 XN-NF-86-35 Revision 3 APPENDIX B INCREASED CORE FLOW OPERATION Analyses were undertaken to support plant operation at up to 105% of rated recirculation fl ow.

Qualification of balance-of-plant systems and the

^

coresident fuel was accomplished by the licensee.

ENC analyses covered the specification of operating limits defined by anticipated and accident conditions included in the power distribution limits in the plant Technical Specifications.

Loss of Coolant Accident (LOCA) analyses were performed at the maximum and E

minimum flow points at which rated thermal power may be reached during

]

operation within the operating power flow map assumption.

LOCA results reported in the body of this document reflect the most adverse consequences L

observed during this flow evaluation.

e l

Plant transients were evaluated at a number of power flow states in the Increased Core Flow region. Analytical results were essentially as expected, and protection of cladding temperature (MCPR) and cladding strain (LHGR) limits is provided by the off-nominal MCPR and LHGR limits identified in the body of this document.

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C-1 XN-NF-86-35 Revision 3 I

APPENDIX C I

EXTENDED LOAD LINE OPERATION Analyses were undertaken to support plant operation above the nominal 100%

flow control line.

Qualification of balance-of-plant systems and the corasident fuel was accomplished by the licensee.

ENC analyses covered the I

specification of operating limits defined by anticipated and accident conditions included in the power distribution limits in the plant Technical Specifications.

Plant transients were evaluated at a number of power-flow states in the Extended Load Line region.

Protection of cladding temperature (MCPR) limits I

and cladding strain (LHGR) limits is provided by the off-nominal MCPR and MAPLHGR limits identified in the body of this document.

E I

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F D-1 XN-NF-86-35 Revision 3

-I APPENDIX D COMBINATION OF LHGR AND MAPLHGR LIMITS Operating limit target values for MAPLHGR and LHGR were selected to allow maintenance ~of design margins throughout Cycle 2.

These limits differ only by a factor equal to the maximum local peaking factor for the limiting bundle at each exposure point.

, The local peaking factors which correlate the MAPtHGR limits identified in this report with the design LHGR limits are given in Figure D-1.

I For added assurance that the design LHGR limits are protected during normal and anticipated operation, the core monitoring software evaluates operating LHGR values against the design limits as a procedural control as described in XN-NF-80-19(A), Volume 4, Revision 1.

I I

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l FIG.

D-1 BURNUP-DEPENDENT LOCAL PERKING 1.25 -

l 1.20 -

a O

c ta aN Zy 1.15 -

c (4.1 Q-1 g

oO

-3 1.10 -

\\

N O

r4 1.05 i

i 0

5 10 15 20 25 30 35 40 45 50 wM RVERRGE PLRNRR (NODRL) EXPOSURE, GWD/MTU i

M M

M M

M M

M M

M M

M M

M M

M M

M i

a

E-1 XN-NF-86-35 Revision 3 I

APPENDIX E CALCULATION OF REDUCED FLOW LHGR LIMITS I

An analysis has been performed to determine the flow dependence associated I

with the rated conditions LHGR limit for ENC fuel in BWR/6 plants.

To be consistent with the operating limits provided by the NSSS supplier, this dependence is administered as a flow-dependent multiplier on the MAPLHGR limit.

I The flow-dependent MAPLHGR factor, or MAPFAC(f), is based on the relative change in the maximum core linear heat generation rate due to a flow increase I

event. The resulting limits are flow-dependent fractions which are applied to the rated conditions operating MAPLHGR limits when operating at reduced flow conditions.

These limits are designed to protect the fuel from exceeding transient design LHGR limits during postulated flow increase transients.

I The MAPFAC(f) factors were determined by simulating BWR/6 flow increase transients with XTGBWR (Ref. 8.1).

Low flow projected operating statepoints I

along the 100% flow control line were used as initial conditions.

Final conditions were determined by increasing the core flow to either 102.5% or 107% of rated flow and iterating on core power until the initial conditions k-effective was obtained.

A total of 147 simulations of BWR/6 flow increase transients were generated I

with XTGBWR.

Initial conditions from which the flow runup transient were simulated included several BWR/6 cycle stepouts, numerous calculations for each rod sequence (A1, A2, 81, B2 rod sequence) and core flows ranging from I

30% to 90% of rated.

The percent increase in the LHGR was fit to a second order polynominal.

A statistical analysis was performed using these 147 data I

I 1

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E-2 XN-NF-86-35 Revision 3 I

points to generate upper bound 95/95 MAPFAC(f) operating limits as a function of core flow. The resulting limits fcr ENC fuel are shown in Figure 5.4.

The MAPFAC(f) limits determined in this fashion, when combined with the g

MAPFAC(p) limits determined during the plant transient analysis, define the 3

maximum allowable MAPLHGR throughout the Cycle 2 operating power-flow map.

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F-1 XN-NF-86-35 Revision 3 APPENDIX F COMBINED LOCA-SEISMIC EVALVATIOM I

The structural response of the ENC XN-18x8 fuel is the same as the structural I

response of the GE6 8x8 fuel it replaces in the Grand Gulf Unit I core.

Therefore, the LOCA-seismic structural response evaluation performed in support of the initial core remains applicable and continues to provide assurance that control blade insertion will not be inhibited following the occurrence of the design basis LOCA-seismic event.

The physical and geometric properties of the ENC and GE 8x8 fuel types are I

summarized in Table F-1.

The close agreement between the important parameters for the two fuel types indicates that the structural response would be very similar for both fuel types.

Similarity in the natural frequencies of the two fuel types is further assured by the stiffness of the fuel assembly channel box.

Both fuel types use the same fuel assembly channel box, and the channel box dominates the overall I

dynamic response of the incore fuel. ENC calculations show that approximately 97% of the stiffness of a BWR fuel assembly is attributable to the stiffness of the channel box.

For this reason, the dynamic structural response of the reload core is essentially that of the initial core, and the original LOCA-seismic analysis remains applicable.

Deformation of the channel to the point that control blade insertion is inhibited is not predicted to occur.

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I F-2 XN-NF-86-35 Revision 3 TABLE F-1 COMPARISON OF PHYSICAL AND GE0 METRIC CHARACTERISTICS OF ENC AND GE 8X8 ASSEMBLIES I

PROPERTY ENC FUEL GE FUEL I

Fuel rod pitch, inch 0.636 0.636 g

Number of spacers 7

7 g

Assembly weight, lb 596 600 Assembly length, inch 176.16 176.16 Fuel rod diameter, inch 0.484 0.483 Cladding thickness, inch 0.035 0.032 I

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'I G-1 XN-NF-86-35 Revision 3 I

1 APPENDIX G XCOBRA-T CONFIRMATION OF COTRANSA HOT CHANNEL RESULTS I

A potential nonconservatism was identified in the formulation of the COTRANSA Hot Channel model used to calculate delta-CPR values.

The effects of this potential nonconservatism for Grand Gulf Unit 1 Cycle 2 were assessed using I

the XCOBRA-T code. Two additional changes in input parameters, main steamline pressure drop and downcomer enthalpy transport time were also included, both of which are in a more conservative direction relative to the methodology qualified in XN-NF-79-71(P), Revision 2, as supplemented. The results of the XCOBRA-T and Hot Channel calculations with the revised parameters are given in

(

Table G-1.

For Grand Gulf the XCOBRA-T calculations confirmed the COTRANSA Hot Channel results as conservative or within 0.01.

All of the recalculated delta-CPRs support the MCPR limits of Cycle 1 which are the MCPR limits being proposed for Cycle 2.

A detailed description of the XCOBRA-T calculations is given in XN-NF-86-36, Revision 3.

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l G-2 XN-NF-86-35 Revision 3 TABLE G-1 XCOBRA-T ANALYSIS RESULTS INITIAL INITIAL REVISED (4)

POWER FLOW COTRANSA XCOBRA-T MARGIN EVENT

(%)

(%)

DELTA-CPR DELTA-CPR TO CHF LRNB(1) 104.2 108 0.08

<0.09 0.02 LRNB 104.2 73.8 0.10

<0.10 0.22 LRNB 70 40 0.16

<0.16 0.79 LRNB 40 108 0.29

<0.29 0.03 LRNB 25 73.8 1.05

<1.05 0.44 LRNB 25 40 0.80

<0.80 0.31 FWCF(2) 104.2 108 0.04

<0.04 0.06 FWCF 104.2 100 0.04

<0.04 0.05 FWCF/NB(3) 104.2 108 0.04 0.04 0.00 FWCF/RT 104.2 100 0.05

<0.05 0.02 I

(1)

LRNB - Generator load rejection without condenser bypass.

(2)

FWCF - Feedwater flow controller failure to maximum demand.

/NB - transient without condenser bypass

/RT - transiant with 100 degree temperature reduction (3) Original analyses did not include this event.

(4) Results of reanalysis at MP&L request.

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I XN-NF-86-35 Revision 3 Issue Date: 8/9/86 l

Distribution

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J. C. Chandler

'R. E. Collingham i

T. P. Currie i

R. A. Decker J. D. Floyd i

S. F. Gaines K. P. Galbraith R. G. Grummer I

R. L. Gulley T. H. Keheley S. E. Jensen T. L. Krysinski I

J. N. Morgan L. A. Nielsen A.

Reparaz I

G. L. Ritter C. J. Volmer G. N. Ward H. E. Williamson I

MP&L/JD Floyd (40)

Document Control (5)

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