ML20086K461

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Cycle 6 Reload Summary Rept
ML20086K461
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 12/31/1991
From:
ENTERGY OPERATIONS, INC.
To:
Shared Package
ML19353B429 List:
References
GNRO-91-00186, GNRO-91-186, NUDOCS 9112130126
Download: ML20086K461 (17)
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Text

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. D.

Grand Gulf Nuclear Station Unit 1 Attachment 4 to Cycle 6 Reload Sunasty Report GNRO 91/00180 Page 1 of 17 GRAND GULF NUCLEAR STATION UNIT 1 B

CYCLE 6 RELOAD

SUMMARY

REPORT DECEMBER 1991

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Grand Gulf Nuclear Station Unit 1 Attachment 4 to Cycle 6 Reload Summary Report GNRO 91/00186 Page 2 of 17 CONTENTS Page 1.0 I NTRO D U CTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.0 CYCLE 6 RELOAD SCOPE . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... 3 3.0 CYCLE S OPERATING HISTORY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4.0 CYCLE 6 CORE DESCRIPilON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5.0 FUEL MECH ANICAL DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 6.0 - THERMAL HYDRAULIC DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 6.1 MC PR S a f ety Limit s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7  ;

6.2 Flo w. depend ent MC PR . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . 7 6.3 Power dependent MCPR . . . . . . . . . . . . . . . . . . . . . ......... 7 6.4 Exposure-dependent MCPR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 6.5 C o r e S t a bilit y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 7.0 N U C LE AR D E SI G N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 7.1 Fuel Dundle Nuclear Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 7.2 Co re R e a c tivit y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 7.3 Spent Fuel Pool Criticality . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 8.0 CORE MONITORING SYSTEM ............................... 10 9.0 - - ANTICIPATED OPERATIONAL OCCURRENCES . . . . . . . . . . . . . . . . . . 10 9.1 C ore-Wid e Tra nsie nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 9.2 Loca l Tra nsients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 9.3 Reduced Flow and Power Operation ......,............... 11 9.4 ASME Overpressurization Analysis ....................... 12 10.0 PO STU LAT E D A CCI D E NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 10.1 Loss-of-Coolant Accident (LOCA) ........................ 13 10.2 Rod Drop Accident . . . ............................... 14 1 1.0 REFU ELING OPERATIO N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 14

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12.0 RE FE R E N C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 i

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Grand Gulf Nuclear Station Unit 1 Attachment 4 to Cycle 6 Reload Summary Report GNRO 91/00186 Page 3 of 17

1.0 INTRODUCTION

This report is a supplernentary document that summarizes the results of the analyses performed in support of GGNS Unit 1 Cycle 6 operation. The fresh fuel to be inserted in this cycle is an SNP 9x9 5 fuel type, it is similar to the 9x9-5 fuelinserted for Cycle 5 except for slightly increased pellet to-clad gap size, increased prepressurization, and differences in enrichment and gadolinia loadings. This fuel has been shown to be compatible with the 8x8 and 9x9-5 fuel types that were inserted during previous reloads and will be resident in the core during Cycle 6 (Reference 1).

The SNP Cycle 6 Reload An-.ysis Report (Reference 1) and the Cycle 6 Plant Transient Analysis Report (Reference 2) serve as the basic tramework for the reload analyses. Where appropriate, reference is made to these and other supporting documents for more detailed information and/or specifics of the applicable analyses. A list of references comprising both the generic and the GGNS-specific documents used in support of the Cycle 6 reload submittalis provided in Section 12.0 of this report.

2.0 CYCLE 6 RELOAD SCOPE During the fifth refueling outage at GGNS Unit 1, depleted SNP 8x8 fuel assemblies will be replaced by SNP 9x9-5 fuel assemblics. Fuel related analyses of the limiting events were performed in support of Cycle 6. This irciud4n 91yzing Cvcle 6 for anticipated transients, the Fuel Misload Error Es ent. U>C \, and ti.s Control Rod Drop Accident. These analyses were pene ened 'a support the safety and operating limits based on SNP methsVogy for both Two Loop and Single Loop Operation. Analyses for norma' operation of the reactor consisted of fuel evaluations in the areas of mechanical, thermal hydraulic, and nuclear design.

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Grand Gulf Nuclear Station Unit 1 Attachment 4 to Cycle 6 Reload Summary Report GNRO 91/00186 Page 4 of 17 Based on SNP's design and safoty analyses of the Cycle 6 reload core, the proposed changes to the GGNS Unit 1 Technical Specifications are as follows:

a. The MCPR Safety Limit values for Two Loop Operation and Single Leop Operation (SLO) are revised,
b. The MAPLHGR multiplier for Single Loop Operation is revised.
c. The flow dependent MCPR limits are revised.
d. The power dependent MCPR limits are revised.
e. The exposure-dependent MCPR limits are revised,
f. The LHGR limits for 8x8 SNP fuel types are revised for average planar exposures beyond 40,000 mwd /MTU.
g. The flow-dependent and power-dependent LHGR multipliers are revised and incorporate fuel type-specific multipliers.

3.0 CYCLE 5 OPERATING HISTORY Cycle 5 core-follow operating data available at the time of the reload design analysis, together with projected plant operation through the end of Cycle 5, was used as a basis for the Cycle 6 core design and as input to the plant safaty analyses. Cycle 5 has continued to operate as expected. No operating anornalics have occurred that would affect the licensing basis for Cycle 6.

The Cycle 6 analyses were performed assuming a nominal Cycle 5 energy of 1698 GWd.

Grand Gulf Nuclear Station Unit 1 Attachment 4 to Cycle 6 Reload Summary Repcrt GNRO 91/00186 Page S of 17 4.0 CYCLE 6 CORE DESCRIPTION I

The Cycle 6 core will consist of 800 fuel assemblics. A breakdown by bundle type / bundle average enrichment is provided in the following table:

Cvele Instited MUmber of Bundles lhDdle Tvos 6 172 SNP 9x0/2.94 w/o U235 6 100 SNr 9x9/3.38 w/o U235 5 284 ENP 9x9/3.42 w/o U235 4 4 SNP 9x9/3.25 w/o U235 4 240 SNP 8x8/3.37 w/o U235 The anticipated Cycle 6 core configuration, topather with additional bundle and core design details, is provided in Section 4.0 of the SNP Cycle 6 Reload Analysis Report (Reference 1). The Cycle 6 core is a conventional c:atter load with the lowest reactivity bundles placed in the peripheral region of the core.

The loading pattern was designed to maximize cycle energy and minimize power peaking factors. Cycle 6 is estimated to provide 1748 GWd of energy based on a Cycle 5 energy output of 1698 GWd.

5.0 FUEL MECHANICAL DESIGN The mechanical design analyses for the SNP 8x8 and 9x9 5 fuel types are described in References 4,5, and 10. The 8x8 fuel assembly design contains 62 prepressurized fuel rods end two water rods, one of which functions as a spacer capture rod. Seven spacers mairitain fuel rod spacing. The 9x9-5 fuel essembly design contains 76 prepressurized fuel rods and five water rods, one of which serves as a spacer capture rod. Seven spacers maintain fuel rod spacing. The diametral pellet to-clad gap on the 9X9 5 fuel rods is smailer on the interior high enrichment rods than on the peripheral rods in order to improve ECCS performance. The Cycle 6 reload batch uses increased prepressurization and gap sizes relative to the Cycle 5 reload batch. The 1

4 Grand Gulf Nuclear Station Unit 1 Attachment 4 to Cyclo 6 Reload Summary Report GNRO 91/00186 Page 6 of 17 additional analyses that were performed tu support these design changes are described in Reference 21.

Mechan! cal design analyses were performed to evaluate cladding steady state strain, transient stresses, f atigue damage, creep collapse, corrosion buildup, hydrogen absorption, fuel rod maximum internal pressure, differential fuel rod growth, creep bow, and grid spacer spring design. These analyses were performed tc support peak assembly discharge burnups of 40 GWd/MTU for both the 8x8 and 9x9-5'_ fuel types. As shown in References 4,5, and 10, all parameters meet their respective design limits; no fuel centerline melting.will

. occur at 120% and 135% overpower conditions for 8x8 fuel and 9x9 5 fuel

types, respectively. The effects of increased prepressurization and larger gap sizes do not impact the available design margins for the Cycle 6 reload batch (Reference 21). - The Cycle 6 core design is bounded by the assumptions used in these analyses. 1 Fuel channels of a design similar to that used for the Cycle 5 reload batch will be used for reload Cycle 6 fuel. As was the case for Cycle 5, fuel channels manufactured by Carpenter Technology Corporation (CarTech) will be used for the Cycle 6 reload batch.

--The mechanical responses of the 8X8 and 9x9 5 SNP assembly designs during seismic-LOCA events for Cycle 6 are essentially the same as for previous  !

cycles because the phyn' cal properties and bundle natural frequencies are similar. - Reference 7 presents the seismic-LOCA analysis for the 8x8 fuel and shows that the resultant loadings do not exceed the fuel design limits.

Reference 23' presents the corresponding seismic-LOCA analysis for 9x9 5 fuel. The applicability of these analyses to the 8x8 and 9x5-5 fuel assemblies

, 'in the Grand Gulf Unit 1 core has been confirmed by SNP (Reference 1).

16.0 THERMAL HYDRAULIC DESIGN XN-NF-80-19(A), Volurne 4,' Revision 1'(Reference 3) discusses the thermal-hydraulic design criteria that are used in the determination of the fe a cladding integrity safety limit and the bypass flow characteristics. SNP

' analyses were performed in accordance with XN-NF-80-19(A), Volume 3,

Grand Gulf Nuclear Station Unit 1 Attachment 4 to Cycle 6 Reload Summary Report GNRO 91/OO186 Pace 7 of 17 Revision 2 (Reference 19) to determino the patameters that demonstrate compliance with these design criteria.

6.1 MCPR Safety Limits The MCPR fuel cladding integrity safety limits are 1.06 and 1.07 for Two Loop Operation and Single Loop Operation (SLO), respectively.

The methodology and generic uncertainties used in the Cycle 6 MCPR safety limit calculation, including the effects of channel bow, are provided in Reference 8.

6.2 Flow-deoendent MCPR The flow-dependent MCPR limits (MCPR,) are revised for Cycle 6. The MCPR, limits are defined for only the Loop Manual mode of operation.

The MCPR, limits are lower than the limits applicable to previous cycles as a result of a lower MCPR safety limit and smaller delta-CPRs for the slow flow runout event for the predominantly 9x9 5 Cycle 6 core. The MCPR,limite are defined over the same range of flows as for Cycle 5.

6.3 Power-deoendent MCPB The power-dependent MCPR limits (MCPRp ) are revised for Cycle 6.

The most limiting events were used as the basis to confirm or establish the acceptability of the MCPRp limits for both Two Loop Operation and SLO during Cycle 6.

6.4 Excosure-deoen@pt MCPR The exposure-Dependent MCPR limits (MCPR,) are revised for Cycle 6.

The MCPR, limits were calculated for two exposure ranges instead of the three ranges used for Cycle 5. The limit is unchanged from the Cycle 5 value for the early part of the cycle and is lower than the Cycle 5 values for the latter part of the cycle.

The most limiting core-wide transients and local events were analyzed b to confirm the acceptability of the MCPR, limits for use in Cycle 6.

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Grand Gulf Nuclear Station Unit 1 Attachment 4 to Cyclo 6 Reload Summary Report GNRO 91/00186 Page 8 of 17 These limits were established consistent with the Cycle 6 operating strategy.

6.5 Core Stab:lity The GGNS 'Jnit 1 Technical Specifications implement the BWROG/GE Interim Recommendations for Stability Actions (IRSA). The IRSA boundaries, which were developed based on GE fuel experience, have been approved for application at GGNS Unit 1 containing SNP Sx8 fuel (Reference 25) and for application ii r Cycle 5 containing the first batch of SNP 9x9 5 fuel (Reference 17).

The Cycle 6 core will contain the second SNP 9x9-5 reload batch.

Confirmatory analyses for ^/cle 6 core stability consisted of a comparative evaluation of the stability characteristics for the Cycle 5 and Cycle 6 cores. In addition, a full 9x9 5 fueled core was analyzed.

The results showed that the core decay ratios for the cycles analyzed are equivalent; the differences in stability performance are comparable to the variations observed for previous cycles. Therefore, the current GGNS-1 stability-related technical specifications are applicable for Cycle 6 operation as well as for a full 9x9-5 fueled core, provided that the core design and cycle operating strategy are not changed significantly. _

7.0 NUCLEAR DESIGN The neutronic methods used for the design and analysis of SNP reloads are described in SNP topical reports (References 9 and 24).

7.1 Euel Bundle Nucioar Desian The Cycle 6 reload fuel utilizes SNP 9x9-5 fuel assemblies. Two basic bundle designs are used with different axially distributed U235 and burnable poison concentrations. For both designs, the top 12 inches and bottom 6 inches of each fuel rod contain natural uranium and the central 132 inch zone of each rod contains enriched uranium at one of eight different enrichments. The central zone of the 2.94 weight I

- Grand Gull Nuclear Station Unit 1 Attachment 4 to  ;

Cycle 6 Reload Summary Report GNRO 91/00180  :

_Page 9 of 17 i

percent (w/o) bundle (identified in Section 4.0) has an average enrichment of 3.25 w/o U235, whereas the 3.38 w/o bundle has a '

central region with an average enrichment of 3.75 w/o U235. The neutronic design parameters and rod enrichmerit distributior are described in Section 4.0 of the Cycle G Reload Analysis Report (Reference 1).

7.2 Core Reactivity The beginning of Cycle 6 (BOC6) cold core K,n with the strongest worth control rod fully withdrawn at cold,68 degrees F reactor

-conditions was calculated to be 0.98914. This corresponds to a ,

- shutdown margin (delta k/k) of 1.10%. BOC + 500 mwd /MTU and BOC+7500 mwd /MTU were determined to be the most limiting conditions with a minimum shutdown margin of 1.03%. Therefore, the difference between the minimum shutdown margin in the cycle and the BOC shutdown margin, R, is 0.07%. The calculated shutdown margin

-is well in excess of the 0.38% delta k/k Technical Specification requirement (Section 3/4.1.1), and will be verified by testing at BOC6 to be greater than or equal to R + 0.38% delta k/k.

- The Standby Liquid Control (SLC) system is designed to inject a quantity of boron that produces a concentration of no less than 660-ppm in the reactor core. Analyses were performed to show that the minimum shutdown margin is at least 3.0% delta k/k with the reactor in a cold, xenon-free state, at the most limiting cycle exposure, and with .

all control rods in their critical full power positions. This assures that the reactor can be brought from full power to a cold, xenon-free shutdown, assuming that none of the withdrawn control rods can be inserted, and confirms the basis of the Technical Specification-requirement for the Cycle 6 reload core.

7.3 Soent Fuel Pool Criticality The most reactive segment of the Cycle 6 fuel at its most reactive point- ,

in life is less reactive than analyzed in Reference 22. Therefore, the Reference 22 analysis is boundi" for the Cycle 6 fuel.

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Grand Gulf Nucle" Station Unit 1 Attachment 4 to Cycle 6 Reload Gummary Report GNRO 91/00188 Page 10 of 17 8.0 CORE MONITORING SYSTEM The POWERP; EX core monitoring system is and will continue to be utilized to monitor reactor parameters at GGNS. The core ac.t ;oring system is fully consiuunt with SNP's nuclear analysis methodobgy as descr: bed in Referewes 9 and 24. In addition, the measured power distribution uncertainties are incorporated into the calculation of the MCPR Safety Limit as described in SNP's Nuclear Critical Power Methodalogy Report (Reference 8).

t 9.0 ANTICIPATED OPERATIONAL OCCURRENCES in ordc. to support the Cych 6 operating limits, eight cate0eries of system transients are considered, as described in SNP's Plant Transient Methodology Report (Reference 11). SNP has provided plant specific analysis results for the following system transients to determine the thermal margin requirements for operation during Cycle 6 (Reference 2):

1) Generator Load Rejection without Bypass (LRNC)
2) Feedwater Controller Failure JFWCF)
3) Loss of Feeseiter Heating (LFWH)
4) Flow Excursion Analyses performed for previour em . ave shown that the other system transients are non-limiting and, mew , are bounded by one of the above. In addition, the Fuel Loading Error was analyzed in accordance with the methodology described in Reference 9. The Control Rod Withdrawal Error (CRWE) transient has been analyzed generically in Reference 18. Single Loop Operation is addressed in the Cycle 6 Transient Analysis Report (Refe ence 2).

9.1 Core-Wide Tranrients The plant transient codes that were used to evaluate the LRNB and FWCF events are SNP's COTRANSA2 (Reference 26) and XCOBRA-T (Referance 20), which incorporate a one dimensionai neutronics mode!

to account for shifts in axial power sh6pe and control rod effectiveness.

Technical Sper.fication scram times (Section 3/4.1.3) were used in the bounding analys!s. The results of the LRN8 and FWCF analyses are 1

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Grand Gulf Nuclear Sta: ion Unit 1 Attachment 4 to Cycle 6 Roload Summary Report ' GNRO 91/00186 Page 11 of 17 provided in the Cycle 6 Plant Transient Analysis Report (Reference 2) and a summary of results is provided in the Cycle 6 Reload Analysis Report (Reference 1). The LFWH event was analyzed by performing quasi-steady state analysis using the MICROBURN-B neutronics code (Reference 24). The LFWH event was analyzed consistent with the MEOD power / flow operating map for actual GGNS operating conditions during Cycles 1 through 5 and for various conditions anticipated during Cycle 6. A sumi.iary of this analysis is provided in Reference 2.

9.2 Local Transient!

The Centrol Rod Withdrawal Error (CRWE) transient has been analyzed generically in Reference 18. The generic analysis provides a statistical evaluation of the consequences of the CRWE transient for BWR/6 plant

~ figurations under conditions that cover the normal operating fer/ flow map, the extended load line region, and the increased core i;o 1 region. This analysis was reevaluated using the ANFB Critical W. er Correlation (Reference 27) and the MICROBURN-B neutronics Oct - (Reference 24). The evaluation demonstrated the continued

.. ,plicability of the generic CRWE analysis results.

9.3 fleduced Flow and Power Ooeration The off sted therma! limits which were established for Cycle 1 MEOD opera +' (Reference 6), were revised appropriately for Cycle 6 operr The power-dependent MCPR operating limits, which are bs . ie results of the Cycle 6 transient analyses and the CRWE ger. m .ialysis, we'e revised for Cycle 6. The flow-dependent MCPR-limits are revised for Cycle 6 based on SNP's Cycle 6 analysis results.

Flow rates used in the analysis are defined in Reference 12. The MCPR limits ensure that potential clad damage resulting from transition boiling is avoided.

Flow-dependent and power-dependent LHGR multipliers that are fuel type-specific were deterrnined for Cycle 6. The flow and power ranges are unchanged from Cycle 5. For the power-dependent LHGR multipliers, bounding limit curves applicable to all core flows were

p R

' Grand Gulf fJuclear Station Unit 1 Attachment 4 to Cycle 6 Reload Summary Report GNRO 91/00186 i

- Page 12 of 17 established using.SNP methodology. The LHGR limits ensure that the fuel mechanical design criteria are satisfied.

Flow-dependent MCPR limits and LHGR multipliers are determined for-only the Loop Manual mode of operation because chanjes in plant configuration have been made to ensure that operation in the Non-Loop Manual mode is not possible (Reference 28).

9.4 ASME Overoressurization Analvsis in order to demonstrate compliance with the ASME Code ,

overpressurization criterion of 110% of vessel design pressure, comparative evaluation of the peak vessel pressures calculated for ,

previous cycles was performed for the limiting event, using an c equivalent set of assumptions. 'The limiting event is the MSIV closure with failure of the MSIV position switch scram. Seven out of twenty safety / relief valves are assumed to be out of service. A conservative 6% tolerance is used for the safety valve setpoints. The results show that the maximum vessel pressure varies over a narrow range and is not

- sensitive to fuel and core design variations; sufficient margin is

-available to the transient pressure limit of 1375 psig (Reference 2). -t 10.0 POSTULATED ACCIDENTS i

in support of Grand Gulf operation, SNP has analyzed the Loss-of-Coolant Accident (LOCA) for Two Loop Operation and for SLO to demonstrate that-MAPLHGR limits for_ Cycle _6 reload fuel coniply with 10CFR50;46 criteria.

Methocology for the LOCA analysis is provided in References 13 through 15.

The Rod Drop Accident (RDA) was analyzed for the Cycle 6 core to demonstrate compliance with the 280 cal /0m Design Limit. Methodology for

- the RDA analysis is described in XN-NF-80-19(A), Volume 1 (Reference 9).

An SNP evaluation shows that the GE analysis of ATWS over-pressurization is

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applicable to SNP fuel and therefore remains valid for Cycle 6.

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Grand Gulf Nuclear Station Unit 1 Attachment 4 to Cycle 6 Reload Summary Report GNRO 91/00186 Page 13 of 17 10.1 Loss-of-Coolant Accident (local The generic BWR/6 LOCA break spectrum analysis (Reference 16) and the LOCA analysis performed in support of the Cycle 2 submittal (Reference 31) remain applicable for Cycle 6. A cycle-specific heatup analysis was performed for Cycle 6. The analysis confirms that the Peak Cladding Temperatures (PCTs) (1691 degrees F for 8X8 fuel and 1713 degrees F for 9X9-5 fuel) remain well below the 10CFR50.46 PCT limit of 2200 degrees F.

A detailed LOCA analysis was performed for Single Loop Operation (SLO) to determine an appropriate multiplier to be applied to the Two Loop Operation MAPLHGRs for the 8X8 and 9X9-5 fuel types (Reference 29). The multiplier wa.s shown to be independent of the flow conditions in the idle loop (Reference 30). The multiplier was selected to ensure that the PCTs resulting from a LOCA during SLO are bounded by the corresponding PCTs for Two Loop Operation. The SLO LOCA analysis determines the highest PCTs over the range of exposures for the 8X8 and 9X9-5 fuel types. The ca'culated PCTs for the SLO LOCA (1631 degrees F for 8X8 fuel and 1609 degrees F for 9X9-5 fuel) are approximately 100 degrees F lower than the corresponding values for the Two Loop Operation LOCA (1738 degrees F for 8X8 fuel and 1713 degrees F for 9X9-5 fuel). The PCT calculated by the Two Loop Operation LOCA analysis for 8X8 fuel (1738 degrees F) is based on the higher (more conservative) MAPLHGR value used in the Reference 31 analysis; this PCT is higher than the Cycle 6 heatup analysis value (1691 degrees F), which is based on the Cycle 6 MAPLHGR limit for 8X8 fuel (Reference 1).

Confirmatory analyses were performed to show that the local Zr-H2O reaction remains below 17% and that the core-wide metal-water reaction (CMWR) remains below 1 % for the limiting LOCA event as required by 10CFR50.46. The *esults of these analyses are presented in Section 6.1 of Reference 1. As stated in the GGNS-1 UFSAR, the hydrogen recombiners have been sized to process the hydrogen released from 0.8% CMWR. Consistent with the Regulatory Guide 1.7 requirements for post-LOCA combustible gas control, the capability of l

i Grand Gulf Nuclear Station Unit 1 Attachtnant 4 to Cycle 6 Reload Summary Report GNRO 91/00186 Page 14 of 17 the hydrogen recombiners to maintain post-LOCA combustible gas concentration below 4 volume percent has been confirmed.

10.2 Bod Droo Accident SNP's methodology for analyzing the Rod Drop Accident (RDA) utilizes a generic parametric analysis that calculates the fuel enthalpy rise during the postulated RDA over a wide range of reactor operating conditions. For Cycle 6, Section 6.2 of Reference 1 shows a value of 166 cal /gm for the maximum deposited fuel rod enthalpy during the worst case postulated RDA. This value is well below the design limit of 280 cal /gm.

11.0 REFUELING OPERATIONS As was done for Cycle 5, refueling operations will be addressed by a 10CFR50.59 Safety Evaluation.

12.0 REFERENCES

1) EMF-91-169, " Grand Gulf Unit 1 Cycle 6 Reload Analysis," Siemens Nuclear Power Ccrporation, October 1991.
2) EMF-91-168, " Grand Gctf Unit 1 Cycle 6 Plant Transient Analysis,"

Siemens Nuclear Power Corporation, October 1991.

3) XN-NF-80-19(P)(A), Volume 4, Revision 1, " Exxon Nuclear Methcdology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Co., June 1986.
4) XN-NF-85-67(P)(A), Revision 1, " Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Exxon Nuclear Co.,

September 1986.

5) ANF-88-152(P), Amendment 1, " Generic Mechanical Design for Advanced Nuclear Fuels 9x9-5 BWR Reload Fuel," Advanced Nuclear Fuels Corporation, September 1989.

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Grand Gulf Nuclear Station Unit 1 Attachment 4 to Cycle 6 Reload Summary Report GNRO 91/00186 Page 15 of 17

6) "GGNS Maximum Extended Operating Domain Analysis," General Electric Company, March 1986.
7) XN NF-81-51(A), "LOCA-Seismic Structural Response of an ENC BWR Jet Pump Fuel Assembly," Exxon Nuclear Co., May 1986.
8) ANF-524(P)(A), Revision 2, " Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors," including Supplements, Advanced Nuclear Fuels Corporation, April 1989.
9) XN-NF-80-19(A), Vame 1, Supplements 1 & 2, " Exxon Nuclear Methodology for Boiling Water Reactors: Neutronics Methods for Design and Analysis," Exxon Nuclear Co., March 1983.
10) ANF-88-183(P), " Grand Gulf Unit 1 Reload XN-1.3 Cycle 4 Mechanical Design," ANF Corporation, Supplement 1, August 1991.
11) XN-NF-79 71(P), Revision 2, " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," Exxon Nuclear Co.,

November 1981.

12) NESDQ-88-003, Revision 0, " Grand Gulf Nuclear Station Unit 1 Revised Flow Dependent Thermal Limits," MSU System Services Inc.,

November 1988.

13) XN-NF-80-19(A), Volumes 2, 2A, 28, & 2C, " Exxon Nuclear Methodology for Boiling Water Reactors: EXEM BWR ECCS Evaluation Model," Exxon Nuclear Co., September 1982.
14) XN-NF-CC-33(A), Revision 1, "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option," Exxon Nuclear Co.,

November 1975. ,

15) XN-NF-82-07(A), Revision 1, " Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Co., November 1982.
16) XN-NF-86-37(P), " Generic LOCA Break Spectrum Analysis for BWR/6 Plants," Exxon Nuclear Co., May 1986.
17) " Issuance of Amendment No.73 to Facility Operating License No.

NPF-29, Grand Gulf Nuclear Station Unit 1, Regarding Fuel Cycle 5 Reload (TAC No. 76992), Letter from L. L. Kintner, USNRC, to W. T.

Cottle, Entergy Operations Inc., dated November 15,1990.

Grand Gulf Nuclear Station Unit 1 Attachreient 4 to Cycle 6 Reload Summary Report GNRO-91/0018S  ;

Page 16 of 17

18) XN-NF-825(P)(A), Supplement 2, "BWR/6 Generic Rod Withdrawal Error l

for All Plant Operations Within the Extended Operating j Analysis, MCPR, Nuclear Company, October 1986.

Domain," Exxon l

19) XN-NF-80-19(P)(A), Volume 3, Revision 2, " Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description," Exxen Nuclear Co., January 1987.
20) XN-NF-84-105(P)(A), Volume 1, "XCOBRA-T: A Computer Code for BWR Transient Thermal Hydraulic Core Analysis," Exxon Nuclear Company, Inc., February 1987.
21) " Reload ANF-1.5 Principal Reload Fuel Design Parameters Document,"

MPEX-90/111, Letter from N. L. Garner, ANF Corporation, to J. B. Lee, Entergy Operations Inc., November 28,1990.

22) ANF-90-060, " Criticality Safety Analysis for the Grand Gulf Spent Fuel Storage Racks with ANF 1.4 Fuel Assemblies," April 1990.
23) XN-NF-84-97(P)(A), "LOCA-Seismic Structural Response of an ENC 9x9 BWR Jet Pump Fuel Assembly," Exxon Nuclear Company Inc.,

August 1986.

24) XN-NF-80-19(P), Volume 1, Supplement 3, "ANF Methodology for BWRs: Benchmark Results for the CASMO 3G/MICROBURN-B Calculation Methr>dology," Advanced Nuclear Fuels Corporation, February 1989, as supplemented by ANF letter RAC:083:90 dated July 20,1990.

l

25) " Issuance of Amendment No. 62 to Facility Operating License No. NPF Grand Gulf Nuclear Station, Unit 1, Regarding Technical Specifications Revisions - Thermal-Hydraulic Stability t

(TAC No. 71808)," Letter from L. L. Kintner, NRC, to W. T. Cottle, l SERI, dated August 31,1989.

l

26) ANF-913, Volume 1, Supplements 1,2, and 3, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analysis,"

l Advanced Nuclear Fuels Corporation, June 1989.

1

27) ANF-1125(P), Supplement 1, "ANFB Critical Power Correlation,"

Advanced Nuclear Fuels Corporation, April 1989.

l

l Grand Gulf Noclear Station Unit 1 Attachment 4 to Cycle 6 Reload Summary Report GNRO-91/00186 Page 17 of 17

28) " Removal of References to Non-Loop Manual Mode of Reactor Recirculation System Flow Control, Proposed Amendment to the Operating License (PCOL-91/13)," GNRO 91/00071, 'N. T. Cottle, to USNRC, May 30,1991.
29) EMF-91-172, " Grand Gulf Unit 1 LOCA Analysis for Single Loop Operation," Siemer's Nuclear Power Corporation, October 1991.
30) GEXI-01391, " Allowed Valve Conditions in the Idle Loop for Grand Gulf Unit 1 Single Loop Operation," Letter, S. L. Leonard, SNP Corporation, to J. B. Lee, Entergy Operations Inc., November 7,1991.
31) XN-NF-86-38, " Grand Gulf Unit 1 LOCA Analysis," Exxon Nuclear Co.,

June 1986.

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