ML20235U789

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Rev 0 to Grand Gulf Nuclear Station Unit 1 Cycle 3 Reload Summary Rept
ML20235U789
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 08/31/1987
From:
SYSTEM ENERGY RESOURCES, INC.
To:
Shared Package
ML20235U774 List:
References
NUDOCS 8710140261
Download: ML20235U789 (27)
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Text

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l GRAND GULF NUCLEAR STATION

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UNIT 1 CYCLE 3 RELOAD

SUMMARY

REPORT s

B'710140261 871oc,9 August 1987 PDR ADOCK 05000416 Revision 0 J10MILC860514

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I CONTENTS Page 1.0 7. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.0 GENERAL DES'CRIPTION OF RELOAD SCOPE. . . . . . . . . . . . . . . . 2 3.0- GGNS UNIT 1 CYCLE 2 OPERATING HISTORY. . . . . . . . . . . . . . . 3 4.0 RELOAD CORE DESCRIPTION. . . . . . . . . . . . . . . . . . . . . . 4 5.0 FUEL MECHANICAL DESIGN . . . . . . . . . . . . . . . . . . . . . . 5 i

6.0 THERMAL HYDRAULIC DESIGN . . . . . . . . . . . . . . . . . ... . . 7 6.1 Hydraulic Compatibili ty . . . . . . . . . . . . . . . . . . . 7 6.2 Safety Limit MCPR . . . . . . . . . . . . . . . . . . . . . . 7 6.3 Core Bypa s s ' Fl ow. . . . . . . . . . . . . . . . . . . . . . . 8 6.4 Core Stabilit; . ....................... 8 7.0 NUCLEAR DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . 8 7.1 Fuel Bundle Nuclear Design. . . . . . . . . . . . . . . . . . 9 7.2 Core Reactivity . . . . . . . . . . . . . . . . . . . . . . 10 7.3 Spent Fuel Pool Criticality . . . . . . . . . . . . . . . . 11 7.3.1 Spent Fuel Pool . . . . . . . . . . . . . . . . . 11 8.0 CORE MONITORING SYSTEM . . . . . . . . . . . . . . . . . . . . . 12 9.0 ANTICIPATED OPERATIONAL OCCURRENCES. . . . . . . . . . . . . . . 12 9.1 Core-Wide Transients. . . . . . . . . . . . . . . . . . . . 14 9.2 Local Transients. . . . . . . . . . . . . . . . . . . . . . 14 9.3 Reduced Flow and Power Operation. . . . . . . . . . . . . . 14 9.4 ASME Overpressurization Analysis. . . . . . . . . . . . . . 16 i

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/6 CONTENTS (Cont'd) l Page 10.0 POSTULATED ACCIDENTS . . . . . . . . . . . . . . . . . . . . . . 17 10.1 Loss-of-Coolant Accident. . . . . . . . . . . . . . . . . . 17 10.2 Rod Drop Accident . . . . . . . . . . . . . . . . . . . . . 18 11.0 REFUELING OPERATIONS . . . . . . . . . . . . . . . . . . . . . . 18 REFERENCES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 l

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1.0 INTRODUCTION

Grand Gulf Nuclear Station (GGNS) Unit 1 Cycle 3 will ' include the second i reload of ANF 8x8 fuel. This report is a supplementary document which provides a general scote and summarizes the results of the reload analyses performed by Advanced Nuclear Fuels (ANF) Corporation, previously known as Exxon Nuclear. Company (ENC), in support of GGNS Unit 1 Cycle 3 operation.

Also addressed is a description of the ANF Cycle 3 reload core design and GE initial core and ANF reload core fuel bundle compatibility.

The ANF Cycle 3 Reload Analysis Report (Reference 1) and the Cycle 3 Plant Transient Analysis Report (Reference 2) serve as the basic framework for the reload analysis. When appropriate, reference is made to these and other supporting documents for more detailed information and/or specifics of the applicable analyses. The ANF Reload Analysis Report is intended to be used in conjunction with ENC topical report XN-NF-80-19(A), Vol. 4 Revision 1 " Exxon Nuclear Methodology for Boiling Water Reactors:

Application of the ENC Methodology to BWR Reloads" (Reference 4), which describes the analyses performed in support of the reload and identifies the methodology used for those analyses. A list of references is provided containing the GGNS specific reload documents and the applicable generic .,

reload documents prepared by ANF (generic methodology previously approved or currently under review) which are being used in support of the Cycle 3 reload submittal.

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t 2.0 GENERAL DESCRIPTION OF RELOAD SCOPE During the second refueling outage at GGNS Unit 1, System Energy Resources, Inc. (SERI) will be replacing approximately one half of the remaining GE initial core fuel assemblies with ANF-8x8 fuel assemblies. The ANF fuel is very similar in design to the GE fuel and analyses have been performed in support of Cycle 2 to account for the slight differences in the mechanical, thermal-hydraulic, and nuclear design of the bundles, and the use of different analysis methodologies. Limiting cases of fuel related analyses previously performed for Cycle 2 were repeated for Cycle 3. This included analyzing Cycle 3 for anticipated operational occurrences to 1

confirm current operating limits, performing LOCA confirmatory analyses for the second reload fuel batch for compliance with 10CFR50.46, and analyzing for the rapid drop of a high worth control rod to assure that excessive energy will not be deposited in the fuel. Analyses for normal operation of the reactor consisted of fuel evaluations in the areas of ,

mechanical, thermal-hydraulic, and nuclear design. These analyses include the expanded power-flow map regions (extended load line and increased core flow) for Cycle 3 operation.

Based on ANF's design and safety analyses of the Cycle 3 reload core, the only change to the GGNS Unit 1 Technical Specifications is the addition of two MAPLHGR curves for the two new reload batch designs. I i

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3.0 GGNS UNIT 1 CYCLE 2 OPERATING HISTORY 1

Actual core-follow operating data at the time of the reload design analysis was used, together with projected plant operation, as a basis for the Cycle 3 core design and as input to the plant safety analyses. Cycle 2 has continued to operate as expected and no operating anomalies have occurred which would affect the licensing basis of the reload core or Cycle 3 performance.

The current end-of-cycle 2 (E0C 2) licensing exposure window ranges from .

, 1 6436 MWD /MTU to 8024 MWD /MTU with a nominal exposure of 7230 MWD /MTV.  ;

This window provides an allowable E0C 2 core average exposure range for which the Cycle 3 plant safety analyses are valid.

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2.0 b l

l 4.0 RELOAD CORE DESCRIPTION The Cycle 3 core will consist of 800 fuel assemblies, which includes 288 fresh ANF-8x8 assemblies (second reload), 264 once burned ANF-8x8 assemblies (first reload), and 248 twice burned GE6 8x8 assemblies. A-breakdown by bundle. type / bundle average enrichment is provided in the following table:

Number of Bundles Bundle Type 204 ANF 8x8/3.01 w/o U235 w/6 rods 4.0 w/o Gd23 0

84 ANF 8x8/3.01 w/o U235 i w/8 rods 4.0 w/o Gd23 0

264 ANF 8x8/2.81 w/o U235 248 GE6 8CR210/2.00 w/o U235 Of the 288 twice burned GE6 8x8 fuel assemblies being discharged at E0C 2, 80'are medium enriched (1.54 w/o U235) and 208 are high enriched (2.00 w/o U235) bundles.

The anticipated Cycle 3 core configuration along with additional core design details is provided in section 4.0 of the ANF Cycle 3 Reload i

AnalysisReport(Reference 1). The reload core is a conventional scatter load with the lowest reactivity bundles placed in the periphery region of the core. The loading pattern was designed to maximize the operating cycle length and minimize power peaking factors. Cycle 3 is estimated to provide 1,420 GWD of energy based on a Cycle 2 energy output of 1,047 GWD.

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21 5.0 FUEL MECHANICAL DESIGN l' c

The mechanical design analyses for the ANF-8x8 fuel (first and second reloads) are described in XN-NF-85-67(A), Revision 1 (Reference 5). The  !

reload fuel assembly design uses 62 fuel rods and two water rods, one of which functions as a spacer capture rod. Seven spacers maintain fuel rod spacing. The fuel rods are pre-pressurized, contain UO2 pellets and use a diametral pellet-to-clad gap which is smaller on the interior high enrichment rods than on the remaining rods in the bundle to improve ECCS margin.

Mechanical design analyses were performed to evaluate cladding steady- l l

state strain, transient stresses, fatigue 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 to support a batch average burnup of 30,000 MWD /MTU, which exceeds the batch average burnup expected for Cycle 3. All parameters meet their respective design limits as shown in Reference 5.

This reference presents the fuel thermal analysis that shows no fuel centerline melting at 120% overpower conditions for all exposures within the design end-of-life exposure.

For the initial cycle, GE provided an LHGR design limit to assure opera-tion within the fuel mechanical design analysis, which was incorporated into the Technical Specifications (TS) as an operating limit. MAPFACpand j

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MAPFAC were introduced in ME0D Cycle 1 and were updated for Cycle 2.

f These MAPFAC and MAPFAC limit factors are retained from Cycle 2 and are p f discussed in Section 9.0 below.

For Reload 2 fuel the design is such that margin to fuel mechanical design limits (e.g., centerline melt, transient stress, etc.) is assured for overpower conditions throughout the life of the fuel as demonstrated by

,e fuel design analyses (Reference 5). As described in Reference 1, the MAPLHGR operating limit has been defined for each fuel design (as required) to ensure conformance with the LHGR mechanical design limit.

For all expect'ed Cycle 3 operations, conformance to the MCPR, MAPLHGR and LHGR operating limits ensures that the power distribution for ANF fuel remains within the assumptions of the fuel design analyses.

The mechanical response of the ANF assembly design during seismic-LOCA events is essentially the same as the response of a GE assembly since the physical properties and bundle natural frequencies are similar. Reference 6 presents the seismic-LOCA analysis for the GE fuel which shows that resultant loadings do not exceed the fuel design limits. Reference 7 presents the seismic-LOCA analysis for ANF fuel in a similar application which showed large design margins for all assembly components. Therefore, based on the similarity between the fuel types and the large margin calculated for ANF fuel in a similar application, the loadings for GGNS Unit 1 do not exceed design limits for ANF fuel assembly components.

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IE L6 0: THERMAL HYDRAULIC' DESIGN-t XN-NF-80-19(A), Volume 4,~ Revision 1_(Reference 4).presentstheprimary.

thermal hydrauli.c design criteria which require analyses;to ' determine:

L (1)j hydraulic compatibility of the ANF and GE fuel bundles,-(2) the,1 fuel' cladding integrity safety' limit, and'(3): bypass flow characteristics.

1 ANF analyses were performed in accordance with XN-NF-80-19(A), ' Volume 3 lRev_ision 2;(Reference 19)-to demonstrate compliance with these design cri teri a'. The analyses performed to determine each of these parameters-are discussed.in this section.

6.1 Hydraulic Compatibility i

Component hydraulic resistances for representative ANF. and GE 8x8 fuel' types have been determined in' single phase flow tests.of full scale assemblies. XN-NF-80-19(A), Volume 4, Revision 1(Reference 4) summarizes the resistances and evaluates.the effects on thermal margin.due to the coresidence of the ANF and GE fuel' bundles. The close similarity between the two representative fuel designs

performance characteristics indicate that ANF and GE 8x8 fuel are sufficiently compatible for coresidence in GGNS Unit 1.

6.2 Safety Limit MCPR The MCPR fuel-cladding integrity safety limit remains 1.06. The methodology and generic uncertainties used in the Cycle 3 MCPR safety J10 MISC 860514 - _ _ - _ _ - _ - _ - _ - _

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.! q g limit calculation are provided in Reference '8. - The GGNS ' Unit 1.

l . specific inputs and-_MCPR safety. limit calculation are provided in' l L Referenhe2.-

~ 6.3 --Core Bypass Flow

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Core bypass flow is calculated using the methodology l described in.

.XN-NF-80-19(A), Volume 3, Revision 2 (Referen'ce 19). The core bypa'ss -

flow fraction, excluding' water rods, for Cycle 3'is 10.6%'(Reference

1) of total core flow which is equal to the Cycle 2 value.

6.41 Core Stability' 4

GGNS Unit 1 Technical--Specifications implements surveillance for

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detecting and suppressing power oscillations. Confirmatory analysis for, Cycle 3(Reference-1)showsthattheCycle2 analysis'resultsare

- m bounding. _Therefore.,GGNS Unit I complies with General-Design

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Criteria 12.

7.0 NUCLEAR DESIGN The neutronic methods used for the design and analysis of ANF follow-on reloads are described in the ANF topical report XN-NF-80-19(A), Vol. I and

. Supplements 1 and 2 (Reference 9). These methods have been reviewed and approved by the Nuclear Regulatory Commission for generic application to ,

i BWR reloads. 1

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7.1 Fuel Bundle' Nuclear Design-

.The second' reload. fuel bundle design is an 8x8: lattice'with two inert

-(water) rods and 62 fuel rods containing 150 inches of_ active fuel.:

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.-The top and bottom six inches of each fuel rod contain natural uranium and'.the central 138 inches (enriched zone) of each rod-contain-enriched uranium ~at one of five different enrichments. The

' fuel bundle burnable poison design- comp' rises .six and eight gadolinia-n-

bearing rods containing 4.0 w/o Gd 230 . These' rods are utilized to reduce the initial: reactivity of.the bundle.

The average enrichment of'the bundle enriched zone is 3.21 w/o U235 and the bundle-average enrichment (including the top and bottoms natural' uranium blankets) is 3.01'w/o U235.' The number of fuel rods at each, enrichment is given below:-

Number of Rods . Enrichment (w/o U235) of Enriched' Zone 1 1.50 5 2.00 16 2.65 (6 or 8 containing'4.0 w/o Gd23 0) 20 3.20 20 '4.05 i The neutronic design parameters and rod enrichment distribution are

.' described in section 4.0 of the Cycle 3 Reload Analysis Report L

(Reference 1).

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7.2 Core Reactivity

The beginning of cycle-3 (B0C-3)' cold core Keff value with all-rods-out was calculated.to be 1.12535. Based on the minimal Cycle 2 length ~of 6,436 MWD /MTV, a minimum Shutdown Margin of 1.16% delta k/k, with the strongest worth control rod-fully withdrawn at cold

-(68 degrees F) reactor conditions, was determined to occur at a Cycle 3 exposure of.7,500 MWD /MTV. The B0C 3 Shutdown Margin was calculated to be 1.18% delta k/k. Therefore, the' difference between the minimum Shutdown Margin in'the cycle and the B0C Shutdown Margin, R, is 0.02%

delta k/k. The calculated Shutdown Margin'is well in excess of the 0.38% delta k/k Technical Specification requirement, and will be i

verified by test at B0C 3 to be greater thanior equal to R + 0.38%

delta k/k.

l The Standby Liquid Control System (SLC System), which is_ designed to inject a quantity of boron that produces a concentration of no less than 660 ppm in the reactor core within approximately 90 to 120 minutes after initiation, was calculated to provide a minimum shutdown l margin of 3.36% delta k/h with the reactor in a cold, xenon free state, all control rods in their critical full power positions, and

.the reactor at the most limiting cycle exposure. These are'the most  !

conservative analytical assumptions and are therefore applicable to l

all plant operating conditions, including the ME0D. 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 j inserted; and thus for the Cycle 3 reload core, confirms the basis of 1

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, ;v r Tschnical' Specification requirement. No credit is taken.for'the SLC modifications to be implemented in Cycle 3.

'7.3L Spent Fuel Pool Criticality b 7 . 3.1 - -Spent Fuel Pool e

.The neutronics analysis for the spent fuel pool was performed by Joseph Oat Corporation (Reference 10). LTheL

. basis of the analysis' assumed the spent fuel pool was loaded with an infinite array of fresh 8x8 fuel assemblies i at a uniform average enrichment of 3.5 w/o U235 containing no burnable poison. The absence of burnable poisons ensures that peak assembly reactivity occurs at~BOL.

With these assumptions, it was calculationally demonstrated' that the spent fuel pool Keff.would always remain below 0.95 as stipulated in the FSAR., Thus, 8x8 fuel assemblies

.can be safely stored as long as the average enrichment of the maximum enriched zone of the assembly is not higher than 3.5 w/o U235. The average enrichment of the Reload 2 fuel assembly enriched zone is 3.21 w/o U235 as described in section 7.1. This enrichment is lower than the 3.5 w/o criterion and thus it is concluded that adequate margin exists to prevent spent fuel pool criticality throughout the Reload 2 fuel assembly lifetime.

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8.0 -CORE MONITORING SYSTEM The POWERPLEX core monitoring system is and will continue to be utilized

-to monitor reactor parameters at GGNS. The POWERPLEX' core monitoring T system incorporates ENC's core simulation methodology and is used for both

-online core monitoring as well as an off-line predictive and backup tool.

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The system is' fully consistent with ANF's n' uclear analysis methodology as;

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described in XN-NF 80-19(A) Volume 1 Supplements 1 & 2 (Reference 9). In.

addition, the measured power distribution uncertainties are incorporated into the calculation of the MCPR Safety Limit as described-in ANF's Nuclear Critical Power Methodology Report XN-NF-524(A) (Reference 8).

9.0 ' ANTICIPATED OPERATIONAL OCCURRENCES' In order to confirm that the Cycle 2 operating limits are applicable to the Cycle 3 fuel, eight categories-of system transients are considered;as described in ANF's Plant Transient Methodology Report XN-NF-79-71(P)

(Reference 11). ANF has provided plant specific analysis results for the following three system transients to determine the operating thermal margin requirements for Cycle 3:

1) Generator Load Rejection without Bypass (LRNB)
2) Feedwater Controller Failure (FWCF)
3) Loss of Feedwater Heating (LFWH) i o As shown in XN-NF-79-71(P) (Reference 11), the other system transients are either non-limiting or bounded by one of the above. In addition, the J10 MISC 860514 - 12 l 1

29 Fuel Loading Error was analyzed in accordance with the methodology f- described in XN-NF-80-19(A) Vol. 1 (Reference 9).

The Loss of Feedwater Heating (LFWH) event was analyzed within the ME00 power / flow operating map for various cycle exposures anticipated during Cycle 2 (Reference 12). It was found that with the conservative assumptions-used the most limiting delta CPR exhibited by this event was 0.11, which justifies a MCPR operating limit of 1.17. A summary of these analyses is provided in Reference 12. A confirmatory analysis was performed for Cycle 3 which upholds the results of the Cycle 2 analyses, which thus bound Cycle 3 operation.

The Control Rod Withdrawal Error (CRWE) transient has been analyzed generically in Reference 18. The Reference 18 analysis provides a statistical evaluation of the consequences of the CRWE transient for BWR/6 plant configurations under conditions which cover the normal operating power flow map and the ELL and ICF regions. The generic conclusions support a power-dependent MCPR limit function as shown in Figure 2.2 of Reference 2. This limit was considered in evaluating the power-dependent Cycle 3 MCPR limits documented in Reference 2. ,

The results of the system and core transient analyses are provided in the Cycle 3 Reload Analysis Report (Reference 1) and in the Cycle 3 Plant Transient Analysis Report (Reference 2). With the conservative assumptions used the Loss of Feedwater Heating event exhibited the most limiting delta CPR of 0.11 which conservatively justifies a MCPR operating limit of 1.17.

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30 9.1 Core-Wide Transients The plant transient codes used to evaluate the Load Reject without Bypass (LRNB) and Feedwater Controller Failure (FWCF) events are the ANF COTRANSA code (Reference 11) and XCOBRA-T code (Reference 20), which incorporate a one-dimensional neutronics model to account for shifts in axial power shape and control rod effectiveness.

Technical Specification scram times were used in the bounding analysis. Therefore, no scram speed adjustment to the rated conditions MCPR operating limit is required for Cycle 3 operation of GGNS Unit 1.

9.2 Local Transients As shown in XN-NF-825(A), Supplement 2 (Reference 18), and in the Cycle 3 Reload Report (Reference 1), the results of the Control Rod Withdrawal Error (CRWE) event are bounded by the TS MCPR operating limit and power dependent MCPR limits (MCPRp ).

I 9.3 Reduced Flow and Power Operation ANF validated the use of the TS flow dependent MCPR for use during Cycle 3 in Reference 2. This validation was based on ANF's analysis of the recirculation pump flow increase event from reduced flow operation and on the results of LRNB and FWCF transients initiated from low flow condition. The operating limit consists of a plot of MCPR versus core flow.

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  • Cycle l MEOD.MCPR was-confirmed ~forCyck$3.LTherefore,the' Cycle 3 f
  • - , flow-dependent MCPR. limit, MCPR , is the:ruximum of the Cycle 3 f ,
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Y rated conditions MCPR operating-limit andithe reduced flow operating'

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.C limit' based on.the recirculation flow: increase event as shown in-  ;,

Figure-2.5 of Reference 2..

x Single Loop Operation'(SLO) is addressed in Sections 1.0, 5.2, 7.1.1, and7.2.1ofthe-Cycle.3'ReloadReport(Reference 1). The single-loop operation analysis for Cycle 2 bounds' operation in Cycle 3..

TheM4PFAC operating limit factor was evaluated for ANF fuel in f

, Grand Gulf through analysis of flow increase transientse lhe results of the ANF analysis which are reported in Reference 2 confirmed ,4 l

the flow-dependest MAPLHGR limit factor given in Figure 2.6 of~

Reference 2. For Single Loop Operation the MAPFAC defined in ,

'I .

Cycle 2 can be conservatively ~used, prcvided the average planar exposure is limited to 28,500 MWD /ST, .(Referance 21). .' ,

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The power-dependent MCPR operating limit was validated forl Cycle 3 operation through the generic CRWE analysis and the 91 ant-specific :j analyses of LRNB and FWCF transients.at representative conditions blanketing the operating power-flow map. Observance of this limit f for each fuel bundle. in the core during operation at less than full power conditions assures that the MCPR Fuel Cladding Integrity Safety Limit' will not be violated during anticipated operational l occurrences. The existing MCPR function is supported by the ANF p

analyses. The power-dependent MCPR limit is given in Figure 2.2 of i

Reference 2 and is applicable to all fuel types.

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The MAPFAC p operating limit factor was validated for Cycle 3 operation of Grand Gulf Unit I through the analysis of the FWCF, LRNB, and CRWE transients. The transient analyses which were used s

to evaluate power-depenaent MCPR limit effects were also used to evaluate power dependent LHGR effects.

The FWCF and LRNB results of the MAPFAC p determination are shown in Table 2.1 of the Cycle 3 Plant Transient Analysis Report (Reference 2)andcomparedwiththeCycle2MAPFAC p limits in Figure 2.3 of Reference 2. The CRWE results of Reference 18 were confirmed to also be bounded in Figure 2.3 of Reference 2. These analyses and comparisons validate the existing MAPFAC p limits for Cycle 3 operation.

9.4 ASME Overpressurization Analysis In order to demonstrate compliance with the ASME Code over-pressurization criterion of 110% of vessel design pressure, the MSIV closure event with failure of the MSIV position switch scram was U analyzed with ANF's LOTRANSA code (Reference 11). The Cycle 3 rr' analysis assumes seven safety relief valves are out of service. The maximum pressure observed in the analysis (at the vessel bottom) is 1271 psig, gMch is well within the 110% design criterion. An ANF l evaluation shows that the GE analysis of ATWS overpressurization is i-applicable to the mixed core.

l The calculated steam dome pressure corresponding to the 1271 psig peak vessel pressure is 1246 psig, for a vessel differential pressure of 25 psi. The current Technical Specification Safety Limit of 1325 l J10 MISC 860514 - 16 L

33 psig is based on dome pressure and conservatively assumes a 50 psi differential across the vessel (1375-1325). Since the calculated vessel differential pressure is.25 psi, the choice of 1325 psig assures compliance with the ASME criterion of 1375 psig peak vessel pressure.

10.0 POSTULATED ACCIDENTS In support of Grand Gulf operation, ANF has analyzed the Loss-of-Coolant Accident (LOCA) to demonstrate that MAPLHGR limits for Reload 2 fuel comply with'10CFR50.46 criteria. The Rod Drop Accident (RDA) was analyzed-for ANF Reload 2 fuel to demonstrate compliance with the 280 cal /gm Design Limic. The results of these analyses are presented in section 6.0 of Y Reference 1. Methodology for the RDA analysis is described in y

XN-NF-80-19(A) Vol. 1 (Reference 9). Methodology for the LOCA analysis is provided in References 13 through 15.

10.1 Loss-of-Coolant Accident (LOCA)

The generic BWR/6 LOCA break spectrum analysis as described in Reference 16 and performed in support of the Cycle 2 submittal is not cycle dependent. The analysis is therefore applicable for the upcoming cycle and no re-analysis is required.

A confirmatory analysis for operation within the current cycle MAPLHGR limits was performed for the upcoming cycle. The analysis confirms that the Reload 2 Peak Cladding Temperature (PCT) remains below the Cycle 1 PCT.

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Zr-H 0-reaction remains! elow 17% and core wide hydrogen-production'

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remains below 1%.for the limiting LOCA event.

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'As' discussed in Section'6.0 ANF fuel is hydraulically compatible with.-

t 1 .GE fuel. . Therefore, the existing.GE LOCA Analysis (which.is described i ,

in' the GGNS FSAR.for Unit-1 and Reference 6 for the expanded power / flow L ...

regibn) and MAPLHGRclimits will remain applicable for GE. fuel during Cycle.3.

10.2. Rod Drop Accident ANF's methodology for analyzing the Rod Drop Accident (RDA) is' described.in Reference 9 and' utilizes a generic parametric analysis which calculates the fuel enthalpy rise during postulated RDA's over a wide range of reactor operating variables. For Cycle 3, Section 6.2 of Reference 1-shows a value of 235 cal /gm for.the maximum

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. deposited fuel rod enthalpy during the worst case postulated RDA.

This-value is well below the design limit of 280 cal /gm. To ensure compliance with the RDA analysis assuinption, control rod sequencing below 20% core thermal power must comply with GE's Banked Position Withdrawal Sequencing constraints (Referer.ce 17).

11.0 Refueling.0perations

~During refueling' for the GGNS Cycle 3 reload plant operations will be restricted to modes 4, 5 and *. Mode

  • is defined in Reference 3 as, I J10 MISC 860514 - 18

35 r.

"when irradiated' fuel-.is being handled in the primary or secondary

-containment.and during CORE ALTERATIONS and' operations with a potential.

' for' draining the reactor. vessel ."

'TheonlyapplicablelFSARaccidents.forrefuelingoperationsarelisted-below:-

a) .Section15.4.1[RodWithdrawalError-lowpower b) .Section 15.'4.3, Control Rod Maloperation (system malfunction or

operatorerror) c) -Section 15.4.7, Misplaced Bundle Accident

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J d) Section 15.7.4, Fuel Handlitig Accident - Auxiliary building e) Section 15.7.6, Fuel Handling Accident - In containment Other FSAR accidents are not considered since the reactor must be

-maintained in a suberitical-condition during modes 4, 5 and *.

Consequently, the accidents are not possible and/or refueling activities have no effect on the cause of the accidents.

Inadvertent criticality either from a Rod Withdrawal Error or Control Rod Maloperation is precluded by plant design and administrative controls when operations are restricted to modes 4, 5 and *. The causes of a Misplaced Bundle Accident are controlled by plant procedures, and RF02 refueling activities and ANF bundle design have no impact on the-effectiveness of these procedures. Fuel handling accidents result from a failure ~of the fuel assembly lifting mechanism. The ANF bundle design is functionally and mechanically equivalent to the initial core GE fuel and Cycle 2 fuel, and is designed to be accommodated by GGNS fuel handling equipment.

48 M SC860514 - 19 _ _ _ _ _ _ _ _ __

36 The functional and mechanical equivalence of the ANF Cycle 3 reload fuel to previously analyzed fuel and the fact that the reactor must be maintained in a subcritical condition during modes 4, 5 and

  • allows typical refueling operations to take place.during RF02 without a reduction in margin to safety.

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=i REFERENCES ll)r ANF-87-67 Revision 1. " Grand Gulf Unit 1. Cycle 3 Reload Analysis",

- Advanced' Nuclear Fuels. Corporation, August.1987.

- 2)l ANF-87-66,~ Revision 1, " Grand Gulf Unit 1 Cycle 3 ' Plant Transient Analysis", Advanced Nuclear Fuels Corporation, August-1987. j Grand. Gulf Unit 1 Technical Specifications, Section 3/4.6. I 13) i i4) XN-NF-80-19(A), Vol. 4, Revision 1,." Exxon Nuclear Methodology.for Boiling Water Reactors: Application _of the ENC Methodology to'BWR Reloads", Exxon y

Nuclear Co., June 1985.

5) .XN-NF-85-67(A), Rev.'1, " Generic Mechanica1LDesign.for' Exxon Nuclear det Pump BWR Reload Fuel", Exxon Nuclear Co., September 1986.
6) ~"GGNS Maximum Extended Operating Domain Analysis", General Electric Company, San Jose, California 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) 'XN-NF-524(A), Rev. 1, " Exxon Nuclear Critical Power Methodology for Boiling Water Reactors", Exxon Nuclear Co., November 1983.

9). XN-NF-80-19(A), Vol. 1 Supplements 1 & 2. " Exxon Nuclear Methodology _for Boiling Water Reactors: Neutronic Methods for Design and Analysis", Exxon Nuclear Co., March 1983.-

l 10)' " Licensing Report on High Density Spent Fuel Racks for Grand Gulf Nuclear Station, Unit 1," Joseph Oat Corporation, Novmeber 1983.

11) XN-NF-79-71(P), Rev. 2, " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors", Exxon. Nuclear Co.', November 1981.
12) AECM-86/0273, Attachment 1, " Loss of Feedwater Heating Grand Gulf Nuclear E Station-1 Cycle 2 Specific Evaluation"
13) XN-NF-80-19(A), Vols. 2, 2A, 2B, & 2C, " Exxon Nuclear Methodology for Boiling Water Reactors: EXEM BWR ECCS Evaluation Model", Exxon Nuclear Co., September 1982.
14) XN-NF-CC-33(A), Rev. 1, "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option", Exxon Nuclear Co., November 1975.

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i 415) XN-NF-82-07(A), Rev. 1, " Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model",LExxon Nuclear Co., November 1982.

i 16). XN-NF-86-37(P),." Generic LOCA~ Break Spectrum Analysis for BWR/6 Plants".

Exxon Nuclear Co., May.1986.

-17) NEDO-21231, ." Banked Position Withdrawal Sequence", General Electric Co.,

January 1977.

18) XN-NF-825(A), Supplement'2,."BWR/6 Generic' Rod' Withdrawal Error Analysis, MCPR for All Plant Operations Within the Extended Operating Domain",

ExxoRNuclear, January 1986.

19): XN-NF-80-19(A), Volume 3, Revision 2, " Exxon Nuclear Methodology.for Boiling Water Reactors THERMEXi Thermal Limits Methodology", Exxon 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) MPEX-81/84, " Grand Gulf Unit 1. Cycle 3 - Single Loop Operation MAPLHGR

' Limits",. September 22,'1987.-

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GGNS UNIT 1 CYCLE 3 PROPOSED STARTUP PHYSICS TESTS AUGUST 1987 i

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Proposed Startup Physics Tests for Cycle 3

1. Core Loading Verification The core will be visually checked to verify conformance to the vendor supplied core loading pattern. Fuel Assembly serial numbers, bundle orientations, and core locations will be recorded. A height check will be performed to assure that all assemblies are properly seated in their respective locations.
2. Control Rod Functional Testing Prior to criticality following the refueling outage, functional testing of the control rods will be performed to assure proper operability. This testing will include coupling verification, withdrawal and insertion timing, and friction testing where required. The subcriticality.of the reloaded core with an individual control rod fully withdrawn will be verified by monitoring the nuclear instrumentation.
3. Shutdown Margin Determination i Control rods will be withdrawn in their standard sequence until criticality is achieved. The shutdown margin of the core will be determined from calculations based upon the critical rod pattern, the reactor period, and the moderator temperature. To assure there is no reactivity anomaly, the actual critical control rod position will be verified to be within 1% dk/k of the predicted critical control rod position.
4. TIP Asymmetry A gross asymmetry check will be performed as part of a detailed statistical uncertainty evaluation of the TIP system. A complete set of TIP data will be obatined at a steady state, equilibrium xenon condition greater than 85% rated power. A total average deviation or uncertainty will be determined for all symmetric TIP pairs as well as the maximum absolute deviation. The results will be evaluated to assure proper operation of the TIP system and symmetry of the core loading.

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