ML20196E939

Cycle 4 Reload Analysis

ML20196E939
Person / Time
Site:	Grand Gulf
Issue date:	11/11/1988
From:	Grummer R, Reynolds R SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML19295G794	List: AECM-88-0241, Application for Amend to License NPF-29,revising Tech Spec to Support Cycle 4 Reload for Facility.Fee Paid.Repts Re Reload forwarded.ANF-88-183(P) Withheld ML20196E919 ML20196E927 ML20196E933 ML20196E939 ML20196E956 ML20196E974
References
ANF-88-149, NUDOCS 8812120145
Download: ML20196E939 (45)

Text

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1 hv ADVANCED NUCLEAR FUELS CORPORATION GRANEs GULF UNIT 1 CYCLE 4 RELOAD ANALYSIS NOVEMBER 1988

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ADVANCED NUCLEAR FUELS CORPORATIO!)

3.NF- 4 :49 Issue Date: 11/11/88 GRAND GULF VNIT 1 CYCLE 4 RELOAD ANALYSIS Prepared by um/dLJ

" " 'R4 S. R~eynolds i$ht BWR(f afety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services I

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.,a G. Grummer I I ~ Y' '

l BWR Neutronics l Neutronics and Fuel Management l

Fuel Engineering and Technical Services November 8, 1988 m

CUSTOMER DISCLAIMER 1 It4PORTANT NOTICS REGARDING CONTENTS ANO USE OF TNIS DOCUMENT PLI.ASE READ CAREFULLY Advanced Nuc$oor Fuets Corocrabon's warranties and reoresentatens con.

comeg the suetect matter of tnis occument are those set forth in the Agreement benwoon Advanced Nu: seer Fuees Corporate and the Customer pursuant to when mis document is issued. Accordegry exceot as otherwee expressty pro-need in suen Agreement, nestner Advanced NucJear Fuets Corporaten nor any person aceng on as bened memes any warranty or recrosentaten. encrossed or eobed men respect to the accuracy. comomeness, or usefulness of the m%r- t menon contamed m mie document. or that the use of any mermaton. accaratus, memod or procese escceed m mis document *di not etnnge pnvateiy owned ngnts: or assumes any nacihtsee utn respect to the use of any eformaten, ao-paretus, memod or procoes escsosed a this occument.

The mformenon corMeesd herem is for the so6e use of Customer.

In order to avoed imoesrment of ngnts of Advanced Nue:*ar Fuees Corporaten m potents or inventons wncn may be included a the eformaten contamed m this document, the rece.ent. by its accoolance of this document agrees Act to puedmen of make putHic use (in the patent use of tne term) of suen mforteaten untd so autnonzed m wntmg Dy Advanced Nue ear Fue4s Corocraten or untd after sia (8) montes fe4lovnng termeaten or espiraten of the aforesa4 Agreement and any

.extensen thereof, uruess otnerwse exotosa#y crowced a tne Agreement. No ngnts or licenses e of to any patents are ireched ey tee fumisrq of tnia cocu-ment.

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ANF-88-149 Page i TABLE OF CONTENTS Section Pag.it

1.0 INTRODUCTION

........................ ... 1 2.0 FUEL MECHANICAL DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . . 5 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . 6 3.2 Hydraulic Characterization . . . . . . . . . . . . . . . . . . . 6 3.2.3 Fuel Centerline Temperature . . . ........... 6 3.2.5 Bypass Flow . . . . . . . . . . . . . . . . . . . . . . . 6 3.3 MCPR Fuel Cladding Integrity Safety Limit ........... 6 3.3.1 Nominal Coolant Condition in Monte Carlo Analysis . . . . 6 3.3.2 Design Basis Radial Power Distribution ......... 6 3.3.3 Design Basis Local Power Distribution . . . . . . . . . . 6 4.0 NUCLEAR DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . 11 4.1 Fuel Bundla Nuclear Design Analysis .............. 11 4.2 Core Nuclear Design Analysis . . . . . . . . . . . . . . . . . . 11 4.2.1 Core Configuration ................... 11 4.2.2 Core Reactivity Characteristics . . . . . . . . . . . . . 12 4.2.4 Core Hydrodynamic Stability . . . . . . . . . . . . . . . 12 5.0 ANTICIPATED OPERATIONAL OCCURRENCES . . . . . . . . . . . . ... . . . 16 5.1 Analysis Of Plant Transients . . . . . . . . . . . . . . '. . . . 16 5.2 . analyses For Reduced Flow Operation .............. 16 5.3 Analyses For Reduced Power Operation . . . . . . . . . . . . . . 16 5.4 ASMF. Overpressurization Analysis . . . . . . . . . . . . . . . . 16 5.5 Control Rod Withdrawal Error . . . . . . . . . . . . . . . . . . 16 5.6 Fuel Load i ng Error . . . . . . . . . . . . . . . . . . . . . . . 17 6.0 POSTULATED ACCIDENTS ........................ 22 6.1 Loss-Of-Coolant Accident . . . . . . . . . . . . . . . . . . . . 22 6.1.1 Break Location Spectrum . . . . . . . . . . . . . . . . . 22 6.1.2 Break Size Spectrum . . . . . . . . . . . . . . . . . . . 22 ,

6.1.3 MAPLHGR Analysis For ANF 8x8 Fuel . . . . . . . . . . . . 22 6.2 Control Rod Drop Accident ................... 23 7.0 TECHNICAL SPECIFICATIONS ...................... 24 7.1 Limiting Safety System Settings ................ 24 7.1.1 MCPR Fuel Cladding Integrity Safety Limit . . . . . . . . 24 7.1.2 Steam Dome Pressure Safety Limit ............ 24 7.2 Limiting Conditions For Operation ............... 24 7.2.1 Average Planar Linear Heat Generation Rate For ANF Fuel . 24 7.2.2 Minimum Critical Power Ratio .............. 25 7.2.3 Linear Heat Generation Rate For ANF Fuel ........ 26

A',4 F 149 Page 11 TABLE OF CONTENTS (Continued)

Section EAgg 7.3 Surveillance Requirements ................... 26 7.3.1 Scram Insertion Time Surveillance . . . . . . . . . . . . 26.

7.3.2 Stability Surveillance ................. 26 i 8.0 METH0COLOGY REFERENCES ....................... 31

9.0 REFERENCES

............................. 32 APPENDIX A SUPPLEMENTARY INFORMATION FOR 9X9-5 LEAD TEST ASSEMBLIES . . . 33

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ANF-88-149 Page iii LIST OF TABLES 14b.l.1 EASA 4.1 Neutronic Design Values .................... 13 LIST OF FIGURES Fiaure Eigt 1.1 Power / Flow Map Used for Grand Gulf Unit 1 ME0D Analysis .... 3 1.2 Grand Gulf Unit 1 Cycle 4 SLO MAPLHGR Limit .......... 4 3.1 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Radial Power Histogram .................... 7 3.2 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local Power Distribution (ANF 1.3 3.61 - 8G4 Fuel) ......... 8 3.3 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local Power Distribution (XN-2 3.21 - 6G4 Fuel) . . . . . . . . . . . 9 .

3.4 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis local Power Distribution (XN-1 2.99-5G3 Fuel) . . . . . . . . . . . . 10 4.1 Grand Gulf Unit 1 Cycle 4 ANF 1.3 3.61-8GZ Enrichment Distribution . . . . . . . . . . . . . . . . . . . . 14 4.2 Grand Gulf Unit 1 Cycle 4 Reference Core Loading Pattern (Quarter Core, Reflective Symmetry) . . . . . . . . . . . . . . 15 5.1 Flow Dependent MCPR Limits for Grand Gulf Unit 1 Cycle 4 . . . . 18 5.2 Power Dependent MCPR Limits for Grand Gulf Unit 1 Cycle 4 ... 19 t

5.3 Flow Dependent MAPFAC Value for Grand Gulf Unit 1 Cycle 4 ... 20 5.4 Power Dependent MAPFAC Value for Grand Gulf Unit 1 Cycle 4 . . . 21 -

7.1 Exposure Dependent Maximum Local Peaking for XN 12.99-5G3 Fuel. 27 i

7.2 Exposure Dependent Maximum Local Peaking for XN 2 3.21 6G4 Fuel ...................... 28 7.3 Exposure Dependent Maximum Local Peaking for XN-2 3.21-8G4 Fuel ............... ...... 29 7.4 Exposure Dependent Maximum Local Peaking for ANF 1.3 3.61 8G4 Fuel . . . . . . . . . . . . . . . . . . . . . 30 A.1 ANF 9x9 5 Lead Test Assembly LHGR Limits . . . . . . . . . . . . 36 A.2 ANF 9x9 5 Lead Test Assembly MAPLHGR Limits .......... 37 i

ANF-88-149

'X - .Page iv ACKNOWLEDGEMENT The authors would like to acknowledge the following individuals for their  ;

contributions to the results reported in this document:

D. A. Adkisson  :

1 D. J. Braun M. E. Byram .

S. J Haynes D. E. Hershberger M. J. Hibbard D. F. Richey S. E. State 1.

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ANF-88-149 Page 1

1.0 INTRODUCTION

This report provides the results of the analyses performed by Advanced Nuclear Fuels Corporation (ANF) in support of the Cycle 4 reload fo' Grand Gulf Unit 1. This report is intended to be used in conjunction with ANF topical report XN-NF-80-19( A), Volume 4, Revision 1, "Applicatie:e of the ENC Methodology to BWR Reloads," which describes the analyses pei formed in support of this reload, identifies the methodology used 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 NSSS vendor performed extensive safety analyses for Grand Gulf Unit 1 in conjunction with the extension of the power / flow operating map to the ME00 in Cycle 1 (Reference 1). These analyres established appropriate operating limits for ME00 operation. The initial reload of ANF fuel in Grand Gulf Unit i occurred in Cycle 2. In support of the initial reload of ANF fuel, extensive additional safety analyses were performed by ANF to either justify the NSSS vendor operating limits or, where necessary, to provide appropriate limits for ANF fuel using ANF methodologies (Reference 2). Subsequent ANF analyses supported an additional reload of ANF fuel in Cycle 3 (Reference 9).

Changes from Cycle 3 to Cycle 4 for Grand Gulf Unit 1 include an addi-tional reload of ANF fuel resulting in a complete core of ANF fuel. The cycle length remains 18 months but with cycle energy increased from 1420 GWd to 1698 GWd. A reload batch design composed of 272 assemblies enriched to 3.37 w/o U235 containing eight rods of axially var, ting Gd 02 3 as well as four (4) Lead Test Assemblies enriched to 3.25 w/o U235 is used to meet the cycle energy requirements. The balance of the core is comp 0 sed of 288 once exposed ANF reload fuel assemblies and 236 twice exposed ANF re'oad fuel assemblies.

The Cycle 4 fuel design increases the maximu,m batch avera;* exposure from 30,000 mwd /MTV to 34,000 mwd /MTV and the maximum assembly exposura from 33,000 mwd /MTV to 39,000 mwd /MTU (Reference 10).

The licensing basis of the four Lead Test Assemblies is described in Appendix A of this report.

ANF-88-149 Page 2 The design and safety analyses reported in this document were based on design and operational assumptions in effect for Grand Gulf Unit I during the Cycle 3 operation and conditions bounding Cycle 4 operation. The MCPR p and MCPRf limits have been revised to reflect ANF calculated limits rather than the NSss generic ME00 limits for this first all-ANF core in Grand Gulf Unit 1.

Analyses were performed in accordance with the existing bases in the plant Technical Specifications, excipt that analysis set points for safety valves have been increased to include a 6% tolerance, and provision has been made in the flow dependent MCPR's for "loop manual" operation as well as "non-loop manual" operation (Reference 11). The analyses also included support of the power / flow operation map for Maximum Extended Operating Domain as shown in Figure 1.1, Monitoring to the plant Technical Specifications presented in this report will be performed using ANF's core monitoring methodology, POWERPLEX* CMSS, in accordance with ANF's thermal limits methodology, THERMEX (Reference 8.6).

The ANF evaluation for Grand Gulf Unit 1 Single Loop Operation (SLO),

operation with feedwater heaters out of service, operation without condenser bypass, and LOCA-seismic considerations were performed for Cycle 2 and confirmed for subsequent cycles. Since the Cycle 4 analyses results are similar to those of Cycle 2, the Cycle 2 antdyses and available margin to limits for these off normal operating conditions assures that these events will continue to be protected. An exception is for SLO MAPLHGR limits for fuel exposures in excess of 28,500 mwd /ST (31,388 mwd /MTV). Since some Cycle 2 fuel will exceed this exposure during Cycle 4, a revised SLO MAPLHGR curve (Figure 1.2) has been conservatively constructed for all fuel types in Cycle 4. The extended curve was constructed to bound all fuel resident in

Cycle 4 by merging previously approved GE and ANF curves. The MAPFACf curve for SLO is unchanged from Cycle 3.

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ANF-88-149 Page 5 2.0 FUEL MECHANICAL DESIfJi ANALYSIS Applicable Fuel Derign Report: Reference 3 Qualification analyses provided in the reference are applicable to the Grand Gulf Unit 1 ANF fuel assemblies. The oxtended burnup design for the Cycle 4 reload is described in Reference 10. This analysis confirms the applicability of fuel mechanical limits for the higher burnup reload fuel design.

Tne expected power history for the fuel to be irradiated during Cycle 4 is bounded by the design LHGR of Figure 3.1 of Reference 3.

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ANF-88-149 Page 6 3.0 THERMAL Hv0RAULIC DESIGN ANAL.YSIS 3.2 liy.drihCharacterization 3.2.3 Fuel Centerline Temoerature Fuel Centerline Melting is protected by the transient LHGR limit given in Reference 3.

3.2.5 Bvoass Flow Calculated Bypass Flow Fraction 10.0%

(Exclusive of Water Rod Flow at 104.2%P/108%F) 3.3 MCPR Fuel Claddina Intearity Safety Limit See Reference 4 1.06*

3.3.1 Nominal Coolant Condition in Monte Carlo Analysis Core Power 4128 MWt Core Inlet Enthalpy 527,9 Etu/lbm Reference Pressure 1050 psia Feedwater Temperat,re 420'F Feedwater Flow Rate 17.74 Mlbm/hr 3.3.2 Desian Basis Radial Power Distribution See Figure 3.1 3.3.3 Desian Basis local Power Distrib dian See Figures 3.2 to 3.4

• For single loop operation the safety limit MCPR increases to 1.07 due to increased uncertainties.

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Figure 3.2 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local Power Distribution (ANF-1,3 3.61-8G4 Fuel) c

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- Gadolinia Location 80 o

1 ANF-88 149 Page 10

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- Gadolinia location

ANF-88-149 Page 11 4.0 NUCLEAR DESIGN ANALYSIS 4.1 Fuel Bundle Nuclear Desian Analysis Assembly Average Enrichment 3.37 w/o Radial Enrichment Distribution Figure 4.1 Axial Enrichment Distribution Uniform 3.61 w/o with 6" natural Uranium at top and bottom Burnable Poisons Figure 4.1 Eqtg: Burnable poisons are not distributed uniformly over the enriched length of the designated rods. Tne natural uranium axial bianket sections do not contain burnable absorber material.

Location of Non-Fueled Rods Figure 4.1 Neutronic Design Parameters Table 4.1 4.2 Core Nuclear Desian Analysis 4.2.1 Core Confiauration Figure 4.2 Core Exposure at E0C3 18104 MWD /MT Core Exposure at B004 10244 MWD /MT Core Exposure at E0C4 22308 MWD /MT Maximum Cycle 4 Licensing Exposure Limit 23130 MWD /MT

i l ANF-88-149 I Page 12 4.2.2 Core Reactivity Charactriif.is1(1),(2)

[4C4 Cold K-effective, All Rods Out 1.13315 .

BOC4 Cold K-effective, All Rods In 0.96019 BOC4 Cold K-effective,

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l Strongest Rod Out 0.98906 l Reactivity Defect /R-Value 0.00% Delta K/K - '

l (Minimum occurs at 0 mwd /MTU)

Stat:dby Liquid Control System Reactivity, 660 PPM i

T.old Conditions, K-effective 0.96215

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(1) Includes calculational bias. '

(2) Evaluated at nominal E003-808 mwd /HTV.

4.2.4 Core Hydrodynamic Stability The results of Cycle 4 core hydrodynamic stability analyses continue to confirm the applicability of the previous cycles analyses. .

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ANF-88-149 Page 13 Table 4.1 Neutronic Design Values Fuel Assemb1v l

II Number of fuel rods 62 Number of inert water rods 2 Fuel rods enrichments Figure 4.1 l

Fuel rod pitch, inches 0.636 l Fuel assembly loading, KgU 175.69 Core Data Number of fuel assemblies 800 Rated thermal power, HWt 3833 Rated core flow, M1bm/hr 112.5 Core inlet subcooling, Btu /lbm 22.2 Moderator tempercture, F 551 Channel thickness, inch 0.120 Fuel assembly pitch, inch 6.0 Sym. water gap thickness, inch 0.545 1

Control Rod Data Absorber material B4C Total blade span, inch 9.804 Total blade support span, inch 1.55 Blade thickness, inch 0.328 Blade face-to-face internal dimension, inch 0.238 Absorber rods per blade (wing) 72 (18)

Absorber rod outside diameter, inch 0.22 Absorber rod inside diameter, inch 0.166 Absorber density, percent of theoretical 70

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• ................................................................. r-

ML  : M  : M*  : H  : H  : M*  : M  : ML  : ..

L  : ML  : M  : M  : M  : M -

ML : L  :

................................................................. p LL R005 ( 1) --

1.50 W/0 U235 L RODS ( 5) --

2.00 W/0 U235 -

ML R005 ( 8) --.

2.90 W/0 U235 M R005 (20) ---

3.48 W/0 U235 ..

H RODS (20) --- 4.57 W/0 U235 M* RODS ( 8) --. 3.48 W/0 U235 + 4.00 W/0 G0203 (Top 24 inches)

M* R005 ( 8) ---

3.18 W/0 U235 + 5.00 W/0 G0203 (Botton 114 inches)

W R005 ( 2) --.

INERT WATER ROD Figure 4.1 Grand Culf Unit 1 Cycle 4 ANF 1.3 3.61-8GZ Enrichment Distribution

\

6

ANF 88-149 Page 15 1 2 3 4 5 6 7 8 9 to 11 12 13 14 15 16

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1

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C 204 AWF Pt3 1N 1.2 3.01 w/o U 235 ecd at 6.0 %

D 272 ANF stb ANF 1.3 3.37 w/o U 235 2;d at . 0 % \ 5 . 0 ?.

E 4 AwF 9u9 AhF 1.3 3.25 w/o U 235 SGd at 5.0'. \ 6.0%

figure 4.2 Grand Gulf Unit 1 Cycle 4 Reference Core Loading Pattern (Quarter Core, Reflective Symetry)

ANF 88-149 Page 16 .

5.0 ANTICIPATED OPERATIONAL OCCURRENCES Applicable Generic Transient Reference 5 Methodology Report 5.1 Analysis Of Plant Transients Reference 4 (Applicable at rated conditions)

Transient ,

Delta CPR LRNB 0.12 LFWH 0.11*

CRWE 0.10**

FWCF 0.04

• Applicable at all conditions.

• " Statistically determined, Ref. 6.

5.2 Analyses For Reduced Flow Ooeration Reference 4 MCPRf Figure 5.1 MAPFACf Figure 5.3 5.3 eaalvses For Reduced Power Ooeration Reference 4 HCPR p Figure 5.2 ,

MAPFAC p Figure 5.4 5.4 A3ME Overoressurization Analysis Reference 4 Limiting Event MSIV Closure  ;

Worst Single Failure HSIV Position Scram Trip M ximum Vessel Pressure 1298 psig -

Maximum Dome Pressure 1280 psig 5.5 Control Rod Withdrawal Error Referunce 6 Values of delta CPR as a function of core power level resulting from a .,

ORWE transient, developed in Reference 6 on a generic casts for BWR/6 class of "

plants including Maximum Extended Operating Ocinain operation, are applicable to Cycle 4 operation. , ,

,x wum

\

ANF-88-149 Page 17 -

5.6 Fuel Loadina Error Refere.ve 8.1 With Correctly Lqajina Error Loaded Core Maximum LHGR 13.90 12.40 l Minimum HCP.". 1.17 1.27 T. ,:,

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ANF 88-149 Page 22 6.0 POSTULATED ACCIDENTS i

6.1 Loss-Of-Coolant Accident 6.1.1 Break Location Soectrum Reference 7 6.1.2 Break Size Soectrum Reference 7 6.1.3 MAPLHGR Analysis For ANF 8x8 Fuel Reference 8 Limiting Break: Double Ended Guillotine Pipe Break in Recirculation Pump Discharge Line with 1.00 Discharge Coefficient (1.0 DEG/RD)

{ Reference l

Analysis ANF 1.3 Peak local Average Planar Analyzed Peak Clad Peak Clad Metal-Water Exoosure MAPLHGR Temoerature Igsperature Reaction 0 GWD/MTV 14.3 kW/ft 1738 F 1663 F 0.3%

5 14.3 1685 1659 0.3 10 14.3 1678 1666 0.3 15 14.3 1687 1679 0.3 20 14.3 1680 1691 0.3 25 13.2 1642 1641 C.3 30 12.1 1575 1577 0.2 35 11.1 1496 1497 0.1 40 10.0 1403 1405 0.1 45 9.0 1321 1328 0.1 50 7.9 -- 1206 0.1 Changes in local peaking in the ANF 1.3 reload fuel, cause the PCT at higher exposures to exceed the reference analysis PCT by uo to ll'F; an 11*F increase in PCI is insignificant due to the fact that the calculated temperatures are over 500*F below the 2200'F limit.

4

k l

ANF-88-149 Page 23 6.2 control Rod Dron Accident -

Reference 8.1 Dropped Control Rod Worth 9.72 mk l Doppler Coefficient -9.5 x 10 6  ;

AK/K/'F Effective Delayed Neutron Fraction 4.5 x 10-3 Four Bundle local Peaking Factor 1.28 Maximum Deposited Fuel Rod Enthalpy 172 Cal /gm

ANF 88-149 Page 24 7.0 TECHNICAL SPECIFICATIONS 7.1 Limitina Safety System Settinos 7.1.1 MCPR Fuel Claddina Intearity Safety Limit Safety Limit MCPR 1.06*

7.1.2 Steam Dome Pressure Safety Limit Pressure Safety Limit 1325 psig 7.2 Limitina Conditions for Ooeration 7.2.1 Averaae Planar Linear Heat Generation Rate For ANF Fuel The following MAPLHGR limits are consistent with the design basis LHGR limits shown in Figure 3.1 of Reference 3. These limits differ only by a factor equal to the maximum local peaking factor at each exposure point. The MAPLHGR limits are made consistent with the LHGR limit so that at reduced power and/or reduced flow the LHGR limit will be protected by the MAPFACf and MAPFAC p multipliers on MAPLHGR. The maximum local peaking factors for the four ANF fuel designs that will be resident in the core during Cycle 4 are shown in Figures 7.1, 7.2, 7.3 and 7.4. The LOCA analysis was performed at conservatively higher MAPLHGR values, Section 6.1.3.

• A safety limit MCPR of 1.07 is to l'e applied during single loop operation.

ANF-88 149 Page 25 Average Planar MAPLHGR*

Exposure ANF299ESG3S8 ANF321E6G4S8 AliERIE8G458 ANF361E8GZSS 0.00 GWd/MTU 13.20 kW/ft 13.33 kW/ft 13.00 kW/ft 12.98 kW/ft 0.25 13.20 13.34 13.00 12.98 1.00 13.38 13.36 13.02 13.01 i o 2.00 13.54 13.40 13.06 13.03 l 4.00 13.89 13.54 13.26 13.13 l 6.00 14.26 13.75 13.59 13.33 8.00 14.26 14.01 13.93 13.60 10.00 14.12 14.03 14.10 13.84 15.00 13.78 13.61 13.87 13.87 20.00 13.30 13.20 13.42 13.38 24.00 13.03 12.92 13.09 13.03 25.00 12.96 12.85 13.01 12.94 1

25.40 12.94 12.82 12.98 12.90 30.00 11.77 11.65 11.75 11.65 35.00 10.48 10.44 10.46 10.31 40.00 9.15 9.17 9.18 9.00 42.00 8.61 8.64 8.64 8.48 MAPLHGR Multipliers for Off Nominal Conditions :

MAPFAC(f)** Figure 5.3 MAPFAC(p) Figure 5.4 7.2.2 Minimum Critical Power Ratio Rated Conditions MCPR Limit 1.18 MCPR(f) Figure 5.1 MCPR(p) Figure 5.2 l

• The MAPLHGR limit of figure 1.2, applicable to all ANF fuel types resident in Cycle 4, is used for SLO.

• • For SLO the Cycle 3 MAPFACf limit is applied to all ANF fuel types resident in Cycle 4.

ANF-88 149 Page 26 7.2.3 Linear Heat Generation Rate For ANF Fuel The current Grand Gulf Unit 1 LHGR limits remain applicable for ANF reload fuel Cycle 4 operation. These limits, which are based on Figure 3.1 of Reference 3, are as follows, .

Averaae Planar Exoosqtt LHGR 0.00 GWd/MT 16.0 kW/ft 25.40 14.1 42.00 9.3 ,

7.3 Surveillance Recuirements  !

l 7.3.1 Scram Insertion Time Surveillance Thermal margins are based on analyses in which scram performance was assumed consistent with the Technical Specification limits. Ne additional surveillance for scram performance is required above that already being done for conformance to Technical Specifications.

7.3.2 Stability Surveillance Submittal regarding stability amendment is being made under separate .

cover by the Licensee.

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l ANF-88-149 Page 31 8.0 METHODOLOGY REFERENCES Section 8 References 8;) through 8.18 are contained in the following '

report:

"Exxon Nu: lear 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).

I Reference 8.6 is superseded by, 8.6 ' Exxon Nuclear Methodology for Boiling Water Reactors THERMEX:

Thermal Limits Methodology Sumary Description," XN-NF 8019f P)f A),

Volume 3, Revision 2 (January 1987).

f i

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ANF-88-149 Page 32

9.0 REFERENCES

i

1. Letter, Lester L. Kintner (USNRC) to 0. D. Kingsley, Jr. (MP&L),

"Technical Specification Changes to Allow Operation with One Recirculation Loop and Extended Operating Domain," August 15, 1986.

2. "Grand Gulf Unit 1 Cycle 2 Reload Analysis," XN-NF-86 35, Revisio1 3, Exxon Nuclear Company, Richland, WA, August 1986.

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

XN-NF-85-67(P)(A), Revision 1 Exxon Nuclear Company, Richland, 'WA, September 1986.

4. "Grand Gul f Unit 1 Cycle 4 Plant Transient Analysis," ANF-88 150, Advanced Nuclear Fuels Corporation. Richland, WA, November 1988.

5. "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors,"

- XN-NF-79-71(P), Revision 2, Exxon Nuclear Company, Richland, WA, November 1981.

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

Nuclear Company, Richland, WA, May 1986, and XN NF-825(P)(A),

Supplerent 2, October 1986.

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

Exxon Nuclear Company, Richland, WA, April 1986.

j I

8. "Grand Gulf Unit 1 LOCA Analysis," XN-NF-86 38, Exxon Nuclear Company, Richland, WA, June 1986, a 9. "Grand Gulf Unit 1 Cycle 3 Reload Analysis," ANF-87 67, Revision 1.

j Advanced Nuclear Fuels Corp., Richland, WA, August 1987.

l 10. "Grand Gulf Unit 1 Reload XN 1,3, Cycle 4 Mechanical Design Report "

! ANF 88-183, Advanced Nuclear Fuels Corporation, Richland, WA, j November 1988.

11. "Grand Gul f Nuclear Station Unit 1 Revised Flow Dependent Thermal i limits," NESDO 88-003, MSU System Services Inc., November 1988.

1 l

l l

I

ANF 88 149 Page 33 APPEN0!X A i

SUPPLEMENTARY INFORMATION FOR r 9X9 5 LEAD TEST ASSEMBLIES A.1 INTRODUCTION ,

Evaluations have been performd consistent with ANF methodology (' Exxon Nuclear Methodology for Boiling Water Reactors, XN NF-8019) to establish a ,

licensing basis for the four (4) ANF 9x9-5 Lead Test Assemblies (LTA's) in the Grand Gulf Cycle 4 core. Justification is provided which demonstrates the applicability of Grand Gulf Cycle 4 operating limits to the LTA's unless J

stated otherwise.

The insertion of only four ANF 9x9-5 LTA's will have negligible effects upon core wide transient performance. However, 9x9 5 specific analyses were performed to assure that the Cycle 4 operating limits also apply to the LTA's.

Fuel type specific limits (LHGR and related MAPLHGR limit) have been developed for the LTA's and are presented in this appendix.

A.2 FUEL MECHANICAL DESIGN A mechanical design analysis has been performed for the 9x9 5 fuel type consistent with ANF's approved methodologies. Fuel design issues related to Anticipated Operational Occurrences (A00's) and accident analysis have been evaluated. These evaluations confirm that the LTA's meet NRC criteria of no centerline melting and less than 1% clad strain ("Generic Mechanical Design for ANF 9x9-5 BWR Reload Fuel," ANF-88 152(P)).

A.3 THERMAL HYDRAULIC DESIGN Cesponent hydraulic resistances have been determined and it is been found that thi 9x9 5 LTA's are hydraulically compatible with the co-resider.t l ANF 8x8 fuel assemblies. Unique design features of the 9x9 5 (two rod l diameters, injection water rod) have been modeled to demonstrate compatibility I

over the full range 'of expected operating conditions. Steady state thermal hydraulic analysis have shown that even though the 9x9-5 design has a some-l i

ANF-88-149 Fage 34 what smaller flow area than the 8x8 design no reduction in thermal margin is experienced i 'i the Cycle 4 core. This is due to the increased thermal performance of the 9x94 5 design and the placement of tne 9x9 5 fuel in non-limiting positions.

A.4 NUCLEAR DESIGN The core wide neutronic impact of replacing four (4) of the 800 fuel assembites in the Grand Gulf Cycle 4 core is negligible. The leads are designed to be neutronically "transparent" relative to the 8x0 fueli that is, reactivity characteristics are similar.

Evaluation of the 9x9 5 LTA's relative to LFWH, Control Rod Drop Accident,#4APFAC f , shutdown margin and Shutdown Liquid Boron Control have been included in the main body of this report. in that they have been explicitly modeled in those calculations. The LTA Misload has been evaluated separately using the XN 3 sorrelation, which has been demonstrated to conservatively i predict critical ptwer in the 9x9-5 design.

A.5 ANTICIPATED OPERATJg @L OCCURRENCES Analyses of limiting transients have shown that the bundle power .2eded to produce boiling transition during transients in the 9x9-5 fuel design is higher than that for the 8x8 fuel design. It has been shown that ANF's approved BWR CHF correlation, XN 3, is conservative when applied to the 9x9 5 CHF data. Therefore, applying 8x8 MCPR operating limits based on XN 3 to the current 9x9 5 LTA's assures lowr bundle powers than would be necessary to reach the 9x9 5 boiling transition.

Because of the neutronic similarity of the 9x9 5 LTA's to the 8x8 assemblies, the consequences of other A00's, such as i.ontrol rod withdrawal error and fuel rotation error are essentially the same as in the case of 8x8 fuel, i

I

ANF-88 149 Page 35 A.6 POSTULATED ACCIDENTS Since heatup is primarily a planar and not an axial phenomena, the ,

appropriate bundle power limit derived from LOCA analyses is the peak bundle planar power, it has been demonstrated that the 9x9 5 LTA's provide better LOCA performance relative to an 8x8 fuel assembly due to the greater surface area provided by the larger number of fuel rods, more inert surface from the tater rods and less stored energy in the rods. The 9x9 5 MAPLHGR limit is based on the LHGR limit provided in "Generic Mechanical Design for ANF 9x9 5 BWR Reload Fuel," (ANF 88152(P)] 4tvided by the maximum local peaking as a function of exposure. Analyses performe3 by ANF demonstrate that this limit l meets 10 CFR 50.46 criteria.

1he conseque,nces of a control rod drop accident are governed primarily by the dropped rrd worth. Since the reactivity of the LTA's is comparable to the coresident 848 fuel and the LTA's are loaded in non limiting locations, no appreciable difference will be experienced due to the LTA's, 4

A.7 TECHNICAL SPECIFJCATIONS All oprational limits used for 8x8 fuel are applicable to the 9x9 5 LTA's except for fuel type specific iMPLHGR limits and a 9x9 5 LHGR limit.

The LHGR limit shown in Figure A-1 is that of ANF 88152. The MAPLHGR limit is shown in Figure A 2 and is consistent with the 9x9 5 LHGR limit. The LTA l

SLO operational limits are bared on the 9x9 5 MAPLHGR multiplied by the smaller of the MAPFACf . MAPFACp , or 0.86.

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Figure A.2 ANF 9x9-5 Lead Test Assembly MAPLHGR ;:mit2

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ANF-88-149 Issue Date: 11/11/88 GRAND GULF UNIT 1 CYCLE 4 RELOAD ANALYSIS

-I Distribution D. A. Adkisson D. J. Braun

0. C. Brown M. E. Byram R. E. Collingham R. A. Copeland W. S. Dunnivant L. J. Federico N. L. Garner R. G. Grummer D. E. Hershberger M. J. Hibbard T. L. Krysinski A. Reparaz R. S. Reynolds G. L. Ritter S. E. State R. B. Stout C. J. Volmer G. N. Ward H. E. Williamson ,

SERI/N. L. Garner (40) [

4 Document Control (10)

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