ML20066D927

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Nonproprietary Rev 2 to ANF-90-022, Grand Gulf Unit 1, Cycle 5 Reload Analysis
ML20066D927
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 08/08/1990
From: Hershberger D, Hibbrad M, Macduff R
SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML19310E502 List:
References
ANF-90-022, ANF-90-022-R02, ANF-90-22, ANF-90-22-R2, NUDOCS 9101170132
Download: ML20066D927 (41)
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Text

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ADVANCEDNUCLEARFUELS CORPORATION ANF 90-022 Revision 2 issue Date: 08/08/90 GRAND GULF UNfT 1 CYCLE 5 RELOAD ANALYSIS Prepared by R. B. Macduff f BWR Fuel Engineepn Fuel Engineering and U sing 0YObv it

/ D.12. Hershberger BWR Fuel Engineering Fuel Engineering and Ucensing

%Uu. W M. J. Hibbard BWR Fuel Engineering Fuel Engineering and Ucensing August 1990 9101170132 910109 DR ADOCK 0500 6

_ _ _ _ , , _ , , _ . , - - - - - , - - ~ ' ' ~ ~

4 1

CUSTOMER OtSCLA14sER NANT le0TICE M^JZ 008titttTS AND USE OF TNG 00CUINNT pt3Ae8 raA0 CAREPULLY Assensed Nuelser Puem "m's werfenses and represemagene som serfunt the sueneet muser of this eseumont are these set form m the Agnesmem between Aevenged Nueteer Puste Corporalem and me Cusemer pursuant to tuften Wee seneurnent e issueEL Aseerenpy, esseet se Omensues empressly pr>

wees in euen Agreement, noener Asveneed Numeer Puses Corporesen nor sny pereen esmg on as beneW meses any womrey or resresemamen, esereseen or 5 Insted, WWI rences le the escuresy eengleteness. Or WashNnees of the trW0r.

mean centenned M tnet encumert er tna the use of any amormemen, apparatus, memed or preesee teentesed in mee desument well not intnnge prweert owned ngnes: er esswrme any temenes won resceu a the use of any wnermanon, ao-perWus, named or process tegetesed in Inne escumert The intermemen sensemed hereen is ter me sein use of Cassemer, in oneer a sweet imperment of ngne of Aevenese Nuesser P#a Corporenon m pelems or inveneens uneen may se sneluesd in me rWormemen somemos m tnse

  • eseument, tne reassent. Dy the assessenes of tnes eseumer% agrees not to puenen or meme punto use (in me pesent use of the term) of Suon rWormeson untti se anstertled in tureMO py Aoveness Nuoteer Puede Corporanen or unet efter eut te menes temowing temunemen or esserseen of me eseresend Agreement ene any essenenen tnerest, imless omennuse eierteely pnnneed an the Agreement, No .

nones er neeness m or e any penne are egned by me fumeneg of ifue occu-ment.

ANF-3145.472A (12/87)

- . . . -- . . . - - = . . - - . - -

ANF 90-022 Revision 2 '

Page1

- TABLE OF CONTENTS Section

.Pggg 1.0 INTR O D U CTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1. . . . . . .

8 2.0 CUEL MECHANICAL DESIGN ANALYSIS . . . . . . . . . , , . . . . . . . . . . . . . . . . . . . . . 4 3.0

- TelERMAL HYDRAUUC DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.2 Hydraulic Characterization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 i 3.2.3 Fuel Centerline Temperature . . . . . . . , , . . . . . . . . . . . . . . . . , , . . . . S-3.2.5 Bypass Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.3 MCPR Fuel Cladding Integrity Safety Umit . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.3.1 Nominal Coolant Condition in Mcate Carlo Analysis . . . . . . . . . . . . . 5 3.3.2. Design Basis Radial Power Distribution . . . . . . . . . . . . . . . . . . . . . . . 5 3.3.3 Design Basis Local Power Distribution . . . . . . . . . . . . . . . . . . . . . . . 5 4.0 NUCLEAR DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 4.1 - Fuel Bundle Nuclear Design Analysis . , . . . . , . . . . . . . . . . . . . . . . . . . . . . 8

, 4.2 . Core Nuclear Design Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 4.2.1 Core Configuration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 4.2.2 Core ReactMty Characteristics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

'4.2.4 Core Hydrodynamic Stability . . . . . . . . . . . . . . . . . . . . . , , . . . . . . . . 9 5.0 ANTICIPATED OPERATIONAL' OCCURRENCES . , . . , , . . . . . . . . . . , . . . . . . . . 15 5.1 Analysis of Plant Transients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

52 Analyses For Reduced Flow Operation . . . . . . . . . . . . . . . . . . . . . . .15. . . . - 15 5.3

- Analyses For Reduced Power Operation . . . . . . . . . . . . . . . . . . . . . . . . . . _15 5.4 ASME Overpressurization Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 5.5 Control Rod Withdrawal Error . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

. 5.6 Fuel loading Error . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 5.7 Determination of Thermal umits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 6.0 POSTULATED ACCIDENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 6.1: Loss Of Coolant Accident - . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . - 22 6.1.1 Break L~'=tian Spectrum . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 6.1.2 Brer. k Size Spectrum . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22

' 6.1.3 M/ >LHGR Analysis For ANF 8x8 and 9x9 5 Fuel r, , . . . . . . , . .-. . .- 22 6.2 Control Rod Drop Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 7.0 TECHNICAL SPECIFICATIONS . . . , . . . . . . . . . , . . . .- ,- . . . . . . . . . . . . . . . . . . 25-7.1 = Umiting Safety System Settings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25-7.1.1 MCPR Fuel Cladding Integrity Safety Umit . . . . . . . . . . . . . . .... 25-7.1.2 Steam Dome Pressure Safety Umit . . . . . . . . . . . . . . . . . . . . . . . . . 25 I-9 +w'we* T e- w ms,w- ,- m --

+p- w g rpr- w ,--y- ym + -w-v vv,--( w

i 1

ANF.90-022 Revision 2 Page 11 '

TABLE OF COPFEN7S (Continued)

Section Eggg 7.2 Umiting Conditions For Operation . . . . . . . -. . . . . . . . . . . . . . . . . . . . . . .. . - 25 7.2.1 Average Planar unear Heat Generation Rate ANF~

F uel . . . . . . . - . . . , r. . . . - . . . . . . . . . . . . . . . ..........

. . . . . . . .for 25

-7.2.2 Minimum Critical Power Ratio . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 7.2.3 Unear Heat Generation Rate For ANF Fuel- . . . . . . . , . . . . . . . . . 26 7,3 Surveillance Requirements . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 7.3.1 ' Scram Insertion Time Surveillance . . . . . . . . . . , ... . . . . . . . .- 27 7.3.2 ~ Stability Surveillance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .....

. . . . 27 8.0 -

METHODOLOGY REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28-

~'

9.0 R E F E RENCE S . . . . . . - . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 -

APPENDIX A SEISMIC /LOCA-ANF 9x9 5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A1 .......

a

1 ANF.90-022 Revision 2 Page lil-UST OF TABLES, I.th!! a f_.ag,q 4.1 Neutronic Design Values . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

- A.1 -

to Fuel Assembly Properties . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -. A.3

- LIST OF FIGURES Flourt f.A22 1.1 Power / Flow Map Used for Grand Gulf Unit 1 MEOD Analysis . . . . . . . . . . . . . . . . . 3 .

3.1_ Grand Gulf Unit _1 Cycle 5 Safety Umit Design Radial Histogram . , , .-. . . . . . . . . . . 6 3.2 ' . Grand Gulf Unit 1 Cycle 5 Safety Umit Design Basis Local Power Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 4.1 Grand Gulf Unit 1 Cycle 5, ANF 1.4 ?.NF380E8GXS95

- Enrichment Distribution . . . . . . . . . . . . . . . . . . - . . . . . . . . . . . . . . . . . . . . . . . . . . 11 4.2 Grand Gulf Unit 1 Cycle 5, ANF.1.4 ANF380E9GXS95 Enrichment Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 4.3 - ! Grand Gulf Unit 1 Cycle 5, ANF 1.4 ANF380E10GXS95-Enrichment Distribution . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 4.4 Grand Gulf Unit 1 Cycle 5 Reference Core Loading Pattern

a. (Quarter Core, Reflective Symmetry) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14

.5.1 . Flow Dependent MCPR Umits for Grand Gulf Unit 1 Cycle 5 . . . . . . . . . . . . . . . . . 17 5.2 Power Dependent MCPR Umits for Grand Gulf Unit .1 Cycle 5.. . . . . . , . . . . ., . . . . 18

'5.3 Flow Dependent LHGRFAC Value for Grand Gulf Unit 1 Cycle 5 . . . . . . . . . . . . . . 19 5.4 --

,_ Power Dependent LHGRFAC Value for Grand Gulf Unit 1 Cycle 5 , . . . . . . . . . . . . 20 5.5 - Exposure Dependent MCPR Umits for Grand Gulf Unit 1 Cycle 5 . . . . . . . . . . . . . 21 8.1 MAPLHGR vs Average Planar Exposure for ANF 8x8 and ANF 9x9-5 Reload Fuel . . . _. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 L

I 1

ANF 90-022

. Revision 2 1 Pagelv l During'.the NRC review of the nuclear peaking uncertainties of the MICROBURN B l methodology, ANF was informed that the proposed TIP asymmetry uncerteinty as presented in l Reference 14 would require further extensive review, ANF was also informed that concurrence

-l to use the currently accepted value would allow the NRC to complete remaining actions l associated with the issuance of the MICROBURN B SER without further technical review by the l NRC staff. ANF agreed to th* eaa of the currently accepted value as stated in Reference 15.

l l The change in uncertainty value required ANF to evaluate the impact upon analyses l performed for the Cycle 5 licensing campaign for Grand Gulf Unit 1 as provided in ANF_90-021 l and ANF 90022.

I l Revision 2 of this report is issued to effect the changes in results associated with die increase

- l _ in TlP asymmetry uncertainty. Text changes from Revision 1 are indicated by revision bars in the l left margin of the report. Figures 5.1,5.2, and 5.5 are also revised.

ANF 90-022 Revision 2 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 5 reload for Grand Gulf Unit 1. This report is intended to be used in conjunction with ANF koical report XN NF-8019(A). Volume 4, Revision 1,

" Application of the ENC Methodology to BW6 Reloads," which describes the analyses performed 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. Methodology used in this report which supersedes XN-NF 80-19(A). Volume 4, Revision 1, is referenced as appropriate.

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 MEOD in Cycle 1 (Reference 1). These analyses established appropriate operating limits for MEOD operation. The initial reload of ANF fuelin Grand Gulf Unit 1 occurred in Cycle 2. In support of the initial reload of ANF fuel, extensive additional safety analyses were perfr.,n<l by ANF to either justify the NSSS vendor operating limi's or, where necessary, to provide appropriate limits for ANF fuel using ANF methodologies (Reference 2). Subsecu^nt ANF analyses supported an additional reload of ANF fuelin Cycle 3 (Reference 9) and again in Cycle 4 (Reference 12).

Changes from Cycle 4 to Cycle 5 for Grand Gulf Unit 1 include an additional reload of ANF fuel resulting in a core comprised of once and twice bumed ANF 8x8 designs, four ANF 9x9-5 LTAs, and fresh ANF 9x9-5 design. The 9x9 5 reload fuel is mechanically, neutronically, and thermal hydraulically compatible with the co-resident 8x8 fuel inserted in previous cycles.

The cycle length remains 18 months and the nominal cycle energy remains 1698 GWd. A reload batch design composed of 284 assemblies enriched to 3.42 w/o U235 containing axially varying Gd23 0 is used to meet the cycle energy requirements. A portion of each assembly contains from eight to ten Gd23 0 rods. The balance of the core is composed of 272 once exposed ANF 8x8 reload fuel assemblies, four once exposed 9x9-5 lead fuel assemblies and 240 twice exposed ANF 8x8 reload fuel assemblies.

ANF 90022 Revision 2 Page 2 The design and safety analyses reported in this document were based on design and operational assumptions in effect for Grand Gulf Unit 1 during Cycle 4 operation and conditions bounding Cycle 5 operation. The MCPRpand MCPR, lim}ts have been vertfled or revised to reflect ANF calculated limits. As in Cycle 4, provision has been made in the flow dependent MCPRs for " loop manual" operation as well as "non loop manual" operation (Reference 11).

Analyses were performed at EOC-2000 mwd /M TU, at EOC, and at EOC + 30 EFPD providing limits for Cycle 5 that are cycle exposure dependent. The analyses also included support of the power / flow operatiorn map for Maximum Extended Operating Domain as shown in Figure 1.1.

MCPR values were determined using the ANFB Critical Power Correlation (Reference 8.9).

Monitoring to the plant thermal limits presented in this report will be performed using ANF's core monitoring methodology, POWERPLEXe CMSS, in accordance with ANF's thermal limits methodology, THERMEX (Reference 8.8).

l l

The ANF evaluation for Grand Gutf Unit 1 $ ingle Loop Operation (SLO) without condenser bfpass and LOCA seismic considerations were confirmed for Cycle 2 and subsequent cycles.

- Since the Cycle 5 SLO anaM are performed using new methodology (References 5 and 8.1 through 8.18), the Cycle 5 results supersede the Cycle 2 results.

t i

S I I I I I I I I I I I g (75. 100) (105. 100)

] ELL Region ICF gE -

g@' Region-oc go6

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E &

.8 - -

l

! V l ES -

/ -

3 (105,42)

@- (34.3. 25) (73.5. 25) -

1 I I I i 1 I I I I I y

% 10 20 30 40 50 60 70 80 90 100 110 120 g Core Flow. X of Rated ui(

un>M Figure 1.1 Power / Flow Map Used for Grand Gulf Unit i MEOD Analysis

~. ._. .. _ _ . . . _ _ _ . _. .. __ ._ ... _ . . ... _. .... _ __ . . .- __ - _ _ _ . . . . _ . . . _ . _ _

ANFWO22 Revision 2 Page 4 2.0 PUEL MECHANICAL DESIGN ANALYSIS Applicable Fuel Design Report: References 3,10, and 13 Qualification anayses provided M the references are applicable to the Grand Gulf Unit 1 ANF fuel assemMos.

T.ie expected power history for the fuel to be irradiated during Cycle 5 is bounded by the design LHGR of Figure 3.1 of References 3 and 13.

4

., m.._,,.m. y -_ _ -. . , _ . , . _ , , _ _._._,y -

ANF 90422 Revision 2 Page 5 3.0 THE4fAAL HYDRAUUC DESIGN ANALYSIS 3.2 Hydrsulle Characterization 3.2.3 Fuel Centerline Tomoerature Fuel Centerline Melting is protected by the transient LHGR limit given in References 3 and 13, 3.2.5 Bvoass Flow Calculated Bypass Flow 10.6% ~

(Exclusive of Water Rod Flow at 104.2%P/108%F) 3.3 MCPR Fuel Claddina intoority Safetv Limit l- See Reference 4 1.09*

3.3.1 Nominal Coolant Condition in Monto Carlo Anaivsis Core Power 4754 MWt Core In* Enthalpy 522.3 Btu /lbm Reference Pressure 1050 psia ,

Feedwater Temperature 420*F Feedwater Flow Rate 20.43 Mlbm/hr 3.3.2 Desion Basis Radial Power DistributiSD See Figure 3.1 3.3.3 Deslan Basis Local Power Distribution See Figure 3.2

] *The 1.09 includes effects for channel bow and single loop operation.

ANF 90 022 Revision 2 Page 6 m

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ANF 90422 Rev'sion 2 Fage 7 i

1 F

l 1.116 1.127 1.108 1.117 1.107 1.116 1.107 1.126 1.116 l 1.127 0.786 1.007 0.973 0.636 0.971 1.004 0.785 1.126 l 1.108 1.007 0.949 0.954 0.976 0.946 0.944 1.004 1.106

l. 1.117 0.973 0.954 0.735 0.000 1.045- 0.946 0.970 1.116 l 1.107 0.636 0.976 0.000 0.000 0.000 0.976 0.633 1.106

' }. 1.116 0.971 0.946 1.045 0.000 0.706 0.956 0.972 1.116 l l'.107 1.004 0.944 0.946 0.976 0.956 0.949 1.006 1.107 1.126 0.785 1.004 0.970 0.633 0.972 1.006 0.784 1.126 1.116 1.126 1.106 1.116 1.106 1.116 1.107 1.126 1.116 t

Figure 3.2 Grand gun UnN 1 Cycle 5 Safety Umit Design Basis Local Power Distribution d:

I

, v - - - - , , - ~ - . - , , ,p.w,,- - ,y,.. , , - - w. .a w. . _ + , . , y,,,, . . . , . - . . , , ,r., ,r_, . . - - - v .

I l

1 ANF 90M2 Revision 2 Page8 4.0 NUCLEAR DESIGN ANALYSIS 4.1 Fuel Bundle Nuclear Deslan Anaksin Assembly Average Enrichment 3.42 w/o Radial Enrichment Distribution Figures 4.1 4.3 Axlal Enrichment Distribution Uniform 3.80 w/o with 12" natural uranium at top and 6* at bottom Burnable Poisons Figures 4.1 4.3 th;tlt: Burnable poisons are not distributed uniformly over the enriched length of the designateo rods. The natural uranium axial blanket sections do not contain burnable absorber material.

Location of Non Fueled Rods Figures 4.1 4.3 Neutronic Design Parameters Table 4.1 4.2 Core Nuclear Deslan Anaksis 4.2.1 Core Conflauration Figure 4,4 Core Exposure at EOC4 23010 mwd /MTU Core Exposure at BOC5 12872 mwd /MTU Core Exposure at EOC5 24948 mwd /MTU Maximum Cycle 5 Ucensing Exposure Umit 25766 mwd /MTU i

l i . .

4 i

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i.

ANF 90-022 Revision 2 Page 9 4.2.2 Core ReactMtv Character!gggU)'M BOC5 Cold K effectNo, M Rods Out 1.12245 BOC5 Cold K effectNo, M Rods in 0.95342 BOC5 Cold K offectNo, Strongest Rod Out 0.98v66 ReactMty Defect /R Value 0.0% Delta K/K (Minimum occurs at 0 mwd /MTU)

Standby Uquid Control System ReactMty, 660 PPM Cold Condhions, K effectNo 0.97065 Ull ncludes calculational bias, MEvaluated at nominal EOC4 825 mwd /MTU.

4.2.4 Core Hydroovnamic StaN!Nv The resuhs of Cycle 5 core hydrodynamic stabilhy analyses continue to confirm the applicabilhy of the previous cycles enalyses results. The presence of 9x9 5 fuelin the Cycle 5 core does not change the conclusions of the stability analysis of the previous cycles, I

i i

ANF 90 022 Revision 2 Page 10 h

Table 4.1 Neutronic Design Values Fuel AssembW (9x9 5)

Number of fuel rods 76 Number of inert water rods 5 Fuel rods enrichments Figures 4.1 4.3 Fuel rod pitch, inches 0.563 Fuel assembly loading, KgU ANF 1,4 H 175.16 ANF 1.4 L 175.59 Core Data Number of fuel assemblies 800 Rated thermal power, kWt 3833 Rated oore flow, Mibm/hr 112.5 Core inlet suboooling, Btu /lbm 22.2 Moderator temperature, F 551 Channel thickness, inch 0.120 Fuel assemoiy pitch, Inch 6.0 Sym. water gap thickness, inch 0.545 Control Rod Data Absorber material B4C Total blade span, inch 9.804 Total blade support span, inch 1.55 Blade thickness, inch 0.326 Blade face-to-face infomal dimension, irJ1 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 A

, m .- _ . - . ~. ,,,-.,..-,y-- -m, . , . .- . .,--..r.,vo -w,,

                    • '**************** ANFM322
  • Revision 2 Page 11 L1 . ML1 M1 MH1 MH1 MH1
  • M1 ML1 L1 ML1 LLl* MH1 H2 LL2* LLl*
  • H2 MH1 Mll M1 MH1 MH1 H2 H2
  • H2 MH1 MH1 M1 MH1 H2 H2 LL2 W
  • H2 H2 H2 MH1 HH1' LL2* H2 W W W H2 LL2* MH1 MH1 H2 H2 H2 W LL2 H2 H2 MH1 M1 MH1 MH1 H2 H2 H2 MH1 MH1 M1
  • h Mll LLl* MH1 H2 LL2* H2 MH1 LLi* ML1 L1 -ML1 M1 MH1- MH1 MH1 M1 Mll L1 i

L1 Rods Mll

--- 2.67 w/o U235 Rods -- 3.33 w/o U235 M1 Rods --- 3.66 w/o U235 MH1 Rods (24) --- -3.98 w/o U235 H2 Rods (22 4.73 w/o U235 LL2 Rods ( 2 --- 2.27 w/o U235 LLl* Rods ( 4 -- 2.27 w/o U235 + 5.5 or 7.0.w/0 Gd23 0 4

LL2* Rods ( 4 -- 2.27 w/o U235 + 5.5 or 7.0 w/0 Gd 0 W Rods ( 5 - Inert Water Rod 23 FIGURE 4.1 GRAND GULF UNIT 1 CYCLE 5, ANF 1.4 ANF3E'E8GXS95 ENRICHMENT DISTRIBUTION

==5 Page 12 L1 HL1 M1 MH1 MH1 MH1 M1 ML1 L1

-Mll LLl* MH1 H2 LL2*

, H2 MH1 LLl* Mll M1 MH1 MH1 H2 H2 H2 MH1 MH1 M1 MH1 H2 H2 LL2 W H2 H2 H2 MH1 MH1 LL2* H2 W W W

, H2 LL2* HH1

  • - MH1 H2 H2 H2 W LL2* H2 H2 MH1 M1 MH1 MH1 H2 H2 H2 MH1 MH1 M1 Mll LLl* MH1 H2 LL2* H2 MH1 LLl* Mll i

L1 Mll M1 MH1 MH1 MH1 M1 ML1 L1 L1 Rods ( 4)

' -- 2.67 w/o U235 Mll Rods ( 8 -- 3.33 w/o U235 M1 Rods ( 8 -- 3.66 w/o U235 MH1 Rods (24 - --- 3.98 w/o U235 H2 Rods 22) --- 4.73 w/o U235

-LL2- Rods 1) --- 2.27 w/o U235 LLl* Rods 4) --- 2.27 w/o U235 + 3.0, 4.5, 5.5 or 7.0 w/0 Gd23 0 LL2* Rods ( 5) --- 2.27 w/o U235 + 3.0, 4.5, 5.5 or 7.0 w/0 Gd 23 0 l W Rods (-5) --- Inert Water Rod L

FIGURE 4.2 GRAND GULF. UNIT 1 CYCLE 5, ANF-1.4 ANF380E9GXS95 ENRICHMENT DISTRIBUTION

-y

  • .****. ***** ... ** ***.* ANF 90 022 Revision 2 Page 13 L1 ML1 M1 MH1

, MH1 MH1 M1 Mt.1 L1 ML1 LLl* MH1 H2 LL2* H2 MH1 LLl* ML1 M1 MH1 MH1 H2 H2 H2 MH1 MH1 M1 MH1 H2 H2 LL2* W

  • H2 H2 H2 MH1 MH1 LL2* H2 W W W H2 LL2*
  • MH1 MH1 H2_ H2 H2 W LL2* H2 H2 MH1 M1 _MH1 MH1. H2 H2 H2 Mn1 MH1 M1 ML1 LLl* MH1 H2 LL2* H2 MH1 LLl* ML1 L1 Mll M1 MH1 MH1 MH1 M1 ML1 L1 L1 Rods 4) 2.67 w/o U235 ML1 Rods 8) 3.33 w/o U235 M1 Rods .3.66 w/o U235 MH1 Rods 2)) - 3.96 w/o U235 H2Rods LLl* Rods 2 ( 4.73 w/o U235

- 2.27 w/o U235 + 4.5, 5.5 or 7.0 w/0 Gd 23 0 LL2* Rods 2.27 w/o U235 + 4.5, 5.5 or 7.0 w/0 Gd23 0 W Rods ( 5 - Inert Water Rod FIGURE 4.3 GRAND GULF UNIT 1 CYCLE 5 ANr.l.4 ANF380E10GXS95 ENRICHMENT DISTRIBUTION

ANF.90-022 Revision 2 Page 14 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 1 A2 C1 E0 C1 to A2 F0 A2 F0 A2 70 A2 70 C1 A2 A2 2 F0 70 C1 A2 C1 E0 C1 70 A2 to C1 F0 C1 C1 B2 A2 3 EO A2 70 C1 B2 C1 to C1 F0 C1 to B2 70 C1 A2 4 C1 70 01 to C1 70 A2 F0 C1 to C1 F0 B2 C1 B2 5 to C1 A2 C1 to C1 C1 A2 70 C1 E0 C1 F0 Cl A2 6 82 to C1 F0 C1 to C1 70 C1 to C1 F0 82 C1 A2 7 F0 C1 to A2 C1 C1 to A2 70 82 F0 C1 F0 C1 A2 8 A2 70 70 C1 A2 F0 A2 E0 C1 to C1 F0 C1 A2 9 F0 A2 F0 C1 70 C1 F0 C1 to C1 70 C1 82 10 A2 to C1 E0 C1 to A2 E0 C1 F0 A2 C1 A2 11 70 Ci to ci to C1 F0 C1 F0 82 C1 B2 A2 12 82 F0 A2 F0 C1 F0 C1 F0 C1 C1 A2 82 13 F0 C1 F0 32 70 82 F0 C1 82 A2 A2 14 C1 C1 C1 C1 C1

.C1 C1 A2 15 A2 82 82 A2 A2 A2 A2 16 52 A2 Xe FUEL fYPE XY Y a CYCLt3 IRRA0lAftp NUNett 0F FUEL AS$tMIllts fYPE (FULL CORE)

.... OtSCRIPfl0W A ...........................................

164

$ 76 ANF 8X8 FMa1.2 3.01 W/0 U 235 6G0 At 4.0%

C 272 ANF 8X8 XN 1.2 3.01 W/0 U+235 800 AT 4.0%

0 4 ANP 8X8 ANF 1.3 3.37 W/0 U 235 8G0 At 4.0% \ 5.0%

t 104 ANF 9X9 ANF 1.3 3.25 W/0 U 235 800 At 5.0% \ 6.0%-

F 180 ANP 9X9 ANP 1.4 3.42 W/0 U 235 1000 AXIALLY ZONED ANP 9X9 ANF 1.4 3.42 W/0 U 235 900 AXIALLY 20Nt0 Figure 4.4 Grand Gulf Unit 1 Cycle 5 Reference Core loading Pattern (Quarter Core, Reflective Symmetry)

,,_,,,,,~,Egy- w

  • 7 * '

ANF 90-022 Revision 2 Pa9e 15 5.0 ANTICIPATED OPERATIONAL OCCURRENCES Appilcable Generic Transient References 5,8.8 Methodology Report 5.1 Analysis of Plant Transients Reference 4 (Applicable at rated conditions)

Transient DeMa CPR*

EOC-2000 mwd /MTU EQQ EOC+30 EFPD .

LRNB 0.06 0.20 0.21 l LFWH** 0.09 0.09 0.09 CRWE*** 0.10 0.10 0.10 FWCFNB 0.08 0.13 -

Umiting values.

,,, Applicable at all conditions.

Statistically determined Reference 6.

Exposure Dependent Umit MCPR, Figure 5.5  !

5.2 Mses For Reduced Flow Operation Reference 4 MCPR, Figure 5.1 LHGRFAC, Figure 5.3 5.3 Analyses For Reduced Power Ooeration Referenco 4 MCPRp Figure 5.2 LHGRFAC p Figure 5.4 '

5.4 ASME Overoressurization Analysis Reference 4 Umiting Event MSIV Closure Worst Single Failure MSIV Position Scram Trip

ANF 90-022 Revision 2 Page 16 Maximum Vessel Pressure 1291 psig Maximum Dome Pressure 1269 psig 5.5 Contro! Rod Withdrawal Error Reference 6 Values of delta-CPR as a function of core power level resulting from a CRWE transient were developed in Reference 6 on a generic basis for BWR/6 class of plants (including Maximum Extended Operating Domain operation). Power dependent limits of MCPR are based on these results as well as the results from the Cycle 5 specific transient analysis (Reference 4).

5.6 Fuel Loadina Error Reference 8.1 With Leadina Error Cderectiv Loaded Com Maximum LHGR 14.33 12.83 Minimum MCPR* 1.21 1.28

  • Determined using ANFB Critical Power Correlation.

5.7 Determination of Thermal Umits The results of the analyses presented in Sections 5.1, 5.2, and 5.3 are used for the determination of the operating limit. Section 5.1 provides the results of analyses at rated conditions, including the operating limit as a function of exposure in the cycle (MCPR,,

Figure 5.5). Sections 5.2 and 5.3 provide for the determination of operating limit at off-rated conditions of reduced flow and reduced power operation (MCPR,, algure 5.1 and MCPR p, Figure 5.2). The highest value of MCPR from among the ones preaanted in these figures for the operating condition of the reactor is to be selected as the operating limit of interest.

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  • 86 Figure 5.1 i Flow Dependent MCPR Limits For Grand Gulf Unit i Cycle 5

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- '8 Figure 5.2 Power Dependent MCPR Limits For Grand Gulf Unit 1 Cycle 5 *~~

-... .- - - - _ - _ - _ _ - _ _ . - . - . . - - . . . . _ - - _ = . - . - . - . - . .. - _.~ - -

ANF 90-022 Revision 2 Page 19 o

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  • 86  ;

Figure 5.4 Power Dependent LHGRFAC Value for Grand Gulf Unit i Cycle 5 8md  !

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ANF 90-022 Revision 2 Page 22 6.0 POSTULATED ACCIDENTS 6.1 Loss Of Coolant Acciden!

6.1.1 Break Location Sooctrum Reference 7 6.1.2 Bigt Size Soectrum leference 7 6.1.3 MAPLHGR Analvsis For ANF 8x8 and 9x9 5 Fuel References 8 and 12 Umiting Break: Double Ended Guillotine Pipe Break in Recirculation Pump Discharge Une with 1.00 Discharge Coefficient (1.0 DEG/RD)

The spray heat transfer coefficients identified in 10CFR50 Appendix K are used for the 9x9-5 fuel in an identical manner as are used for the ANF 9x9 2 fuel design. This includes the 2

use of 5 BTU /hr ft ,.F for all of the unheated surfaces including the five water rods.

MAPLHGR results for the two reload fuel types are reported below:

Peak Local Maximum Metal Water PCT (*F) Reaction (%)

8x8 Fuels 1891 0.3 9x9 Fuels 1896 0.4 The core wide metal water reaction is less than 0.1%.

The MAPLHGR limits for 8x8 and 9x9 5 are shown in Figure 6.1.- These are bounding limits. The 9x9 5 limits are bounding for the LTA. The 8x8 limits are provided in Reference 8.

For single-loop operation, a reduction factor of 0.80 is applied to the two-loop MAPLHGR limits I L

L f

i .

l

_ ._ _ _. _._. _ _ _ _ _____. _ _- _ .,. _ _ . - _ . . ._ . . _ . . . . _ . . . ~ . . . . _ _

ANF 90-022 Revision 2 Page 23 i

shown in Figure 6.1. Application of this reduction factor ensures that the PCT for a single loop operation LOCA is bounded by the two-loop LOCA analysis.

6.2 Control Rod Droo Accident Reference 8.1 Dropped Control Rod Worth 8.8 mk Doppler Coefficient 10.4 x 10 4 AK/KrF Effective Delayed Neutron Fraction 5.47 x 104 i Four-Bundle Local Peaking Factor 1.439 Maximum Deposited Fuel Rod Enthalpy 192 cal /g The Control Rod Drop ~ Accident analysis is unaffected by the lowering of the BPWS operability requirement from 20% power to 10% power. .

1 I

L t

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- ._-. . . - _ _ . - - _.- - _ ~ -. - - -_ -_._ .. -__. . . . = _ _ - - _ _ . -

ANF 90 022 Revision 2 Page 24 I I O

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- ANF 90-022 Revision 2 - l Page 25 l 7.0 TEC:lNICAL SPECIFICATIONS 7.1 Umitina Safety System SettlDQt 7.1.1 MQPR Fuel Claddina intearity Safetv Umit l Safety Umst MCPR 1.09*

7.1.2 Steam Dome Presauro Safety Umit Pressure Safety'Umit 1325 psig 7.2 umitina Conditions For Ooor@n 7.2.1 Aversoe Planar Unear Heat Generation Rate for ANF Fuel The following MAPLHGR limits are consistent with 10CFR50.46 requirements. Unlike previous cycles, the MAPLHGR limit is not usesd to protect the design basis LHGR limits for the fuel types co-resident in Cycle 5.- <

Average Planar MAPLHGR MAPLHGR Exoosure 8x8 9x9 5 0.0 GWd/MTU 14.3 kW/ft 12.5 kW/ft 20.0 14.3 12.5 50.0 7.9 ,

9.5 55.0 -

9.0 For single-loop operation, a reduction factor of 0.8 is applied to the above two loop .

MAPLHGR limits.

1 l *The 1.09 safety limit acc nts for channel bow and single loop operation.

. - - - ~ _ . - . - - . . .- . . . _.

ANF 90-022 Revision 2 Page 26 7.2.2 Minimum Crktcal Power Ratio MCPR(f) Figure 5.1 MCPR(p) Figure 5.2 MCPR(e) Figure 5.5 7.2.3 Unear Heat Generation Rate FEANF Fuel

- The LHGR limits for Grand Gulf I as p wlously anayzed remain applicable for ANF 8x8 fuel during Cycle 5 operation. These limits are er ended to cover the exposure range for Cycle 5.

These limits, which are based on Figure 3.1 of Reference 3, are as follows:

Averaos Planar Exoosure LHGR 0.00 GWd/MTU 16.0 kW/ft 25.40 14.1 50.00 6.96 The LHGR limits for 9x9-5 fuel, based on Figure 3.1 of Reference 13 for ANF reload fuel during Cycle 5, operation are as follows:

Averaos Planar Exoosure {.HGR 0.00 GWd/MTU 13.1 kW/ft 15.50 13.1

-55.00 8.0 LHGRFAC, and LHGRFACp multJpliers are applied directly to the Technical Specification LHGR limits for each fuel type at reduced power and/or flow conditions to ensure protection of the limits.

l LHGRFAC Multipliers for Off Nominal Conditions:

! LHGRFAC(f) Figure 5.3 LHGRFAC(p) Figure 5.4 l

i

-s-.--, . , , - - , - . - - - - - . . - - - - , - . - - . , - - - , _ .- ,-, ,. .-,,-,-

ANF 90022 Revision 2 Page 27 7.3 Surveillance Recuirements 7.3.1 Scram Insertion Time Surveillance Thermal margins are based on anah/ ses in which scram performance was assumed consistent with the Technical Specifiestion limits.

No additional surveillance for scram performance is required above that already being done for conformance to Technical Specmcations.

7.3.2 Gtabilltv Surveillance Core stability surveillances have been addressed by the Ucensee in TS 3/4, 4.1.1 (Technical Specmcation Amendment No. 62).

l-i u _ . . . - - - - - . . - - - - - -- --- -- -

ANF 90422 Revision 2 Page 28 8.0 METHODOLOGY REFERENCES Section 8 References 8,1 through 8.18 are contained in the following report:

' Exxon Nuc4 ear Methodology for Bolling Water Reactors: Application of the ENC Methodology to BWR Reloads,' XN NF 8019(A), Volume 4, Revision 1. Exxon Nuclear Company, Richland, Washir,gton (March 1985),

Reference 8.6 is superseded by:

8.6 ' Exxon Nuclear Methodology for Bolling Water Reactors THERMEX: Thermal Umits Methodology Summary Description,' XN-NF4019(P)(A). Volume 3, Revision 2 (January 1987),

References 8.9 and 8.18 are superseded by:

8.9

'ANFB Critical Power Correlation,' ANF 1125. Supplement 1(P)(A) (April 1990),

Reference 8.10 is superseded by:

8.10

' Advanced Nuclear Fuels Corporation Critical Power Methodology for Bolling Water Reactors,' ANF 524(P). Revision 2, and Supplements, April 1989.

ANF 90-022 Revision 2 Page 29

9.0 REFERENCES

1.

Letter, Lester L Kint,wr (USNRC) to O. D. Nngsley, 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 Anahsis,' XN NF-86 3 Revision 3, Exxon Nuclear Company, Richland, WA, August 1986.

3.

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

XN NF-85-67(PifA). Revision 1, Econ Nuclear Company, Richland, WA, September 1986.

4. ' Grand Gulf Unn 1 Cycle 5 Plant Transient AnaYsis,' ANF 90021. Revision 2, Advanced Nuclear Fuels Corporation, Richland, WA, August 1990.

5.

'COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analysis,"

ANF 913. Volume 1, Supplements 1,2, and 3.

6. 'BWR/6 Generic Rod WRhdrawal Error Anaysis, MCPRp,' 2L NE 02E61, Exxon Nuclear Company, Richland, WA, May 1986, and XN NF 825(P)(A). Supplernent 2, October 1986.
7. ' Generic LOCA Break Spectrum Anaysis for BWR/6 Plants,' XN-NF 86 37(P). Exxon Nuclear Company, Richland, WA, April 1986.

8.

' Grand Gulf Unit 1 LOCA Anahsis,' XN NF-86-38. Exxon Nuclear Company, Richland, WA, June 1906.

- 9. ' Grand Gulf Unn 1 Cycle 3 Reload Analysis,' ANF-87 67. Revision 1, Advanced Nuclear Fuels Corp., Richland, WA, August 1987, t

i 10. ' Grand Gulf Unn 1 Reload ANF 1.4, Cycle 5 Mechanical, Thermal Hydraulle, and Neutronic Design for Advanced Nuclear Fuels 9x9 5 Fuel Assembinas,' ANF 8917_1.(P) volumes 1 and 2, Advanced Nuclear Fuels Corporation, Richland, WA, January 1990.

11. ' Grand Gulf Nuclear Station Unit - 1 Revised - Flow. Depene ont Thermal Umits,'

NESDQ-88-001 MSU System Services Inc., November 1988.

12. ' Grand Gulf Unit 1 Cycle 4 Reload Analysis,' ANF-88149. Advanced Nuclear Fuels Corporation, Richland, WA, November 1988.
13.
  • Generic Mechanical Design for Advanced Nuclear Fuels 9x9-S BWR Reload Fuel,' _

ANF-8B-152fP). Amendrnent 1, September 1989, Advanced Nuclear Fuels Corporation, Richland, WA.

1

1 1

ANF 90-022 i

Revision 2 -

Page 20 i

14. Letter, R. A. Copeland (ANF) to Director, NRR (NRC), " Submittal of MICROBURN 8,* dated March 8,1989 (RAC
0N:90),
15. -

Letter, R. A. Copeland (ANF) to Lambros Lois (NRC), "TIP Asymmetry Uncertainty,' dated l JuY 20,1990 (RAC:083:90),

, . , . , u.._..---,,...,; -. . ,.-_.-_____;_.-._--_--..-

ANF 90422 Revision 2 Page A 1 APPENDIX A SEISMIC /LOCA ANF 9x9-5 The acceptability for Grand Gulf Unit 1 of the ANF 9x9-5 fuel selsmic LOCA performance is qualified by its similartty to the GE 8x8 fuel originalty licensed to operate in Grand Gulf Unit 1.

The 9x9 5 fuel will exhibit essentialty the same static and dynamic response as the GE 8x8 since it has essentially the same dynamic and hydraulic characteristics as identified below and is subjected to the same dynamic excitation.

The dynamic input to the reload fuel will be the same as that for the existing fuel since it will be installed at the same location and there are no significant changes which would affect the overall response of the reactor pressure vessel (RPV) and its pedestal. The dynamic response of the assemblies is dependent on the mass and stiffness propertles of the fuel elements which determine their natural frequencies.

Table A.1 presents, for comparison, fuel assembly properties for the GE 8x8, ANF 8x8, and ANF 9x9 fuel. Based on the data presented, the important dynamic characteristics for the various fuel bundles are similar.

The channeled fuel assembfy dynamic response is primarily a function of the channel.

Because channels of a sim;lar design are used for both the 8x8 and 9x9 fuel, then, the in-reactor dynamic characteristics of the channeled fuel assembly for these fuel types would essentially be identical. This is confirmed in the analysis documented in the Susquehanna Unit 2 Cycle 2 reload analysis, XN NF-86-60 Appendix B, where the soismic LOCA performance of the 8x8 and 9x9 assemblies are compared. The ANF analysis reported in XN NF 8151(P)(A),'LOCA Seismic Structural Response of an Exxon Nuclear Company BWR Jet Pump Fuel Assembly," dated May 1986, used a channel allowable stress of 24,000 psi at 545'F. The NRC has concluded that the ANF value of 24,000 psl is conservative relative to the GE channel faulted allowable stress.

The Cartech channel uses the same material as GE and has a limiting faulted allowable stress

ANF 90022 Revision 2 Page A 2 ~

(1.2 x oyp) greater than 28,380 pel at 545 'F. Thus, design margin exists when either GE or Cartech channels are used.

The pressure drop of different fuel designs can be compared from calculations performed for typical full core loadings of the respective designs at the rated condPjons of flow and power.

- The results can ba considered in terms of overall pressure drop and in terms of fuel assembly drop. The overall pressure drop considers the pressure drop from the orifice inlet to the top of the upper tie plate wh!ie the fuel assembly pressure drop suotracts out the orifice pressure drop.

The results of the typical BWR4 analysis show that the ANF 8x8 and the ANF 9x9 5 fuel designs have lower pressure drops than the comparable GE 8x8 fuel design,

- In comparing overall pressure drops, the ANF 8x8 fuel shows an 8% lower pressure drop than the GE 8x8 fuel, The ANF 9x9 5 fuel shows a 2% lower pressure drop than the GE 8x8 fuel.

Focusing on,the fuel assembly pressure drop, the ANF 8x8 fuel shows a 12% lower pressure drop than the GE 8x8 fuel while the ANF 9xS5 fuel shows a 3% lower pressure drop than the GE 8x8 fuel.

In summary [the 9x9 5 dynamic and hydraulic characteristics are essentially the same as

' those of the fuel it replaces. - Therefore, the results of previous analyses are applicable to the 9x9-5; l

l-

,4

ANF 90022 Revision 2 Page A 3 Table A.1 Fuel Assembly Properties Procertv QEQx.g* M 8x8 9x9-5 Active Fuel Length (in) 150.0 150.0 150.0 150.0 Fuel Rod OD (in) 0.483 0.424 0.484 0.443/0.417 Pellet OD (in) 0.410 0.3565 0.405 0.375/0.353 Fuel Rod Pitch (in) 0.636 0.563 0.636 0.563 Spacer Pitch (in) 20.15 20.15 20.15 20.15 Number of Water Rods 2 2 2 S Fuel Assembly Weight (Ib) 600 574 585 583 Channel Length (in) 167.4 167.4 167.4 137.4 Channel Wall Thickness 0.120 0.120 0.120 0.120 (in)

Channel Weight (Ib) 98.7 98.7 98.7 98.7 Channel Minimum inside Envelope (in) 5.205 5.205 5.205 5.205

  • Estimated value.

I ANF.90 022 Revision 2 issue Date: 08/08/90 i

i i

GRAND GULF UNIT 1 CYCLE 5 RELOAD ANALYSIS Distribution O. C. Brown R. A. Copeland  !-

- W. S. Dunnivant L J. Federloo

- c.

N. L Garner-M. E. Grirstt

-D. E. Hershberger M. J. Hibbard T. L Krysinski - l R. B. Macduff R. S. Reynolds -

S. E. State R. B. Stout C. J. Volmer G. N. Ward -

H. E. Williamson SERl/N. L Gamer (40) rf Document Control (5)

L i

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