ML20246B316

Forwards Rev 2 to Current Cycle Safety Analysis as Source for Current Accident & Transient Analyses

ML20246B316
Person / Time
Site:	Grand Gulf
Issue date:	06/30/1989
From:	Cottle W SYSTEM ENERGY RESOURCES, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
AECM-89-0110, AECM-89-110, NUDOCS 8907070279
Download: ML20246B316 (8)

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• Nuciety Cimfons June 30, 1989 U.S. Nuclear Regulatory Commission Mail Station P1-137 Washington, D.C. 20555 Attention: Document Control Desk Gentlemen:

SUBJECT:

Grand Gulf Nuclear Station Unit 1 Docket No. 50-416 License No. NPF-29 Current Cycle Safety Analysis, '

Revision 2 AECM-89/0110 Pursuant to 10CFR 50.71(e), System Energy Resources, Inc. (SERI) hereby transmits one signed original and ten copies of Revision 2 to the Current Cycle Safety Analysis (CCSA), for Grand Gulf Nuclear Station (GGNS).

The CCSA provides information on the current GGNS fuel cycle operations and contains analyses supporting operation and the current fuel reload.

Analyses included in the CCSA have been approved by the NRC either through specific review of SERI reload applications or as topical reports submitted by vendors.

The CCSA was developed to provide a convenient source for current accident and transient analyses. The CCSA is considered part of the Final Safety Analysis Report (FSAR), and is updated at least annually in accordance j with 10CFR 50.71(e). Revision 1 was submitted November 21, 1988 and reflected analyses to support GGNS Cycle 3. This update to CCSA reflects the fuel reload ,

accomplished during the third GGNS refueling outage performed in early 1989 (cycle 4).

As required by 10CFR 50.71(e)(2) and as authorized by SERI, I hereby certify, to the best of my knowledge, information and belief, that the

-information given in the attached CCSA, Revision 2, accurately presents changes made since the previous submittal, necessary to reflect information and analyses cubmitted to the NRC or prepared pursuant to NRC requirement.

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'If'you have any question, please contact this office.  :

1 Yours truly, j e.o Y W WTC:tkm Attachment cc: Mr. J. G. Cesare (w/0)

Mr. ' T ~. H. Cloninger (w/o)

Mr. R. B. McGehee (w/o)

Mr. N. S. Reynolds (w/o)

Mr. H. L. Thomas (w/o) i Mr. H. O. Christensen (w/o Mr. Stewart D. Ebneter (w/o)

Regional Administrator U.S. Nuclear Regulatory Commission Region II 101 Marietta St. , N.W. , Suite 2900 Atlanta, Georgia 30323 Mr, L. L. Kintner, Project Manager (w/o)

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 14B20 Washington, D.C. 20555 4

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L C A System Energy Resources, Inc.

Grand Gulf Nuclear Station Unit 1 Document No. 50-416 Revision 2 L

Instruction For Filing Revision 2 Insert the Revision 2 tab, transmittal letter and this instruction sheet to the back of the CCSA Volume.

Remove and insert the pages and topical reports listed below. Dashes (~~~)

in the remove or insert column indicate no action required.

REMOVE INSERT Page 2 (Table of Contents) Page 2 (Table of Contents)

Page 6 (Introduction) Page 6 (Introduction)

Page 7 (Introduction) Page 7 (Introduction)

Page8(Introduction) Page 8 (Introduction)

Page 9 (Introduction) Page 9 (Introduction)

,m, ANF-87-67, Rev 1 (all pages) ANF-88-149 (all pages)

.(j ANF-87-66, Rev 1 (all pages) ANF-88-150 (all pages)

MPEX-86/92 (all pages) ANF-88-183 (all pages)

Tab 8

[_~_~_~[ NESDQ-88-003 1

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CURRENT CYCLE SAFETY ANALYSIS (CCSA) 1 Table of Contents

Tab Description 1 Introduction 2 ANF-88-149 GGNS Unit 1 Cycle 4 Reload Analysis 2.

3 ANF-88-150 GGNS Unit 1 Cycle 4 Plant Transient Analysis

'4 XN-NF-86-37 (P) Generic LOCA Break Spectrum Analysis for BWR/6 Plants 5 XN-NF-80-19(P)(A) Exxon Nuclear Methodology Boiling Volume 4 Water Reactors: Application of the Revision 1- ENC Methodology to BWR Reloads 6 XN-NF-825 Sup. 2 BWR/6 Generic Rod Withdrawal Error Analysis, MCPR(P) for Plant Operational Within the Extended Operating Domain 7- ANF-88-183 Grand Gulf Unit 1 Reload XN-1.3, Cycle 4 Mechanical Design Report 2

8 NESDQ-88-003 GGNS Unit 1 Revised Flow Dependent Thermal Units Revision 0 AECM-87/0234 Transmittal letter Revision 1 AECM-88/0188 Transmittal letter and Revision 1 Insert Instructions Revision 2 AECM-89/0110 Transmittal letter and Revision 2 2 Insert Instructions O ,

NLSMISC89060801 - 2 Rev. 2 7/89 2 I"

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CPR -

Critical Power Ratio -

CS -

. Core Stability

. ELL. -

Extended Load Line FLE -

Fuel Loading Error FWCF -

Feedwater Controller Failure-FWHOS -

Feedwater Heater Out of Service ICF -

Increased Core Flow LOFWH -

Loss of Feedwater' Heating ~ transient LRNB - - Load Reject No Bypass (also known as GLR for Generator Load Rejection) .

RDA- - Rod Drop Accident RWE -

Rod Withdrawal Error SLMCPR - Safety Limit MCPR SLO -

Single Loop Operation 3.0 C_CSA Attachments The following lists attachments found in the CCSA.

1. ANF-88-149 Grand Gulf Unit 1 Cycle 4 Reload Analysis Event: SLO; LOCA; CS 2-

2. ANF-88-150 Grand Gulf Unit 1 Cycle 4 Plant Transient Analysis Event: LOFWH; FWCF; LRNB

3. XN-NF-86-37(P) Generic LOCA Break Spectrum Analysis for BWR/6 Plants Event: LOCA

4. XN-NF-80-19(P)(A) Exxon Nuclear Methodology Boiling Water Reactors:

Volume 4 Application of the ENC Methodology to BWR Reloads Revision 1 Event: FLE; f RDA NLSMISC89060801 - 6 Rev. 2 7/89 2

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/ \ 5. XN-NF-825 Supp 2 BWR/6 Generic Rod Withdrawal Error Analysis. j kJe MCPR(p) for Plant Operations within the Extended Operating Domain Event: RWE

6. ANF-88-183 Grand Gulf Unit 1 Reload XN-1.3, Cycle 4 Mechanical Design Report 2

7. NESDQ-88-003 GGNS Unit 1 Revised Flow Dependent Thermal Limits I

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i l i ) TABLE 1 Summary of Analyzed Events UFSAR SECTION EVENT PRIMARY PURPOSE CCSA ATTACHMENT 4.1 Methods General analysis ANF-88-149 techniques, summary XN-NF-80-19 Vol. 4 fuel description 4.2 Methods Fuel mechanical ANF-88-149 design description XN-NF-80-19 Vol. 4 ANF-88-183 4.3 Methods Nuclear design ANF-88-149 description XN-NF-80-10 Vol. 4 4.4 Methods Thermal-hydraulic ANF-88-149 design description XN-NF-80-19 Vol. 4 4.4.4.6 CS Define detect and ANF-88-149 em suppress region b,

5.2.2 Overpressure Overpressure ANF-88-150 protection SA ASME over- MSIV closure Max ANF-88-149 pressurization pressure ANF-88-150 6.3.3 ECCS Peak Clad ANF-88-149 Performance Temperature 15.0.3.3 Safety Limit SLMCPR ANF-88-150 2

15.1.1 LOFWH CPR with reduced FW ANF-88-150 temperature 2 15.1.2 FWCF CPR at rated ANF-88-150 150 FWCF CPR with ICF ANF-88-150 15D FWCF CPR with ELL ANF-88-150 1

15D FWCF CPR at Power below ANF-88-150 {

40% (w/o direct scram)

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,, ISD FWCF CPR - w/o bypass ANF-88-150 V)

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15D FWCF CPR with FWH05 ANF-88-150 I

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TABLE 1

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Summary of Analyzed Events - Continued UF5AR SECTION EVENT PRIMARY PURPOSE CCSA ATTACHMENT 15.2.2 LRNB CPR at rated ANF-88-150 15D LRNB CPR with ICF ANF,-88-150 15D LRNB CPR with ELL ANF-88-150 15D LRNB CPR at power below ANF-88-150 40%(w/odirectscram) 15D LRNB CPR with FWHOS ANF-88-149  !

Section 1.0 15C SLO Operation with one ANF-88-149 loop out of service 2-15.4.1, 2 CPR vs. Power XN-NF-825 Supp. 2

.e3 RWE 15.4.7 FLE CPR - Misloaded ANF-88-149 bundle XN-NF-80-19 Vol. 4 15.4,9 RDA Enthalpy ANF-88-149 deposition XN-NF-80-19 Vol. 4 15.6.5 LOCA Determine break XN-NF-86-37 location, limiting break size ,

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150 Flow Runout CPR & MAPLHGR ANF-88-150 l vs. Flow NESDQ-88-003 a

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l '\ \b;y y ADVANCEDNUCLEARFUELSCORPORATION GRAND GULF UNIT 1 RELOAD XN-1.3, CYCLE 4 MECHANICAL DESIGN REPORT O

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GRAND GULF UNIT 1 RELOAD IN-1.3. CYCLE 4 MECHANICAL DESIGN REPORT PREPARED BY: S. A /-9-E"6 W. S. Dunnivant Date Project Engineer CONCURRED BY: N A.'Repar , Mana r' Date

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APPROVED BY: - /[6!P'7 G. J. Dussv1 man, Manager Date Fuel Design

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GRAND GULF UNIT 1 Rn 04D IN-1.3. CYCLE 4

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MECHANICAL DESIGN REPORT r .

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l'. TABLE OF CONTENTS

'1 Seetion li.t]f .Pggg

1.0 INTRODUCTION

............................ 1 2.0

SUMMARY

. . . . . . . . . . . . . . . . . . . . . . . . . . . .. 2 2.1 Design Description Summary ..................... 2 2.2 Mechanical Design Summary . . . . . . . . . . . . . . . . . . . . . . 2 3.0 DESIGN CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

.4.0 MECHANICAL DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4.1 Fuel Rod Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4.1.1 Maximum Cladding Strain During. Steady State Operation ...... 6 4.1.2 Maximum Cladding Stress During Steady State Operation ...... 6 4.1.3 Anticipated Operational Occurrences Analysis . . . . . . . . . . . 7 4.1.4 Fuel Rod Internal Pressure . . . . . . . . . . . . . . . . . . . . 7 i

l 4.1.5 Fuel Pellet Centerline Temperature . . . . . . . . . . . . . . . . 7 4.1.6 Fuel Rod Cladding Fatigue .................... 8 4.1.7 Cladding Collapse ........................ 8 i 4.1.8 Fuel Rod Spacing . . . . . . . . . . . . . . . . . . . . . . . . . 8 4.1.9 Cladding Corrosion and Hydrogen Concentration .......... 9 o

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l ANF-88-183(NP), Rev. O Page 1 1

4 GRAND GULF UNIT 1 RELQAD IN-1.3. CYCLE 4 MECHANICAL DESIGN REPORT

1.0 INTRODUCTION

This report is a nonproprietary version of ANF-88-183. It has been edited to remove information proprietary to Advanced Nuclear Fuels Corp.

(ANF). It provides a summary discussion and references the detailed discussion of the design description, design criteria, technical bases, supporting analyses, and test results for the Advanced Nuclear Fuels i

Corporation Jet Pump Boiling Water Reactor Type 6 reload fuel for the Grand Gulf Unit 1 Nuclear Power Reactor.

r~N This report extends the assembly exposure limit of the Grand Gulf I XN-1.3 8x8 fuel to 39,000 mwd /MTU. The mechanical design of Grand Gulf 1 XN-1.3 is essentially the same as the generic ANF Type 4/5/6 design; thus, the majority of the mechanical design related sections of this report are covered l by specific references to generic mechanical design reports. Where applicable, the analysis has been extended, consistent with ANF's generically I I approved methodology, to cover the increased burnup of 39,000 mwd /MTU. .

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2.0 SupMARY The ANF 8x8 fuel design for Grand Gulf I XN-1.3 has been evaluated to allow operation up to a peak assembly exposure greater than 39,000 mwd /MTV.

The results of the evaluation indicate that the Design Criteria are met'. The fuel description mechanical design is summarized below.

I 2.1 Desian Description Summary The ANF 8x8 assembly design for Grand Gulf 1 XN-1.3 reload uses 62 fuel l

-rods and two centrally located water rods, one of which functions as a spacer j capture rod. Seven spacers maintain fuel rod spacing. The design uses a f quick-removable upper tie plate design to facilitate fuel inspection and i bundle reconstitution of irradiated assemblies.

The fuel rods are Zircaloy-2 cladding. The rods are pressurized, and contain either UO 2-Gd23 0 or UO2 . Natural uranium axial fuel blanketing, at the top and the bottom of the fuel column, is provided for greater neutron economy.

Two small modifications to the previous fuel design for Grand Gulf I have  ;

j been implemented, due to the increased fuel exposure of 39,000 mwd /MTU. Thus, the redesigned assembly maintains the same design margins at 39,000 mwd /MTU of

! those that exist at 35,000 mwd /MTV.

2.2 Mechanical Desian Summary The Mechanical Design Analyses were performed to evaluate cladding steady-state strain and stress, transient strain and stress, fatigue damage, i creep collapse, corrosion, hydrogen absorption, fuel rod internal pressure, differential fuel rod growth, creep bow, and spacer grid design. The analyses i 4

, justify irradiation to 39,000 mwd /MT peak assembly burnup.

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.- ANF-88-183(NP), Rev. O Page 4 3.0 DESIGN CRITERIA

-The ' detailed design criteria for the Advanced Nuclear Fuels Corporation

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Jet Pump Boiling Water Reactor for Grand Gulf 1 XN-1.3 reload fuel is given in XN-NF-85-67, Rev. 1, " Generic Mechanical Design for Nuclear Jet Pump BWR Reload Fuel".

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1 4.0 MECHANICAL DESIGN Two reports have already been issued to document the mechanical design analyses for the Grand Gulf 1 ANF 8x8 fuel. These reports are XN NF-83 25, Rev 1, " Grand Gulf 1 XN-1 Design Report Mechanical, Thermal Hydraulic, and Neutronic Design for Exxon Nuclear JP BWR/6 Fuel Assemblies", issued in August 1983, and XN-NF-85-67, Rev. 1, " Generic Mechanical Design For Exxon Nuclear Jet Pump BWR Reload Fuel", issued in September of 1986. The analyses in the first report were performed with the RODEX2 computer code and justified irradiation up to 33,000 mwd /MTU assembly burnup.

Analyses reported in the Generic Mechanical Design Report were performed using the computer code RODEX2A and justified irradiation up to 35,000 mwd /MTU assembly exposure. Both RODEX and RODEX2A codes have been approved for generic application by the NRC. The Generic Mechanical Design Report was fm submitted and approved for generic use by the NRC in 1986.

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This document reports the results of design calculations performed to support higher fuel assembly exposure than that reported previously. The calculations in this report used the RODEX2A computer code.

l The fuel assembly has been analyzed to a peak assembly exposure of 39,000

} mwd /MTV. The analyses have been performed assuming a design power history I identical to that used in XN-NF-85 67, Rev.1. At higher exposures, the power

, history was extended in such a way that the LHGR limit at higher exposure is i

! linearly extrapolated from that defined in XN-NF-85-67, Rev.1. I I

4.1 Fuel Rod Analyses f

Fuel rod analyses, where required, have been performed to verify adequate {

l performance of the fuel to 39,000 mwd /MTU assembly exposure. The exposures l assumed are conservative estimates of the maximum exposures to be reached with i the Grand Gulf 1 XN-1.3 8x8 reload fuel. The design power history used in I l 0 .

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ANF-88-183(NP), Rev. O Page 7

, tube wall is obtained at beginning-of-life (BOL). This conservative assumption leads to conservative stress and is also applicable to the 39,000 e < mwd /MTU assembly burnup. Consequently, the analysis results reported in Table 3.3 of XN-NF-85-67, Rev. I are applicable.

4.1.3 Anticipated Doerational Occurrences Analysis Two criteria are' imposed on the fuel rod to avoid fuel failure during power changes caused by anticipated. operational occurrences (AAO's). These are to limit the. cladding strain to less than 1% and to maintain the maximum pellet. temperature below melting. The A00's are assumed to produce a maximum nodal power equal to those defined in Figure 3.4 of XN NF-85-67, Rev.1. The analysis ' consists of calculating the. cladding strain and fuel centerline temperature at' the power levels defined in Figure 3.4 and verify that they remain below the design criteria.

' The calculations performed in support of XN-NF 85-67, Rev. I have been reviewed to determine if the higher exposure of the Grand Gulf 1 XN-1.3 8x8 fuel requires a reanalysis. It has been determined that the burnup at which the margin to the design criteria is the lowest is not at EOL, consequently, the analysis performed in support of XN NF-85-67, Rev. I are applicable to this design.

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I 4.1.4 Fuel Rod Internal Pressure The fuel rod- internal. pressure is limited to the design criteria pressure. The analysis in XN-NF-85-67, Rev. I have been extended up to a conservative estimate of the maximum assembly exposure of 39,000 mwd /MTU. The l

, analysis indicate that the maximum internal pressure is below the design criteria requirement.

4.1.5 Fuel Pellet Centerline Temperature The fuel pellet centerline temperature calculation performed in support of the results reported in XN NF-85-67, Rev. I has been reviewed. The review

_j indicates that the minimum margin against fuel melting, accounting for the

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4 ANF-88-183(NP), Rev. O Pane 9 rod bow has been evaluated for applicability at higher exposures. The correlation used by ANF to calculate fuel rod bow is exposure dependent. A small incremental increase in rod bow is calculated to occur between 35,000 j mwd /MTU and 39,000 mwd /MTU; the maximum fuel rod channel closure at 39,000 I mwd /MTU, however, provides ample margin to the channel closure that could affect the thermal performance. -

4.1.9 Claddina Corrosion and Hydrocen Concentration The current ANF design criteria is to limit the metal loss due to ]

corrosion. Hydrogen absorption is also limited by design criteria. The analysis performed in Reference I have been evaluated and the effects of l increasing the burnup to 39,000 MWD /MTU have been obtained.

The evaluation indicates that at the revised exposure, the cladding corrosion and hydrogen absorption will remain well below the design criteria.

I Figures 3.11 and 3.14 of Reference I provide the information for residence times consistent with 35,000 mwd /MTU assembly exposure. Assuming the residence time is increased by 12% to accun' alate the new design exposure, the evaluation indicates the design criteria are met.

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4.2 Fuel Assemb1v Evaluation

] The performance of the fuel assembly at 39,000 mwd /MTU has been I The structural strength, spacer design, and assembly growth have evaluated.

g been investigated. The results are as follows.

t 4.2.1 Structural Strenath The structural strength of tie plates, locking mechanism, and tie rods is not decreased with exposure. The analysis and test results previously reported in XN NF-85-67, Rev.1 are applicable.

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ANF-88-183(NP), Rev. 0 l Issue Date: 1/6/89 {

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SRAND GULF UNIT 1 J RELOAD XN-1.3. CYCLE 4 ECHANICAL DESIGN REPORT DISTRIBUTION I l

W. S. Dunnivant N. L. Garner (8) .

T. L. Krysinski l A. Reparaz Document Control (5) b 1

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s^ ! ADVANCED NUCLEAR FUELS CORPORATION i

GRAND GULF UNIT 1 CYCLE 4 PLANT TRANSIENT ANALYSIS I 4

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,~ ADWU4CEDNUCLEAR FUELSCORPORATION ANF-88-150 Issue Date: 11/11/G8 GRAND GULF UNIT 1 CYCLE 4 PLANT TRANSIENT ANALYSIS Prepared by s//AN R. A' Reynolds

1 BWR $fety Analysis i Licensing and Safety Engineering f Fuel Engineering and Technical Services Lu L / 'R. G. Grummer n-r-er 8WR Neutronics Neutronics and Fuel Management Fuel Engineering and Technical Services November 8, 1988

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CUSTOMER DISC 1. AIMER mIPORTANT MmCE REGAAONeG CO8mpf75 ANO USE OF TH18 DOCURENT M READ CAREPULLY I

Acwenced Nuoteer Fuele Corporeson's warrertnee and representetsons con.

oormng me suspect meser of mis occument are moos est form in the Agreemer.:

Denseen Advenoeg Nusteer Puede CorporeDon and tne Cuesomer pursuant to when vue assument e esues. Adoorergy, essest es omerwise escreesty pro-visse in euen Agreement. neener Aewenses Nuelser Fuste Corporanon nor any porgen ensig on es genest meses any warranty or represemenon, escreened or wnsees, men ressest a the assuracy, comosseeness, or usefuanees of me infor.

rneson sonomes a the oocument or that tne use of any intermanon. epoerorue.

memed or presses encloses an tmo eseumont well not intnnge artveteey ownec

.j, regnes; or assumes spy liegetene usin respect a the use of arty informenon, so-paranus, enames or presses disposee en the soeumerit.

The sitermellen conseined herein e for the este use of Cumomer in arcer a sweel impawment of ngMIe of Aewenced Nuclear Puese Corporenon in $

pelenes or inwonesne wman rney to inclusse wi the witormenon comewiec in mis secument, tne receMont. Dy its asospeance of Inse document. egrees not !c puamen er rnaise pushc use (in the gegent use of the term) of sucn informecon untd i 80 euinensee in wneng Oy Aewences Nucesar Fuses Corporenon or urmi efter six '

(g) rnonme tenseeng termanenen er espersoon of the slotseems Agrooment ene any essensen tnereof, unsees oenerwies espressay answiced e the Agrooment. No j ngnie or licenese wt or to any pamnte are impleed Dy me fumemng of mio cocu-mont.

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/,s Page i As). q TABLE OF CONTENTS Section .P.JLqi

1.0 INTRODUCTION

............................ 1 2.0 SUMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 3.0 THERMAL LIMITS ANALYSIS . . . . . . . .............. 12 3.1 Introduction . . . . . . . . ................. 12 3.2 System Transients .......... ............ 12 3.2.1 Design Basis ...................... 13 3.2.2 Anticipated Transients ................. 13 3.2.2.1 Loss Of Feedwater Heating ........... 13 3.2.2.2 Load Rejection No Bypass . . . . . . . . . . . . 14 3.2.2.3 Feedwater Controller Failure . . . . . . . . . . 15 3.2.2.4 Control Rod Withdrawal Error . . . . . . . . . . 16 3.3 Flow Excursion Analysis .................... 16 3.4 Safety Limit . . . . . . . . . . . . . . . . . . . . . . . . . . 17 3.5 Summary of Results . . . . . . . . . . . . . . . . . . . . . . . 17 3.5.1 Power Dependent Thermal Limits and Values . . . .. 18 O

3.5.2 Flow Dependent Thermal Limits and Values ........ 18 V

4.0 MXIMUM OVERPRESSURIZATION ..................... 29 4.1 Design Basis . . . . . . . . . . . . . . . . . . . . . . . . . . 29 4.2 Maximum Pressurization Transients ............... 29 4.3 Results ............................ 30

5.0 REFERENCES

............................. 33 i

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ANF-88-150 c Page ii LIST OF TABLES Table pjtgg 2.1 Results of Analyses ...................... 6 2.2 Operating Limit Coordinates .................. I 7

3.1 Grand Gulf Unit 1 Cycle 4 LFWH Data Summary . . . . . . . . . 19 LIST OF FIGURES Fiaure f,ggg 1.1 Power / Flow Map Used For Grand Gulf Unit 1 ME00 Analysis .... 3 2.1 Power Dependent MCPR Limits For Grand Gulf Unit 1 Cycle 4 ... 8 2.2 Power Dependent MAPFAC Factor For Grand Gulf Unit 1 Cycle 4 .. 9 2.3 Flow Dependent MCPR Limits For Grand Gulf Unit 1 Cycle 4 . . . 10 L 2.4 Flow Dependent MAPFAC Factor for Grand Gulf Unit 1 Cycle 4 . . 11 3.1 Analysis of LFWH Initial MCPR Versus Final MCPR .......20 3.2 Load Rejection Without Bypass (Power and Flows) . . . . . . . 21 3.3 Load Rejection Without Bypass (Vessel Pressure and Level) . . .'22 3.4 feedwater Controller Failure (Power and Flows) . . . . . . . . 23 3.5 Feedwater Controller Failure (Vessel Pressure and Level) . . . 24 3.6 Design Basis Radial Power Distribution . . . . . . . . . . . . 25 3.7 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis local Power Distribution (XN-12.99-5G3 Fuel) . . . . . . . . 26 3.8 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local Power Distribution (XN-2 3.21-6G4 Fuel) . . . . . . . . 27 3.9 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local Power Distribution (ANF-1.3 3.61-8G4 Fuel) . . . . . . 28 4.1 MSIV Closure Without Direct Scras (Power and Flows) . . . . . 31 4.2 MSIV Closure Without Direct Scram (Vessel Pressure and Level) .........................32 O

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ANF-88-150 Page iii ACKNOWLEDGEMENT The authors would like to acknowledge the following individuals for their contributions to the results reported in this document:

D. J. Braun ]

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

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1.0 INTRODUCTION

This report presents the results of analyses performed by Advanced Nuclear Fuels Corporation (ANF) for reload fuel in Grand Gulf Unit 1 Cycle 4 for operation within the Maximum Extended Operating Domain (ME00). The NSSS vendor performed extensive transient 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 analyses established conservative operating limits for MEOD operation. The initial reload of ANF fuel in Grand Gulf Unit 1 occurred in Cycle 2. In support of the initial reload of ANF fuel, extensive additional transient 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).

The objective of these analyses was to confirm the applicability of the Grand Gulf Unit 'l Cycle 3 Technical Specification MCPR at rated conditions,

, establish MAPLHGR limits for Cycle 4 operation, and establish revised thermal limits for off-rated conditions for the all-ANF core. An additionEl objective was to demonstrate that vessel integrity is protected during the most limiting Cycle 4 pressurization event.

Changes from : Cycle 3 to Cycle 4 for Grand Gulf Unit 1 include the discharge of remaining GE fuel, an additional reload of ANF fuel and an increase in cycle energy from 1420 GWd to 1698 GWd while maintaining the cycle length at 18 months. The reload fuel for Cycle 4 is the same as that for Cycles 2 and 3 except for changes in enrichment, the number of rods per bundle containing gadolinia, and the gadolinia concentration (Reference 3).

The Cycle 4 transient analysis consists of recalculation of the limiting transients at state points having the least margin to operating limits to confim that the effects of the Cycle 4 changes on transient results are small and establish appropriate limits. Rem'ysis of the limiting transients for Cycle 4 assures that the less limiting transients which were previously addressed will continue ta be protected by the established operating limits for Cycle 4. The power / flow conditions analyzed in Cycle 3 and Cycle 4 are presented in Figure 1.1.

1,.

ANF-88-150

• Page 2 L

The MCPR p , MCPRf , and MAPFACf limits have been revised to reflect ANF calculated limits using ANF methodology. The Grand Gulf Unit 1 power and flow dependent MCPR analyses for Cycle 4 were performed at limiting power / flow conditions. Flow dependent MAPFAC analyses were performed on the 100% rod line with the initial core flow varying from 40% to 80% of rated flow. )

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, q Page 4 ,

V 2.0 SUMARY

)

The results of the Grand Gulf Unit 1 Cycle 4 transient analyses support I appropriate thermal limits for the first Grand Gulf all-ANF core. ANF thermal

)

limits have been provided for MCPRp above 40% power that are based on generic ANF Control Rod Withdrawal Error (CRWE) analyses (Reference 4). Additionally, ,

MCPRf limits and MAPFACf values (Reference 12) have been confirmed for both

" loop manual" and "non-loop manual" operation. j Minor differences in the maximum local peaking as a function of exposure l for the different ANF fuel types require that different MAPLHGR limits be monitored. These MAPLHGR limits are consistent (differ by maximum local peaking factor) with the LHGR limits so that at reduced power and/or reduced flow the LHGR limit will be protected by the MAPFACf and MAPFACp multipliers l on MAPLHGR. Cycle 4 reload fuel MAPLHGR limits are included because of the slight changes in the local peaking factor.

t0

'd Table 2.1 summarizes the transient analyses results applicable to Grand Gulf Unit 1 Cycle 4. These results, together with the Grand Gulf Unit 1 Cycle 4 calculated safety limit MCPR of 1.06, support continued use of the existing 1.18 MCPR operating limit (at rated conditions) for Cycle 4 operation.

The plant transient and safety limit analyses results reported herein support revising the Cycle 3 power dependent Minimum Critical Power Ratio (MCPR p ) so that it is based on the generic CRWE results of Reference 4 above 40% power and supports the continued use of Cycle 3 limits below 40% power.

The power dependent Maximum Average Planar Linear Heat Generation Factor (MAPFAC p ) for Cycle 3 is confirmed for Cycle 4 operation. The revised MCPR p limits, the MAPFAC p confinnation, and the re si;! + < cf ANF's analyses are presented in Figures 2.1 and 2.2, respectively.

The flow dependent Minimum Critical Power Ratio (MCPR f ) and the results ,

n of ANF's analysis are presented in Figure 2.3. The flow dependent Maximum )

b Average Planar Linear Heat Generation Rate Factor (MAPFAC f ) is presented in I

p.

E m

4: 9 uk, /., w ANF-88-I (s .y wa "1 Pace h

..q 7; e, m; = , ,3 7 ,,

Figure 2.4h :Thesejflow dependent MAPFACf values and MCPRf limits have eer revised from Cycle 3 to support Cycle 4 in both the ' loop manual

• and *-

"non-loop manual" mode of operation. These curves are based on conserva; maximum core flow rates. Table 2.2 shows the coordinates used to construc Figures 2.1 through 2.4.

.i The results of the maximum system pressurization transient analysis e presented in Table 2.1. The safety valve pressure setpoint tolerances

C this analysis. have been increased to 65 for Cycle 4; the results show th:

+ the Grand ' Gulf Unit I safety valves have sufficient capacity and perfefman::

l, with tho' increased setpoint tolerances to protect the vessel pressurei,l,*safet limit of'1375 psig during cycle 4. _ %R M

wm The fuel related Technical Specification limits for Cycle '4 operaticr.

n. . +.

. included in the reload analysis report (Reference 3).

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\ }. Table 2.1 Results of Analyses THERMAL LIMITS j Transient Delta CPR Loss of Feedwater Heating (all conditions) 0.11 Control Rod Withdrawal Error (1005 power, Ref. 4) 0.10

, Feedwater Controller Failure (104.2/108)* 0.04 4 Load Rejection Without Bypass ,

1 Power /1 Care Flow 104.2/100* 0.12 104.2/73.g,- 0.02 40/100***' 0.15 {

O -

40/100 25/73.00 0.32 0.93 0.69 25/40. ..

104.2/1002 0.09 92.5/67 t 0.02 70/40i 0.04 55/40 g

• 0.03

, 40/40 ,' O.03 MAXIMUM SYSTEM PRES $URIZATION Mw /4:..p 3.. 7 Core 1"^gn; Flow Vassal Lower Plenum Steam Dome

.. _ gif iy:gRM MSIV; Closure 1 G 9 p 194;2/106,. '

1298 psig 1271 psig s N e;I4 m # 104.2/73.8 1297 psig 1280 psig 3 4,g+gifj(

?

r.: w 3J.e a,

• 104.2% power /1005 core flow is used for~ the Reload Licensing Analysis (RLA) conditions tc conservatively bound 1001 power /10SE core flow.

• • Direct scram'on turbine trip enabled

• • • Direct scram M t'arbine trip disabled.

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ANF-88-150 j (g) 4;-

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~ Table 2.2 Operating Limit Coordinates 3 1 i

i GRAND GULF UNIT 1 CYCLE (

MCPRin) Limits MAPFACfo) Limits (Figure 2.1) (Figure 2.2)

Percent of Rated Percent of Rated Core Power MCPRin) Core Power MAPFAC(o) 100 1.18 100 1.0 -

70 1.24 40 0.60* -

70 1.40 40 0.69*** {

40 1.48* 24.4 ' 0.57**3 40 1.85 24.4 O.61^ O;. .

40 2.10** <

25

• Core Flow' > 505 .; 4 2.05**

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• Core Flow < 505 -

• • Core Flow I 505 ,,

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McMIfl L' a ts HAMhtIf) ~ Limits (Figure U.:l) -(F gare 2.4)

Percent'of Loop- Non-loop Percent of- ~ - Loop Non-loop Rated Corei Rated Core tManual Manual Manual. ~ . Manual F1 ogL,,,,, McMff1 EPR(f) F1ow  % MPFACf f) MAPFACff) 30 1.55 1.73 110 11.00- 1.00 40 -1.41 1.57 91.0 1.00 1.00 50 1.31- 1.44 90.0. >

1.00 .992 60 a 1.24 84.3 :1.00~ -

70 .N s % I. h;D, %w 1.35 80.0 # .977 .904 i- 70.0 .928 .827

'73.4Mii3tIW(k'h(Y1.27 4

80 'D 4:R.;1;M7l* w 1.21 60.0 .880 .757 i, 86.3 Q W l.N 50.0 .837 .695 105 , !Q 4 1.18 6@ h l.18 -

40.0. , . 794 .638 c 9??hy 2;rl.18 30.0 .752 .586 .

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4 ANF-88-150 ipD Page 12 q) .

3.0 THERMAL LIMITS ANALYSIS 3.1 Introduction The scope of the thermal limits analysis includes system transients, localized core events, and safety limit analysis. Results of these analyses are used to confirm / establish power and flow dependent MCPR and MAPFAC values, COTRANSA (Reference 5), XCOBRA-T (Reference 6), and XTGBWR (Reference 7) are the major codes used in the thermal limits analyses as described in ANF's THERMEX Methodology Report (Reference 8) and Neutronics Methodology Report (Reference 7). COTRANSA is a transient system simulation code which includes an axial one-dimensional neutronics model. XCOBRA-T is a transient thermal-  !

hydraulic code used in the analysis of thermal margins of the limiting fuel assembly. XTGBWR is a three-dimensional steady state core simulation code which is used for Control Rod Withdrawal Error (CRWE), Loss of Feedwater

( Heating (LFWH), and flow excursion events. 4 3.2 System Transients Reviscd thermal limits were established for the all-ANF Cycle 4 core.

. Figure 1.1 shows the ten power / flow conditions that were analyzed in support  !

of the Cycle 4 reload. These state points were esiyzed for Grand Gulf Unit 1 Cycle 4 using COTRANSA. The Load Reject No Bypass (LRNB) pressurization l,

transient analysis was performed at each of the ten state points. The feedwater Controller Failure (FWCF) analysis was performed at 104.2% power and 108% flow. ASME pressurization analyses were performed at state points 104.2%/108% and 104.2%/73.8%. LFWH analyses were performed with XTGBWR at f five different state points for eight exposures. The generic analysis for f control rod withdrawal error is applicable to Grand Gulf Unit 1 Cycle 4.

l These analyses show less restrictive results or little change from the Cycle 3 f analyses due to Cycle 4 changes, thus justifying that the less limiting transients not analyzed for Cycle 4 will continue to be protected.

l lO

\v,

ANF-88-150 g .

Page 13-3.2.1 Desian Basis The LRN8 and FWCF transients have been determined to be most limiting at end of full power capability when control rods are fully withdrawn from the core. 'The delta CPR calculated for end of full power conditions is conserva-tive for cases where control rods are partially inserted. The analysis for Grand Gulf Unit I with ME00 was performed using conservative analytical limits for trips and setpoints. Events initiated at core powers below 40% rated were analyzed with the direct scram due to turbine control and stop valve fast closure disallowed. and with the recirculation pump high to low speed transfer disabled.~ The Loss of Feedwater Heating (LFWH) transient has been analyzed throughout the cycle at state points which bound the ME00 operating map.

3.2.2 Anticipated Transients ANF's tranMent methodology report for jet pump BWRs (Refe-ence 5)

. considered eight categories of anticipated transients. The most limiting transients were evaluated at various power / flow points within ME00 to verify the power dependent' thermal margin for Grand Gulf Unit 1 Cycle 4.

. The limiting transients analyzed for Grand Gulf Unit 1 Cycle 4 were:

o Loss of Feedwater Heating. ,

o Load Rejection No Bypass I o Feedwater Controller Failure Other transients are inherently non-limiting or bounded by one of the above as shown in the NSSS vendor ME00 analyses for Cycle 1 and the ANF Grand Gulf Unit 1 Cycle 2 analyses. Control Rod Withdrawal Error is an exception in that it has been analyzed generically. 4 3.2.2.1 Loss of Feedwater Heatina Analysis of the loss of feedwater heating event was performed to reflect reactor operation over the ME00 operating power versus flow map and conditions

. anticipated during actual Grand Gulf reactor operation. l'

ANF-88-150 Page 14

'v)

?

Calculations performed for Cycle 4 assumed a conservative reduction of ,

100'F in the feedwater temperature. Table 3.1 provides the conditions of each )

case analyzed in terms of cycle exposure, core power, and core flow. The initial and final MCPR values are presented for each case.

Analysis of the data revealed a strong correlation between the initial and final MCPR. A least squares fit of these data resulted in a linear relationship such that:

MCPR(initial) = -0.0514 + 1.1130

• MCPR(final).

In order to conservatively bound all of the calculated data, the largest deviation between the calculated and fitted results were applied to the least squares fit such that the LFWH MCPR operating limit is defined by OLMCPR(LFWH) - -0.0112 + 1.1130

• SLMCPR w .

This bounding relationship is presented in Figure 3.1. Substituting the SLMCPR of 1.06, the MCPR operating limit for the LFWH event for all operating conditions is 1.17.

3.2.2.2 Load Reiection No Bvoass The Load Rejection No Bypass (LRNB) event is the most limiting of the class of transients characterized by rapid vessel pressurization for Grand Gulf Unit 1. The load rejection causes a fast closure of the turbine control valves. The resulting compression wave travels through the steam lines into the vessel and creates the rapid pressurization condition. A reactor scram is initiated by the fast closure of the control valves as well as the recircula-tion pump high to low speed transfer. Condenser bypass flow, which can mitigate the pressurization effect, is not allowed. The excursion of the core power due to void collapse is primarily terminated by reactor scram and void l

growth due to the recirculation pump high to low speed transfer. ]

O i

t i

ANF-88-150

/ Page 15 3 Figures 3.2 and 3.3 present the response of various reactor and plant-  !

parameters to the LRNB event initiated at the Reload Licensed Analysis condi-tion (104.2% power /108% core flow). Table 2.1 lists the delta CPRs- for this transient at the other power / flow conditions analyzed for Grand Gulf Unit 1.

3.2.2.3 Feedwater Controller Failure The failure of the feedwater controller to maximum demand (FVCF) is the most limiting of the vessel inventory increase transients. Failure of the j feedwater control system to maximum demand would result in an increase in the coolant level in the reactor vessel. Increased feedwater flow results in lower temperatures at the core inlet, which in turn cause an increase in core power level. If the feedwater flow stabilizes at the increased value, the core power will stabilize a: a new, higher value. If the flow increase continues, the water level in the downcomer will eventually reach the high level setpoint, at which time the turbine stop valve is closed to avoid damage to the turbine from excessive liquid inventory in the steamline. The high water level t.4p also initiates reactor scram, and recirculation pum: trip.

Turbine bypass is assumed to function for this analysis, mitigating the consequences to some extent. The core power excursion is terminated by the same mechanisms that end the LRNB transient.

Figures 3.4 and 3.5 present the response of various reactor and plant parameters to the FWCF event initiated at the Reload Licensed Analysis condi-tion (104.2% power /108% core flow) . The delta-CPR for this event was calculated to be 0.0/,, indicating a MCPR operating limit requirement of 1.10 for the event. In support of the Cycle 2 reload, FWCF transients were also analyzed without condenser bypass and with a 100*F reduction in feedwater temperature. it was shown that these conditions had a minor impact on the delta-CPR and that significant margin exists to limits (Reference 2). Since the FWCF transient analyzed for Cycle 4 results in a delta-CPR similar to that obtained for Cycle 3, it is not necessary to repeat the other FWCF transients for the Cycle 4 reload. In Reference 2, the LRNB transient was shown to bound a'll FWCF transients at rated and off-rated conditions.

4 ANF-88-150 p Page 16

'Q,]

3.2.2.4 Control Rod Withdrawal Error Reference 4 documents ANF's generic CRWE analysis for Grand Gulf Unit 1-operation within the ME00. This generic analysis is applicable to Cycle 4.

The results from Reference 4 and the results of the Cycle 4 system transient confirmatory analyses show that the CRWE is limiting above 40% of rated power.

The Grand Gulf Cycle 4 CRWE based limits, and analysis results are presented in Figure 2.1. These data demonstrate that the CRWE limits may be used as a basis for the Grand Gulf Unit 1 MCPR p Technical Specification limits in Cycle 4 above 40% power. The rated condition MCPR operating limit remains unchanged at 1.18.

3.3 Flow Excursion Analysis The flow excursion transient is analyzed to detemine the flow dependent thermal limits and values (MCPRf and MAPFACf ). This transient is analyzed by assuming a failure of the recirculation flow control system such that the

( recirculation flow increases slowly to the physical maximum attainable by the equipment. Two modes of oper: tion are analyzed for Grand Gulf Unit 1 Cycle ?,

" loop manual" and "non-loop manual." These two modes of operation correspur.d to a single recirculation loop flow excursion event and a dual recirculation loop flow excursion event, respectively.

For both flow excursion events, the Cycle 4 MCPRf confirmation analysis of the power ascension associated with a flow increase was determined to be conservative when compared to the MCPRf operating limit (Reference 12). In the confirmation calculation the change in critical power along the ascension path was calculated with XCOBRA (Reference 8). Peaking factors were selected such that the bundle with the least margin would reach the safety limit MCPR of 1.06 at the maximum flow. Figure 2.3 presents the MCPRf limits for maximum achievable core flows for both events, assuming that the recirculation system equipment is capable of 110% of rated.

The Cycle 4 MAPFACf confirmation analysis of the power ascension associated with a flow increase was determined to be conservative when compared to MAPFACf limits. Confirmation calcula'tions were performed for

'Y

't fj ANF-88-150 Page 17

.both " loop manual" and "non-loop manual"~ modes of operation with XTG8WR. .The Cycle 2. analysis of reduced flow LHGR limits (MAPFACf values) was performed statistically based upon a wide variety of initial conditions.

~

2' For "non-loop-

. manual" operation, _ confirmatory calculations were performed for Cycle 4 at 0.5, 2.0, 3.5, 6.0, 8.0, 9.5, and 11.0 GWd/MTU with XTG8WR to simulate the flow runup event from 40% of rated flow. For " loop manual" operation, these analyses simulated a flow runup where the initial flow was varied from 40% to 80% _of rated. . Final flow was established based on the mode of operation

-(Reference 12). Figure 2.4 presents the MAPFACf limits for maximum achievable core flows for both events as well as the results of the Cycle 4 analysis.

3.4 Safety Limit-The safety limit MCPR is defined as the minimum value of the critical power ratio at which the fuel could be operated, with the expected number of rods in boiling transition not exceeding 0.1% of the fuel rods in the core.

O The safety limit is the minimum critical power ratio which would be permitted to occur duringlthe limiting anticipated operational occurrence. The safety limit MCPR for all fuel types in Grand Gulf Unit 1 Cycle 4 operation was confirmed to remain at 1.06 using the methodology presented in References 9 and 11.

The input parameter values for uncertainties used in the safety limit MCPR analysis are unchanged from the Cycle 2 analysis presented in Reference 2. Cycle 4 specific design basis radial and local power distribu-tions are shown in Figures 3.6 to 3.9.

3.5 Summary of Results The results of the Grand Gulf Unit 1 Cycle 4 thermal limits analysis

+

confirm the Cycle 3 safety limit MCPR of 1.06 and a MCPR operating limit of 1.18 at rated conditions.

.O l

ANF-88-150 f

A Page 18 l l

3.5.1 Power Denendent Thermal Limits and Values The power dependent MCPR limit (MCPR p ) protects against exceeding the srfety limit MCPR during anticipated operational occurrences from off-rated power conditions. Tha MCPR p limit is determined by adding the delta CPR for j the limiting event to the calculated safety limit MCPR.

The ' power dependent MAPFAC (MAPFAC p ) is used to protect against both fuel melting and 1% clad strain during anticipated system transients from off-rated power conditions. The conservative LHGR values for protection against fuel failure during anticipated operational occurrences are given in Reference 10.

The results are then presented in a fractional form for application to the MAPLHGR value. The MAPLHGR is developed to be consistent with the LHGR limit through consideration of the maximum local peaking factor.

The MCPR p limits and MAPFACp values for Cycle 4 are shown to bound the results of ANF's analysis in Figures 2.1 and 2.2, respectively. Above 40%

power the MCPR p limit is based on the ANF CRWE limit of Reference 4. Below 40% power, the Cycle 3 MCPR p limit remains applicable to Cycle 4. The Cycle 4 MAPFAC p value remains unchanged from Cycle 3.

3.5.2 fA' Pwendent Thermal Limits and Values The ' i dependent MCPR limit (MCPR f ) protects against exceeding the safety lim, MCPR for flow excursion events. The results of the MCPRf analysis for Grand Gulf Unit 1 Cycle 4 are presented in Figure 2.3. The flow dependent MAPFAC (MAPFAC f ) protects against both fuel melting and 1% clad strain. The MAPFACf values to be used in Cycle 4 are presented in Figure 2.4.

The flow dependent thermal limits were confirmed to be conservative for Cycle 4 operation.

l LO o _ - - - -

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,m ANF-88-150 Page 19

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1 Table 3.1 Grand Gulf Unit 1 Cycle 4 LFWH Data Summary l I

Initial State Final State 1

Cycle Total Core Total Core Core Total Core Total Core Core l Exposure Power Flow Minimum Power Flow Minimum )

(GWd/MT) MWt J, hlb /hr) CPR MWt (M1b/hr) CPR 0.500 3833.0 118.13 1.35 4369.2 118.13 1.27 0.500 3066.4 118.13 1.68 3506.1 118.13 1.57 1 0.500 3833.0 84.38 1.20 4356.2 84.38 1.14 0.500 2376.5 34.88 1.41 2736.8 34.88 1.31 0.500 1533.2 112.50 3.17 1735.8 112.50 2.91 2.000 3833.0 118.13 1.35 4387.9 118.13 1.25 2.000 3066.4 118.13 1.68 3514.5 118.13 1.54 2.000 3833.0 84.38 1.24 4379.7 84.38 1.16 2.000 2376.5 34.88 1.41 2753.2 34.88 1.29 2.000 1533.2 112.50 3.25 1743.8 112.50 2.93 3.500 3833.0 118.13 1.36 4364.9 118.13 1.28 3.500 3066.4 118.13 1.69 3500.2 118.13 1.57

,-m 3.500 3833.0 84.38 1.19 4360.2 84.38 1.13 s

'-) 3.500 3.500 2376.5 1533.2 34.88 112.50 1.37 3.27 2733.9 1728.0 34.88 112.50 1.27 3.00 5.000 3833.0 118.13 1.31 4359.8 118.13 1.23 5.000 3066.4 118.13 1.63 3498.4 118.13 1.51 5.000 3833.0 84.38 1.18 4358.6 84.38 1.12 5.000 2376.5 34.88 1.36 2735.0 34.88 1.26 5.000 1533.2 112.50 3.17 1725.1 112.50 2.89 6.500 3833.0 118.13 1.29 4357.0 118.13 1.21 6.500 3066.4 118.13 1.60 3490.5 118.13 1.49 6.500 3833.0 84.38 1.15 4349.7 84.38 1.09 6.500 2376.5 34.88 1.36 2731.3 34.88 1.24 6.500 1533.2 112.50 3.02 1728.3 112.50 2.76 8.000 3833.0 118.13 1.29 4348.8 118.13 1.20 8.000 3066.4 118.13 1.59 3486.4 118.13 1.47 8.000 3833.0 84.38 1.13 4345.5 84.38 1.07 8.000 2376.5 34.88 1.35 2729.3 34.88 1.24 8.000 1533.2 112.50 2.95 1723.9 112.50 2.70 9.500 3833.0 118.13 1.32 4337.3 118.13 1.23 9.500 3066.4 118.13 1.63 3471.8 118.13 1.51 9.500 3833.0 84.38 1.16 4336.3 84.38 1.09 9.500 2376.5 34.88 1.39 2718.6 34.88 1.30 9.500 1533.2 112.50 3.02 1716.7 112.50 -2.78 11.000 3833.0 118.13 1.30 4329.4 118.13 1.21 11.000 3066.4 118.13 1.60 3467.1 118.13 1.43 11.000 3833.0 84.38 1.15 4337.1 84.38 1.08 rw 11.000 2376.5 34.88 1.40 2713.7 34.88 1.30

't) 11.000 1533.2 112.50 2.96 1718.7 112.50 2.71

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LL L  : ML  : M  : M

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1.04  :

1.02 : 0.00 : 0.95 : 1.02  : 1.05 : 1.01  :

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g

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ML  : M  :  :

H H  : H  : H  : MLl* : ML  :

0.99 : 0.99 : 1.03  : 1.02 : 1.03 : 1.04 : 0.94  : 1.00 :

L  : MLl* : M  :

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Figure 31 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local Power Distribution (XN.1 2.99 5G3 Fuel) f

• - Gadolinia Location i N

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M  : M  : ML - -  !

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Figure 3.8 5 9n asis Local Grand Gulf Unit 1 C !fN 2 3 216G4 F el)

PowerDistributionf

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ANF.88-150 Page 28 3A . . . .......***.******.........

LL  : L  : ML  : M  : M  : M  : ML  : L  :

0.87  : 0.89 : 0.98 : 1.02 : 1.02 : 1.04 : 1.01 : 0. 94 :

L  : M  : M*  : H  : H  : M*  : M  : ML  :

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

0.98  : 0.96  : 1.04  : 1.01  : 1.00 : 1.03  : 0.95  : 1.04 -

. M  : H  : H  : W  : M  : H . H  : M  :

1.02  : 1.06  : 1.01  : 0.00 : 0.88 : 1.00 : 1.06 : 1.02  :

M  : H  :. H  : M  : W  : H  : H  : M -

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1.02 .

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M  : M*  : H  : H  : H  : H  : M*  : M  :

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

1.01  : 1.01  : 0.95  : 1.06 : 1.06  : 0.95 -

1.02  : 1.01 -

L . ML  : M  : M  : M  : M Mt . L

, 0.94  : 1.01  : 'J.04  : 1.02  : 1.02  : 1.04 1.01 0.94 Figure 3.9 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local Power Distribution (ANF-1.3 3.61-8G4 Fuel)

• - Gadolinia Location i i

! U

l

-s ANF-88-150

( ') Page 29 v

4.0 MAXIMUM OVERPRESSURIZATION Maximum system pressure has been calculated for the containment isolation event (rapid closure of all main steam isolation valves) with an adverse scenario as specified ir, the ASME Pressure Yessel Code. This analysis showed that the Grand Gulf Unit 1 safety valves have sufficient capacity and performance to prevent pressure from reaching the established transient pressure safety limit of 110% of design pressure (1.1 x 1250 - 1375 psig).

The maximum vessel pressures at the most limiting power / flow point (104.2% power /108% flow) are shown in Table 2.1.

4.1 Desion Basis ,

During the transient, the most critical active component (direct scram on MSIV closure) was assumed to fail. The event was terminated by the high flux scram. Credit war taken for actuation of only 13 of the 20 safety / relief valvesi 6 in the relief mode and 7 in the safety mode. The calculation was Q,, performed with ANF's plant simulation code, COTRANSA, which includes e axial one-dimensional neutronics model. The safety valve analysis setpoints for this calculation included a 6% tolerance. Relief valve setpoints for this analysis remain unchanged from Cycle 3.

4.2 Maximum Pressurization Transients Scoping analyses described in Reference 5 found the closure of all main steam isolation valves (MSIVs) without direct scram to be limiting. The MSIV closure was found to be limiting when all transients are evaluated on the same basis (without direct scram) because of the smaller steam line volume associated with MSIV closure. Though the closure rate of the MSIVs is substantially slower than turbine stop or control valves, the compressibility of the additional fluid in the s. team lines associated with a turbine isolation causes these faster closures to be less severe. Once the containment is isolated, the subsequent core power production must be absorbed in a smaller volume compared to that of a turbine isolation resulting in higher vessel  !

pressures.

(

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- - - - - 4

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ANF-88-150 j .. Page 30 -

4.3 Results The results of the maximum system pressurization analysis are presented in Table 2.1. Figures 4.1 and 4.2 present the response of various reactor i

and plant parameters during the MSIV closure event from 104.2% power /108% '

l flow. Thes.e results show that the Grand Gulf Unit I safety valves have sufficient capacity and performance to protect the previously established maximum vessel pressure safety limit of 1375 psig for Cycle 4. Two state points were analyzed in order to cover the ME00 range for full power operation.

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u m 5'.0; REFERENCES li? Lester L. Kintner, USNRC, Letter to 0. D. Kingsley, Jr., MP&L, " Technical Specification Changes to Allow Operation with One Recirculation Loop and

~

T. Extended Operating Domain," August 15, 1986. j I. " Grand Gulf Unit 1 Cycle 2 Plant Transient Analysis," XN-U-86 -36, Revision 3, Exxon Nuclear Company, Inc., Richland, WA, August 1986.

3 .' " Grand Gulf Unit 1 Cycle 4 Reload Analysis," ANF-88-149, Advanced Nuclear 1 -Fuels Corporation, Richland, WA, November 1988. )

. . w i 4B "BWR/6 Generic Rod Withdrawal Error Analysis; MCPR for Plant Operations

% within tho' Extended Operation Domain," XN-NF-82!rf P)(A), Supplement 2, S- Exxon Nuclear Company, Inc., Richland, WA, October 1986.

.A ]

'5$ " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactor,"

1 $ 'XN-NF-79-71(P), Revision 2, including Supplements 1, 2 & 3(A), Exxon .

@ Nuclear Company, Inc., Richland, WA, November 1981.

k"XCOBRA-T: A Computer Code for BWR. Transient Thermal Hydraulic Core yp Analysis," IN-NF-84-105f Pif A), Volume,1. Exxon Nuclear Company, Inc.,

y gg'Richland,WA, February 1987.

3:,

TNA " Exxon Nuclear Methodology for BoilingEWater Reactors: Neutronics y b Methods for Design and Analysis," XN-NF-Al-19fA), Volume 1, Exxon Nuclear W Company, Inc., Richland, WA,- March 1983;-. as supplemented by Letter,

.4 R. A. Copeland, Advanced Nuclear Fuels, to M. W. Hodges, USNRC, " Void Q History Correlation," RAC:058:88, September,13, 1988.

.p.

18A ." Exxon Nuclear. Methodology for Boiling Water Reactors THERMEX: Thermal m Limits Methodology Summary Description," 10-NF-80-19fP)(A), Volume 3, f Revision 2. Exxon Nuclear Company, Inc., Rich'and, WA, January 1987.

ESP '" Exxon Nuclear Critical Power Methodology: for Boiling Water Reactor,"

$ XN-NF-524f P' (A14evision 1, Exxon ' Nuclear Company, Inc., Richland, WA, qWW November..1903 +e t $ n -

- 10$ " Generic Mechanical' Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"

Q11N-NF-85-67(P (A%? Revision 1, Exxon Nuclear Company, Inc., Richland, WA, 1 5 September 1980.wa.t~t g

110 " Exxon Nuclear Meiihodology for Boiling" Water Reactors: Application of the y ENC Methodology to BWR Reloads, IN-0F-80-19fP)(A), Volume 4, P . Revision 1, Exxon Nuclear Company, Inc., Rich'and, WA, June 1986.

t

- 12. c " Grand Gulf Nuclear Station Unit 1 Revised Flow Dependent Thermal

, Limits," NESD0 88-003, MSU System Services Inc., November 1988.

y 44 1 s,

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n ANF-88-150 Issue Date: 11/11/88 GRAND GULF UNIT 1 CYCLE 4 PLANT TRANSIENT ANALYSIS 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

{q g . M. J. Hibbard T. L. Krysinski-A. Reparaz R. S. Reynolds S. E. State' '

R. B. Stout C. J. Volmer G. N. Ward:

H. E. Williamson:

4 SERI/N. L. Garner (40)

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GRAND GULF UNIT 1 CYCLE 4 RELOAD ANALYSIS O

NOVEMBER 1988 5

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ADWWCEDhRICLEARFUELS CORPORATION ANF-88-149.

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

Prepared by 1,& $/tff[

~~ ~R4 5. Reynolds *'

BWR(JafetyAnalysis Licensing and Safety Engineering Fuel Engineering and Technical Services h/

,... . w R.' G. Grummer l I 'lE I BWR Neutronics Neutronics and Fuel Management Fuel Engineering and Technical Services November 8, 1988 i

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custoutn otscLAiuEn NANT 880TICE S*T Z CONTENTS ANO USE OF THis DOCUuSNT PLEASE READ CAMEPULLY Aavenone Nuceer Fume Corooreson's warrermee anc reoresematens con.

commg tne g menst of me secument are thoes set form in the Agrooment tenmoon Aewences Nucear Puo. Corporeson ene the Caserner pursuant to '

unen the assument e tenuse. Aeoonsney, except as omensee engrosary pro.

visse a cuan Ayooment. neener Aswances Numeer Fuse Corporeson nor any pereen among on se nones meses any warremy or reoressmason, enerossee w impeles. uset ressent to the enouresy, compostenses, or usefumees of me efor.

mean eeroamme en mas escument. or the me use of any iriformason, asceratus, l

memos or seneses essesses a me escument weil not mfrmge onvetery ownea regnes: or soeumes any heesene men reasset a me use of any esormuen, ao.

paresus, nesmos or peceso essesse in tne secument The weermeson semesres herein a ter the sees use of Casomer i l

in weer e ame meermem a none of Aeveness Nummer Fue Corooreson in .

pasense or wwensons wnsen eney as mesumes e the vitermenon comunee m me eseumont. the resseent. Dy e acessenwe of me escument agrees not to pienen er mese suche use 6n me pasent use 9f the term) of sucn miormenon until so enanenses a wnnng Oy Aewenome Nuneser Fusse Careersson or unni after six i

(c) merene inesunng termmason or espresen of the aforeams AG,w ano any essenmon thereof, umees onnenuse empresor proviese in me Agreement. No ngme er 16eenees e or to any nesenes are imesse ey me fumeneg of mio accu-rnent.

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.. 1 ANF-88-149 ic) i (,/.

Page i TABLE OF CONTENTS Section .P_iLqt

1.0 INTRODUCTION

............................ I 2.0 FUEL MECHANICAL DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . . 5 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . 6 3.2 Hydraulic Characterization . . . . . . . . . . . . . . . . . . . 6 3.2.3 Fuel Centerline Tem erature . . . . . . . . . . . . . . . 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 NUCLEA R DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . 11 4.1 Feel Bundle Nuclear Design Analysis .............. 11 4.2 Corc Nuclear Design Analysis . . . . . . . . . . . . . . . . . . 11 4.2.1 Core Configuration ................... 11 4.2.2 Cora Reactivity Characteristics . . . . . . . . . . . . . 12 4.2.4 Core Hydrodynamic Stability . . . . . . . . . . , . . . . 12 5.0 ANTICIPATED OPERATIONAL OCCURRENCES . . . . . . . . . . . . . . . . . 16 5.1 Analysis Of Plant Transients . . . . . . . . . . . . . . . . . . 16 S.2 Analyses For Reduced Flow Operation .............. 16 5.3 Analyses For Reduced Power Operation . . . . . . . . . . . . . . 16 5.4 ASME Overpressurizatior Analysis . . . . . . . . . . . . . . . . 16 5.5 Control Rod Withdrawal Error . . . . . . . . . . . . . . . . . . 16 5.6 Fuel Loading 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 Fcr ANF Fuel . 24 7.2.2 Minimum Critical Power Ratio .............. 25 7.2.3 Linear Heat Generation Rate For ANF Fuel ........ 26

_._____.__.m. _ _ . __. _ _ _ _ . _ _ _ _ _ _ _ . _ . . _ _

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,- ANF-88-149 i Page 11

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TABLE OF CONTENTS (Continued) 'l Section  !

. f321 7.3 Surveillance Requirements . . . . . . . . . . . . . . . ... . . 26 7.3.1 Scram Insertion Time Survet!1ance . . . . . . . . . . . . 26 7.3.2 Stability Surveillance .....,............ 26 8.0 METHODOLOGY REFERENCES ....................... 31 9.0. REFERENCES . . . . . . . ... . . . . . . . . . . . . . . ... . . . . ~32 APPENDIX A SUPPLEMENTARY INFORMATION FOR 9X9-5 LEAD TEST ASSEMBLIES . . . 33 O

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ANF-88-149 1

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Page 111 f LIST OF TABLFS liklg E.iLqt 4.1 Neutronic Design Values .................... 13 I 1

LIST OF FIGURES j

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Fiaure Pagg ]

1.1 Power / Flow Map Used for Grand Gulf Unit 1 ME00 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 g 3.2 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local {

( 3.3 Power Distribution (ANF 1.3 3.61 - 8G4 Fuel) .........

Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local 8 )

i Power Distribution (XN-2 3.21 - 6G4 Fuel) . . . . . . . . . . . 9 l

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 1 5.3 Flow Dependent MAPFAC Value for Grand Gulf Unit 1 Cycle 4 ... 20 1 5.4 Power Dependent MAPFAC Value for Grand Gulf Unit 1 Cycle 4 . . . 21 I 7.1 Exposure Dependent Maximum Local Peaking for XN-12.99-5G3 Fuel. 27 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 l ANF 1.3 3.61-8G4 Fuel . . . . . . . . . . . . . . . . . . . . . 30 A.1 AhF 9x9-5 Lead Test Assembly LHGR Limits . . . . . . . . . . . . 36 )

A.2 ANF 9x9-5 Lead Test Assembly MAPLHGR Limits .......... 37 l 1

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9e ACKNOWLEDGEMENT The authors would like to acknowledge the following individuals for their contributions to the results reported in this document:

D. A. Adkisson D. J. Braun H. E. Byram-S. J Haynes D. E. Hershberger M. J. Hibbard D. F. Richey S. E. State l

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ANF-88-149 rm Page 1 N]

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1.0 INTRODUCTION

This report provides the results of the analysn performed by Advanced Nuclear Fuels Corporation (ANF) in support of the Cyc.4 4 reload for Grand Gulf Unit 1. This report is intended to be used in conjunction with ANF l topical report XN-NF-80-19( A), Volume 4, Revision 1, " Application of the ENC l Methodology to BWR 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-80-19fA), 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 uperating map to the ME0D in Cycle 1 (Reference 1). These analyses established appropriate operating limits for ME00 operation. The initial reload of ANF fuel in Grand Gulf Unit 1 occurred in Cycle 2. In support of the initial reload of ANF fuel,

/O extensive additional safety analyses were performed by ANF to either justify

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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 varying Gdp03 as well as four ]

(4) Lead Test Assemblies enriched, to 3.25 w/o U235 is used to meet the cycle j I

energy requirements. The balance of the core is composed of 288 once exposed ANF reload fuel assemblies and 236 twice exposed ANF reload fuel assemblies.

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

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

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l 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 i

MCPRf limits have been revisod 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 w "- ." dance with the existing bases in the plant  !

Technical Specifications, except that analysis set points for safety valves have been increased to include a 6% tolerance, and provision has been made in the flow depenJent- 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 vill be performed using ANF's core monitoring methodology, POWERPLEX* CMSS, in accordance with ANF's thermal limits methodology, THERMEX

- (Reference 8.6).

O 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 confinned for subsequent cycles. Since the Cycle 4 analyses results are similar to those of Cycle 2, the Cycle 2 analyses 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 /MTU). 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 til fuel types in Cycle 4. The extended curve was constructed to bound all Tuel 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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.iG 2.0 FUEL MECHANICAL DESIGN ANALYSIS Applicable Fuel Design Report: Reference 3 Qualification analyses provided in the reference are applicable to the Grand Gulf Unit 1 ANF fuel assemblies. 'The extended 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.

The 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 HYDRAULIC DESIGN ANALYSIS l

3.2 Hydraulic Characterization 3.2.3 Fuel Centerline Temperature Fuel Centerline Melting is protected by the t ansient 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 Btu /lbm Reference Pressure 1050 psia Feedwater Temperature 420*F Feedwater Flow Rate 17.74 M1bm/hr 3.3.2 Desian Basis Radial Power Distribution See Figure 3.1 3.3.3 Desian Basis Local Power Distribution See Figures 3.2 to 3.4 l

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

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lr .*  : LL  : L-  : ML  : M'  : M  :  :  ;

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• : M.  : H  : H.  : W  : M  : H  : H  : M  :

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M  : H  : H  : M  : W  : H  : H  : M'  :

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x * . . . . . .

y *  :-  :  :  :  :  :  :  :  :

t-  : M  : M*  : H  : H  : H  : H  : M*  : M  :

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

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L  : ML  : M  : M  : M  : M  : ML  : L  :

0.94 : 1.01  : 1.04 : 1.02 : 1.02 : 1.04 : 1.01  : 0.94  :

Figure 3.2 Grand Gulf Unit I Cycle 4 Safety Limit Design Basis Local s Power Distribution (ANF.1.3 3.61-8G4 fuel)

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4

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ANF-88-149

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

1.00 : 0.94  : 0.99 : 1.05 : 1.06  : 0.91  : 0.97  : 1.00 :

L  : ML  : M  : M  : M  : M  : ML  : L  :

0.98 : 1.00 : 1.03  : 1.02  : 1.02  : 1.03 : 1.00 : 0.98 :

l Figure 3.3 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis local Power Distribution (XN-2 3.21 6G4 Fuel)

Gadolinia Location

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Page 10

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0.99 : 0.99 :  :  :  :  : 1.00 :

L  : MLl* : M  : H  : H  : MLl* : M  : L  :

0.97  : 0.96  : 0.99  : 1.05  : 1.05 : 0.94 : 1.01  : 0.97  :

L  : L  : ML  : M  : M  : ML  : L  : L  :

1.00 : 0.97  : 0.99  : 1.01  : 1.01  : 1.00 : 0.97  : 1.00 :

Figure 3.4 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis local Power Distribution (XN-1 2.99-5G3 Fuel) l

• - Gadolinia Location

ANF-88-149 J 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

. N_ gig: Burnable poisons are not distributed uniformly over the )

enriched length of the designated rods. The natural uranium axial blanket sections do not contain burnable absorber material.

r.

.( Location of Non-Fueled Rods Figure 4.1 Neutronic Design Parameters Table 4.1 4.2 Core Nuclear Desian Analysis 4.2.1 Core Configuration Figure 4.2 Core Exposure at E0C3 18104 MWD /MT Core Exposure at B004 10244 MWD /MT Core Exposure at EOC4 22308 MWD /MT Maximum Cycle 4 Licensing Exposure Limit 23130 MWD /MT O

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4.2.2 Core Reactivity Characteristics (l)e(2)  !

BOC4 Cold K-effective, All Rods Out 1.13315 BOC4 Cold K-effective, All Rods In 0.96019 BOC4 Cold K-effective, i Strongest Rod Out 0.989,06 Reactivity Defect /R-Value 0.00% Delta K/K (Minimum occurs at 0 mwd /MTU)

Standby Liquid Control l System Reactivity, 660 PPM Cold Conditions, K-effective 0.96215 (1) Includes calculational bias.

(2) Evaluated at nominal EOC3-808 mwd /MTU.

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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( s Table 4.1 Neutronic Design Values ,

Fuel Assemb1v .

Number of fuel rods 62 Number of inert water rods 2 Fuel rods enrichments Figure 4.1 Fuel rod pitch, inches 0.636 Fuel assembly loading, Kgu 175.69 ,

l Core Data Number of fuel assemblies 800 Rated thermal power, MWt 3833 Rated core flow, Mlbm/hr 112.5 Core inlet subcooling, Btu /lbm 22.2 Moderator temperature, F 551 Channel thickness, inch 0.120 Fuel assembly 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 0.328 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 O

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M*- RODS ( 8) ---

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W RODS ( 2) -.-

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f' Figure 4.1 Grand Gulf Unit 1 Cycle 4 ANF 1.3 3.61-8GZ Enrichment Distribution s

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1 2 3 4 5 6 7 8 9 to 11 12 14 13 15 16 1 A2 C1 00 C1 00 A2 00 C1 00 C1 00 C1 00 C1 A2 A2 l

l 2 C1 00 A2 00 A2 DO C1 00 A2 00 C1 DO A2 C1 A2 A2 3 00 A2 D0 C1 Bt 81 00 C1 DO C1 00 81 00 C1 A2 4 C1 DO C1 00 #2 DO A2 00 l

A2 00 81 00 A2 C1 A2 5 DO A2 81 A2 DO C1 C1 81 DO C1 DO C1 DO C1 A2 6 A2 DO B1 00 C1 00* A2 00 00 C1 C1 00 A2 C1 A2 7 D0 C1 00 A2 C1 A2 00 81 D0 C1 00 81 81 A2 A2 8 C1 00 C1 00 81 00; 81 00 81 00 81 00 A2 A2 9 00 A2 00 A2 00 C1. 00 81 00 A2 DO C1 A2 10 C1 D0 C1 DO C1 00 C1 DO A2 E0 C1 C1 A2 1  ; 11 00 C1 DO 81 00 00 L ,' C1 81 DO C1 A2 A2 A2 12 C1 00 81 00 C1 00 81 00 C1 C1 A2 A2 1

13 Do A2 00 A2 DO A2 81 A2 #2 A2 A2 14 81 C1 C1 C1 C1 C1 A2 A2 15 A2 A2 A2 A2 A2 A2 A2 X s fuet fype 16 A2 A2 Xv v = Cyctes Irredieted Number of Fuel Assemblies Type (Full Core) Description A 236 ANF $X8 XN 1.1 2.81 w/o U 235 SCd at 3.0 %

8 84 ANF 8X8 XN l.2 3.01 w/o U 235 SGd at 4.0 %

C 204 ANF 8X8 XN 1.2 3.01 w/o U 235 6Cd at 4.0 %

D 272 ANF 8X8 ANF 1.3 3.37 w/o U 235 BGd at 4.0% \ 5.0%

E 4 ANF 9X9 ANF 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)

l

(' ANF-88-149 l

• 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) . l Transient Delta CPR LRNB 0.12 LFWH 0.11*

CRWE 0.10**

FWCF 0.04

• Applicable at all conditions. 4

• • Statistically determined, Ref. 6.

5.2 Analyses For Reduced Flow Ooeration Reference 4 MCPRf Figure 5.1 MAPFACf Figure 5.3 5.3 Analyses For Reduced Power Ooeration Reference 4 MCPR p Figure 5.2 MAPFAC p Figure 5.4 5.4 MME Overoressurization Analysis Reference 4 Limiting Event MSIV Closure Worst Single Failure MSIV Position Scram Trip Maximum Vessel Pressure 1298 psig Maximum Dome Pressure 1280 psig 5.5 Control Rod Withdrawal Error Reference 6 Values of delta CPR as a function of core power level resulting from a CRWE transient, developed in Reference 6 on a generic basis for BWR/6 class of plants including Maximum Extended Operating Domain operation, are applicable to Cycle 4 operation.

r ANF-88-149 Page 17 5.6 Fuel Loadina Error Reference 8.1 With Correctly Loadina Error loaded Cotg Maximum LHGR 13.90 12.40 Minimus MCPR 1.17 1.27 s

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6.0 POSTULATED ACCIDENTS 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 Analysis ANF 1.3 Peak Local Average Planar Analyzed Peak Clad Peak Clad Metal-Water

.( ) Exoosure MAPLHGR Temperature Temperature Reaction 0 GWD/MTU 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 0.3 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 up to ll'F; an 11*F increase in PCT is insignificant due to the fact that the calculated temperatures are over 500*F below the 2200*F limit.

O

,9 ANF-88-149 Page 23

.g.

6.2 Control Rod Dron Accident Referende 8.1 Dropped Control Rod Worth 9.72 mk Doppler Coefficient . -9.5 x 10-6  ;

AK/K/*F Effective Delayed Neutron Fraction A.5 x 10-3 Four-Bundle local Peaking Factor 1.28 Maximus Deposited Fuel Rod Enthalpy 172 Cal /gm 1.

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ANF-88-149 Page 24 1:

l- 7.0 TECHNICAL SPECIFICATIONS 7.1 Limitina' Safety System Settinas 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 Averace Planar Linear Heat Generation Rate For ANF Fut].

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 8 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 analysi: was performed at conservatively higher MAPLHGR values, Section 6.1.3.

Q *A safety limit MCPR o 1.07 is to be applied during single loop operation.

1

ANF-88-149 Page 25

~( .

-1 Average Planar MAPLHGR* ,

Exoosure ANF299E5G3S8 ANF321E6G458 ANF321E8G458 ANF361E8G118 0.00 GWd/MTU 13.20 kW/ft 13.33 kW/ft 13.00 kW/ft 12.98 kW/ft j 0.25 -

13.20 13.34 13.00 12.98 j 1.00 13.38 13.36 13.02 13.01 2.00 13.54 13.40 13.06 13.03 .

4.00 13.89 13.54 13.26 13.13 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 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 : e MAPFAC(f)** Figure 5.3 MAPFAC(p) Figure 5.4 7.2.2 Minimum Critical Power Rat'io Rated Conditions MCPR Limit 1.18 MCPR(f) Fiture 5.1 MCPR(p) Figure 5.2

• 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 D-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 Pland Exoosure LHGR 0.00 GWd/MT- 16.0 kW/ft 25.40 14.1 42.00 9.3 7.3 Surveillance Requirements 7.3.1 Scram Insertion Time Surveillaqqg Thermal margins are based on analyses in which scram performance was assumed consistent with the Technical Specification limits. No 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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ANF-88-149 A Page 31 Q.f 8.0 METHODOLOGY REFERENCES Section 8 References 8.1 through 8.18 are contained in the following

. report:

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

• Revision 1, Exxon Nuclear Company, Richland, Washington (March 1985).

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

Thermal Limits Methodology Summary Description," XN-NF-80-19(P)( A),

Volume 3. Revision 2 (January 1987).

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. ANF-88-149 Page 32

9.0 REFERENCES

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, 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(P)(A), Revision 1 Exxon Nuclear Company, Richland, WA, September 1986.

4. " Grand Gulf 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 Nuclear Company, Richland, WA, May 1986, and XN-NF-825(P)(A),

Supplement 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.

8. " Grand Gulf Unit 1 LOCA Analysis," XN-NF-86-38, Exxon Nuclear Company, Richland, WA, June 1986.

9. " Grand Gulf Unit 1 Cycle 3 Reload Analysis," ANF-87-67, Revision 1, Advanced Nuclear Fuels Corp., Richland, WA, August 1987.

10. " Grand Gulf Unit 1 Reload XN-1.3, Cycle 4 Mechanical Design Report,"

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

11. " Grand Gulf Nuclear Station Unit 1 Revised Flow Dependent Thermal Limits," NESD0-88-003, MSU System Services Inc., November 1988.

Vp ~

. ANF-88-149 Page 33

%f APPENDIX A SUPPLEMENTARY INFORMATION FOR 9X9-5 LEAD TEST ASSEMBLIES A.1 INTRODUCTION Evaluations have been performed consistent with ANF methodology (" Exxon Nuclear Methodology for Boiling Water Reactors, XN-NF-80-19) 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 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 relt,ted MAFLn6A M mit) 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 i 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 HYORAULIC DESIGN Component hydraulic resistances have been determined and it has been found that the 9x9-5 LTA's are hydraulically compatible with the co-resident ANF 8x8 fuel assemblies. Unique design features of the 9x9-5 (two rod diameters, injection water rod) have been modeled to demonstrate compatibility

/ 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-

ANF-88-149 O Pagd 34 U

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

A.4 NUCLEAR DESIGN The core wide neutronic impact of replacing four (4) of the 800 fuel assemblies in the Grand Gulf Cycle 4 core is negligible. The leads are designed to be neutronically " transparent" relative to the 8x8 fuel; that*is, reactivity characteristics are similar.

Evaluatien of the 9x9-5 LTA's relative to LFWH, Control Rod Drop Accident, MAPFAC 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 Hisload has been evaluated separately using the XN-3 correlation, which has been demonstrated .to conservatively predict critical power in the 9x9-5 design.

A.5 ANTICIPATED OPERATIONAL OCCURRENCES Analyses of limiting transients have shown that the bundle power needed 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 lower 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 control rod withdrawal error and fuel rotation error are essentially the same as in the case of 8x8 fuel.

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. ANF-88-149 Page 35

\J 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 l 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 water rods 7.nd less stored energy in the rods. The 9x9-5 MAPLHGR limit is based on the LHGR limit provided in " Generic Methanical Design for ANF 9x9-5 BWR Reload Fuel," [ANF-88-152(P)] divided by the maximum local peaking as a function of exposure. Analyses performed by ANF demonstrate that this limit meets 10 CFR 50.46 criteria.

The consequences of a control rod drop accident are governed primarily by the dropped rod worth. Since the reactivity of the LTA's is comparable to the coresident 8x8 fuel and the LTA's are loaded in non-limiting locations, no appreciable difference will be experienced due to the LTA's.

'A.7 TECHNICAL SPECIFICATIONS All operational limits used for 8x8 fuel are applicable to the 9x9-5 LTA's except fcr fuel type specific MAPLHGR limits and a 9x9-5 LHGR limit.

The LHGR limit shown in Figure A-1 is that of ANF-88-152. The MAPLHGR limit is shown in Figure A-2 and is consistent with the 9x9-5 LHGR limit. The LTA SLO operational limits are based on the 9x9-5 MAPLHGR multiplied by the smaller of the MAPFACf , MAPFAC p , or 0.86.

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ANF-88-149 Issue Datt: 11/11/88 GRAND GULF UNIT _1 CYCLE 4 RELOAD ANALYSIS 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 s 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)

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