ML20247B173
| ML20247B173 | |
| 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) |
| Shared Package | |
| ML20247B179 | List: |
| References | |
| AECM-89-0110, AECM-89-110, NUDOCS 8907240059 | |
| Download: ML20247B173 (8) | |
Text
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. 1 CORRECTED COPY To correct for missing yey attachment.
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(~)- HESDEJUGE RW".
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vce Pescent l wm coecons June 30, 1989-U.S. Nuclear Regulatory Commission Mail Station P1-137 Washington, D.C. 20555 ;
i Attehtion: Document Control Desk Gentlemen:
SUBJECT:
Grand Gulf Nuclear Station Unit 1 I Docket No. 50-416 License No. NPF-29 Current Cycle Safety Analysis, Revision 2 AECM-89/0110
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() 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 -l and contains analyses supporting operation and the current fuel reload, i Analyses included in the CCSA hae 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 f accident and transient analyses. The CCSA is considered part of the Final i Safety Analysis Report (FSAR), and is updated at least annually in accordance ]
with 10CFR 50.71(e). Revision I was submitted November 21, 1988 and reflected j analyses to support GGNS Cycle 3. This update to CCSA reflect: the fuel reload j accomplished during the third GGNS refueling outage performed in early 1989
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(cycle 4). ,
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As required by 10CFR 50.71(e)(2) and as a nnorized by SERI, I hereby !
certify, to the Sst of my knowledge, information and 'velief, that the information given in the attact.ed CCSA, Revision 2, accurately presents changes made since the previous submittal, necessary to reflect information ,
and analyses submitted to the NRC or prepared pursuant to NRC requirement.' ;
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cc: Mr. J. G. Cesare (w/o)
Mr. T. H. Cloninger (w/o)
Mr. R. B. McGehee (w/o) 'j Mr. N. S. Reynolds (w/o Mr.H.L. Thomas (w/o)) ,
Mr. H. 0. .Christensen (w/a)
Mr. Stewart D. Ebneter . (w/a)
Regional Administrator U.S. Nuclear Regulatory Comission l
Region II .
101 Marietta St. , N.W., Suite 2900 Atlanta, Georgia 30323-Mr. L. L. Kintner, Project Mcnager (w/o)
. Office of Nuclear Reactor Regulation
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Grand Gulf Nuclear Station Unit 1 Document No. 50-416 Revision 2 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 Page6(Introduction)
Page 7 introduction Page 7 Introduction)
Page 8 Introduction Page 8 Introduction)-
Page 9 Introduction) Page 9 Introduction)
ANF-87-67,Rev1(allpages) ANF-88-149 all'pages)
ANF-87-66, Rev 1 (all pages) ANF-88-150 all pages)
^O MPEX-86/92 (all pages) ANF-80-183 all pages)
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7 C'URRENT CYCLE AFETY ANALYSIS.
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. Table of Contents Tab- Description t
1 Introduction 2 ANF-88-149 GGNS Unit'l Cycle 4 Reload' Analysis 2'
3 ANF-88-150 GGNS i; nit 1 Cycle 4 Plant Trantient '
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 j Revision 1 AECM-88/0188 Transmittal letter and Revision 1 .)
Insert Instructions q Revision 2 AECM-89/0110 Transmittal letter and Revision 2 2 Insert Instructions O
V.
NLSMISC89060801 - 2 Rev. 2 7/89 2
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2.3.2. -The:following defines acronyms.used in Table 1.
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CPR .
Critical Power Ratio .
CS -
Core Stability i'
ELL -
' Extended Load Line Fuel Loading Error
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, FLE -
FWCF -
Feedwater Controller Failure FWHOS '-
Feedwater Heater Out of Servica ICF -
Increased Core Flow LOFWH' -
Loss of Feedwater Heating transient LRNB - - Load: Reject No Bypass (also known as GLR for Generator Load Fejection)
RDA -
Rod Drop Accident RWE -
Rod Withdrawal Error SLMCPR -
Single Loop Operation
'3.0 .CCSA Attachments The.following lists attachments found in.the CCSA.
- 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; y RDA NLSMISC89060801 - 6 Rev. 2 7/89 2
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- 5. XN-NF-825 Supp 21 BWR/6 Generic Rod Withdrawal Error Analysis.
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MCPR(p) for Plant Operations withie the Extended' Operating Domain Event: RWE
'6. ANF-88-183 Grand Gulf Unit:1 Reload XN-1.3, Cycle 4'. 21 m Mechanical Design Report-2-
7.2 .NESDQ-88-003 GGNS Unit:1' Revised Flow Dependent Themal Limits 1
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,g TABLE 1 Summary of Analyzed Events UFSAR SECTION EVENT PRIMARY PURPOSE CCSA ATTACHMENT l
4.1 Methods General analysis ANF-88-149 techniques, summary XN-NF-80-19 Vol. 4 fuel description
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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 suppress region
( 5.2.2 Overpressure Overpressure protection ANF-88-150-5A A">ME 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 15.1.2 FWCF CPR at rated ANF-88-150 15D FWCF CPR with ICF / >N F-88-150 ISD FWCF CPR with ELL ANF-88-150 150 FWCF CPR at Power below MF-88-150 40% (w/o direct scram)
ISD FWCF CPR - w/o bypass ANF-88-150
(' 150 FWCF CPR with FWH0S ANF-88-150 L
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-SECTION- EVENT : PRIMARY PURPOSE : CCSA ATTACHMENT'
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'15.2.2: LRNB- .CPR at rated ANF-88-150.
150 LRNB CPR with,ICF ANF,-88-150 ,
'150 1RNB CPR with ELL ANF-88-150~ M 11SD LRNB. CPR at power below ANF-88-150 j- 40%:(w/o direct scram)
ISD LRNB CPR with FWHOS ANF-88-149 Section 1.0' 15C ' . SLO- Operation with one ANf-88-149 loop out of service .
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- f. 15.4.1, 2 RWE CPR vs. Power XN-NF-825'Supp. 2" 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 2
15D Flow Runout CPR & MAPLHGR ANF-88-150 vs. Flow NESDQ-88-003 O
NLSMISC89060801 - 9 Rev. 2 7/89 2
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E ANF-88-149 ney.y ADWUMCEDNUCLEARFUELSCORPORATION GRAND GULF UNIT 1 CYCLE 4 RELOAD ANALYSIS O
NOVEMBER 1988 y -NiW~W O
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R_) kOVANCBDNUCLEAR FUELS CORPORATION ANF-88-149 Issue Date: 11/11/88 GRAND GULF UNIT 1 CYCLE 4 RELOAD ANALYSIS i
Prepared by l
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-( Licensing and Safety Engineering Fuel Engineering and Technical Services h)I ,. .w
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/ RJ G. Grumer BWR Neutronics.
Neutronics and Fuel Management 1 Fuel Engineering and Technical Services l
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~j RIPORTANT NCmCE S'-.Z GGidenFw Af80 USE Off THIS DOCUtdENT PL3Aag MEAD CAREFUL 4,Y Advemose Nusher Fuse Cofporebon e warrantes anc represemacons con.
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corning me suopoet menor of me oscument are mese set form in me Agreement genusen Aevences Nucoser Puses Corporesen and me Cuesomer pursuant to wruch Stas eneument e issuse. Acoerengly, encept as otnerwee expressly pr1> -
vesse e suon Agreement, neener Asvenose Nussear Fues Corporepon nor any pereen spong on as noned meses any warranny or repressmenon escreesse or imates, won resseet to tne escunecy, comomeonees, or usefumese of me mvor.
mason caneense m tPe comument. or met me use of any mtormenon, acoermus, memes or proprAs diesesses e tnee coeument 'and not sninnge privateey ownee nones: or assumes any nesones won resseet e ine use of any insormaten. an-parass, memos or penmaa st6essesse m Ins escument.
The intermeen comesnee Perest to Ior me sees use of Cuesomer in sfeer e ausse imperment of ngnie of Aswenese Nusteer Pusu Corporate 9 in gesense or wwensons wruen may be enmusse e the eformenon contesnee in the document, tne reggeant. By de tegeWenes of flus occumer.t. agrees not to f puthen er messe pubhc use (in tne pasent use of the term)of euen arHormacon unDi so aumenese an wnting Dy Advanese Nusseer Fuses Corporenen or uttal after six :
(e) menes tomouang termmason or esowenen of the aforeeme Agreement and any sesensen tnereof. uneses omerwee espreessy proviese in me Agreemem. No
- 3pque or thionese m or to any peerses are ernched by tne fumert#ng of mig occu-ment.
ANF-3145 472A (12 87)
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rN ANF-88-149
. !. f Page'i TABLE OF CONTENTS J,1ction hgg
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 Temperature . . . . . . . . . . . . . . . 6 3.2.5 Bypass Flow . . . . . . . . . . . . . . . . . . . . . . . 6 3.3 MCPR Fuel Cladding Integrity Safety Limit ........... 6 3.3.1 Nominal Coolant Condition in Monte Carlo Analysis . . . . 6 3.3.2 Design Basis Radial Power Distribution ......... 6 3.3.3 Design Basis Local Power Distribution . . . . . . . . . . 6 4.0 NUCLEAR DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . 11 4.1 Fuel Bundle Nuclear Design Analysis .............. 11 4.2 Core Nuclear Design Analysis . . . . . . . . . . . . . . . . . . 11 q 4.2.1 Core Configuration ................... 11 I Q 4.2.2 Core Reactivity Characteristics . . . . . . . . . . . . .
4.2.4 Core Hydrodynamic Stability . . . . . . . . . . . . . . .
12 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 Overpressurization Analysis . . . . . . . . . . . . . . . . 16 5.5 Control Rod Withdrawal Error . . . . . . . . . . . . . . . . . . 16 5.6 Fuel Loading Error . . . . . . . . . . . . . . . . . . . . . . . 17 6.0 POSTULATED ACCIDENTS ........................ 22 6.1 Loss-Of-Cool ant 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 i 7.1.1 MCPR Fuel Cladding Integrity Safety Limit . . . . . . . . 24 7.1.2 Steam Dome Pressure Safety Limit ............ 24 7.2 Limiting Conditions For Operation ............... 24 7.2.1 Average Planar Linear Heat Generation Rate For ANF Fuel . 24 l 7.2.2 Minimum Critical Power Ratio .............. 25 O
d 7.2.3 Linear Heat Generation Rate For ANF Fuel ........ 26 1
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J i l ANF-88 149-Page ii i TABLE OF CONTENTS (Continued) !
i Section p
__ggg 7.3 Surveillance Requirements . . . . . . . . . . . . . . . ... . . 26 7.3.1 Scram Insertion Time Surveillance . . . . . . . . . . . . 26 7.3.2 Stability Surveillance ............. ... 26 8.0 METHODOLOGY REFERENCES ....................... 31
9.0 REFERENCES
............................. 32 APPENDIX A SUPPLEMENTARY !NFORMATION FOR 9X9-5 LEAD TEST ASSEMBLIES . . . 33 O
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.V Page lii-LIST OF TABLES likll! EiL9.it
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l 4.1 Neutronic Design Values .................... 13 l
LIST OF FIGURES i
Fiaure Eggt 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
~T Radial Power Histogram .................... 7 j (V 3.2 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local Power Distribution (ANF 1.3 3.61 - 8G4 Fuel) . . . . . . ... 8 J
3.3 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local Power Distribution (XN-2 3.21 - 6G4 Fuel) . . . . . . . . . . . 9 l 3.4 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local 1 Power Distribution (XN-1 2.99-5G3 Fuel) . . . . . . . . . . . . 10 4.1 Grand Gulf ll nit 1 Cycle 4 ANF 1.3 3.61-8GZ Enrichment Distribution . . . . . . . . . . . .. . . . . . . . . 14 ;
4.2 Grand Gulf Unit 1 Cycle 4 Reference Core Loading Pattern _l (Quarter Core, Reflective fyrmetry) . . . . . . . . . . . . . . 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 5.3 Flow Dependent MAPFAC Value for Grand Gulf Unit 1 Cycle 4 ... 20 5.4 Power Dependent MAPFAC Value for Grand Gulf Unit 1 Cycle 4 . . . 21 7.1 Exposure Dependent Maximum Local Peaking for XN-12.99-5G3 Fuel. 27 7.2 Exposure Dependent Maximum Local Peaking for ,
i
. XN-2 3.21-6G4 Fuel ...................... 28
,.1 Exposure Dependent Maximum Local Peaking for !
, XN-2 3.21-8G4 Fuel ...................... 29 7.4 Exposure Dependent Maximum Local Peaking for ANF 1.3 3.61-8G4 Fuel . . . . . . . . . . . . . . . . . . . . . 30 A.1 ANF 9x9-5 Lead Test Assembly LHGR Limits . . . . . . . . . . . . 36 i A.2 ANF 9x9-5 Lead Test Assembly MAPLHGR Limits . . . . . . . ... 37
. ll ANF-88-149 Page'iv .
i 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 M. E. Byram S. J Haynes D. E. Hershberger M. J. Hibbard D. F. Richey S. E. State I
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I l f'. ANF-88-149 h Page 1 l.0- INTRODUCTION This report provides the results of the analyses performed by Advanced Nuclear Fuels Corporation (ANF) in support of the Cycle 4 reload for Grand Gulf Unit 1. This report is intended to be used in conjunction with ANF topical report XN-NF-80-19(A), Volume 4, Revision 1, " Application of the ENC Methodology to BWR Reloads," which describes the analyses performed in support of this reload, identifies the methodology 'used for those analyses, and I provides a generic reference list. Section numbers in this report are the l same as corresponding section numbers in XN-NF-80-19(A), Volume 4, Revision 1.
1 The NSSS vendor performed extensive safety analyses for Grand Gulf Unit 1 in conjunction with the extension of the power / flow operating map to the ME00 in Cycle 1 (Reference 1). These 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, extensive additional safety analyses were performed by ANF to either justify the NSSS vendor operating limits or, where necessary, to provide appropriate limits for ANF fuel using ANF methodologies (Reference 2). Subsequent ANF analyses supported an additional reload of ANF fuel in Cycle 3 (Reference 9).
Changes from Cycle 3 to Cycle 4 for Grand Gulf Unit 1 include an addi-tional reload of ANF fuel resulting in a complete core of ANF fuel. The cycle length remains 18 months but with cycle energy increased from 1420 GWd to 1698 GWd. A reload batch design composed of 272 assemblies enriched to 3.37 w/o U235 containing eight rods of axially varying Gd 02 3 as well as four (4) Lead Test Assemblies enriched, to 3.25 w/o U235 is used to meet the cycle energy requirements. The balance of the core is 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 /MTU and the maximum assembly exposure from 33,000 mwd /MTU to 39,000 mwd /MTV (Reference 10).
The licensing basis of the four Lead Test Assemblies is described in Appendix A of this report.
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I' 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 MCPRp and j
.MCPR f - limits have' been revised ' to reflect- ANF calculated limits rather' than ]
the NSSS generic ME00 limits for this first al'-ANF core in. Grand Gulf Unit 1. J Analyses' were performed ' in accordance with' the exi. sting bases . in the plant L- Technical Specifications, . except that analysis set points for safety' valves j' have been-increased to include a 6% tolerance, and provision has been made in' !
the flow . dependent MCPR's . for " loop manual" operation as well . as "non-loop '
manual" operation (Reference 11). The analyses also included support of- the.
power / flow operation; map for Maximum Extended Operating Domain as shown in Figure 1.1. Monito-ing to the plant Technical Specifications presented in this report will be performed using ANF's core monitoring methodology, POWERPLEX* CMSS . in accordance with ANF's thermal limits methodology, THERMEX (Reference 8.6).
The ' ANF evaluation for Grand Gulf Unit 1 Single Loop Operation (SLO),
operation with feedwater heaters out of service, operation without condenser bypass, and LOCA-seismic considerations were performed for Cycle 2 and confirmed for subsequent cycles. Since the Cycle 4 analyses results are similar to those of Cycle 2, the Cycle 2 analyses and available margin to limits for these off normal operating c6nditions 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 (31,368 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 all fuel types in Cycle 4. The extended curve was constructed to bound all fuel resident in Cycle 4 by merging previously approved GE and ANF curves. The MAPFACf curve for SLO is unchanged from Cycle 3.
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L 2.0 FUEL MECHANICAL DESIGN ANALYSIS J Applicable fuel Design Report: Reference 3 ';
i 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 p, applicability of fuel mechanical limits for the ~ higher burnup reload fuel
? design.
l 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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3.0 THERMAL HYDRAULIC DESIGN ANALYSIS 3.2'hdraulicCharacterization 3.2.3 Fuel Centerline Temperature Fuel Centerline Melting is protected by the transient LHGR limit given in-Reference 3.
3.2.5 Sygan.f. log Calculated Bypass Flow Fraction 10.0%
(Exclusive of Water Rod Flow at 104.2%P/108%F) 3.3 MCPR Fuel Claddina Inteority Safety Limit See Reference 4 1.06*
3.3.1 Nominal Coolant Condi_on 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 Rite 17.74 M1bm/hr 3.3.2 Desian Basis Radial Power Distrib 1t_tga See Figure 3.1 3.3.3 Desian Basis local Power Distribut.ign See Figures 3.2 to 3.4 l
- For single loop operation the safety limit MCPR increases to 1.07 due to increased uncertainties.
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- 0.94 : 1.01 : 1.04' : 1.02 : 1.02 : 1.04 - : 1.01 : 0.94 :
Figure 3.'2 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local Power Distribution (ANF-1.3 3.61-8G4 Fuel)
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- 0.97 1.00 :
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Figure 3.3 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis local Power Distribution (XN-2 3.21-6G4 Fuel) (
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- 1.00 : 0.97 : 0.99 : 1.01 : 1.01 : 1.00 : 0.97 : 1.00 : 1 Figure 3.4 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local Power Distribution (XN-1 2.99-5G3 Fuel) i, s .
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l(j-s) ANF-88-149 Page 11 iT 4' 0 WJCLEAR DESIGN ANALYSIS-j 4.1 Fuel Bundle Nuclear Desian An11vsis )
1 Assembly. Average Enrichment 3.37 is/o 1 Radial Enrichment' Distribution Figure 4.1:
, Axial Enrichment Distribution Uniform 3.61 w/o j with-6" natural ]
Uranium at top l
~
and bottom Burnable Poisons Figrte 4.1
- H21g:- Burnable poisons are .not distributed uniformly over the enriched leng+.h of the designated rods. . The natural uranium axial blanket sections do not contain burnable absorber material.
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Location of Kon-Fueled Rods figure 4.1 Neutronic Design Parameters Table 4.1
, I 4.2 Core Nuclear Desian Analysis
.I 4.2.1 Core Configuration Figure 4.2 Core Exposure at EOC3 18104 MWD /MT Core Exposure at BOC4 10244 MWD /MT Core Exposure at EOC4 22308 MWD /MT Maximum Cycle 4 Licensing Exposure Limit 23130 MWD /MT O
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.c 4.2.2 Core Reactj.vity Characteristics (1),(2)
BOC4 Cold K-effective, All Rods Out 1.13315 j BOC4 Cold K-effective, All Rods In' O.96019- !
80C4 Cold K-effective, I Strongest Rod Out 0.98906 Rcactivity Defect /R-Value 0.00% Delta K/K (Minimum occurs'at 0 mwd /MTU)
Standby Liquid Control 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 Stabiht.y The results of Cycle 4 core hydrodynamic stability analyses continue to confirm the applicability of the previous cycles analyses. ;
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Table 4.1- Neutronic Design Values Fuel Assembiv .
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, Xgu 175.69
)
Core Data '
Number of fuel assemblies 800
.' Rated thermal power, MWt 3833 Rated core flow, Mlbm/hr 112.5 Core inlet subcooling, 8tu/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 81ade 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 e
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ANF-88-149 4
- 'Page 14'.
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LL R005 ( 1) ---
1.50 W/0 U235 L RODS ( 5) ---
2.00 W/0 U235
'ML RODS ( 8) ---
2.90 W/0 U235 M 'R005 (20) - --
3.48 W/0 U235 H RODS (20) -- - 4.57 W/0 U235 M* R005 ( 8) --- 3.48 W/0 U235 + 4.00 W/0 GD203 (Top 24 inches)
M* RODS ( 8) - --
3.48 W/0 U235 + 5.00 W/0 GD203 (Botton 114 inches)
W RODS ( 2) ---
INERT WATER ROD n
U Figure 4.1 Grand Gulf Unit 1 Cycle 4 ANF 1.3 3.61-8GZ Enrichment Distribution
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- ANF-88-149 T
Page 15 1 2 3 4 5 6 7 8 9 10 11 12
(
13 14 - 15 16 1 A2 C1 DO C1 00 A2 00 C1 00 C1 00 C1 00 C1' A2 A2 2 C1 D0 A2 00 A2 00 C1 DO A2 00 C1 00 A2 C1- A2 42 '!
3 00 A2 00 C1 89 .81 00 C1 DO C1 00 81 00 C1 A2
~~
4 C1 00 C1 00 A2 00 A2 ' 00 A2 00 1 00 A2 C1 A2 5 00 A2 81 A2 00 C1 C1 81 00 C1 00 C1 00 C1 A2
'6. A2 DO 81 00 C1 00* 42 00 C1 00 C1 00 A2 C1 A2 7' 00 C1 00 'A2 C1 A2 00 81 00 C1 00 81 81 A2 A2 6 C1 00 Ci 00 81 00 81 00 81 00 81 00 A2 A2 ,
l 9 00 A2 00 A2 00 C1. 00 00 81 A2 00 C1 A2
/.N C1 00 C1 D0 C1 00 C1 00 A2 E0 C1 C1 A2 11 00 C1 00 81 00 C1 00 81 00 C1 A2 A2 A2
.12 C1 00 81 D0 C1 00 91 DO C1 C1 A2 A2
'13 DO A2 DO A2 00 A2 81 A2 A2 A2 A2 14 81 C1 C1 C1 C1 C1 A2 A2 15 A2 A2 A2 A2 A2 A2 A2 x = Fuel Type 16 A2 A2 XY T = Cycles Irradiated Number of Fuel Assembtles Type tFull Core) Description 4.. .. ..... . . . . . . .. . . . ..
A 236 ANF 8X8 XNe1.1 2.81 w/o U 235 SCd at 3.0 %
8 $4 ANF 8x0 XN 1 2 3.01 w/o U 235 SGd at 4.0 t 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 8Cd at 4.0% \ 5.0%
E 4 ANF 9x9 ANF 1.3 3.25 w/o U 235 OGd at 5.0% \ 6.0%
O Figure 4.2 Grand Gulf Unit 1 Cycle 4 Reference Core Loading Pattern (Quarter Core, Reflective Symetry) i
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Page 16 5.0 ANTICIPATED OPEMTIONAL OCCURRENCES Applicable Generic Transiert Reference 5 i Methodology Report 1
1 5.1 Analysis of Plant Transients -
Reference 4 (Applicable at rated conditions) .
Transient Delta CPR LRNB 0.12 LFWH 0.11*
- CRWE 0.10**
FWCF 0.04
- Applicable at all conditions. e
- 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 MCPRp figure 5.2 MAPFAC p Figure 5.4 5.4 ASME 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 sontrol 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 gaeric basis for BWR/6 class of plants including Maximum Extended Operating Domain operation, are applicable to Cycle 4 operation.
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'3( ANF-88-149 Page 17 q
, 5.6 Fuel Loadina Error Reference 8.1-With Correctly 3 Loadina Error- Loaded Core Maximum LHGR~ 13.90 12.40 =l
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Minimum MCPR 1.17 1.27 1
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/ ANF-88-149 L Page 22 6.0 POSTULATED ACCIDENTS 6.1 Loss-Of-Coolant Accident 6.1.1 f}riak 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 Ret.irculation Pump Discharge Line with 1.00 Discharge Coefficient (1.0 DEG/RD)
Reference
,o Analysis ANF 1.3 Peak Local
'~' /
is Average Planar Analyzed Peak Clad Peak Clad Metal-Water Exoosure MAPLHGR, Temperature Temoeratqtg 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 l 30 12.1 1575 1577 0.2 I 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 i
i 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 11*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. l 1
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ANF-88-149-Page 23 .,
.,.1 6.2 gontrol Rod Droo Accident- Referende 8.1 Dropped Control-Rod Worth 9.72 mk.
!!.i. Doppler Coefficient -9.5 x 10-6 ,
i AK/K/*F.
Effective Delayed Neutron Fraction 'A.5 x 10*'3 Four-Bundle Local Peaking Factor 1.28 ,
i Maximum Deposited Fuel Rod Enthalpy 172 Cal /gm i
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Page 24
, 7.0 TECHNICAL SPECIFICATIONS l
7.1 Limiti.po Safety System Setting 1-7,1,1 MCPR Fuel Claddina Intearity Safety 1,1311 e ' Safety Limit MCPR 1.06*
7.1.2 Steam Dome Pressure Safety Limit Pressure Safety Limit 1325 psig 7.2 Limitino Conditions For Operation 7.2.1 Averaae Planar Linear Heat Generation Rate For ANF Fuel The following MAPLHGR limits are consistent with the design basis ' LHGR
(~~ limits shown in Figure 3.1 of Reference 3. These limits differ only by a
\- - factor equal- to the maximum local peaking factor at each exposure point. The MAPLHGR limits are made consistent with the LHGR limit Eso that at reduced !
. power and/or reduced flow the LHGR limit will be protected by the MAPFACf and' i MAPFAC p multipliers on MAPLHGR. The maximum local peaking factors for the four ANF fuel designs that will be resident in the core during Cycle 4 are shown in Figures 7.1, 7.2, 7.3 and 7.4. The LOCA analysis was performed at conservatively higher MAPLHGR values, Section 6.1.3.
i I
O *A safety limit MCPR of 1.07 is to be applied during single loop operation.
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ANF-88-149 i Page 25
' Average Planar' MAPLHGR*
Exoosure ~ANF299E52 18 ANF321E6G4S8 ANF321E8G458 ANF361E8GZS8 O.00 GWa/MTU 13.20 kW/ft 13.33 kW/ft 13.00 kW/ft 12.98 kW/ft 0.25 -
13.20 13.34 13.00 ' 12.98 1.00 13.38 13.36 13.02 13.01 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 13.20 13~.30 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 12.93 12.90 30.00 11.77 11.65 11.7$ 11.65 35.00 10.48 10.44 1 01 46 10.31 40.00 9.15 9.17 9.18 9.00 '
42.00 8.61 8.64 8.64 8.46 MA?tHGR Multipliers for Off-Nominal Conditions :
MAPFAC(f)** Figure 5.3 MAPFAC(p) Figure 5.4 7.2.2 Minimum Critical Power Ratio Rated Conditions MCPR Limit 1.18 MCPR(f) Figure 5.1 MCPR(p) Figure 5.2
- The MAPLHGR limit of Figure 1.2, applicable to all ANF fuel types resident in Cycle 4, is used for SLO.
O V **For SLO the Cycle 3 MAPFACf limit is applied to all ANF fuel types resident in Cycle 4.
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[~$ ' ANF-88 149
-(
.Page 26 7.2.3 Linear Heat Generation Rate For ANF Fuel The current Grand Gulf Unit 1 LHGR limits remain applicable for ANF reload fuel Cycle 4 operation. These limits, which are based on Figure 3.1 of Reference 3, are as follows, 1
Averaae Pland Excosure LHGR 0.00 GWd/MT 16.0 kW/ft 25.40 14.1 42.00 9.3 7.3 Surveillance Rea irements 7.3.1 Scram Insertion Time Surveillance Thermal margins are based on analyses in which scram performance was g- 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. l 7.3.2 Stability Surveillance Submittal regarding stability amendment is being made ender separate cover by the Licensee.
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p ANF-88-149 Page 31 j 8.0 METHODOLOGY REFERENCES Section 8 References 8.1 through 8.18 are contained in the following
- report: !
a
" 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 Sumary Description," XN-NF-80-19(P)( A),
Volume 3, Revision'2 (January 1987).
O
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ANF-88-149 Page 32
0.0 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 1980.
- 5. " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors,"
XN-NF-79-71(P), Revision 2, Exxon Nuclear Company, Richland, WA, November 1981.
p 6. "BWR/6 Generic Rod Withdrawal Error Analysis, MCPRp," XN-NF-825(A), Exxon
-Q 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-88-181, Advanced Nuclear Fuels Corporation, Richland, WA, November 1988.
- 11. " Grand Gulf Nuclear Station Unit 1 Revised Flow Dependent Thermal Limits," NESDO-88-003, MSU System Services Inc., November 1988.
l m _ _ _ . ___ _____ _ _ _ _ _ _ _ . _ _ ___
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. ANF-88-149 V Page 33 APPENDIX A SUPPLEMENTARY INFORMATION FOR 9X9-5 LEAD TEST ASSEMBLIES A.I INIR000CIl0H Evaluations have been performad 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 uper. core-wide transient performance. However, 9x9-5 specific analyses were performed to assure that the Cycle 4 operating limits also apply'to the LTA's.
Fuel type specific limits (LHGR and related MAPLHGR limit) have been de coloped for the LTA's and are presented in this appendix.
A.2 FUEL MECHANICAL DESIGN A mechanical design analysis has been performed for the 9x9-5. fuel type consistent with ANF's approved methodologies. Fuel design issues related to Anticipated Operational Occurrences (A00's) and accident analysis have been evaluated. These evaluations confirm that the LTA's meet NRC criteria of no centerline melting and less than 1% clad strain [* Generic Mechanical Design for ANF 9x9-5 BWR Reload Fuel," ANF-88-152(P)].
A.3 THERMAL HYDRAULIC DESIGN 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-
7 rr t
,fj ANF-88-149 >
(); Pagd 34 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.
Evgluation of the 9x9-5 LTA's relative to LFWH, Control Rod Drop Accident, MAFFAC f , shutdown margin and Shutdown Liquid Boron Control have been included in the nain body of this report in that they have been explicitly-
- modeled in those calculations. The LTA Misload has been evaluated separately using the XN-3 correlation, which has been demonstrated ' to conservatively predict critical power in the 9x9-5 design.
A.:i MIICIPATfu 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 6x8 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.
- v. ,
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. ANF-88-149 4 Page 35
- A.6 EQLTULATED ACCIDENTS Since. heatup is primarily a planar and not an axial phenomena, the
- appropriate bundle power limit derived from LOCA analyses is the peak bundle planar power. It has been ' demonstrated that the 9x9-5 LTA's provide better LOCA performance relative to. an. 8x8 fuel assembly due to the greater surface area provided by the larger number of fuel rodr, more inert surface from the water rods and. less stored energy in the rodi. The 9x9-5 MAPLHGR limit is
. based on. the LHGR-limit _ provided in '" Generic Mt, .hanical 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-coreside-t 8x8 fuel and th'e 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 'ex8 fuel are applicable to the 9x9-5 LTA's except for 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, i
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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-O- D. E. Hershberger I 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)
Document Control (10) i i
O mignammemum u- I
_ - . , _ . _ . _ _ _.__.__.______._.._______._________________________m_.,_
._,_____-.___.__,q 4-ANF-88-150
-] ,
5"h y, k l, ADVANCED NUCLEAR FUELS CORPORAT GRAND GULF UNIT 1 CYCLE 4 PLANT TRANSIENT ANALYSIS O
l l
NOVEMBER 1988 o 4+m-wn,e,
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7 ADIGMCCDNUCLEARFUELSCORPORATION ANF-88-150 Issue Date: 11/11/08 GRAND GULF llNIT 1 CYCLE 4 PLANT TRANSIENT ANALYSIS Prepared by 1'
__ ///8N K. J. Reynolds /'
BWR %fety Analysis t Licensing and Safety Engineering '
s Fuel Engineering and Technical Services i
% ll-f'ff
/ 'R. G. Grunner BWR Neutronics Neutronics and Fuel Management Fuel Engineering and Tec'nical Services November 8, 1988 ;
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I CUSTOMRR OtSCLAIMER SIPORTANT NOTICE REGAR0fMS CONTENTS ANO 098 0F THtt N
PLEASE READ CAREPULLY ,
Aevenese Nue6eer Fuele C&geremon's warrermes and representatens con.
cemsig the Guggest mener of the dooumem are mese est form in me Agreement esswoon Aevenose Nummer Pues Corporamen ene me C esomer pursuant to onen sus assument e ==m8 Aeograngfy, essent as emerwee escreamy pro.
vuest an euen Agreement. neeer Ads'entos Nucoser Fuses Corporamen nor any person among en as nonest menes any wee ney or rearementanen. esoreseed or emphed, wWI respect tlt the accuracy, eGrrJestences. or usefumees of the infor.
megen genusned at the document. er that the use of any intermemen. asseretus, memed or pagesse cdesseems wt One secument usel ret intnnge pnvasesy ownee rtpIW: er asumes igny W wel reseset to me use of any artformassen, an.
pomas, momed er asusses essened in tne secument ,
The wWarmamen eeneurice namn a ter the sees use of Cuesomer l
In erger to goes snaarment of nghie of Adveneet Nuchaer Pusse Corneremos in pEenW or WTweneene unsen may De sneluned WI the artformenon corttelned in mio -
desument, the resonant. Dy its asessence of tnse document. 30rees not to puneen er messe puetic use On the puent use of the term) of euen artformepon untif j es etanerged WI wnting Dy Advenoeg Nucteer Puses Corporanen er urval after tu d (6) meses fuseustg terminaison er esperenen of the saerseesd Agreemem and any sementen thereof, unsees oInerweg espressly prevised in me Agreement No nght., or stoonose en or to any paiones are innesise Dy tne turmening of mis docu.
ment.
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ANF-88-150 Page i TABLE OF CONTENTS l Section EA21 1~
1.0 INTRODUCTION
............................ I 2.0
SUMMARY
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 S umma ry o f Re s u l t s . . . . . . . . . . . . . . . . . . . . . . . 17 3.5.1 Power Dependent Thermal Limits and Values . . . . . . . . 18
~~s 3.5.2 Flow Dependent-Thermal Limits and Values ........ 18 (y
4.0 MAXIMUM OVERPRESSURIZATION ..................... 29 4.1 Design Basis . . . . . . . . . . . . . . . . . . . ...... 29 4.2 Maximum Pressurization Translents ............... 29 4.3 Results ............................ 30
5.0 REFERENCES
............................. 33
(,
t r
l ANF-88-150 Page ii LIST OF TABLES lihl1 EAgg.
2.1 Results of Analyses ....................... 6 l 2.2 Operating Limit Coordinates .................. 7 i 3.1 Grand Gulf Unit 1 Cycle 4 LFWH Data Summary . . . . . . . . . 19 i
l '
LIST OF FIGURES Fiaure Eggg 1.1 Power /Flote Map Used For Grand Gulf Unit 1 MEOD 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
(., 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 Dasign 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 (XH-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 Scram (Power and Flows) . . . . . 31 4.2 MSIV Closure Without Direct Scram (Vessel Pressure and Level) .........................32 O
..g 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 0.'F. Richef S. E. State O .
O
I l
ANF-88-150 Page 1
1.0 INTRODUCTION
l 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 witMa the Maximum Extended Operating Domain (MEOD). The NSSS vendor perfomed 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 -l 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, I extensive additional transient 4nalyses 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 (L ference 2).
The objective of these analyses was to confirm the applicability of the Grand Gulf Unit 1 Cycle 3 Technical Specification HCPR at rated conditions, p v establish MAPLHGR limits for Cycle 4 operation, and establish revised thermal limits for off-rated conditions fer the all-ANF core. An additional 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 pointr having the least margin to operating limits to confirm that the effects of the Cycle 4 changes on transient results are small and establish appropriate limits. Reanelysis of the limiting transients for Cycle 4 assures that the less limiting transients which were previously addressed will continue to be protected by the established operating limits p for Cycle 4. The power / flow conditions analyzed in Cycle 3 and Cycle 4 are presented in Figure 1.1.
72 m ,
ANF-88-150 Page 2
\
, The MCPR p , MCPRf , and MAPFACf limits have been . revised to reflect ANF calculated limits r'og ANF methodology. The Grand Gulf Unit I power and flow
. dependent MCit knalyses 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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. 2.0 - SUIDIARY The results of the Grand Gulf Unit 1 Cycle 4 transient analyses support 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 MAPFACr values (Reference 17.) have been confirmed for both
" loop manual" and "non-loop manual" operation.
Minor differences in the maximum local peaking as a function of exposure for the different ANF fuel types require that different MAPLHGR limits be monitored. These MAPLHCR 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 on MAPLHGR. Cycle 4 reload fuel MAPLHGR limits are included because of the 1 slight changes in the local peaking factor.
V Table 2a summarizes the transient analyses results applicable to Grand Gulf Unit 1 Cycle 4. These' results, together with the Grand Gulf Unit 1 j Cycle 4 calculated safety limit MCPR of 1.06, support (.ontinued use of the ]
existing 1.18 MCPR operating limit (at rated conditions) for Cycle 4 l operation.
]
The plant transient and safety limit analyses results reported herein support - revising the Cycle 3 power dependent Minimu:s Critical Power Ratio (MCPR p ) so that it is besed on the generic CRWE results of Reference 4 above ;
401 power and scpports the continued use of Cycle 3 limits below 40% power. !
The power dependent Maximum Average Planar Linear Heat Generation Factur (MAPFAC p ) for Cycle 3 is confirmed for Cycle 4 operation. The revised MCPR p limits, the MAPFACp confirmation, ard the results of ANF's analyses are {
presented in Figures 2.1 and 2.2, respectively.
The flow dependent Minimum Critical Power Ratio (MCPRr) and the results of ANF's analysis are presented in Figure 2.3. The flow dependent Maximum
-Q Average Planar Linear Heat Generation Rate Facto * (MAPFAC f ) is presented in q l
l
1 8.'
ANF-88-150 I Page 5 i
Figure 2.4. These flow dependent MAPFACf values and MCPRf limits have beer.
- revised from Cycle 3 to support Cycle 4 in both the " loop manual" and the-
'non-loop manual' mode of operation. These curves are based on conservative maximum core flow rates. Table 2.2 shows the coordinates usad to construct k Figures 2.1 through 2.4.
The results of the maximum system pressurization transient analysis are j presented . in Table - 2.1. The safety valve pressure setpoint tolerances in this analysis have been increased to % for Cycle 4; the results show that )
the Grand Gulf Unit I safety valves hays sufficient capacity and perforr.ance l with the increated setpoint tolerances to protect the vessel pressure safety u limit of 1375 psig during Cycle 4.
The fuel related Technical. Specification limits for Cycle 4 operation are
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included in the reload analysis report (Reference 3).
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, ANF-88-150 Pag ~e 6 O
Table 2.1 Results.of Analyses THERMAL. LIMITS Transient Delta CPR.
Loss of Feedwater Heating (all con'ditions) 0.11 Control Rod Withdrawal Error (100% power, Ref. 4) 0.10 Feedwater Controller Failure (104.2/108)* 0.04 a
1 Load Rejection Without Bypass
% Power /% Core Flow 104.2/108* 0.12 104.2/73.g, 0.02 40/108*** 0.15 40/108 0.32-
-'O . 25/73.8- 0.93 25/40 0.69 104.2/100 0.09 92.5/67 0.02 70/40 0.04 55/40 0.03 40/40** 0.03 MAXIMUM SYSTEM PRESSM IZATION Transient % Power /% Core Flow Vessel Lower Plenum Steam Domg l MS1V Closure 104.2/108* 1298 psig 1271 psig 104.2/73.8 1297 psig 1280 psig
- 104.2% power /108% core flow is used for the Reload Licensing Analysis (RLA) conditions to conservatively bound 100% power /105% core flow.
- Direct scram on turbine trip enabled
- Direct scram ,on turbine trip disabled
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Table 2.2 Operating Limit Coordinates.
j [d/Mg),_ GULF UNIT 1 CYC1Lj, <
Jij tg Rfn) Limits ]gPFAC(n) Liq 11.1 (Figure 2.1) (Figure-2.2)-
Percent of Rated Percent of Rated
,_.far_, Power l1(2Elg.). _,,,_{g.cg;f_QMir._. MAPFACfo1- ;
100 1.18 100 1.0 70 1.24 40 0.60*' u 0.69'*'
.70- Ic40 40 c 40 '1.48* 24.4 ?0.S7**
40 1.85 24.4 0.61
- 40 2.10***
25- 2.05**
- Core Flow > 50%
25 2.15 ** Core Flow s 50%
- Core Flow < 50%
- Core Flow'i 50%
h l j- MC79ff'> Limits 12ffACf f) Lird.t,1 (Figure 2.3) (Figure 2.4)
Percent of Loop Non-loop Fercent.of Loop Non-loop Rated Core ' Manual Manual P.ated Core Manual Manual .'
Flow' MCPRff) MCPRff) f*l,gg_,__ fBfff,CLfj.: MAPFAC(fl 30 1.55 1.73 110 1.00 1.00-40- 1.41 1.57- 91.0 1.00 1.00. j 50 1.31 1.44 90.0 1.00 .992 i 60 1.24 1.35 84.3 1.00 - -
70 1.19 1.27 80.0 ,977 .904 ,
73.4 1.18 - 70,0 .928 .827 80 1.18 1.21 60.0 . 880 .757 86.3 'l,18 1.18 50.0 .837 .695 105 1.18 1.18 40.0 .794 .636 1 30.0 .752 .586 , !
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x 3.0 THERMAL LIMITS ANALYSIS 3.1 Introductig,n 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 on9-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.
3.2 System T.tu i.i.tnia Revised 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 analyzed for Grand Gulf Unit 1 Cycle 4 - using COTRANSA. The load Reject No Bypass (LRNB) pressurization transient analysis was perforud at each of the ten state points. The feedwater Controller Failure (FWCF) analysis was performed at 104.2% power and 10S% flow. ASNE pressurization analyses were performed at state points 104.2%/108% and 104.2%/73.8%. LFWH analyses were performed with XTGBkR at five different stato points for eight exposures. The generic analysis for control rod withdrawal error is applicable to Grand Gulf Unit 1 Cycle 4.
These analyses show less restrictive results or little change from tt.e Cycle 3 analyses due to Cycle 4 changes, thus justifying that the less limiting transients not analyzed for Cycle 4 will continue to be protected.
O
- I ANF-88-150
/ s Page 13 O j 3.2.1 Desian Basis The LRNB and FWCF transients have been determined to be most limiting at end of full power capability when control rods are fully withdrawn from the l 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 l
Grand Gulf Unit I with ME00 was performed using conservative analytical limits l for trips and satpoints. 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 j throughout the cycle at state points which bound the ME00 operating map. j l
3.2.2 Anticipated Transients ANF's transient methodology report for jet pump BWRs (Reference 5) j 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-l o Loss of Feedwater Heating i o Load Rejection No Bypass o Feedwater Controller Failure Other transients are inherently non-limiting or bounded by one of the above as shosm in the NSSS vendor HEOD 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.
3.2.2.1 Lgjls Of Feedwater Heatina Analysis of the loss of feedwater heating event was performed to reflect reacter operation over thn MEOD operating power versus flow map and conditions anticipated during actual Grand Gulf reactor operation.
1 I i i
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l ANF-88-150
-. Page 14 Calculations performed for Cycle 4 assumed a conservative reduction of I 100'F in the feedwater temperature. Table 3.1 provides the conditions of each I case analyzed in terms of cycle exposure, core power, and core flow. The initial and final MCPR values are presented for each case, l
1 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:
l 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 1.FWH MCPR operating limit is defined by i i
p OLMCPR(LFWH) = -0.0112 + 1.1130
- SLMCPR i 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.
1 3.2.2.2 Load Re.iection No Byoass 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 reacter scram is initiated by the fast closure of the centrol valves as well as the recircula-tion pump high to low speed transfer. Condenser bypass fl ow, which can mitigate the pressurization effect, is not allowed. The excursion of the core porer due to void collapse is primarily terminated by reactor scram and void growth due to the recirculation pump high to low speed transfer.
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,4 ANF-88-150 Page 15 1
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Figures 3.2 and 3.3 present the response of various reactor and plant j parameters to the LRNB event initiated at the Reload Licensed /aalysis condi-tion (104.2% power /1085 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 (FWCF) is the l 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 l 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 at 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 (f to the turbine from excessive liquid inventory in the steamline. The high water level trip also initiates reactor scram, and recirculation pump 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 /1085 core flow) . The delta-CPR for this event was calculated to be 0.04, 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
. all FWCF transients at rated and off-rated conditions.
l ANF-88-150 jey Page 16 kJ' l
3.2.2.4 Control Rod Withdrawal Error i Reference 4 documents ANF!s generic CRWE analysis for Grand Gulf Unit 1 operation within the HE00. 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 i l
unchanged at 1.18.
3.3 Flow Exrgrsion Analvris The flow excursion transient is analyzed to determine 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 (G recirculation flow increases slowly to the physical maximum attainable by the equipment. Two modes of operation are analyzed for Grand Gulf Unit 1 Cycle 4,
" loop manual" and "non-loop manual." These two modes of operation correspond 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 operatir.g limit (Reference 12). In the confimtion calculation the channe 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 bn.s. events, assuming that the recirculation system equipment is capable of 110% of rated.
The Cycle 4 MAPFACf confirmation analysis of the power ascension fm associated with a flow increase was determined to be conservative when k compared to MAPFACf limits. Confirmation calculations were performed for
=h l' W
f ANF-88-150 Page 17 y
both " loop manual" and "non-loop manual" modes of operation with XTGBWR. The Cycle 2 analysis of reduced flow LHGR limits (MAPFACf values) was performed statistically based upon a wide variety of. initial conditions. 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 XTGBWR to simulate the flow renup event from 40% of rated flow.. For " loop manual" operation, these.
analyses. simulated a flaw 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 j core flows for both events as well as the results of the Cycle 4 analysis. 'I
~
3.4 Safety Limit The safety limit EPR is defined as the minimum value of the critical power ratio at which tbo fuel could be operated. with the expected number of rods 'in botling transitian not exceeding 0.1% of the fuel rods in the core.
The safety limit is the minimum critical power ratio which would be permitted to occur during the limiting anticipated operational occurrence. The safety i 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 unchangad 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 S rv of Results )
The results of the Grand Gulf Unit 1 Cycle 4 thermal limits analysis )
confim the Cycle 3 safety limit MCPR of 1.06 and a MCPR operating limit of 1.18 at rated conditions.
O
ANF-88-150 0- Page 18 L
3.5.1 Power Dependent Thermal Limits and Values The power dependent MCPR limit (MCPR p ) protects against exceeding the safety limit MCPR during anticipated opeic ticcal occurrences from off-rated power conditions. The MCPRp limit is determ1i.d Sy ad4;ng the delta CPR for the limiting event to the calculated safety limit A.S.
The power dependent MAPFAC (MAPFAC p ) is used to protect against both fuel melting and 1% clad strain during anticipated syste' 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 MAPFAC p 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. Delow 40% powe , the Cyc.le 3 MCPR p limit remains applicable to Cycle 4. The Cycle 4 MAPFACp value remains unchanged from Cycle 3.
3.5.2 Flow Denendent Thermal limits and Values The flow dependent MCPR limit (MCPRr) protects against exceeding the safety limit MCPR for flow excursion events. The results of the MCPRr 4 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 MAiTACf values to be used in Cycle 4 are presented in Figure 2.4.
The flow dependant thermal limits were confirmed to be conservative for Cycle 4 operation.
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ANF-88-150 f Y Page 19 J
Table 3.1 Grand Gulf Unit 1 Cycle 4 LFWH Data Summary Initial State Final State Cycle Total Core Total Core Core Total Core Total Core Core Exposure Power Flow Minimum Power Flow Minimum (GWd/MT) MWt (Mlb/hr) CPR MWt (Mlb/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 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. 112.50 2.91 2.000 3833.0 118.13 1.35 432" s 118.13 1.25 2.000 3066.4 118.13 1.68 3bl4,5 118.13 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 3.500 3833.0 84.38 1.19 4360.2 84.38 1.13 3.500 2376.5 34.88 1.37 2733.9 34.88 1.27
/~^)
(- ~ 3.500 1533.2 112.50 3.27 1728.0 112.50 3.00 5.000 3833.0 118.13 1.31 4359.8 118.13 1.23 5.000 3066.4 118.13 1.63 3496.4 118.13 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 i 6.000 3333.0 118.13 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 1.15 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 1.48 11.000 3833.0 84.38 1.15 4337.1 84.38 1.08
, 11.000 2376.5 34.88 1.40 2713.7 34.88 1.30
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- 11.000 1533.2 112.50 2.96 1718.7 112.50 2.71 m
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Figure 3.7 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local Power Distribution (XN-1 2.99-5G3 Fuel)
- - Gado11nia Location
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Figure 3.8 Grand Gulf Unit 1 Cycle 4 Safety Limit Design Basis Local !
Power Distribution (XN.2 3.21 6G4 Fuel)
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i ANF-88 150
,a Page 29 O
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 in the ASME Pressure Vessel Code. This analysis showed that the Grand Gulf Unit I 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 Desian 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 r.s 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
') performed with ANF's plant simulation code, COTRANSA, which includes an 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 valver, the compressibility of the additional fluid in the steam lines associated witn a turbine isolation causes these faster closures to be less severe. Oace the containmen+ is isolated, the subsequent core power production must be absorbed in a aaller volume compared to that of a turbine isolation resulting in higher vessel pressures.
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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 and plant parameters during the MSIV closure event from 104.2% power /108% i flow. These results show that the Grand Gulf Unit 1 safety valves have sufficient capacity and performance to protect the previously established q maximum vessel pressure safety limit of 1375 psig for Cycle 4. Two state {
points were r.nalyzed in or'er d to cover the MEOD range for full power operation.
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5.0 REFERENCES
L
- 1. 'Lester L. Kintner, USNRC, Letter 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 Plant Transient Analysis," XN-NF-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 Fuels Corporation, Richland, WA, November 1988.
- 4. "BWR/6 Generic Rod Withdrawal Error Analysis;' MCPR 3 for Plant Operations within ithe Extended Operation Domain," XN-NF-82E(P)(A), Supplement 2, -
Exxon Nuclear Company, Inc., Richland, WA, October 1986.
5.- " Exxon Nuclear Plant Transient Methodology for. Boiling Water ' Reactor,"
XN-NF-79-71(P),' Revision 2, including Supplements 1 ~, 2 & 3 ( A) , Exxon Nuclear Company,~Inc., Richland, WA, November 1981.
- 6. "XCOBRA-T: A Computer Code for BWR Transient Thermal Hydraulic Core Analysis," XN-NF-84-105(P)( A), Volume 1, Exxon Nuclear Company, Inc.,
Richland, WA, February 1987.
- 7. " Exxon : Nuclear Methodology for Boiling Water Reactors:' Neutronics o . Methods for Design and Analysis,* XN-NF-80-19fA), Volume 1, Exxon Nuclcar Company, Inc., Richland, WA, March 1983; as supplemented by Letter, R. A. Copeland, Advanced Nuclear Fuels, to M. W. Hodges, USNRC, " Void History Correlation," RAC:058:88, September 13, 1988. j
- 8. " Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal .j Limits Methodology Summary Description," XN-NF-80-19(P)(A), Volume 3, '
Revision 2, Exxon Nuclear Company, Inc., Richland, WA, January 1987. )
- 9. '" Exxon Nuclear Critical Power Methodology for Boiling Water Reactor,"
XN-NF-524(P)(A), Revision - 1, Exxon Nuclear Company, Inc., Richland, WA, November 1983. .
- 10. " Generic Mechanical Design for Exxon Nuclear det Pump BWR Reload Fuel,"
XN-NF-85-67(P)(A), Revision 1, Exxon Nuclear Company, Inc., Richland, WA, September 1986.
- 11. " Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," XN-NF-80-19(P)(A), Volume 4, Revision I, Exxon Nuclear Company, Inc., Richland, WA, June 1986.
- 12. " Grand Gulf Nuclear Station Unit 1 Revised Flow Dependent Thermal Limits," NESD0-88-003, MSU System Services Inc., November 1988.
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I ANF-88-150 Issue Date: 11/11/88-l 1.
l GRAND GULF UNIT 1 CYCLE 4 PLANT TRANSIENT ANALYSIS ,
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l' Distribution D. A. Adkisson D. J. Braun O. C. Brown M. E. Byram R. E. Collingham R. A. Copeland W. S. Dunnivant L. J. Federico N. L. Garner R. G. Grunner D. E. Hershberger M.-J. Hibbard
.Os T. L. Krysinski A. Reparaz R. S. Reynolds S. E. State R. B. Stout C. J. Volmer G. N. Ward H. E. Williamson SERI/N. L. Garner (40)
Document Control (5) i O
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1 ANF-88-183(NP)
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ADVANCEDNUCLEARFUELS CORPORATION l
1 GRAND GULF UNIT 1 RELOAD XN-1.3, CYCLE 4 MECHANICAL DESIGN REPORT l
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I JANUARY 1989 i
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.ANF-88-183(NP), Rev. O i
Issue Date: 1/6/89 GRAND GULF UNIT 1 RELCAD IN-1.3. CYCLE 4 MECHANICAL DESIGN REPORT PREPARED BY: S. -__- S /- 4-E6 W. S. Dunnivant Date Project Engineer CONCURRED BY: M N'
,A.'Reparpf, Manager' D4te BWR Enpfneerin ,/
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APPROVED BY: //6!P'7 G. J. Buss 1(Iman, Manager Date Fuel Design J . Y. t $c iftl Y G. L. Ritter, Manager D' ate g
Fuel Engineering & Technical Services
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< ANF-88-183(NP), Rev. O GRAND GULF UNIT 1 -
ELQ$0 IN-1.3. CYCLE 4 .
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TABLE OF CONTEL'Il J 1
i Section Ittle gggg
1.0 INTRODUCTION
. . . . ............s ........... 1 2.0
SUMMARY
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.1 Design Description Summary ..................... 2 I
i 2.2 Mechanical Design Summary . . . . . . . . . . . . . . . . . . . . . . 2 3.0 DESIGN CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4.0 MECHANICAL DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . 5 (3 m) l 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 l 4.1.4 Fuel Rod Internal Pressure . . . . . . . . . . . . . . . . . . . . 7 f
- 4.1.5 Fuel Pellet Centerline Temperature . . . . . . . . . . . . . . . . 7 ;
4.1.6 fuel Rod Cladding Fatigue .................... 8 l 1
4.1.7 Cladding Collapse ........................ 8 i
i 4.1.8 Fuel Rod Spacing . . . . . . . . . . . . . . . . . . . . . . . . . 8 4.1.9 Cladding Corrosion and Hydrogen Concentration .......... 9 j 1
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' i GRAND GULF UNIT 1 .!
RELQAD IN l.3. CYCLE 4 MECHANICAL DESIGN REPORT
1.0 INTRODUCTION
j 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, j supporting analyses, and test results for the Advanced Nuclear fuel s i
! Corporation Jet Pump Boiling Water Reactor Type 6 reload fuel for the Grand Gulf Unit 1 Nuclear Power Reactor.
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 I XN-1.3 is essentially the sanie as the generic ANF Type 4/5/6 designt 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 ;
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 1 XN-1.3 has been evah.ated to allow operation up to a peak assembly exposure greater than 39,000 mwd /MTU.
The results of the evaluation indicate that the Design Criteria are met' The fuel description mechanical design is summarized below.
2.1 Desian Description Summarv The ANF 8x8 assembly design for Grand Gulf 1 XN-1.3 reload uses 62 fuel I rods and two centrally located water rods, one of which functions as a spacer capture rod. Seven spacers maintain fuel rod spac'ing. The design uses a quick-removable upper tie plate design to facilitate fuel inspection and bundle reconstitution of irradiated assemblies.
The fuel rods are Zircaloy-2 cladding. The rc,ds are pressurized, and contain either UOp -Gd 230 or U02. Natural uranium axial fuel blanketina. at
/
the top and the bottom of the fuel column, is provided for greater neutron economy.
Two smail modifications to the previous fuel design for Grand Gulf I have been implemented, due to the increased fuel exposure of 39,000 mwd /MTU. Thus, the redesigned assembly maintains the same design margins at 39,00.0 mwd /MTV of
! ttiose that exist at 35,000 mwd /MTU.
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
, justify irradiation to 39,000 mwd /MT peak assembly burnup.
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.A, l-ANF-88-183(NP), Rev. 0 Page 4 3.0 DESIGN CLITERIA 1he detailed design criteria fc: the Advanced Nuclear Fuels Corporation Jet '/ ump Boiling Water coactor 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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l 4.0 MEQlANICAL DESIGN Two reports have already been issued to document the mechanical tiesign analyses for the Grand Gulf 1 ANF 8x8 fuel. These reports are XN-NF 83-25, Rev. 1 " Grand Gulf I 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.
l 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 submitted and approved for generic use by the NRC in 1986.
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.
The fuel assembly has been analyzed to a peak assembly exposure of 39,000 l mwd /MTU. The analyses have been performed assuming a design power history 1 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 exposurd is
! linearly extrapolated from that defined in XN-NF-85-67, Rev.1.
4.1 Fuel Rod Analyses f
Fuel rod analyses, where required, have been performed to verify adequate i performance of the fuel to 39,000 mwd /MTU assembly exposure. The exposures assumed are conservative estimates of t'n- maximum exposures to be reached with the Grand Gulf 1 XN-1.3 8x8 reload fut The design power history used in l
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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 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 Operational 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 reviewer 1 to determine if the higher exposure of the Grand Gulf 1 XN-1,3 8x8 fuel rcquires a reanalysis. It has been determined tnti 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. 1.re applicable to this design.
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' j 4.1.4 Fuel Rod Internal PreHER The fuel rod internal pre:isure is limited to the design criteria pressure. The analysis in XN-NF-85-67, Rev. I have been extended up to a j conservative estimate of the maximum assembly exposure of 39,000 mwd /MTV. The J l
, enalysis indicate that the maximum internal pressure is below the design criteria requirement.
4.1.5 Fuel pellet Centerline Temperature j 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 i indicates that the minimum margin against fuel melting, accounting for the s
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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 mwd /MTU and 39,000 mwd /MTU; the maximum fuel rod channel closure at 39,000 mwd /MTU, however. provides ample margin to the channel closure that could affect the thermal performance.
4.1.9 Claddina corrosion and Hydroaen Concentratiori 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 1 have been evaluated and the effects of 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.
\ Figures 3.11 and 3.14 of Reference 1 provide the information for residence times consistent with 35,000 mwd /MTU assembly exposure. Assuming the residence time is increased by 12% to accumulate the new design exposure, the evaluation indicates the design criteria are met.
1 i
4.2 Fuel Assembiv Evaluation l 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. I are applicable.
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ANF-88-183(NP), Rev. 0
, Issue Date: 1/6/89 i
i EPh e GULF UNIT 1 RELGAD IN-1.3. CYCLE 4 K CHANICAL DESIGN REPORT DISTRIBUTION W. S. Dunnivant N. L. Garner (8)
T. L. Krysinski A. Reparaz Document Control (5) 1 I
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