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{{#Wiki_filter:0KN-N-79-103PE(lllklkPlljlllLEkk PILklll'NllllLE JIBMFET7kkkL7SIIS PEPBP'll'IIYM Iw3IIXEBQXIIBEbhSSEliNBILIIES DECEMBER1979RICHLAND, NA99352 EI'II-II XN-NF-79-103 0:IR/R1479R.E.GINNANUCLEARPLANTCYCLE10.SAFETYANALYSISREPORTWITHMIXEDOXIDEASSEMBLIES Prepared:
{{#Wiki_filter:0 KN-N-79-103 P E (lllklk PlljlllLEkk PILklll'NllllLE JIB MFET7 kkkL7SIIS PEPBP'll'IIYM Iw3IIXEB QXIIBE bhSSEliNBILIIES DECEMBER 1979 RICHLAND, NA 99352 E I'I I-I I XN-NF-79-103 0: IR/R14 79 R.E.GINNA NUCLEAR PLANT CYCLE 10.SAFETY ANALYSIS REPORT WITH MIXED OXIDE ASSEMBLIES Prepared: G.J.Buss man, Manager Neutronics and Fuel Management Approved: G.A.Sofe nager Nuclear Fue s Engineering Concurred:
G.J.Bussman,ManagerNeutronics andFuelManagement Approved:
J.N.Morgan, Manage Licensing and Safety Engineering Concurred:
G.A.SofenagerNuclearFuesEngineering Concurred:
L.J.Federico, Manager Nuclear Fuels Project l'~/7 E)j(ON NUCLEAR COMPANY, Inc.
J.N.Morgan,ManageLicensing andSafetyEngineering Concurred:
NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT Ir PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Exxon Nuclear Company, Inc.It is being Sub.mitted by Exxon Nuclear to the USNRC as part of a technical contri.bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear.fabricated reioarl fuel or other teclmical services provided by Exxon Nuclear for lieht water power reactors an<I it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief.The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNQC whicII are customers ot Exxon Nuclear in their demonstration of compliance wIth the USNRC's regulations.
L.J.Federico, ManagerNuclearFuelsProjectl'~/7E)j(ONNUCLEARCOMPANY,Inc.
Without derogating from the foregoing, neither Exxon Nuclear nor any person acting on its behalf: A.Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained,in this document, or that the use ot any information, apparatus, method, or process disclosed in this document will not Itffringe privately owned rights;or 8.Assumes any liabilities with respect to the use of, or for dan'ages resulting from the use of, any information, ap.paratus, method, or process disclosed in this document.XN-NF-FQO, 766 XN-NF-79-103 TABLE OF CONTENTS'ection
NUCLEARREGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICEREGARDING CONTENTSANDUSEOFTHISDOCUMENTIrPLEASEREADCAREFULLY Thistechnical reportwasderivedthroughresearchanddevelopment programssponsored byExxonNuclearCompany,Inc.ItisbeingSub.mittedbyExxonNucleartotheUSNRCaspartofatechnical contri.butiontofacilitate safetyanalysesbylicensees oftheUSNRCwhichutilizeExxonNuclear.fabricated reioarlfuelorotherteclmical servicesprovidedbyExxonNuclearforliehtwaterpowerreactorsan<IitistrueandcorrecttothebestofExxonNuclear's knowledge, information, andbelief.Theinformation contained hereinmaybeusedbytheUSNRCinitsreviewofthisreport,andbylicensees orapplicants beforetheUSNQCwhicIIarecustomers otExxonNuclearintheirdemonstration ofcompliance wIththeUSNRC'sregulations.
Withoutderogating fromtheforegoing, neitherExxonNuclearnoranypersonactingonitsbehalf:A.Makesanywarranty, expressorimplied,withrespecttotheaccuracy, completeness, orusefulness oftheinfor-mationcontained,in thisdocument, orthattheuseotanyinformation, apparatus, method,orprocessdisclosed inthisdocumentwillnotItffringe privately ownedrights;or8.Assumesanyliabilities withrespecttotheuseof,orfordan'agesresulting fromtheuseof,anyinformation, ap.paratus,method,orprocessdisclosed inthisdocument.
XN-NF-FQO,766 XN-NF-79-103 TABLEOFCONTENTS'ection


==1.0INTRODUCTION==
==1.0 INTRODUCTION==


ANDSUMMARY.2.0OPERATING HISTORYOFTHEREFERENCE CYCLE.3.0GENERALDESCRIPTION 4.0FUELSYSTEMDESIGN.~Pae2595.0NUCLEARDESIGN.................'........
AND
105.1PHYSICSCHARACTERISTICS 5.1.1POWERDISTRIBUTION CONSIDERATIONS.
5.1.2CONTROLRODREACTIVITY REQUIREMENTS.
125.1.3MODERATOR TEMPERATURE COEFFICIENT CONSIDERATIONS
..135.2ANALYTICAL METHODOLOGY.
136.0THERMALHYDRAULIC DESIGN....................
207.0ACCIDENTANDTRANSIENT ANALYSIS.
7.1PLANTTRANSIENT ANDECCSANALYSESFORR.E.GINNA.7.2RODEJECTIONANALYSISFORR.E.GINNACYCLE10.


==8.0REFERENCES==
==SUMMARY==
.2.0 OPERATING HISTORY OF THE REFERENCE CYCLE.3.0 GENERAL DESCRIPTION 4.0 FUEL SYSTEM DESIGN.~Pa e 2 5 9 5.0 NUCLEAR DESIGN.................'........
10 5.1 PHYSICS CHARACTERISTICS 5.1.1 POWER DISTRIBUTION CONSIDERATIONS.
5.1.2 CONTROL ROD REACTIVITY REQUIREMENTS.
12 5.1.3 MODERATOR TEMPERATURE COEFFICIENT CONSIDERATIONS
..13 5.2 ANALYTICAL METHODOLOGY.
13 6.0 THERMAL HYDRAULIC DESIGN....................
20 7.0 ACCIDENT AND TRANSIENT ANALYSIS.7.1 PLANT TRANSIENT AND ECCS ANALYSES FOR R.E.GINNA.7.2 ROD EJECTION ANALYSIS FOR R.E.GINNA CYCLE 10.


.~~21212225
==8.0 REFERENCES==
~~~.l~~~t~~t XN-NF-79-103 LISTOFTABLESTable~Pae3.1R.E.GINNACYCLE10FUELASSEMBLYDESIGNPARAMETERS
-.....65.1R.E.GINNANEUTRONICS CHARACTERISTICS OFCYCLE10COMPAREDWITHCYCLE9DATA....................
155.2R.E.GINNACONTROLRODSHUTDOWNMARGINSANDREQUIREMENTS FORCYCLE10~~~~~~~~~~~~~~~~~~~~~~~~II7.1R.E.GINNAKINETICPARAMETERS
.16~~237.2EJECTEDRODWORTHANDPEAKINGFACTORS..............24
\~~~~~~~~~l~~iI~~~~~
XN-NF-79-103 LISTOFFIGURES~Fiure2.1R.E.GINNACYCLE9CRITICALBORONCURVE,PREDICTED VS.MEASUREDe~~~~~~~~~~~~~~~~~~~~~~~~2.2R.E.GINNAPOWERDISTRIBUTION COMPARISON TOMAPIX-245)505MWD/MTe~~~~~~~~~~~~~~~~~~~~~~3.1R.E.GINNACYCLE10LOADINGPATTERN.3.2R.E.GINNABOC10QUARTERCOREEXPOSUREDISTRIBUTION ANDREGIONID@~~~~~~~~~~~~~~~~~~~~~~~~5.1R.E.GINNACYCLE10AROCRITICALBORONCONCENTRATION VS.EXPOSURE~~~~~~~~~~~~~~~~~~~~~~~~~5.2R.E.GINNACYCLE10POWERDISTRIBUTION HFP,0MWD/MT1)254PPM5.3R.E.GINNACYCLE10POWERDISTRIBUTION HFP,9,500MWD/MTPPM~~~~~~~~~~~~~~~~~~~~~~~~~~7~Pae~~3478171818 t1~~~~~~I~li~~~~l XN-NF-79-103 R.E.GINNANUCLEARPLANTCYCLE10SAFETYANALYSISREPORTkgl


==1.0INTRODUCTION==
.~~21 21 22 25
~~~.l~~~t~~t XN-NF-79-103 LIST OF TABLES Tabl e~Pa e 3.1 R.E.GINNA CYCLE 10 FUEL ASSEMBLY DESIGN PARAMETERS
-.....6 5.1 R.E.GINNA NEUTRONICS CHARACTERISTICS OF CYCLE 10 COMPARED WITH CYCLE 9 DATA....................
15 5.2 R.E.GINNA CONTROL ROD SHUTDOWN MARGINS AND REQUIREMENTS FOR CYCLE 1 0~~~~~~~~~~~~~~~~~~~~~~~~II 7.1 R.E.GINNA KINETIC PARAMETERS
.16~~23 7.2 EJECTED ROD WORTH AND PEAKING FACTORS..............24
\~~~~~~~~~l~~i I~~~~~
XN-NF-79-103 LIST OF FIGURES~Fi ure 2.1 R.E.GINNA CYCLE 9 CRITICAL BORON CURVE, PREDICTED VS.MEASURED e~~~~~~~~~~~~~~~~~~~~~~~~2.2 R.E.GINNA POWER DISTRIBUTION COMPARISON TO MAP IX-24 5)505 MWD/MT e~~~~~~~~~~~~~~~~~~~~~~3.1 R.E.GINNA CYCLE 10 LOADING PATTERN.3.2 R.E.GINNA BOC10 QUARTER CORE EXPOSURE DISTRIBUTION AND REGION ID@~~~~~~~~~~~~~~~~~~~~~~~~5.1 R.E.GINNA CYCLE 10 ARO CRITICAL BORON CONCENTRATION VS.EXPOSURE~~~~~~~~~~~~~~~~~~~~~~~~~5.2 R.E.GINNA CYCLE 10 POWER DISTRIBUTION HFP, 0 MWD/MT 1)254 PPM 5.3 R.E.GINNA CYCLE 10 POWER DISTRIBUTION HFP, 9,500 MWD/MT PPM~~~~~~~~~~~~~~~~~~~~~~~~~~7~Pa e~~3 4 7 8 17 18 18 t 1~~~~~~I~l i~~~~l XN-NF-79-103 R.E.GINNA NUCLEAR PLANT CYCLE 10 SAFETY ANALYSIS REPORT k gl


ANDSUMMARYTheR.E.GinnaNuclearplantwilloperateinCycle10beginning inearly1980withthreeregionsoffuelsuppliedbyExxonNuclearCompany(ENC).Theloadingwillconsistof32ENCassemblies inRegion12and4Westinghouse mixoxide(MOX)assemblies.
==1.0 INTRODUCTION==
Theremainder ofthecorecontains40once-burnt and32twice-burnt ENCassemblies and13exposedWestinghouse suppliedassemblies.
 
Thecharacteristics ofthefuelandofthereloadedcoreresultinconformance withexistingTechnical Specification limitsregarding shutdownmarginprovisions andthermallimits.Thisdocumentprovidestheneutronic analysisfortheplantduringCycle10operation andthecontrolrodejectionanalysis.
AND
TheENCfueldesign.isunchanged fromthefueldesignusedinthe"Cycle 8and9ENCfuelreloads.ThepreviousPlantTransient Analysis'(2)remainsvalidforCycle10.TheECCSanalysisisapplicable toCycle10operation.
 
Theconsequences oftherodejectionaccidentforCycle10areslightlylessseverethanthosecalculated forCycles8and9.The(4)(5)introduction ofthe4MOXassemblies intothereactorcoreleadstosmallchangesinthecoreaveragekineticparameters resulting inminimaleffectstothepreviousanalysesperformed forCycles8'and9(1,2,4)(1,3,5)  
==SUMMARY==
The R.E.Ginna Nuclear plant will operate in Cycle 10 beginning in early 1980 with three regions of fuel supplied by Exxon Nuclear Company (ENC).The loading will consist of 32 ENC assemblies in Region 12 and 4 Westinghouse mix oxide (MOX)assemblies.
The remainder of the core contains 40 once-burnt and 32 twice-burnt ENC assemblies and 13 exposed Westinghouse supplied assemblies.
The characteristics of the fuel and of the reloaded core result in conformance with existing Technical Specification limits regarding shutdown margin provisions and thermal limits.This document provides the neutronic analysis for the plant during Cycle 10 operation and the control rod ejection analysis.The ENC fuel design.is unchanged from the fuel design used in the"Cycle 8 and 9 ENC fuel reloads.The previous Plant Transient Analysis'(2)remains valid for Cycle 10.The ECCS analysis is applicable to Cycle 10 operation.
The consequences of the rod ejection accident for Cycle 10 are slightly less severe than those calculated for Cycles 8 and 9.The (4)(5)introduction of the 4 MOX assemblies into the reactor core leads to small changes in the core average kinetic parameters resulting in minimal effects to the previous analyses performed for Cycles 8'and 9 (1,2,4)(1,3,5)  
~~~~~l~~~}i~~~~~~~
~~~~~l~~~}i~~~~~~~
XN-NF-79-103 2.0 OPERATING HISTORY OF THE REFERENCE CYCLE R.E.Ginna Cycle 9 has been chosen as the reference cycle with respect to Cycle 10 due to the close resemblance of the neutronic characteristics between these two cycles.The Cycle 9 operation began on April 3, 1979, and as of November 31, 1979 the core had accrued about 6,714 MWD/MT.The Cycle 9 loading included 40 fresh ENC fuel assemblies with 32 exposed ENC assemblies and 49 exposed Westinghouse assemblies.
The measured power peaking factors at hot-full-power, equilibrium xenon conditions, have remained considerably below the Technical Specification ,limits throughout Cycle 9.The total nuclear peaking factors, F , and the radial nuclear pin peaking factor, F H, have remained below 1.75 and 1.45, respectively.
Cycle 9 operation has typically been rod'free with the D control bank positioned in the range of 218 to 222 steps, 225 steps being fully withdrawn.
It is anticipated that similbr control bank insertions will be seen in Cycle 10.The critical boron concentration as calculated by ENC for Cycle 9 has agreed to within about 8 ppm compared to the observed values (see Figure 2.1).Also the predicted power distributions have typically agreed to within+3 percent of the, measured values (see Figure 2.2 for typical com-parison).
-~~--~~-~~=-:LL~=*~-:..t=~I-~~-~t 44 C-l:-f~I~~t*'I t j=~~~++4 I~~=~~~~-~~~~~-~~-~*t-Ut'e.~----I.fl GMO'NT.}'~-:.-::F.-T''re
~t~-:Cri tTca t Boron Cu ve,,=P=red:.~-~-~-
XN-NF-79-103
XN-NF-79-103
 
.968.995-2.71 1.110 1.129-1.68.920.938-1.92.949..965-1.66 1.189 1.178'93.996.963 3.43.794.782 1.53 1.108 1.131-2.03.986 1.013-2.67 1.029 1.048-1.81 1.099 1.103-.36 1.189 1.168 1.80 1.074 1.051 2.19.663.645 2.79.918.947-3.06 1.030 1.051-2.00.996 1.006-.99 1.203 1.191 1.01 1.079 1.055 2.27;977.974.31.953.968-1.55 1.098 1.105-.63 1.191 1.192-.08 1.036 1.021 1.47 1.178 1.149 2.52.714.711.42 10 1;195 1.179 l.36 1.188 1.169 1.63 1.072 1.055 1.61 1.173 1.149 2.09.804.798.75.985.963 2.28.772.782-1.28 1.071 1.051 1.90.651 ,.645.93.977.975.21.715.712.42 Measured Assembly Power Calculated (XTGPWR)x 100 c 12 13 Calculated Measured%Difference N Fq 1.528 1.564 2.33 F~H 1.351 F 1.105 1.337 1.154-1.07 4.42 Figure 2.2 R.E.Ginna Power Distribution Comparison, To Map IX-24, HFP, 5,505 MWD/MT I-l XN-NF-79-103 3.0 GENERAL DESCRIPTION The R.E.Ginna reactor consists of 121 assemblies, each having a 14xl4 fuel rod array.Each assembly contains 179 fuel rods, 16 RCC guide tubes, and 1 instrumentation tube.The fuel rods consist of slightly enriched U02 pellets inserted into zircaloy tubes.The RCC guide tubes and the instr umen-tation tube are made of SS-304L.Each ENC assembly contains nine zircaloy spacers with Inconel springs;eight of the spacers are located within the active fuel region.Four of the 121 assemblies contain Mixed Oxide (Pu02 plus U02)bearing fuel rods.The MOX assemblies consist of three enrichment zones of Pu02 utilizing natural U02 as the diluent.The projected Cycle 10 loading pattern is shown in Figure 3.1 with the assemblies identified by their Fabrication ID's and Region ID's.The initial enrichments of the various regions are listed in Table 3.1.BOC10 exposures, based on an EOC9 exposure of 9,570 MHD/MT, along with Region ID's are shown in Figure 3.2.The core consists of 32 fresh ENC assemblies at 3.45 w/o and 4 fresh Westinghouse MOX assemblies loaded on the periphery with 72 ENC and 13 Westinghouse exposed assemblies scatter-loaded in the center portion of the core.Pertinent fuel assembly parameters for the Cycle 10 fuel are depicted in Table 3.1.The transuranic elements, including Am-241, have been\accounted for up to the time of the anticipated reactor startup.
==2.0 OPERATING==
XN-NF-79-103 Table 3.1 R.E.Ginna Cycle 10 Fuel Assembly Design Parameters 10 Region 12 MOX Enrichment, wtX U-235 Number of Assemblies Pellet Densi'ty, X TD Pellet-to-Clad Diametrical Gap, Mil Fuel Stack Height, inch 7.5 7.5 141.4 142.0 3,103 3,100 13 32 95,0 94,0 40 94.0 32 4 94.0 95.0 7.5 7.5 7.5 142.0 142.0 141.4 3.200 3.450 2.626*Region Average Burnup at BOC10, MWD/MT Nominal Assembly Weight, KgU 24,339 17,885 8,335 0 0 392.56 373.78 373,78 373.78 395.91***wtX Pu (based on assembly average)*" in Kg HM XNrNF-79-103 K J I H.G F E 0 C B A 12 HOX 12 12 12 N09'12 L14 12 L09 M14 K03 L31 M39 12 12 L01 H01 12 12 M02 L06 N17 L19 M28 L26 N36 L05 N12 12 12 L02 M33 L21 M23 K05 N30 L24 M20 L12 12 12 MOX L32 M40 L27 12 K09 M25 M31 K13 N07 K20 N08 K28 K19 N06 N22 L18 H13 K18 M27 K27 L13 12 12 MOX 12 L15 M15 L20 M24 K26 M05 K17 H29 L25 M38 L30 12 12 Ll 0 M18 L22 N32 K25 M21 L23 M35 L04 12 12 Ml 0 L07 12 M03 N34 L28 N26 L03 M37 K14 L17 H16 M19 L08 M04 Lll Mll 12 12 12 12 L29 12 L16 12 12 12 MOX 12 Fabri cation or New Fuel Region Identification Figure 3.1 R.E.Ginna Cycle 10 Loading Pattern XN-NF-79-103 D C B 24,736 7,517 24,117 7,809 24,708 0 12 MOX 7;522 24,093 11,341 17,578 9,544 10 19,816 10 12 24,117 11,342 17,506 6,926 16,061 10 10 12 7,809 17,574 6,928 10 18,677 10 6,393 0 12]n 24,708 9,549 16,061 6,198 10 12 12 19,809 10 0 a 12 12 BOC10 Exposure MWD/MT Region ID*12 0 MOX 12 13*See Table 3.1 for Region definitions f Figure 3.2 R.E.Ginna BOC10 quarter Core Exposure Distribution and Region ID XN-NF-79-103 4.0 FUEL.SYSTEM DESIGN A description of the Exxon Nuclear supplied fuel design and design methods is contained in Reference 1.This fuel has been specifically designed to be compatible to the resident fuel supplied by Westinghouse.
HISTORYOFTHEREFERENCE CYCLER.E.GinnaCycle9hasbeenchosenasthereference cyclewithrespecttoCycle10duetothecloseresemblance oftheneutronic characteristics betweenthesetwocycles.TheCycle9operation beganonApril3,1979,andasofNovember31,1979thecorehadaccruedabout6,714MWD/MT.TheCycle9loadingincluded40freshENCfuelassemblies with32exposedENCassemblies and49exposedWestinghouse assemblies.
I 10 XN-NF-79-103 5.0 NUCLEAR DESIGN The neutronic charact ristics of the projected Cycle 10 core are quite similar to those of the Cycle 9 core (see Section 5.1).The nuclear design bases for the Cycle 10 core are as follows: 1)The design shall permit operation within the Technical Specifications for the R.E.Ginna plant.2)The length of Cycle 10 shall be determined on the basis of an assumed Cycle 9 length of 9,570 MWD/MT.3)The Cycle 10 loading pattern shall be optimized to achieve power distributions and control rod reactivity worths according to the following constraints:
Themeasuredpowerpeakingfactorsathot-full-power, equilibrium xenonconditions, haveremainedconsiderably belowtheTechnical Specification
a)The peak F~shall not exceed 2.32 and the peak F H shall not exceed 1.66 (including uncertainties) in any single fuel rod.through the cycle under nominal full power operation condi-tions.b)The scram worth of all rods minus the most reactive shall exceed BOC and EOC shutdown requirements.
,limitsthroughout Cycle9.Thetotalnuclearpeakingfactors,F,andtheradialnuclearpinpeakingfactor,FH,haveremainedbelow1.75and1.45,respectively.
4)The Cycle 10 core shall have a negative power coefficient.
Cycle9operation hastypically beenrod'freewiththeDcontrolbankpositioned intherangeof218to222steps,225stepsbeingfullywithdrawn.
5)The MOX assemblies shall be located in a region of the reactor core as to minimize the effects on shutdown margin provisions and thermal limits.The neutronic design methods utilized to ensure the above requirements are consistent with those described in References 6, 7, and 8.
Itisanticipated thatsimilbrcontrolbankinsertions willbeseeninCycle10.Thecriticalboronconcentration ascalculated byENCforCycle9hasagreedtowithinabout8ppmcomparedtotheobservedvalues(seeFigure2.1).Alsothepredicted powerdistributions havetypically agreedtowithin+3percentofthe,measuredvalues(seeFigure2.2fortypicalcom-parison).  
XN-NF-79-103 5.1 PHYSICS CHARACTERISTICS The neutronic characteristics of the Cycle 10 core are compared with those of Cycle 9 and are presented in Table 5.1.The data presented in the table indicate the'eutronic similarity between Cycles 9 and 10.The Cycle 10 loading pattern is applicable for Cycle 9 lengths of+700 MWD/MT and-800 MWD/MT about the nominal length of 9,570 MWD/MT.The calculated boron letdown curve for Cycle 10 is shown in Figure 5.1.The curve indicates a BOC10, no xenon, critical boron concentration of 1,254 ppm.At 150 MWD/MT, equilibrium xenon, the critical boron concentration is 921 ppm.The Cycle 10 length is projected to be 9,500+300 MWD/MT with 7 ppm of boron at EOC.5.1.1 Power Distribution Considerations P Representative predicted power maps for Cycle 10 are shown in Figures 5.2 and 5.3 for BOC and EOC conditions, respectively.
-~~--~~-~~=-:LL~=*~-:..t=~I-~~-~t44C-l:-f~I~~t*'Itj=~~~++4I~~=~~~~-~~~~~-~~-~*t-Ut'e.~----I.flGMO'NT.}'
The power distributions were obtained from a three-dimensional model with moderator I, density and Doppler feedback effects incorporated.
~-:.-::F.-T''re
For the projected Cycle 10 loading pattern the calculated BOC nuclear power peaking factors, F~, N N N F, and Fz, are,l.745, 1.433, and 1.201, respectively.
~t~-:CritTcatBoronCuve,,=P=red:.~-~-~-
At EOC conditions the corresponding values are 1.517, 1.358, and 1.098.The Technical Specifi-cation limits relative to F~and F>H, with the measurement uncertainties N N backed out, are 2.15 and 1.60.Additionally the predicted axial F distri-butions are well below the axially dependent Technical Specification limits on F~.The BOC F value of 1.745 compares with the measured Cycle 9 value N in Table 5.1 of 1.758.
XN-NF-79-103
12 XN-NF 79-,103 The control of the core power distribution is accomplished by following the procedures as discussed in the report, XN-76-40,"Exxon Nuclear Power Distribution Control for Pressurized Water Reactors", September 1976 and its addendum.The results reported in these documents demonstrate that the Power Distribution Control (PDC)procedures defined in the report will protect an axially dependent F limit with a peak value of 2.30.The Technical Specification limit for R.E.Ginna has a peak of 2.32 and an axial dependence identical to that supported by the procedures.
.968.995-2.711.1101.129-1.68.920.938-1.92.949..965-1.661.1891.178'93.996.9633.43.794.7821.531.1081.131-2.03.9861.013-2.671.0291.048-1.811.0991.103-.361.1891.1681.801.0741.0512.19.663.6452.79.918.947-3.061.0301.051-2.00.9961.006-.991.2031.1911.011.0791.0552.27;977.974.31.953.968-1.551.0981.105-.631.1911.192-.081.0361.0211.471.1781.1492.52.714.711.42101;1951.179l.361.1881.1691.631.0721.0551.611.1731.1492.09.804.798.75.985.9632.28.772.782-1.281.0711.0511.90.651,.645.93.977.975.21.715.712.42MeasuredAssemblyPowerCalculated (XTGPWR)x100c1213Calculated Measured%Difference NFq1.5281.5642.33F~H1.351F1.1051.3371.154-1.074.42Figure2.2R.E.GinnaPowerDistribution Comparison, ToMapIX-24,HFP,5,505MWD/MT I-l XN-NF-79-103
The physics characteristics of the Ginna Cycle 10 core are similar to those utilized in the PDC supporting analysis.The Ginna Technical Specification limits on F can therefore be protected by operation under the PDC procedures as stated in XN-76-40.5.1.2 Control Rod Reactivit Re uirements Detailed calculations of shutdown margins for Cycle 10 are compared with Cycle 9 data in Table 5.2.The ENC Plant Transient Simulation (PTS)Analysis indicates that the minimum required shutdown margin is 1,800 pcm based upon the steamline break accident analyzed for ENC fuel at the EOC conditions.
 
A value of 1,900 pcm is used at EOC in the evaluation of the shutdown margin to be consistent with'the Technical Specifications.
==3.0 GENERALDESCRIPTION==
The Cycle 10 analysis indicates excess shutdown margins of 1,414 pcm at the BOC and 344 pcm at the EOC.The Cycle 9 analysis indicates excess shut-down margins for that cycle of 1,795 pcm at the BOC and 393 pcm at the EOC.The slightly lower Cycle 10 excess shutdown margins, when compared to the Cycle 9 values, are due to slightly lower calculated rod worths.
TheR.E.Ginnareactorconsistsof121assemblies, eachhavinga14xl4fuelrodarray.Eachassemblycontains179fuelrods,16RCCguidetubes,and1instrumentation tube.ThefuelrodsconsistofslightlyenrichedU02pelletsinsertedintozircaloytubes.TheRCCguidetubesandtheinstrumen-tationtubearemadeofSS-304L.EachENCassemblycontainsninezircaloyspacerswithInconelsprings;eightofthespacersarelocatedwithintheactivefuelregion.Fourofthe121assemblies containMixedOxide(Pu02plusU02)bearingfuelrods.TheMOXassemblies consistofthreeenrichment zonesofPu02utilizing naturalU02asthediluent.Theprojected Cycle10loadingpatternisshowninFigure3.1withtheassemblies identified bytheirFabrication ID'sandRegionID's.Theinitialenrichments ofthevariousregionsarelistedinTable3.1.BOC10exposures, basedonanEOC9exposureof9,570MHD/MT,alongwithRegionID'sareshowninFigure3.2.Thecoreconsistsof32freshENCassemblies at3.45w/oand4freshWestinghouse MOXassemblies loadedontheperiphery with72ENCand13Westinghouse exposedassemblies scatter-loaded inthecenterportionofthecore.Pertinent fuelassemblyparameters fortheCycle10fuelaredepictedinTable3.1.Thetransuranic
13 XN-NF-79 103 The control-rod groups and insertion limits for Cycle 10 will remain unchanged from Cycle 9.With these limits the'nominal worth of the control bank, D-bank, inserted to the insertion limits'at HFP is 122 pcm at, BOC and'70 pcm at EOC.The control rod shutdown requirements in Table 1 5.2 allow for a HFP D-bank insertion equivalent to 300 pcm for both BOC and EOC.5.1.3 Moderator Tem erature Coefficient Considerations The reference Cycle 10 design calculations indicate that the moderator temperature coefficient is negative at all times during the cycle as shown in Table 5.1.This meets the Technical Specification requirement that the moderator temperature coefficient be negative at all times during power operation and the design criteria that the power coefficient be nega-tive.The least negative moderator temperature coefficient occurs at BOC HZP and is-2.0+2pcm/
: elements, including Am-241,havebeen\accounted foruptothetimeoftheanticipated reactorstartup.
F.This compares with the BOC9 HZP value of-2.0 pcm/F.5.2 ANALYTICAL METHODOLOGY The methods used in the Cycle 10 core analyses are described in References 6, 7, and 8.These methods have been verified for both U02 and Pu02-U02 lattices.In summary, the reference neutronic design analysis of the reload core was performed using the XTG (Reference 9)reactor simulator system-.The input exposure data were based on quarter core depletion calcu-lations performed from Cycle 5 to Cycle 9 using the XTG code.The BOC5 exposure distribution was obtained from plant data.The fuel shuffling between c'ycles was accounted for in the calculations.
XN-NF-79-103 Table3.1R.E.GinnaCycle10FuelAssemblyDesignParameters 10Region12MOXEnrichment, wtXU-235NumberofAssemblies PelletDensi'ty, XTDPellet-to
14 XN-NF-79-103 Predicted values of F~, Fx , and F were studied, with the XTG reactor model.The calculational thermal-hydraulic feedback and axial exposure distribution effects on power shapes, rod worths, and cycle lifetime are explicitly included in the analysis.
-CladDiametrical Gap,MilFuelStackHeight,inch7.57.5141.4142.03,1033,100133295,094,04094.032494.095.07.57.57.5142.0142.0141.43.2003.4502.626*RegionAverageBurnupatBOC10,MWD/MTNominalAssemblyWeight,KgU24,33917,8858,33500392.56373.78373,78373.78395.91***wtXPu(basedonassemblyaverage)*"inKgHM XNrNF-79-103 KJIH.GFE0CBA12HOX121212N09'12L1412L09M14K03L31M391212L01H011212M02L06N17L19M28L26N36L05N121212L02M33L21M23K05N30L24M20L121212MOXL32M40L2712K09M25M31K13N07K20N08K28K19N06N22L18H13K18M27K27L131212MOX12L15M15L20M24K26M05K17H29L25M38L301212Ll0M18L22N32K25M21L23M35L041212Ml0L0712M03N34L28N26L03M37K14L17H16M19L08M04LllMll12121212L2912L16121212MOX12FabricationorNewFuelRegionIdentification Figure3.1R.E.GinnaCycle10LoadingPattern XN-NF-79-103 DCB24,7367,51724,1177,80924,708012MOX7;52224,09311,34117,5789,5441019,816101224,11711,34217,5066,92616,0611010127,80917,5746,9281018,677106,393012]n24,7089,54916,0616,19810121219,809100a1212BOC10ExposureMWD/MTRegionID*120MOX1213*SeeTable3.1forRegiondefinitions fFigure3.2R.E.GinnaBOC10quarterCoreExposureDistribution andRegionID XN-NF-79-103 4.0FUEL.SYSTEMDESIGNAdescription oftheExxonNuclearsuppliedfueldesignanddesignmethodsiscontained inReference 1.Thisfuelhasbeenspecifically designedtobecompatible totheresidentfuelsuppliedbyWestinghouse.
15 XN-NF-79-103 Table 5.1 R.E.Ginna Neutronics Characteristics of Cycle 10 Compared with Cycle 9 Data BOC C cle 9 EOC C cle 10 BOC EOC (Critical Boron HFP, ARO, Equilibrium Xenon (ppm)HZP, ARO No Xenon (ppm)Moderator Temperature Coefficient HFP, (pcm/oF)HZP, (pcm/oF)Doppler Coefficient, (pcm/F)Boron Worth, (pcm/ppm)HFP HZP Total Nuclear Peaking Factor Fq, HFP 961')12(1,41 0(2)-8.12-8.58-8.72 1.758(-7.6(2)-30.4-2.0-21.5-1.25 to-2.0 921 1,414-8.1-2.0-30.4-21.6-7.95 ,-8.62 1.745 1.517-,1.35-1.84 Delayed Neutron Fraction.0061.0051.0058.0052 Control Rod Worth of All Rods In Minus Most Reactive Rod, HZP, (pcm)Excess Shutdown Margin (pcm)Moderator Pressure Coefficient (pcm/psi)5,751)5,821.1,795()393(0.35 5';341 5,696 1,414 344 0.35 (1)Extrapolated from measured data (2)Measured Data (3)70/Power Map (4)Reference 5 16 XN-NF-79-103 Table 5.2, R.E.Ginna Control Rod Shutdown Margins and Requirements for Cycle 10 BOC C cle 9**EOC C cle 10 BOC EOC Control Rod Worth HZP , cm All Rods Inserted (ARI)ARI less most reactive (N-1)N-1 less lOX allowance L(N-1)*9l Reactivit Insertion cm 5,176 5,239 6,407 6,634 5,751 5,821 5,949 6,420 5,341 5,696 4,807 5,125 Moderator plus Doppler Flux Redistribution Void Sum of the above three Rod Insertion Allowance Total Requirements 1,431 600 50 2,081 300 2,381 1,996 600 50 2,'646 300 2,946 1,443 1,932 600 600 50 50 2,093 2,582 300 300 2,393 2,882 Shutdown.Margin (N-l)*.9-Total Requirements Required Shutdown Margin*Excess Shutdown Margin 2,795 2,293 1,000 1,900 1,795 393 1,414 344 2,414 2,244 10000 1,900*Technical Specification 3.10"*Calculated values from Reference 5  
I 10XN-NF-79-103
~-~.~~'-~~*~-~*~I~>>.~=~~~=~~-~-t~=~~4 h~g~~=~=Wt WI.~'~~\-~*~4~t~~~->>*~~~~=-~=~-~-~-~~-Figure 5tl R.E.Ginna Cycle 10 ARO Critical Boron Concentration vs.Exposure r 4 el~~~~~~~~~~~~~~~~~~~~~~~I~~J Ll I I~~
 
==5.0 NUCLEARDESIGNTheneutronic==
charactristicsoftheprojected Cycle10corearequitesimilartothoseoftheCycle9core(seeSection5.1).ThenucleardesignbasesfortheCycle10coreareasfollows:1)Thedesignshallpermitoperation withintheTechnical Specifications fortheR.E.Ginnaplant.2)ThelengthofCycle10shallbedetermined onthebasisofanassumedCycle9lengthof9,570MWD/MT.3)TheCycle10loadingpatternshallbeoptimized toachievepowerdistributions andcontrolrodreactivity worthsaccording tothefollowing constraints:
a)ThepeakF~shallnotexceed2.32andthepeakFHshallnotexceed1.66(including uncertainties) inanysinglefuelrod.throughthecycleundernominalfullpoweroperation condi-tions.b)ThescramworthofallrodsminusthemostreactiveshallexceedBOCandEOCshutdownrequirements.
4)TheCycle10coreshallhaveanegativepowercoefficient.
5)TheMOXassemblies shallbelocatedinaregionofthereactorcoreastominimizetheeffectsonshutdownmarginprovisions andthermallimits.Theneutronic designmethodsutilizedtoensuretheaboverequirements areconsistent withthosedescribed inReferences 6,7,and8.
XN-NF-79-103 5.1PHYSICSCHARACTERISTICS Theneutronic characteristics oftheCycle10corearecomparedwiththoseofCycle9andarepresented inTable5.1.Thedatapresented inthetableindicatethe'eutronic similarity betweenCycles9and10.TheCycle10loadingpatternisapplicable forCycle9lengthsof+700MWD/MTand-800MWD/MTaboutthenominallengthof9,570MWD/MT.Thecalculated boronletdowncurveforCycle10isshowninFigure5.1.Thecurveindicates aBOC10,noxenon,criticalboronconcentration of1,254ppm.At150MWD/MT,equilibrium xenon,thecriticalboronconcentration is921ppm.TheCycle10lengthisprojected tobe9,500+300 MWD/MTwith7ppmofboronatEOC.5.1.1PowerDistribution Considerations PRepresentative predicted powermapsforCycle10areshowninFigures5.2and5.3forBOCandEOCconditions, respectively.
Thepowerdistributions wereobtainedfromathree-dimensional modelwithmoderator I,densityandDopplerfeedbackeffectsincorporated.
Fortheprojected Cycle10loadingpatternthecalculated BOCnuclearpowerpeakingfactors,F~,NNNF,andFz,are,l.745, 1.433,and1.201,respectively.
AtEOCconditions thecorresponding valuesare1.517,1.358,and1.098.TheTechnical Specifi-cationlimitsrelativetoF~andF>H,withthemeasurement uncertainties NNbackedout,are2.15and1.60.Additionally thepredicted axialFdistri-butionsarewellbelowtheaxiallydependent Technical Specification limitsonF~.TheBOCFvalueof1.745compareswiththemeasuredCycle9valueNinTable5.1of1.758.
12XN-NF79-,103Thecontrolofthecorepowerdistribution isaccomplished byfollowing theprocedures asdiscussed inthereport,XN-76-40, "ExxonNuclearPowerDistribution ControlforPressurized WaterReactors",
September 1976anditsaddendum.
Theresultsreportedinthesedocuments demonstrate thatthePowerDistribution Control(PDC)procedures definedinthereportwillprotectanaxiallydependent Flimitwithapeakvalueof2.30.TheTechnical Specification limitforR.E.Ginnahasapeakof2.32andanaxialdependence identical tothatsupported bytheprocedures.
Thephysicscharacteristics oftheGinnaCycle10corearesimilartothoseutilizedinthePDCsupporting analysis.
TheGinnaTechnical Specification limitsonFcantherefore beprotected byoperation underthePDCprocedures asstatedinXN-76-40.
5.1.2ControlRodReactivit Reuirements Detailedcalculations ofshutdownmarginsforCycle10arecomparedwithCycle9datainTable5.2.TheENCPlantTransient Simulation (PTS)Analysisindicates thattheminimumrequiredshutdownmarginis1,800pcmbaseduponthesteamline breakaccidentanalyzedforENCfuelattheEOCconditions.
Avalueof1,900pcmisusedatEOCintheevaluation oftheshutdownmargintobeconsistent with'theTechnical Specifications.
TheCycle10analysisindicates excessshutdownmarginsof1,414pcmattheBOCand344pcmattheEOC.TheCycle9analysisindicates excessshut-downmarginsforthatcycleof1,795pcmattheBOCand393pcmattheEOC.TheslightlylowerCycle10excessshutdownmargins,whencomparedtotheCycle9values,areduetoslightlylowercalculated rodworths.
13XN-NF-79103Thecontrol-rodgroupsandinsertion limitsforCycle10willremainunchanged fromCycle9.Withtheselimitsthe'nominal worthofthecontrolbank,D-bank,insertedtotheinsertion limits'at HFPis122pcmat,BOCand'70pcmatEOC.Thecontrolrodshutdownrequirements inTable15.2allowforaHFPD-bankinsertion equivalent to300pcmforbothBOCandEOC.5.1.3Moderator TemeratureCoefficient Considerations Thereference Cycle10designcalculations indicatethatthemoderator temperature coefficient isnegativeatalltimesduringthecycleasshowninTable5.1.ThismeetstheTechnical Specification requirement thatthemoderator temperature coefficient benegativeatalltimesduringpoweroperation andthedesigncriteriathatthepowercoefficient benega-tive.Theleastnegativemoderator temperature coefficient occursatBOCHZPandis-2.0+2pcm/
F.ThiscompareswiththeBOC9HZPvalueof-2.0pcm/F.5.2ANALYTICAL METHODOLOGY ThemethodsusedintheCycle10coreanalysesaredescribed inReferences 6,7,and8.ThesemethodshavebeenverifiedforbothU02andPu02-U02lattices.
Insummary,thereference neutronic designanalysisofthereloadcorewasperformed usingtheXTG(Reference 9)reactorsimulator system-.Theinputexposuredatawerebasedonquartercoredepletion calcu-lationsperformed fromCycle5toCycle9usingtheXTGcode.TheBOC5exposuredistribution wasobtainedfromplantdata.Thefuelshuffling betweenc'ycleswasaccounted forinthecalculations.
14XN-NF-79-103 Predicted valuesofF~,Fx,andFwerestudied,withtheXTGreactormodel.Thecalculational thermal-hydraulic feedbackandaxialexposuredistribution effectsonpowershapes,rodworths,andcyclelifetimeareexplicitly includedintheanalysis.
15XN-NF-79-103 Table5.1R.E.GinnaNeutronics Characteristics ofCycle10ComparedwithCycle9DataBOCCcle9EOCCcle10BOCEOC(CriticalBoronHFP,ARO,Equilibrium Xenon(ppm)HZP,ARONoXenon(ppm)Moderator Temperature Coefficient HFP,(pcm/oF)HZP,(pcm/oF)DopplerCoefficient, (pcm/F)BoronWorth,(pcm/ppm)
HFPHZPTotalNuclearPeakingFactorFq,HFP961')12(1,410(2)-8.12-8.58-8.721.758(-7.6(2)-30.4-2.0-21.5-1.25to-2.09211,414-8.1-2.0-30.4-21.6-7.95,-8.621.7451.517-,1.35-1.84DelayedNeutronFraction.0061.0051.0058.0052ControlRodWorthofAllRodsInMinusMostReactiveRod,HZP,(pcm)ExcessShutdownMargin(pcm)Moderator PressureCoefficient (pcm/psi) 5,751)5,821.1,795()393(0.355';3415,6961,4143440.35(1)Extrapolated frommeasureddata(2)MeasuredData(3)70/PowerMap(4)Reference 5
16XN-NF-79-103 Table5.2,R.E.GinnaControlRodShutdownMarginsandRequirements forCycle10BOCCcle9**EOCCcle10BOCEOCControlRodWorthHZP,cmAllRodsInserted(ARI)ARIlessmostreactive(N-1)N-1lesslOXallowance L(N-1)*9lReactivit Insertion cm5,1765,2396,4076,6345,7515,8215,9496,4205,3415,6964,8075,125Moderator plusDopplerFluxRedistribution VoidSumoftheabovethreeRodInsertion Allowance TotalRequirements 1,431600502,0813002,3811,996600502,'6463002,9461,4431,93260060050502,0932,5823003002,3932,882Shutdown.
Margin(N-l)*.9-TotalRequirements RequiredShutdownMargin*ExcessShutdownMargin2,7952,2931,0001,9001,7953931,4143442,4142,244100001,900*Technical Specification 3.10"*Calculated valuesfromReference 5  
~-~.~~'-~~*~-~*~I~>>.~=~~~=~~-~-t~=~~4h~g~~=~=WtWI.~'~~\-~*~4~t~~~->>*~~~~=-~=~-~-~-~~-Figure5tlR.E.GinnaCycle10AROCriticalBoronConcentration vs.Exposurer4 el~~~~~~~~~~~~~~~~~~~~~~~I~~JLlII~~
I~~~~0~~I~I~I~~~~~~~~~~
I~~~~0~~I~I~I~~~~~~~~~~
20XN-NF-79-103
20 XN-NF-79-103 6.0 THERMAL HYDRAULIC DESIGN The thermal and hydraulic considerations in the Region 12 design are unchanged from those presented in Reference 4 for Region 10 fuel.
 
l~I j 21 XN-NF-79-103 7.0 ACCIDENT AND TRANSIENT ANALYSIS 7.1 PLANT TRANSIENT AND ECCS ANALYSES fOR R.E.GINNA The ECCS analysis provided in Reference 3 is applicable to all ENC fuel residing in the core during Cycle 10 operation.
==6.0 THERMALHYDRAULIC==
The Plant Transient Analysis reported in XN-NF-77-40 for the R.(2)E.Ginna plant was intended to cover all anticipated ranges of values for all significant fuel dependent plant parameters for Cycle 8 and for all future 7 reloads.Table 7.1 presents a comparison of the kinetic parameters used in the Plant Transient Analysis and the parameters calculated specifically for Cycle 10.Due to the introduction of the 4 MOX assemblies the reactivity worth of the-boric acid used by the HPSIS (High Pressure Safety Injection System)and the BOC delayed neutron fraction have been calculated to be outside the range reported in the XN-NF-77-40 analysis.The analysis was reviewed and it was found that the change in boric acid worth affects the smal,l and large steamline break transients and that the delayed neutron fraction most affects the fast uncontrolled rod withdrawal transient.
DESIGNThethermalandhydraulic considerations intheRegion12designareunchanged fromthosepresented inReference 4forRegion10fuel.
The enveloping data for both steamline breaks are the EOC data and for the fast uncontrolled rod withdrawal are BOC data.The impact of the Cycle 10 parameters (see Table 7.1)have been evaluated for each of the transients.
l~Ij 21XN-NF-79-103
The results of the evaluation for the transients were found to be nearly equivalent to the previous results and that the figure of merit for the transients were not violated, i.e.for the small steamline break the system does not go critical, for the large steamline break the MDNBR is greater than the 1.30 limit and for the uncontrolled rod withdrawal the MDNBR margin is not altered.
 
22 XN-NF-79-103 7.2 ROD EJECTION ANALYSIS FOR R.E.GINNA CYCLE 10 A Control Rod Ejection Accident is defined as the mechanical fail-ure of a control rod mechanism pressure housing, resulting'in the ejection of a Rod Cluster Control Assembly (RCCA)and drive shaft.The consequence of this mechanical failure is a rapid reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel damage.~The rod ejection accident analysis presented in the document XN-NF-78-53 is still applicable to Cycle 10 operation.
==7.0 ACCIDENTANDTRANSIENT==
The location of the 4 MOX assemblies introduces minimal effects on ejected rod worths and hot pellet peaking factors.The e'jected rod worths and hot pellet peaking factors are cal'culated using the XTG code.No credit was taken for the powei flattening effects, of Doppler or moderator feedback in the calculation of ejected rod worths'r peaking'factors.The calculations made for Cycle 10 using XTG were two-dimensional (x-y)with appropriate axial buckling correc-tion term's.The total'eaking factors (F~)were determined as the product of the radial peaking facto'r (as calculated using XTG)and a conservative axial peaking factor;The pellet energy deposition resulting from an ejected rod was evaluated to be less than the r'esults reported in References 4 and 5.The rod ejection accident was found to result in energy deposition of less than 280 cal/gm st'ated in Regulatory Guide 1.77'and provides a greater energy deposition marg'in than that determined by Reference 4.The results of the control rod ejection transient for this case are presented in Table 7.2 along with results'from References 4 and 5.
ANALYSIS7.1PLANTTRANSIENT ANDECCSANALYSESfORR.E.GINNATheECCSanalysisprovidedinReference 3isapplicable toallENCfuelresidinginthecoreduringCycle10operation.
23 XN-NF-79-103 7.1 R.E.Ginna Kinetic Parameters Parameters
ThePlantTransient AnalysisreportedinXN-NF-77-40 fortheR.(2)E.Ginnaplantwasintendedtocoverallanticipated rangesofvaluesforallsignificant fueldependent plantparameters forCycle8andforallfuture7reloads.Table7.1presentsacomparison ofthekineticparameters usedinthePlantTransient Analysisandtheparameters calculated specifically forCycle10.Duetotheintroduction ofthe4MOXassemblies thereactivity worthofthe-boricacidusedbytheHPSIS(HighPressureSafetyInjection System)andtheBOCdelayedneutronfractionhavebeencalculated tobeoutsidetherangereportedintheXN-NF-77-40 analysis.
, Moderator Temperature Coefficient (pcm/oF)Moderator Pressure Coefficient (pcm/psia)
Theanalysiswasreviewedanditwasfoundthatthechangeinboricacidworthaffectsthesmal,landlargesteamline breaktransients andthatthedelayedneutronfractionmostaffectsthefastuncontrolled rodwithdrawal transient.
Reference Cycle (1)BOC E C 0.0-35.0+.25+.35 Cycle 10 B C-8.1+.09 E C-30.4+.35 Moderator Density Coefficient (pcm/gm/cm3)
Theenveloping dataforbothsteamline breaksaretheEOCdataandforthefastuncontrolled rodwithdrawal areBOCdata.TheimpactoftheCycle10parameters (seeTable7.1)havebeenevaluated foreachofthetransients.
Doppler Coeffi:cient (pcm/F)-1.25-2.00-1.35-1.84 0.0+29635.0+6858.0+25740.0 Boron Worth Coefficient (pcm/ppm)Delayed Neutron Fraction-8.75-8.72.0061.0051-7.95.0058-8.62.0052 Reference 2 XN-NF-7.9.-103 Table 7.2 Ejected Rod Worth and Peaking Factors'~Before Ejection~C1 8()~C1 9()~C1 10(HFP HZP HFP HZP HFP HZP 2,25 2.82 2.24 2.62 2.15 2.59 F~After Ejection N 4.36'.30 2.96 5.59 g g4()6 01(Maximum Rod Worth from a Full Inserted Bank (X hp)0.470 0.640 0.362 0.553 0.280 0.435 Energy Deposition (cal/gm)171 37 (1)Includes a conservative estimate of F at HFP of 1.4 and at HZP of 1.8.(2)Reference 4, calculated with XTRAN.(3)Reference 5, calculated with XTGPWR.(4)Calculated with XTGPWR.
Theresultsoftheevaluation forthetransients werefoundtobenearlyequivalent tothepreviousresultsandthatthefigureofmeritforthetransients werenotviolated, i.e.forthesmallsteamline breakthesystemdoesnotgocritical, forthelargesteamline breaktheMDNBRisgreaterthanthe1.30limitandfortheuncontrolled rodwithdrawal theMDNBRmarginisnotaltered.
25 XN-NF-79-103
22XN-NF-79-103 7.2RODEJECTIONANALYSISFORR.E.GINNACYCLE10AControlRodEjectionAccidentisdefinedasthemechanical fail-ureofacontrolrodmechanism pressurehousing,resulting'in theejectionofaRodClusterControlAssembly(RCCA)anddriveshaft.Theconsequence ofthismechanical failureisarapidreactivity insertion togetherwithanadversecorepowerdistribution, possiblyleadingtolocalized fueldamage.~Therodejectionaccidentanalysispresented inthedocumentXN-NF-78-53isstillapplicable toCycle10operation.
Thelocationofthe4MOXassemblies introduces minimaleffectsonejectedrodworthsandhotpelletpeakingfactors.Thee'jectedrodworthsandhotpelletpeakingfactorsarecal'culated usingtheXTGcode.Nocreditwastakenforthepoweiflattening effects,ofDopplerormoderator feedbackinthecalculation ofejectedrodworths'rpeaking'factors.
Thecalculations madeforCycle10usingXTGweretwo-dimensional (x-y)withappropriate axialbucklingcorrec-tionterm's.Thetotal'eaking factors(F~)weredetermined astheproductoftheradialpeakingfacto'r(ascalculated usingXTG)andaconservative axialpeakingfactor;Thepelletenergydeposition resulting fromanejectedrodwasevaluated tobelessthanther'esultsreportedinReferences 4and5.Therodejectionaccidentwasfoundtoresultinenergydeposition oflessthan280cal/gmst'atedinRegulatory Guide1.77'andprovidesagreaterenergydeposition marg'inthanthatdetermined byReference 4.Theresultsofthecontrolrodejectiontransient forthiscasearepresented inTable7.2alongwithresults'fromReferences 4and5.
23XN-NF-79-103 7.1R.E.GinnaKineticParameters Parameters
,Moderator Temperature Coefficient (pcm/oF)Moderator PressureCoefficient (pcm/psia)
Reference Cycle(1)BOCEC0.0-35.0+.25+.35Cycle10BC-8.1+.09EC-30.4+.35Moderator DensityCoefficient (pcm/gm/cm3)
DopplerCoeffi:cient (pcm/F)-1.25-2.00-1.35-1.840.0+29635.0+6858.0+25740.0BoronWorthCoefficient (pcm/ppm)
DelayedNeutronFraction-8.75-8.72.0061.0051-7.95.0058-8.62.0052Reference 2
XN-NF-7.9.-103 Table7.2EjectedRodWorthandPeakingFactors'~
BeforeEjection~C18()~C19()~C110(HFPHZPHFPHZPHFPHZP2,252.822.242.622.152.59F~AfterEjectionN4.36'.302.965.59gg4()601(MaximumRodWorthfromaFullInsertedBank(Xhp)0.4700.6400.3620.5530.2800.435EnergyDeposition (cal/gm)17137(1)Includesaconservative estimateofFatHFPof1.4andatHZPof1.8.(2)Reference 4,calculated withXTRAN.(3)Reference 5,calculated withXTGPWR.(4)Calculated withXTGPWR.
25XN-NF-79-103


==8.0REFERENCES==
==8.0 REFERENCES==


1.XN-NF-77-52, "R.E.GinnaReloadFuelDesign",November, 1977.2.3.4.5.6.'XN-NF-77-40, "PlantTransient AnalysisforR.E.Ginna,Unit1NuclearPowerPlant",Revision1,July,1979.XN-NF-77-58, "ECCSAnalysisfortheR.E.GinnaReactorwithENCWREM-IIPWREvaluation Model",December, 1977.XN-NF-77-53, "R.E.GinnaNuclearPlantCycle8SafetyAnalysisReport",December, 1977.XN-NF-78-50,
1.XN-NF-77-52,"R.E.Ginna Reload Fuel Design", November, 1977.2.3.4.5.6.'XN-NF-77-40,"Plant Transient Analysis for R.E.Ginna, Unit 1 Nuclear Power Plant", Revision 1, July, 1979.XN-NF-77-58,"ECCS Analysis for the R.E.Ginna Reactor with ENC WREM-II PWR Evaluation Model", December, 1977.XN-NF-77-53,"R.E.Ginna Nuclear Plant Cycle 8 Safety Analysis Report", December, 1977.XN-NF-78-50,="R.E.Ginna Cycle 9 Safety Analysis Report," December, 1978.F.B.Skogen,"Exxon Nuclear Neutronics Design Methods for Pres-surized Water Reactors", XN-75-27(A), Exxon Nuclear Company, April, 1977.7.XN-75-27(A), Supplement 1 to Reference 6, April, 1977.8.XN-75-27, Supplement 2 to Reference 6, December, 1977.9.XN-CC-28, Rev.3,"XTG: A Two Group Three-Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing (PWR Version)", January, 1975.
="R.E.GinnaCycle9SafetyAnalysisReport,"December, 1978.F.B.Skogen,"ExxonNuclearNeutronics DesignMethodsforPres-surizedWaterReactors",
I l x I 26 XN-NF-79-103 R.E.GINNA CYCLE 10 RELOAD SAFETY ANALYSIS REPORT WITH MIXED OXIDE ASSEMBLIES DISTRIBUTION K.H.Blank G.J.Busselman L.J.Federico R.L.Feuerbacher R.G.Grummer B.L.Johnson (2)M.R.Killgore T.L.Krysinski C.E.Leach J.N.Morgan W.S.Nechodom L.A.Nielsen G.F.Owsley J.F.Patterson A.W.Prichard F.B.Skogen G.A.Sofer A.V.Wojchouski C.H.Wu RG&E/L.J.Federico (80)Document Control (10) l I 1}}
XN-75-27(A),
ExxonNuclearCompany,April,1977.7.XN-75-27(A),
Supplement 1toReference 6,April,1977.8.XN-75-27, Supplement 2toReference 6,December, 1977.9.XN-CC-28, Rev.3,"XTG:ATwoGroupThree-Dimensional ReactorSimulator Utilizing CoarseMeshSpacing(PWRVersion)",
January,1975.
IlxI 26XN-NF-79-103 R.E.GINNACYCLE10RELOADSAFETYANALYSISREPORTWITHMIXEDOXIDEASSEMBLIES DISTRIBUTION K.H.BlankG.J.Busselman L.J.FedericoR.L.Feuerbacher R.G.GrummerB.L.Johnson(2)M.R.KillgoreT.L.Krysinski C.E.LeachJ.N.MorganW.S.NechodomL.A.NielsenG.F.OwsleyJ.F.Patterson A.W.PrichardF.B.SkogenG.A.SoferA.V.Wojchouski C.H.WuRG&E/L.J.Federico(80)DocumentControl(10) lI1}}

Revision as of 15:03, 7 July 2018

Cycle 10 SAR W/Mixed Oxide Assemblies.
ML17249A374
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Issue date: 12/14/1979
From: BUSSELMAN G J, JOHNSON B L, SOFER G A
SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
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0 KN-N-79-103 P E (lllklk PlljlllLEkk PILklll'NllllLE JIB MFET7 kkkL7SIIS PEPBP'll'IIYM Iw3IIXEB QXIIBE bhSSEliNBILIIES DECEMBER 1979 RICHLAND, NA 99352 E I'I I-I I XN-NF-79-103 0: IR/R14 79 R.E.GINNA NUCLEAR PLANT CYCLE 10.SAFETY ANALYSIS REPORT WITH MIXED OXIDE ASSEMBLIES Prepared: G.J.Buss man, Manager Neutronics and Fuel Management Approved: G.A.Sofe nager Nuclear Fue s Engineering Concurred:

J.N.Morgan, Manage Licensing and Safety Engineering Concurred:

L.J.Federico, Manager Nuclear Fuels Project l'~/7 E)j(ON NUCLEAR COMPANY, Inc.

NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT Ir PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Exxon Nuclear Company, Inc.It is being Sub.mitted by Exxon Nuclear to the USNRC as part of a technical contri.bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear.fabricated reioarl fuel or other teclmical services provided by Exxon Nuclear for lieht water power reactors anH, with the measurement uncertainties N N backed out, are 2.15 and 1.60.Additionally the predicted axial F distri-butions are well below the axially dependent Technical Specification limits on F~.The BOC F value of 1.745 compares with the measured Cycle 9 value N in Table 5.1 of 1.758.

12 XN-NF 79-,103 The control of the core power distribution is accomplished by following the procedures as discussed in the report, XN-76-40,"Exxon Nuclear Power Distribution Control for Pressurized Water Reactors", September 1976 and its addendum.The results reported in these documents demonstrate that the Power Distribution Control (PDC)procedures defined in the report will protect an axially dependent F limit with a peak value of 2.30.The Technical Specification limit for R.E.Ginna has a peak of 2.32 and an axial dependence identical to that supported by the procedures.

The physics characteristics of the Ginna Cycle 10 core are similar to those utilized in the PDC supporting analysis.The Ginna Technical Specification limits on F can therefore be protected by operation under the PDC procedures as stated in XN-76-40.5.1.2 Control Rod Reactivit Re uirements Detailed calculations of shutdown margins for Cycle 10 are compared with Cycle 9 data in Table 5.2.The ENC Plant Transient Simulation (PTS)Analysis indicates that the minimum required shutdown margin is 1,800 pcm based upon the steamline break accident analyzed for ENC fuel at the EOC conditions.

A value of 1,900 pcm is used at EOC in the evaluation of the shutdown margin to be consistent with'the Technical Specifications.

The Cycle 10 analysis indicates excess shutdown margins of 1,414 pcm at the BOC and 344 pcm at the EOC.The Cycle 9 analysis indicates excess shut-down margins for that cycle of 1,795 pcm at the BOC and 393 pcm at the EOC.The slightly lower Cycle 10 excess shutdown margins, when compared to the Cycle 9 values, are due to slightly lower calculated rod worths.

13 XN-NF-79 103 The control-rod groups and insertion limits for Cycle 10 will remain unchanged from Cycle 9.With these limits the'nominal worth of the control bank, D-bank, inserted to the insertion limits'at HFP is 122 pcm at, BOC and'70 pcm at EOC.The control rod shutdown requirements in Table 1 5.2 allow for a HFP D-bank insertion equivalent to 300 pcm for both BOC and EOC.5.1.3 Moderator Tem erature Coefficient Considerations The reference Cycle 10 design calculations indicate that the moderator temperature coefficient is negative at all times during the cycle as shown in Table 5.1.This meets the Technical Specification requirement that the moderator temperature coefficient be negative at all times during power operation and the design criteria that the power coefficient be nega-tive.The least negative moderator temperature coefficient occurs at BOC HZP and is-2.0+2pcm/

F.This compares with the BOC9 HZP value of-2.0 pcm/F.5.2 ANALYTICAL METHODOLOGY The methods used in the Cycle 10 core analyses are described in References 6, 7, and 8.These methods have been verified for both U02 and Pu02-U02 lattices.In summary, the reference neutronic design analysis of the reload core was performed using the XTG (Reference 9)reactor simulator system-.The input exposure data were based on quarter core depletion calcu-lations performed from Cycle 5 to Cycle 9 using the XTG code.The BOC5 exposure distribution was obtained from plant data.The fuel shuffling between c'ycles was accounted for in the calculations.

14 XN-NF-79-103 Predicted values of F~, Fx , and F were studied, with the XTG reactor model.The calculational thermal-hydraulic feedback and axial exposure distribution effects on power shapes, rod worths, and cycle lifetime are explicitly included in the analysis.

15 XN-NF-79-103 Table 5.1 R.E.Ginna Neutronics Characteristics of Cycle 10 Compared with Cycle 9 Data BOC C cle 9 EOC C cle 10 BOC EOC (Critical Boron HFP, ARO, Equilibrium Xenon (ppm)HZP, ARO No Xenon (ppm)Moderator Temperature Coefficient HFP, (pcm/oF)HZP, (pcm/oF)Doppler Coefficient, (pcm/F)Boron Worth, (pcm/ppm)HFP HZP Total Nuclear Peaking Factor Fq, HFP 961')12(1,41 0(2)-8.12-8.58-8.72 1.758(-7.6(2)-30.4-2.0-21.5-1.25 to-2.0 921 1,414-8.1-2.0-30.4-21.6-7.95 ,-8.62 1.745 1.517-,1.35-1.84 Delayed Neutron Fraction.0061.0051.0058.0052 Control Rod Worth of All Rods In Minus Most Reactive Rod, HZP, (pcm)Excess Shutdown Margin (pcm)Moderator Pressure Coefficient (pcm/psi)5,751)5,821.1,795()393(0.35 5';341 5,696 1,414 344 0.35 (1)Extrapolated from measured data (2)Measured Data (3)70/Power Map (4)Reference 5 16 XN-NF-79-103 Table 5.2, R.E.Ginna Control Rod Shutdown Margins and Requirements for Cycle 10 BOC C cle 9**EOC C cle 10 BOC EOC Control Rod Worth HZP , cm All Rods Inserted (ARI)ARI less most reactive (N-1)N-1 less lOX allowance L(N-1)*9l Reactivit Insertion cm 5,176 5,239 6,407 6,634 5,751 5,821 5,949 6,420 5,341 5,696 4,807 5,125 Moderator plus Doppler Flux Redistribution Void Sum of the above three Rod Insertion Allowance Total Requirements 1,431 600 50 2,081 300 2,381 1,996 600 50 2,'646 300 2,946 1,443 1,932 600 600 50 50 2,093 2,582 300 300 2,393 2,882 Shutdown.Margin (N-l)*.9-Total Requirements Required Shutdown Margin*Excess Shutdown Margin 2,795 2,293 1,000 1,900 1,795 393 1,414 344 2,414 2,244 10000 1,900*Technical Specification 3.10"*Calculated values from Reference 5

~-~.~~'-~~*~-~*~I~>>.~=~~~=~~-~-t~=~~4 h~g~~=~=Wt WI.~'~~\-~*~4~t~~~->>*~~~~=-~=~-~-~-~~-Figure 5tl R.E.Ginna Cycle 10 ARO Critical Boron Concentration vs.Exposure r 4 el~~~~~~~~~~~~~~~~~~~~~~~I~~J Ll I I~~

I~~~~0~~I~I~I~~~~~~~~~~

20 XN-NF-79-103 6.0 THERMAL HYDRAULIC DESIGN The thermal and hydraulic considerations in the Region 12 design are unchanged from those presented in Reference 4 for Region 10 fuel.

l~I j 21 XN-NF-79-103 7.0 ACCIDENT AND TRANSIENT ANALYSIS 7.1 PLANT TRANSIENT AND ECCS ANALYSES fOR R.E.GINNA The ECCS analysis provided in Reference 3 is applicable to all ENC fuel residing in the core during Cycle 10 operation.

The Plant Transient Analysis reported in XN-NF-77-40 for the R.(2)E.Ginna plant was intended to cover all anticipated ranges of values for all significant fuel dependent plant parameters for Cycle 8 and for all future 7 reloads.Table 7.1 presents a comparison of the kinetic parameters used in the Plant Transient Analysis and the parameters calculated specifically for Cycle 10.Due to the introduction of the 4 MOX assemblies the reactivity worth of the-boric acid used by the HPSIS (High Pressure Safety Injection System)and the BOC delayed neutron fraction have been calculated to be outside the range reported in the XN-NF-77-40 analysis.The analysis was reviewed and it was found that the change in boric acid worth affects the smal,l and large steamline break transients and that the delayed neutron fraction most affects the fast uncontrolled rod withdrawal transient.

The enveloping data for both steamline breaks are the EOC data and for the fast uncontrolled rod withdrawal are BOC data.The impact of the Cycle 10 parameters (see Table 7.1)have been evaluated for each of the transients.

The results of the evaluation for the transients were found to be nearly equivalent to the previous results and that the figure of merit for the transients were not violated, i.e.for the small steamline break the system does not go critical, for the large steamline break the MDNBR is greater than the 1.30 limit and for the uncontrolled rod withdrawal the MDNBR margin is not altered.

22 XN-NF-79-103 7.2 ROD EJECTION ANALYSIS FOR R.E.GINNA CYCLE 10 A Control Rod Ejection Accident is defined as the mechanical fail-ure of a control rod mechanism pressure housing, resulting'in the ejection of a Rod Cluster Control Assembly (RCCA)and drive shaft.The consequence of this mechanical failure is a rapid reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel damage.~The rod ejection accident analysis presented in the document XN-NF-78-53 is still applicable to Cycle 10 operation.

The location of the 4 MOX assemblies introduces minimal effects on ejected rod worths and hot pellet peaking factors.The e'jected rod worths and hot pellet peaking factors are cal'culated using the XTG code.No credit was taken for the powei flattening effects, of Doppler or moderator feedback in the calculation of ejected rod worths'r peaking'factors.The calculations made for Cycle 10 using XTG were two-dimensional (x-y)with appropriate axial buckling correc-tion term's.The total'eaking factors (F~)were determined as the product of the radial peaking facto'r (as calculated using XTG)and a conservative axial peaking factor;The pellet energy deposition resulting from an ejected rod was evaluated to be less than the r'esults reported in References 4 and 5.The rod ejection accident was found to result in energy deposition of less than 280 cal/gm st'ated in Regulatory Guide 1.77'and provides a greater energy deposition marg'in than that determined by Reference 4.The results of the control rod ejection transient for this case are presented in Table 7.2 along with results'from References 4 and 5.

23 XN-NF-79-103 7.1 R.E.Ginna Kinetic Parameters Parameters

, Moderator Temperature Coefficient (pcm/oF)Moderator Pressure Coefficient (pcm/psia)

Reference Cycle (1)BOC E C 0.0-35.0+.25+.35 Cycle 10 B C-8.1+.09 E C-30.4+.35 Moderator Density Coefficient (pcm/gm/cm3)

Doppler Coeffi:cient (pcm/F)-1.25-2.00-1.35-1.84 0.0+29635.0+6858.0+25740.0 Boron Worth Coefficient (pcm/ppm)Delayed Neutron Fraction-8.75-8.72.0061.0051-7.95.0058-8.62.0052 Reference 2 XN-NF-7.9.-103 Table 7.2 Ejected Rod Worth and Peaking Factors'~Before Ejection~C1 8()~C1 9()~C1 10(HFP HZP HFP HZP HFP HZP 2,25 2.82 2.24 2.62 2.15 2.59 F~After Ejection N 4.36'.30 2.96 5.59 g g4()6 01(Maximum Rod Worth from a Full Inserted Bank (X hp)0.470 0.640 0.362 0.553 0.280 0.435 Energy Deposition (cal/gm)171 37 (1)Includes a conservative estimate of F at HFP of 1.4 and at HZP of 1.8.(2)Reference 4, calculated with XTRAN.(3)Reference 5, calculated with XTGPWR.(4)Calculated with XTGPWR.

25 XN-NF-79-103

8.0 REFERENCES

1.XN-NF-77-52,"R.E.Ginna Reload Fuel Design", November, 1977.2.3.4.5.6.'XN-NF-77-40,"Plant Transient Analysis for R.E.Ginna, Unit 1 Nuclear Power Plant", Revision 1, July, 1979.XN-NF-77-58,"ECCS Analysis for the R.E.Ginna Reactor with ENC WREM-II PWR Evaluation Model", December, 1977.XN-NF-77-53,"R.E.Ginna Nuclear Plant Cycle 8 Safety Analysis Report", December, 1977.XN-NF-78-50,="R.E.Ginna Cycle 9 Safety Analysis Report," December, 1978.F.B.Skogen,"Exxon Nuclear Neutronics Design Methods for Pres-surized Water Reactors", XN-75-27(A), Exxon Nuclear Company, April, 1977.7.XN-75-27(A), Supplement 1 to Reference 6, April, 1977.8.XN-75-27, Supplement 2 to Reference 6, December, 1977.9.XN-CC-28, Rev.3,"XTG: A Two Group Three-Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing (PWR Version)", January, 1975.

I l x I 26 XN-NF-79-103 R.E.GINNA CYCLE 10 RELOAD SAFETY ANALYSIS REPORT WITH MIXED OXIDE ASSEMBLIES DISTRIBUTION K.H.Blank G.J.Busselman L.J.Federico R.L.Feuerbacher R.G.Grummer B.L.Johnson (2)M.R.Killgore T.L.Krysinski C.E.Leach J.N.Morgan W.S.Nechodom L.A.Nielsen G.F.Owsley J.F.Patterson A.W.Prichard F.B.Skogen G.A.Sofer A.V.Wojchouski C.H.Wu RG&E/L.J.Federico (80)Document Control (10) l I 1

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