U-8.軽水炉次世代燃料ベンチマーク問題(MOXピンセル)  問題識別番号:PSMC-RA-C02  (丸山 博見) MVP-BURNの入力データ   LWR-WP Benchmark Pin Cell Problem (MOX) * *--------+---------+------ MVP-BURN INPUT START -+---------+---------+-- *$$MVPBURN ** ************************************************************************ ** MVP-BURN : LWR-WP Benchmark Problem (Unit Pin Cell MOX) ** by Keisuke OKUMURA (JAERI) ************************************************************************ ** *$BURNUP * TITLE1('LWR-WP Benchmark') * TITLE2('Case : Unit Pin Cell, MOX Fuel') * CASEID( PMOX ) /* 4-characters without blanck * NSTEP (16) * POWERL( 16(1.790E-04) ) /* Power level in MW/cm(2D) * MWDT ( 0.10E+03 0.50E+03 1.00E+03 2.50E+03 5.00E+03 * 10.00E+03 15.00E+03 20.00E+03 25.0E+03 30.00E+03 * 35.00E+03 40.00E+03 50.00E+03 60.0E+03 70.00E+03 * 73.00E+03 ) /* (MWD/TON) * PC( 16(1) ) /* PC-method for all steps * START(1) /* Start step # (0:restart/1...NSTEP) * SAVE-MVP-OUTPUT ( <%NSTEP>(1011) ) ** N = N1*1000 + N2*100 + N3*100 + N4 ** N1 = 1 : keep standard input file for MVP (caseVIxx) ** N2 = 1 : keep binary fission site file of MVP (caseVSxx) ** N3 = 1 : keep binary output file of MVP (caseVRxx) ** N4 = 1 : keep standard output of MVP for all steps(xx) (caseVPxx) *$END BURNUP *$$END MVPBURN *--------+---------+------ END of MVP-BURN INPUT +---------+---------+-- * NO-RESTART FISSION EIGEN-VALUE FLUX-PRINT EDIT-MACROSCOPIC-DATA(04404040) EDIT-MICROSCOPIC-DATA(04404040) DYNAMIC-MEMORY(30000000) ******************************************************** * P : Cell pitch (cm) * HT: Cell height (arbitrary because of reflective B.C.) * C : Outer radius of fuel * D : Outer radius of cladding * Boron(Nat.) = 0ppm through burn-up ******************************************************** % P = 1.265 , HT = 1000.0 % C = 0.412 , D = 0.476 % NG= 1 NGROUP() NMEMO(10) TCPU(120.0) IRAND(20001213) NPART(500000) NHIST(10000) NBANK(<%NHIST*1.1>) NFBANK(<%NHIST>) ETOP(2.0000E+7) ETHMAX(4.5) AMLIM(300.) NSKIP(10) EWCUT(4.0) ********************************************************************** % MFUEL = 1 , MCLAD = 2 , MH2O = 3 ************************************************** for burnup ******** % PI = 3.1415926536 , VOLF = PI*C*C , FTEMP = 900.0 ********************************************************************** $XSEC * FUEL & IDMAT() *MVPBURN VOLM( ) TRGNAM(@PINFUEL) TEMP( ) PU800900( 8.3986E-05 ) PU900900( 2.1706E-03 ) PU000900( 9.9154E-04 ) PU100900( 3.6732E-04 ) PU200900( 2.5174E-04 ) AM100900( 1.0664E-04 ) U0500900( 3.8879E-05 ) U0800900( 1.9159E-02 ) O0600900( 4.6330E-02 ) * CLAD & IDMAT() ZRN00600( 4.3107E-02 ) * H2O & IDMAT() H01H0600( 4.4148E-02 ) O0600600( 2.2074E-02 ) $END XSEC *************GEOMETRY DATA ***************************************** $GEOM RPP ( 1111 <-P/2.0>

<-P/2.0>

0.0 ) RCC ( 1 0.0 0.0 0.0 0.0 0.0 ) RCC ( 11 0.0 0.0 0.0 0.0 0.0 ) END C21 : : -2000 : -1111 C22 :FUEL :: 1 C23 :CLAD :: 11 -1 C24 :MODE :: 1111 -11 ************* TALLY REGION DATA *********************************** #TALLY REGION DEFINE @PINFUEL( FUEL ) @ALLCELL( * ) $END GEOM ****** INITIAL SOURCE ********************************************** $SOURCE & NEUTRON RATIO( 1.0 ) @E = #FISSION(PU900900 1.00E-01) ; @(X Y) = #DISC ( 0.0 ) ; @Z = ; $END SOURCE ************* VARIANCE REDUCTION PARAMETERS *********************** % NR = %NREG, NRG=NR*NG WKIL( ( 0.2 ) ) WSRV( ( 1.0 ) ) *********** FISSION NEUTRON GENERATION *************************** WGTF( (0.8) ) ************* TALLY ENERGY BOUNDARIES ***************************** ENGYB( 2.0000E+7 1.000E-5 ) / ******************* * input for CGVIEW ******************* * TITLE(Pin Cell MOX) PAPER( <-P/2.> <-P/2.> 1.0 0.0 0.0 0.0 1.0 0.0 200) XMAX(

) LEVEL(-1) *Color Drawing for Zone/Region/Material STYLE(0) SPTYP(0) /