Revised on April 6,1994 SS Bulk Shield Benchmark Experiment at FNG Compilation for IAEA/NDS M. Martone, M. Angelone, P. Batistoni, M. Pillon, V. Rado Neutronics Division - Fusion Department ENEA - Ente per le Nuove tecnologie, l'Energia e l'Ambiente C. R. Frascati - I - 00044 FRASCATI (Italy) List of files Description of contents README.DOC This text INFO.DOC General information on the benchmark experiment. SOURCE.DATA Source neutron spectrum averaged over a spherical cap of 60 degree aperture. Format of the data: Upper energy of the group in MeV; I=1,172 Current spectrum ín/sr/source I=1,173 (Total in 173') Errors (fractions) in percent. I=1,172 (Error on total in 173') Spectrum normalized to one; I=1,172 (Total in 173') SOURCE.FORT Fortran routine used by MCNP.4 to calculate the FNG 14-MeV neutron source with the proper energy/angle distribution. GEOM.DATA Geometrical data of the experiment given in MCNP input format. DETEC.DATA List of activation reactions used, foil locations inside the block and foil size. Information about detectors and measurements. EXPE.DATA Experimental data given for every reaction as follows: Foil position (penetration depth inside the block); in cms. E = measured reaction rate, 10.0E+24 ractions/source random error on E, (fraction) in % systematic error on E, (fraction) in % total error on E, (quadratic sum of random and systematic error, fraction) in %. CALC.DATA Calculated data given for every reaction as follows Foil position (penetration depth inside the block), in cms C = calculated reaction rate 10.0E+24 ractions/source error on C due to MCNP statistics, (fraction in percent) error on C due to uncertainty on activation cross section, (fraction in percent) total error on C, (quadatic sum of two previous errors, fraction in percent) C/E ratios, total errors on C/E ratios (quadratic sum of total error on C and of total error on E, absolute) INFO.DOC STARTS **************************************************** Measurements The experiment consists of the irradiation of a stainless steel (SS) block by 14-MeV neutrons. The experiment is carried out at FNG (Frascati Neutron Generator) facility [1]. The absolute value of the neutron source intensity is measured by the associated particle method, by means of a SSB detector. The accuracy of the measurement improved during the experimental period from +4.4% to +1.6% . The block lateral size is 100x100 cm2, the thickness (in the z direction,where the z axis corresponds to the axis of the accelerator tube) is 70 cm. The SS is AISI316 type; the chemical composition is given in MCNP format in GEOM.DATA (M1). The block is located at 5.3 cm distance (along the z axis) from the neutron source. Of these 5.3 cm, 5 cm are in air and the remaining 0.3 cm are due to the target support structure ( 1 mm Cu, 1 mm H2O, and 1 mm SS). The assembly geometry is described in MCNP format in GEOM.DATA, the origin of coordinates is located on the neutron source. Te block is positioned over an aluminum support which is included in GEOM.DATA, as well as the bunker walls. The current spectrum of the neutron source is given in SOURCE.DATA. This current spectrum is calculated with MCNP and represents the current averaged over a solid angle DW/4P=0.25 in the z direction toward the block; this spectrum was calculated by MCNP including the target zones (TARGET ZONES only in GEOM.DATA) and using the subroutine SOURCE.FORT . In the experiment, activation foils were located in different locations inside the block along the block central (horizontal) axis, corresponding to the z axis. Only foils of the same material were irradiated during every single irradiation. The foil location is everywhere specified (EXPE.DATA, CALC.DATA) by the penetration depth, i.e. the distance of the centre of the foil from the block surface exposed to the neutron source, in cm. The activation reaction used and the foil dimensions are given in DETEC.DATA. The nuclear data used are also given in DETEC.DATA and are taken from [2]. Soon after each irradiation, the foil activity is recorded by a set of calibrated HPGe detectors, also described in DETEC.DATA. The measured reaction rates are given in EXPE.DATA in units of 1024 reactions/(source neutron). Here, the random errors include the g-ray counting statistics and the uncertainty on HPGe calibration. Other errors (f.i. on foil mass and nuclear data) are negligible. The systematic error is the uncertainty on the neutron source intensity. Calculations The calculated reaction rates are given in CALC.DATA in units of 1024 reactions/(source neutron). These data were obtained with MCNP.4, with the geometry input given in GEOM.DATA. The source was provided by the subroutine SOURCE.FORT, with the proper angle/energy distribution, i.e. taking into account the reaction kinematics and the slowing down of beam deuterons in the tritium/titanium target. Transport cross sections from EFF.1 were used; the activation cross sections (for foil detectors) were taken from IRDF.90, with the exception of 55Mn(n,g) which was taken from EFF.2. The track length tally (F4 in MCNP, flux averaged over a cell) was used to calculate the reaction rates. The detector responses were calculated in a MCNP model including everything but "BUNKER AND WALLS" in the input GEOM.DATA. The room background contribution to the detector responses was estimated by means of a surface source at the block external boundaries. This surface source was recorded once in a dedicated MCNP run including the bunker walls in the geometry. The finding is that the background is not negligible on Mn and Au foils located close to the block front surface and amounts to 8% at 5 cm, to 6% at 10 cm for 197Au(n,g), and to 5% at 5 cm and to 4% at 10 cm for 55Mn(n,g). This background contributions are included in the corresponding C values. The error on reaction rate due to uncertainty on activation cross sections was calculated by variance analysis using the IRDF-90 covariance data in 175 energy group format, except for 55Mn(n,g) for which the EFF.2 covariance file has been used. The random errors on calculated reaction rates, due to the statistics in the MCNP calculations and to uncertainty on activation cross sections, are separately given in CALC.DATA.The total error is the quadratic sum of the two. C/E values in CALC.DATA are the ratio of calculated over measured reaction rates. The total error on C/E is the quadratic sum of the total error on C and of the total error on E. References [1] M. Martone, M. Angelone, M. Pillon, "The 14-MeV Frascati Neutron Generator (FNG)", ENEA Report RT/ERG/FUS/93/65 [2] J. H. Baard, W. L. Zijp, H. Nolthenius, "Nuclear Data Guide for Reactor Neutron Metrology", Kluwer Academic Publishers for the Commission of the European Community, (1989). INFO.DOC ENDS******************************************************* SOURCE.DATA STARTS******** TARGET SOURCE SPECTRUM FOR FNG BENCHMARK ANALYSIS UPPER LIMITS OF 172 ENERGY GROUPS (MeV) 1.0000E-07 4.1399E-07 5.3158E-07 6.8256E-07 8.7642E-07 1.1254E-06 1.4450E-06 1.8554E-06 2.3824E-06 3.0590E-06 3.9279E-06 5.0435E-06 6.4760E-06 8.3153E-06 1.0677E-05 1.3710E-05 1.7603E-05 2.2603E-05 2.9023E-05 3.7267E-05 4.7851E-05 6.1442E-05 7.8893E-05 1.0130E-04 1.3007E-04 1.6702E-04 2.1445E-04 2.7536E-04 3.5358E-04 4.5400E-04 5.8295E-04 7.4852E-04 9.6112E-04 1.2341E-03 1.5846E-03 2.0347E-03 2.2487E-03 2.4852E-03 2.6126E-03 2.7465E-03 3.0354E-03 3.3546E-03 3.7074E-03 4.3074E-03 5.5308E-03 7.1017E-03 9.1188E-03 1.0595E-02 1.1709E-02 1.5034E-02 1.9305E-02 2.1875E-02 2.3579E-02 2.4176E-02 2.4788E-02 2.6058E-02 2.7000E-02 2.8500E-02 3.1828E-02 3.4307E-02 4.0868E-02 4.6309E-02 5.2475E-02 5.6562E-02 6.7379E-02 7.2000E-02 7.9500E-02 8.2500E-02 8.6517E-02 9.8037E-02 1.1109E-01 1.1679E-01 1.2277E-01 1.2907E-01 1.3569E-01 1.4264E-01 1.4996E-01 1.5764E-01 1.6573E-01 1.7422E-01 1.8316E-01 1.9255E-01 2.0242E-01 2.1280E-01 2.2371E-01 2.3518E-01 2.4724E-01 2.7324E-01 2.8725E-01 2.9452E-01 2.9720E-01 2.9850E-01 3.0197E-01 3.3373E-01 3.6883E-01 3.8774E-01 4.0762E-01 4.5049E-01 4.9787E-01 5.2340E-01 5.5023E-01 5.7844E-01 6.0810E-01 6.3928E-01 6.7206E-01 7.0651E-01 7.4274E-01 7.8082E-01 8.2085E-01 8.6294E-01 9.0718E-01 9.6164E-01 1.0026E+00 1.1080E+00 1.1648E+00 1.2246E+00 1.2873E+00 1.3534E+00 1.4227E+00 1.4957E+00 1.5724E+00 1.6530E+00 1.7377E+00 1.8268E+00 1.9205E+00 2.0190E+00 2.1225E+00 2.2313E+00 2.3069E+00 2.3457E+00 2.3653E+00 2.3852E+00 2.4660E+00 2.5924E+00 2.7253E+00 2.8650E+00 3.0119E+00 3.1664E+00 3.3287E+00 3.6788E+00 4.0657E+00 4.4933E+00 4.7237E+00 4.9659E+00 5.2205E+00 5.4881E+00 5.7695E+00 6.0653E+00 6.3763E+00 6.5924E+00 6.7032E+00 7.0469E+00 7.4082E+00 7.7880E+00 8.1873E+00 8.6071E+00 9.0484E+00 9.5123E+00 1.0000E+01 1.0513E+01 1.1052E+01 1.1618E+01 1.2214E+01 1.2523E+01 1.2840E+01 1.3499E+01 1.3840E+01 1.4191E+01 1.4550E+01 1.4918E+01 1.5683E+01 2.0000E+01 NEUTRON CURRENT SPECTRUM IN A FORWARD CONE (60 DEGREES) NORMALIZED TO ONE T(d,n) SOURCE NEUTRON -TOTAL IN 173' 6.5591E-06 4.4179E-06 8.0958E-07 9.1309E-07 9.1884E-07 9.6500E-07 1.0114E-06 1.1154E-06 1.2237E-06 1.3108E-06 1.4112E-06 1.4838E-06 1.5535E-06 1.7038E-06 1.8762E-06 1.9407E-06 2.1355E-06 2.3048E-06 2.4412E-06 2.8678E-06 2.8678E-06 3.1126E-06 3.4252E-06 4.0426E-06 4.2461E-06 4.4325E-06 4.7454E-06 5.4300E-06 5.4716E-06 6.1622E-06 7.0677E-06 8.1440E-06 8.7192E-06 8.9465E-06 1.0464E-05 1.1719E-05 5.4402E-06 5.4693E-06 2.7270E-06 2.6946E-06 5.4509E-06 6.4994E-06 6.7301E-06 1.0624E-05 2.0435E-05 2.3428E-05 2.3463E-05 1.7701E-05 1.1754E-05 3.5744E-05 3.8062E-05 2.3105E-05 1.3763E-05 4.4783E-06 6.0150E-06 1.3327E-05 9.8470E-06 9.9675E-06 2.2808E-05 1.9506E-05 4.7725E-05 3.5946E-05 4.4245E-05 2.8025E-05 6.8648E-05 2.9789E-05 4.7308E-05 2.1352E-05 2.3171E-05 7.3814E-05 1.0944E-04 4.6592E-05 5.5823E-05 5.7413E-05 6.5609E-05 6.4925E-05 5.6205E-05 6.6986E-05 7.2167E-05 7.5154E-05 8.2986E-05 8.6404E-05 8.8387E-05 1.0009E-04 1.0211E-04 1.0993E-04 1.1259E-04 2.5728E-04 1.3888E-04 6.5248E-05 2.6246E-05 1.1403E-05 3.3778E-05 3.1236E-04 3.3420E-04 1.9414E-04 1.9619E-04 3.9677E-04 4.6825E-04 2.5423E-04 2.5724E-04 2.7686E-04 2.8432E-04 2.7109E-04 3.0109E-04 3.0842E-04 3.2908E-04 3.5276E-04 3.4309E-04 3.5131E-04 3.6087E-04 4.4522E-04 3.1041E-04 7.4156E-04 4.0087E-04 4.2835E-04 4.3828E-04 4.2187E-04 4.4656E-04 4.4736E-04 4.2258E-04 4.2621E-04 4.2545E-04 4.4223E-04 4.2258E-04 4.2672E-04 4.1197E-04 4.3274E-04 2.8650E-04 1.4580E-04 7.5776E-05 7.8547E-05 2.9492E-04 4.2068E-04 4.1583E-04 4.1225E-04 4.2165E-04 3.9958E-04 3.9918E-04 7.6998E-04 7.3158E-04 6.9208E-04 3.0011E-04 3.1544E-04 3.0621E-04 2.9132E-04 2.6869E-04 2.9419E-04 3.0359E-04 1.9828E-04 1.0978E-04 3.1146E-04 2.8778E-04 3.2377E-04 3.0590E-04 2.8780E-04 2.8675E-04 2.9709E-04 2.4957E-04 2.4505E-04 2.3459E-04 2.4696E-04 3.0490E-04 2.8192E-04 5.1734E-04 9.3428E-04 7.6304E-04 1.3539E-03 5.8329E-02 1.4869E-01 4.0895E-02 0.0000E+00 2.7894E-01 FRACTIONAL STANDARD DEVIATIONS 0.012 0.012 0.017 0.045 0.019 0.019 0.020 0.021 0.022 0.021 0.022 0.021 0.021 0.021 0.021 0.025 0.023 0.023 0.023 0.028 0.023 0.025 0.027 0.048 0.028 0.046 0.032 0.038 0.027 0.030 0.036 0.043 0.041 0.033 0.036 0.038 0.059 0.056 0.072 0.080 0.048 0.065 0.059 0.041 0.033 0.034 0.033 0.042 0.052 0.034 0.034 0.047 0.061 0.099 0.102 0.062 0.062 0.063 0.046 0.054 0.035 0.041 0.037 0.048 0.032 0.050 0.040 0.057 0.056 0.031 0.031 0.048 0.044 0.044 0.041 0.040 0.043 0.040 0.040 0.039 0.037 0.035 0.036 0.034 0.034 0.034 0.032 0.022 0.030 0.044 0.068 0.104 0.060 0.020 0.020 0.026 0.026 0.018 0.017 0.023 0.023 0.023 0.023 0.023 0.022 0.021 0.021 0.020 0.020 0.020 0.020 0.018 0.022 0.015 0.020 0.019 0.019 0.019 0.019 0.019 0.020 0.019 0.019 0.019 0.020 0.020 0.020 0.019 0.024 0.033 0.046 0.046 0.024 0.020 0.020 0.020 0.020 0.020 0.020 0.015 0.015 0.016 0.024 0.023 0.023 0.024 0.025 0.024 0.024 0.029 0.039 0.023 0.024 0.023 0.024 0.025 0.025 0.024 0.027 0.027 0.028 0.027 0.024 0.025 0.019 0.014 0.016 0.012 0.002 0.001 0.002 0.000 0.001 SPECTRUM NORMALIZED TO 1.0 2.3514E-05 1.5838E-05 2.9023E-06 3.2734E-06 3.2940E-06 3.4595E-06 3.6259E-06 3.9987E-06 4.3870E-06 4.6991E-06 5.0590E-06 5.3195E-06 5.5691E-06 6.1082E-06 6.7262E-06 6.9575E-06 7.6558E-06 8.2626E-06 8.7515E-06 1.0281E-05 1.0281E-05 1.1159E-05 1.2279E-05 1.4492E-05 1.5222E-05 1.5890E-05 1.7012E-05 1.9466E-05 1.9615E-05 2.2091E-05 2.5337E-05 2.9196E-05 3.1258E-05 3.2073E-05 3.7513E-05 4.2011E-05 1.9503E-05 1.9607E-05 9.7762E-06 9.6599E-06 1.9541E-05 2.3300E-05 2.4127E-05 3.8087E-05 7.3259E-05 8.3989E-05 8.4113E-05 6.3456E-05 4.2139E-05 1.2814E-04 1.3645E-04 8.2831E-05 4.9339E-05 1.6055E-05 2.1563E-05 4.7775E-05 3.5301E-05 3.5733E-05 8.1765E-05 6.9928E-05 1.7109E-04 1.2886E-04 1.5862E-04 1.0047E-04 2.4610E-04 1.0679E-04 1.6960E-04 7.6546E-05 8.3069E-05 2.6462E-04 3.9233E-04 1.6703E-04 2.0012E-04 2.0582E-04 2.3521E-04 2.3275E-04 2.0149E-04 2.4014E-04 2.5872E-04 2.6942E-04 2.9750E-04 3.0975E-04 3.1686E-04 3.5881E-04 3.6607E-04 3.9408E-04 4.0365E-04 9.2234E-04 4.9788E-04 2.3391E-04 9.4090E-05 4.0878E-05 1.2109E-04 1.1198E-03 1.1981E-03 6.9597E-04 7.0332E-04 1.4224E-03 1.6786E-03 9.1139E-04 9.2221E-04 9.9254E-04 1.0193E-03 9.7185E-04 1.0794E-03 1.1057E-03 1.1797E-03 1.2646E-03 1.2300E-03 1.2594E-03 1.2937E-03 1.5961E-03 1.1128E-03 2.6585E-03 1.4371E-03 1.5356E-03 1.5712E-03 1.5124E-03 1.6009E-03 1.6037E-03 1.5149E-03 1.5279E-03 1.5252E-03 1.5854E-03 1.5149E-03 1.5298E-03 1.4769E-03 1.5513E-03 1.0271E-03 5.2268E-04 2.7165E-04 2.8159E-04 1.0573E-03 1.5081E-03 1.4907E-03 1.4779E-03 1.5116E-03 1.4325E-03 1.4310E-03 2.7603E-03 2.6227E-03 2.4811E-03 1.0759E-03 1.1308E-03 1.0977E-03 1.0444E-03 9.6323E-04 1.0547E-03 1.0884E-03 7.1084E-04 3.9355E-04 1.1166E-03 1.0317E-03 1.1607E-03 1.0966E-03 1.0317E-03 1.0280E-03 1.0651E-03 8.9468E-04 8.7850E-04 8.4100E-04 8.8532E-04 1.0931E-03 1.0107E-03 1.8547E-03 3.3494E-03 2.7355E-03 4.8536E-03 2.0911E-01 5.3305E-01 1.4661E-01 0.0000E+00 1.0000E+00 SOURCE.DATA ENDS****************************************************** SOURCE.FORT STARTS********************************************************** USED BY MCNP.4 TO GENERATE 14 MeV NEUTRONS WITH THE PROPER ENERGY /ANGLE DISTRIBUTION. FOR THE BENCHMARK ANALYSIS, USED INPUT PARAMETER VALUES ARE EB=0.230 (BEAM ENERGY IN MeV), XT=1.5 (TRITIUM ATOMS/TITANIUM ATOMS) AND ARE GIVEN IN THE RDUM CARD IN THE MCNP INPUT FILE SUBROUTINE SOURCE C DUMMY SUBROUTINE. ABORTS JOB IF SOURCE SUBROUTINE IS MISSING. C IF NSR=0, SUBROUTINE SOURCE MUST BE FURNISHED BY THE USER. C AT ENTRANCE, A RANDOM SET OF UUU,VVV,WWW HAS BEEN DEFINED.THE C FOLLOWING VARIABLES MUST BE DEFINED WITHIN THE SUBROUTINE: C XXX,YYY,ZZZ,ICL,JSU,ERG,WGT,TME AND POSSIBLY IPT,UUU,VVV,WWW. C SUBROUTINE SRCDX MAY ALSO BE NEEDED. C IMPLICIT DOUBLE PRECISION (A-H,O-Z) C PARAMETER (MAXF=16,MAXI=34,MAXV=19,MAXW=3,MCPU=32,MINK=200,MIPT=3, 1 MJSF=9,MKFT=9,MKTC=22,MLGC=100,MPB=5,MPNG=21,MRKP=100,MSEB=301, 2 MSPARE=3,MTOP=49,MWNG=25,MXDT=20,MXDX=5,MXLV=10,NBMX=100,NDEF=1 4, 3 NOVR=5,IUI=31,IUO=32,IUR=33,IUX=34,IUD=35,IUB=60,IUP=37,IUS=38, 4 IU1=39,IU2=40,IUSW=41,IUSR=42,IUSC=43,IUC=44,IUT=45,IUZ=46, 5 IUK=47,IU3=48,IU4=49,ZERO=0.,ONE=1.,THIRD=ONE/3., 6 PIE=3.1415926535898D0,AVGDN=.59703109D0,SLITE=299.7925, 7 FSCON=137.0393) C C ------------------------------------------------------------------- --- C C C VARIABLE COMMON -- VARIABLE BUT REQUIRED FOR A CONTINUE RUN. C ARRAYS THAT ARE BACKED UP WHEN A TRACK IS LOST. COMMON /VARCOM/ CPK,CTS,DBCN(20),DMP,EACC(4),FEBL(2,16),OSUM(3), 1 OSUM2(3,3),PAX(6,16,MIPT),PRN,RANI,RANJ,RDUM(50),RIJK,RKK, 2 RLT(2),RNR,RSUM(2),RSUM2(2,2),SMUL(3),SUMK(3),TMAV(MIPT,3), 3 TWAC,TWSS,WCS1(MIPT),WCS2(MIPT),WGTS(2),WT0,WSSI(7), 4 ZVARCM, 5 IDUM(50),INIF,IST,IST0,IXAK,IXAK0,JRAD,KCSF,KCT,KCY,KNOD, 6 KSDEF,KZKF,LOST(2),NBAL(MCPU),NBHWM,NBOV,NBT(MIPT),NBY,NCT(MIPT), 7 NDMP,NERR,NETB(2),NFER,NPC(20),NPD,NPNM,NPP,NPPM,NPS,NPSOUT,NPSR, 8 NQSS,NQSW,NRNH(3),NRRS,NRSW,NSA,NSA0,NSKK,NSS,NSS0,NSSI(8),NTC, 9 NTC1,NTSS,NWER,NWSB,NWSE,NWSG(2),NWST,NWWS(2,99),NZIP,NZIX, 1 NZIY(8,MXDX,MIPT), 2 MVARCM COMMON /PBLCOM/ XXX,YYY,ZZZ,UUU,VVV,WWW,ERG,WGT,TME,VEL,DLS, 1 DXL,DTC,ELC(MIPT),FIML(MIPT),FISMG,WTFASV,RNK,SPARE(MSPARE), 2 ZPBLCM, 3 XXX9(MPB),YYY9(MPB),ZZZ9(MPB),UUU9(MPB),VVV9(MPB),WWW9(MPB), 4 ERG9(MPB),WGT9(MPB),TME9(MPB),VEL9(MPB),DLS9(MPB),DXL9(MPB), 5 DTC9(MPB),ELC9(MPB,MIPT),FIML9(MPB,MIPT),FISMG9(MPB), 6 WTFAS9(MPB),RNK9(MPB),SPARE9(MPB,MSPARE), 7 ZPB9CM(MPB), 1 NPA,ICL,JSU,IPT,IEX,NODE,IDX,NCP,JGP,LEV,III,JJJ,KKK,IAP, 2 MPBLCM, 3 NPA9(MPB),ICL9(MPB),JSU9(MPB),IPT9(MPB),IEX9(MPB),NODE9(MPB), 4 IDX9(MPB),NCP9(MPB),JGP9(MPB),LEV9(MPB),III9(MPB),JJJ9(MPB), 5 KKK9(MPB),IAP9(MPB), 6 MPB9CM(MPB) C COMMON /MARIO/ DEDX(100),EMIN,EB,XT,SML,NENE C ******************************************************************* *** C STANDARD DATA C ******************************************************************* *** DIMENSION CDEDX(100),SIG(100),TDEDX(100),PL(100) DIMENSION ED(100),SUML(3601),TH(3601) REAL MD,MT,MN,MA DATA MD/2.01410219/ DATA MT/3.01602994/ DATA MN/1.00866544/ DATA MA/4.00260361/ DATA AUX/25.20734546/ DATA Q/17.589/ C C INPUT DATA FOR THIS PROGRAM ARE: C TRITIUM ATOM PER ONE TITANIUM ATOM C DEUTERON / TRITONS ENERGY IN MeV, UP TO 0.5 MEV; C ******************************************************************* *** C DATA C ******************************************************************* *** DATA NPTS/50/ DATA ED/0.010,0.020,0.030,0.040, *0.050,0.060,0.070,0.080,0.090,0.100,0.110,0.120,0.130,0.140, *0.150,0.160,0.170,0.180,0.190,0.200,0.210,0.220,0.230,0.240,0.250, *0.260,0.270,0.280,0.290,0.300,0.310,0.320,0.330,0.340,0.350,0.360, *0.370,0.380,0.390,0.400,0.410,0.420,0.430,0.440,0.450,0.460,0.470, *0.480,0.490,0.500,50*0.0/ C C T(d,n) DEUTERON ON TRITIUM CROSS SECTION C DATA SIG/1.0E-4,4.3E-3,0.0196,0.0529,0.106,0.175,0.250,0.315, *0.367,0.394,0.399,0.387,0.367,0.339,0.317,0.286,0.262,0.236,0.215, *0.199,0.181,0.167,0.153,0.142,0.133,0.122,0.114,0.106,0.100,0.0952 *,0.0911,0.0873,0.0838,0.0806,0.0775,0.0743,0.0713,0.0686,0.0659, *0.0635,0.0611,0.0589,0.0568,0.0549,0.0530,0.0513,0.0498,0.0483, *0.0469,0.0455,50*0.0/ C C ENERGY LOSS OF DEUTERONS IN TITANIUM AND TRITIUM C DATA CDEDX/0.142,0.1929,0.2299,0.2592,0.2835,0.3038,0.3209, *0.3354,0.3475,0.3577,0.3660,0.3728,0.3781,0.3821,0.3851,0.3870, *0.3880,0.3883,0.3879,0.3869,0.3854,0.3834,0.3811,0.3785,0.3756, *0.3725,0.3692,0.3658,0.3622,0.3586,0.3550,0.3513,0.3476,0.3439, *0.3402,0.3365,0.3329,0.3293,0.3258,0.3223,0.3189,0.3155,0.3122, *0.3089,0.3058,0.3027,0.2996,0.2966,0.2937,0.2909,50*0.0/ DATA TDEDX/0.5959,0.8014,0.9420,1.046,1.124,1.182,1.225,1.255, *1.275,1.287,1.291,1.290,1.285,1.275,1.262,1.247,1.231,1.212, *1.193,1.173,1.152,1.132,1.111,1.090,1.069,1.048,1.028,1.008, *0.9886,0.9695,0.9509,0.9327,0.9150,0.8978,0.8811,0.8649,0.8491, *0.8338,0.8190,0.8046,0.7906,0.7771,0.7640,0.7513,0.7390,0.7270, *0.7155,0.7042,0.6934,0.6828,50*0.0/ C C ******************************************************************* *** IF(SML.GT.0) GO TO 1111 EMIN=0.010 C BEAM ENERGY IN MeV, UP TO 0.5 MEV; EB=RDUM(1) C TRITIUM -TITANIUM RATIO XT=RDUM(2) C SOURCE COORDINATES XX=RDUM(3) YY=RDUM(4) ZZ=RDUM(5) C ******************************************************************* DO 100 I=1,NPTS C STOPPING POWER 100 DEDX(I)=CDEDX(I)*48./(48.+3.*XT)+TDEDX(I)*3.*XT/(48.+3.*XT) THL=-0.05 DO 1000 L=1,3601 THL=THL+0.05 TH(L)=THL*3.1415927/180. CTH=COS(TH(L)) NENE=0 DO 250 I=1,NPTS IF(ED(I).LE.EB) NENE=I ET=ED(I)+Q B=MD*MN*ED(I)/ET/AUX D=MT*MA/AUX*(1.+MD*Q/(MT*ET)) E=SQRT(D/B-1.+CTH*CTH) EN=ET*B*(CTH+E)**2 G=SQRT(B*D)*E/(EN/ET) 250 PL(I)=SIG(I)/G/DEDX(I) CALL XINT(ED,PL,NPTS,EMIN,EB,XR) SUML(L)=XR*SIN(TH(L)) 1000 CONTINUE CALL XINT(TH,SUML,3601,TH(1),TH(3601),TI) SML=TI 1111 ICL=1 JSU=0 IPT=1 XXX=XX YYY=YY ZZZ=ZZ TME=0.0 C 60 PMAX=1.1 60 VVV= 2.*RANG()-1. I=INT(NENE*RANG())+1 DELTA= ED(I+1)-ED(I) DELTE=RANG()*DELTA ET=ED(I)+DELTE+Q C**** ET=ED(I)+Q C PMAX=PMAX*RANG() PMAX=RANG() B=MD*MN*(ED(I)+DELTE)/ET/AUX C**** B=MD*MN*ED(I)/ET/AUX D=MT*MA/AUX*(1.+MD*Q/(MT*ET)) E=SQRT(D/B-1.+VVV*VVV) EN=ET*B*(VVV+E)**2 G=SQRT(B*D)*E/(EN/ET) XIN=ED(I)+DELTE CALL INTERP(ED,SIG,NPTS,XIN,SIGOUT) CALL INTERP(ED,DEDX,NPTS,XIN,DEDXOT) P=SIGOUT/G/DEDXOT C**** P=SIG(I)/G/DEDX(I) C****************************************************************** ***** C ENEA C WRITE(6,*) NENE,P C ENEA IF(P.GE.1.0) WRITE(6,32) 32 FORMAT(5X,' A T T E N Z I O N E P .GE. 1.0') IF(PMAX.GT.P) GO TO 60 ERG=EN ST=SQRT(1.-VVV*VVV) 70 X1=2.*RANG()-1. X2=2.*RANG()-1. X11=X1*X1 X22=X2*X2 RO=X11+X22 IF(RO.GT.1.0) GO TO 70 SF=2*X1*X2/RO CF=(X11-X22)/RO UUU=ST*SF WWW=ST*CF WGT=1. RETURN END C SUBROUTINE FOR INTERPOLATION SUBROUTINE INTERP(X,Y,NPTS,XIN,YOUT) IMPLICIT DOUBLE PRECISION (A-H,O-Z) C ------------------------------------------------------------------- --- C DIMENSION X(NPTS),Y(NPTS) I=1 IF(X(I)-XIN)2,10,12 2 DO 1 I=2,NPTS IF(X(I)-XIN)1,10,11 1 CONTINUE GO TO 12 11 YOUT=Y(I-1)+(Y(I)-Y(I-1))/(X(I)-X(I-1))*ABS(X(I)-XIN) RETURN 10 YOUT=Y(I) RETURN 12 WRITE(6,100) 100 FORMAT(20X,'ERROR INTERPOLATION '/) STOP END C SUBROUTINE FOR NUMERICAL INTEGRATION SUBROUTINE XINT(E,F,NPTS,E1,E2,SUM) IMPLICIT DOUBLE PRECISION (A-H,O-Z) C C ------------------------------------------------------------------- --- C DIMENSION E(NPTS),F(NPTS) SUM=0 L=1 IF(E2-E1)50,55,60 50 A=E2 E2=E1 E1=A L=-1 60 I=1 IF(E(I)-E1)2,10,12 2 DO 1 I=2,NPTS IF(E(I)-E1)1,10,11 1 CONTINUE GO TO 12 11 CALL INTERP(E,F,NPTS,E1,FOUT) SUM=SUM+(E(I)-E1)*(F(I)+FOUT)*0.5 10 KMIN=I J=1 IF(E(J)-E2)3,13,12 3 DO 4 J=2,NPTS IF(E(J)-E2)4,13,15 4 CONTINUE GO TO 12 15 CALL INTERP(E,F,NPTS,E2,FOUT) SUM=SUM+(E2-E(J-1))*(F(J-1)+FOUT)*0.5 J=J-1 13 KMAX=J-1 IF(KMIN.GT.KMAX)GO TO 6 DO 5 K=KMIN,KMAX 5 SUM=SUM+(E(K+1)-E(K))*(F(K+1)+F(K))*0.5 6 SUM=SUM*L 55 RETURN 12 WRITE(6,100) 100 FORMAT(20X,'ERROR INTEGRAL '/) STOP END SOURCE.FORT ENDS************************************************************ GEOM.DATA STARTS************************************************************ GEOMETRICAL DATA FOR "SS BULK SHIELD EXPERIMENT AT FNG" THE NEUTRON SOURCE HAS COORDINATES X=0, Y=0, Z=0 INPUT MCNP4 (GEOMETRY) C TARGET ZONES 1 3 -8.94 6 (1:7) -2 -8 2 2 -1.0 6 8 -2 -9 3 1 -7.954 6 9 -5 -10 4 1 -7.954 5 -2 9 -11 5 2 -1.0 2 -3 -11 6 1 -7.954 3 -4 -11 7 1 -7.954 -6 17 7 -10 8 0 -1 17 -7 9 1 -7.954 -6 18 10 -11 10 1 -7.954 11 -16 12 -13 11 1 -7.954 11 -16 -14 15 12 2 -1.0 11 -16 13 14 -19 13 4 4.614E-5 6 -5 10 -11 14 4 4.614E-5 17 -18 10 -11 15 4 4.614E-5 11 -16 (-12:-15) 16 4 4.614E-5 4 -100 -11 17 4 4.614E-5 6 -100 11 16 -19 18 4 4.614E-5 -6 17 11 20 21 -19 19 2 -1.0 -6 17 -20 20 2 -1.0 -6 17 -21 21 4 4.614E-5 17 -100 102 -103 104 -105 19 C SS BLOCK 100 1 -7.954 100 -101 102 -103 104 -105 C DRIFT TUBE AND OTHER ACCELERATOR STRUCTURES 22 1 -7.954 36 (22:23) -17 -19 7 23 0 22 -17 -7 24 1 -7.954 -22 26 24 -25 25 0 -22 26 -24 26 1 -0.07954 -22 26 (-23 36:-19 -36) 25 27 7 -8.4 -26 27 -29 28 4 4.614E-5 (-17 104:-36 -104) 27 34 -35 -105 400 (19 26:29 -26) C ALUMINUM SUPPORT OF THE BLOCK 29 4 4.614E-5 17 -40 34 -35 104 -105 (-102:103:101) 30 5 -2.7 36 -39 30 -31 41 -104 31 4 4.614E-5 36 -40 34 -35 41 -104 (39:-30:31) 32 5 -0.73 37 -38 32 -33 -41 +42 33 4 4.614E-5 36 -40 34 -35 -41 +42 (-37:38:-32:33) 34 5 -0.52 36 -40 34 -35 -42 +43 35 5 -2.7 -43 400 -44 45 36 4 4.614E-5 -43 400 -45 37 5 -2.7 -43 400 -46 47 38 4 4.614E-5 -43 400 -47 39 5 -2.7 -43 400 -48 49 40 4 4.614E-5 -43 400 -49 41 5 -2.7 -43 400 -50 51 42 4 4.614E-5 -43 400 -51 43 4 4.614E-5 36 -40 34 -35 -43 400 44 46 48 50 C BUNKER AND WALLS 44 4 .00004614 300 -310 200 -210 400 -410 (40:-27:-34:35:105) 45 6 -2.6 (-300:310:-200:210:-400:410) 309 -319 209 -219 409 -419 46 0 -309:319:-209:219:-409:419 C TARGET ZONES 1 PY 0.0 2 PY 0.1 3 PY 0.2 4 PY 0.32 5 PY -0.02 6 PY -1.9 7 CY 1.5 8 CY 1.6 9 CY 1.7 10 CY 1.8 11 CY 2.4 12 Y -0.02 2.4 -1.45 6.4 13 Y .1 2.4 -1.33 6.4 14 Y .2 2.4 1.63 6.4 15 Y .32 2.4 1.75 6.4 16 2 C/X .15 0.0 1.6 17 PY -12.4 18 PY -4. 19 CY 17.7 20 2 C/Y -16. 0. 1.6 21 2 C/Y 16. 0. 1.6 22 PY -13.0 23 CY 17.4 24 CY 5.35 25 CY 5.75 26 PY -165. 27 PY -255. 29 CY 30. 100 PY 5.3 101 PY 71.8 102 PX -49.5 103 PX +49.5 104 PZ -49.2 105 PZ +49.2 C ALUMINUM SUPPORT OF THE BLOCK 30 PX -51.5 31 PX +51.5 32 PX -57.5 33 PX +57.5 34 PX -65.0 35 PX +65.0 36 PY -42.2 37 PY -26.2 38 PY +88.8 39 PY +122.0 40 PY +177.8 41 PZ -53.2 42 PZ -137.2 43 PZ -173.2 44 C/Z -53. -7.2 11. 45 C/Z -53. -7.2 10.2 46 C/Z 53. -7.2 11. 47 C/Z 53. -7.2 10.2 48 C/Z -53. 142.8 11. 49 C/Z -53. 142.8 10.2 50 C/Z 53. 142.8 11. 51 C/Z 53. 142.8 10.2 C BUNKER AND WALLS 200 1 PX -570 209 1 PX -620 210 1 PX 570 219 1 PX 620 300 1 PY -760 309 1 PY -810 310 1 PY 480 319 1 PY 530 400 PZ -406 409 PZ -456 410 PZ 530 419 PZ 580 MODE N *TR1 0 0 0 45 135 90 45 45 90 90 90 0 *TR2 0 0 0 45 90 45 90 0 90 135 90 45 C STAINLESS STEEL (AISI 316) M1 5010.89C -2.8E-4 5011.89C -2.87E-3 6000.89C -0.04 14000.89C -0.41 23000.89C -0.16 24000.89C -16.8 25055.89C -1.14 26000.91C -68.1 27059.35C -0.14 28000.89C -10.7 42000.89C -2.12 29000.89C -0.09 C H2O M2 1001.89C 2. 8016.89C 1. C COPPER M3 29000.89C 1. C AIR M4 7014.04C .788903 8016.89C .211097 C ALUMINUM M5 13027.89C 1. C CONCRETE + 5% FE DENS. 2.6 M6 1001.89C -0.005358 8016.89C -0.474193 11023 -0.016312 13027.89C -0.043491 14000.89C -0.299658 19000.35C -0.018031 26000.91C -0.04 C COPPER 50% IRON 50% M7 26000.91C .5 29000.89C .5 C SOURCE SPECIFICATIONS RDUM .230 1.5 0.0 0.001 0.0 GEOM.DATA ENDS***************************************************** DETEC.DATA STARTS************************************************** I. ACTIVATION REACTIONS AND NUCLEAR DATA EMPLOYED ******************************************************************** Reaction Half-life Isotopic g-ray energy g-ray abundance(%) (keV) branching(%) ******************************************************************** 27Al(n,a)24Na 14.96h 100.0 1368.6 100.0 56Fe(n,p)56Mn 2.577h 91.72 846.8 98.87 58Ni(n,2n)57Ni 1.503d 68.27 1377.6 80.0 58Ni(n,p)58Co 70.92d 68.27 810.8 99.44 115In(n,n')115mIn 4.486h 95.7 336.24 45.9 55 Mn(n,g)56M(n,g) 2.577h 100.0 846.8 98.87 197Au(n,g)198Au 2.696d 100.0 411.8 95.56 ******************************************************************** II. ACTIVATION MEASUREMENT CONDITIONS ******************************************************************** Reaction Foil diameter Foil thickness mm Penetration depth < 25 cm > 25 cm ******************************************************************** 58Ni(n,2n)57Ni 18 1 mm 2 mm 27Al(n,a)24Al 18 1 mm 2 mm 56Fe(n,p)56Mn 18 1 mm 2 mm 58Ni(n,p)58Co 18 1 mm 2 mm 115In(n,n')115mIn 18 1 mm 2 mm 55 Mn(n,g)56Mn 18 200 microm 200 microm 197Au(n,g)198Au 18 50 microm 50 microm ********************************************************************* III. GAMMA RAY MEASUREMENTS - DETECTORS AND CALIBRATION Induced g-rays emission was measured by three HPGe detectors. Activation reaction rates were deduced from the measured gamma-ray peak counts, using the Standard Absolute Radiometric Technique. The absolute efficiency of the HPGe detectors is determined up to 1836 keV by using standard point-like gamma-ray sources, whose intensity is known within less than +-1% for mono-gamma ray sources, and within +-2.5% for multi-gamma ray sources. The experimental data are fitted using the least square method, and the resulting uncertainty on the calibration curve is +-2% at one sigma level. A routine check of the stability of HPGe efficiency is performed using 137Cs and 60Co sources. Intercalibration of the detectors is also performed by comparing activated samples: the comparison typically shows an agreement among the detectors within quoted +-2% uncertainty. DETEC.DATA ENDS**************************************************** EXPE.DATA STARTS ************************************************* ____________________________________________________ 27Al(n,a)24Na Depth E Random error(%) Systematic error (%) Total error (%) (cm)) (RE) (SE) (TE) 4.95 5.94E-5 3.5 4.4 5.6 10.05 1.52E-5 3.5 4.4 5.6 20.15 1.70E-6 3.5 4.4 5.6 30.30 2.29E-7 3.8 4.4 5.8 40.50 3.66E-8 3.8 4.4 5.8 50.70 6.65E-9 4.7 4.4 6.4 55.90 2.60E-9 8.9 4.4 10.0 ______________________________________________________________________ 56Fe(n,p)56Mn Depth (cm) E RE (%) SE (%) TE (%) 4.95 5.09E-5 3.3 4.4 5.5 10.05 1.28E-5 3.3 4.4 5.5 20.15 1.48E-6 4.5 4.4 6.3 30.30 1.93E-7 5.1 4.4 6.7 40.50 3.24E-8 4.5 4.4 6.3 50.70 5.61E-9 5.6 4.4 7.1 60.90 1.21E-9 15.4 4.4 16.0 ____________________________________________________ 115In(n,n')115mIn Depth (cm) E RE (%) SE (%) TE (%) 4.95 1.67E-4 2.2 2.5 3.3 10.05 6.81E-5 2.2 2.5 3.3 20.15 1.35E-5 2.3 2.5 3.4 30.30 2.86E-6 3.0 2.5 3.9 40.50 6.81E-7 3.7 2.5 4.5 50.60 1.57E-7 3.9 2.5 4.6 60.90 4.02E-8 4.1 2.5 4.8 ____________________________________________________ 197Au(n,g)197Au Depth (cm) E RE (%) SE (%) TE (%) 5.0 7.79E-4 1.9 4.4 4.8 10.0 8.88E-4 2.4 4.4 5.0 20.0 9.29E-4 2.4 4.4 5.0 30.0 7.76E-4 2.4 4.4 5.0 40.0 5.33E-4 3.3 4.4 5.5 50.0 3.11E-4 3.3 4.4 5.5 60.0 1.17E-4 3.3 4.4 5.5 ____________________________________________________ 55Mn(n,g)56Mn Depth (cm) E RE (%) SE (%) TE (%) 5.0 2.66E-5 2.4 4.4 5.0 10.0 3.24E-5 2.4 4.4 5.0 20.0 3.33E-5 2.4 4.4 5.0 30.0 2.40E-5 2.4 4.4 5.0 40.0 1.58E-5 2.4 4.4 5.0 50.0 1.00E-5 2.4 4.4 5.0 60.0 3.89E-6 2.4 4.4 5.0 ____________________________________________________ 58Ni(n,p)58Co Depth (cm) E RE (%) SE (%) TE (%) 4.95 2.17E-4 2.8 1.6 3.2 10.05 6.66E-5 2.8 1.6 3.2 20.15 9.21E-6 2.9 1.6 3.3 30.30 1.41E-6 3.3 1.6 3.7 40.50 2.45E-7 4.1 1.6 4.4 50.70 3.85E-8 5.3 1.6 5.5 60.90 7.10E-9 6.3 1.6 6.5 ____________________________________________________ 58Ni(n,2n)57Ni Depth (cm) E RE (%) SE (%) TE (%) 4.95 1.78E-5 3.1 1.6 3.5 10.05 4.32E-6 3.7 1.6 4.0 20.15 4.42E-7 3.9 1.6 4.2 30.30 5.98E-8 6.2 1.6 6.4 ____________________________________________________________ EXPE.DATA ENDS **************************************************** CALC.DATA STARTS**************************************************** The six columns are: Depth (cm), C, MCNP statistics %, Error due to activ. react.% Tot. error on C in %, C/E and Total error on C/E ______________________________________________________ 27Al(n,a)24Na 4.95 5.71E-5 2.0 0.6 2.1 0.96 0.06 10.05 1.40E-5 2.0 0.6 2.1 0.92 0.06 20.15 1.48E-6 2.0 0.6 2.1 0.87 0.06 30.30 2.09E-7 2.0 0.6 2.1 0.91 0.06 40.50 3.40E-8 2.0 0.6 2.1 0.93 0.06 50.70 5.93E-9 2.0 0.6 2.1 0.89 0.06 55.90 2.53E-9 2.0 0.6 2.1 0.97 0.10 ______________________________________________________ 56Fe(n,p)56Mn 4.95 5.39E-5 2.0 1.3 2.4 1.06 0.06 10.05 1.29E-5 2.0 1.3 2.4 1.01 0.06 20.15 1.38E-6 2.0 1.3 2.4 0.93 0.07 30.30 1.94E-7 2.0 1.3 2.4 1.01 0.07 40.50 3.07E-8 2.0 1.3 2.4 0.95 0.07 50.70 5.07E-9 2.0 1.3 2.4 0.90 0.07 60.90 8.94E-10 2.0 1.3 2.4 0.74 0.12 _______________________________________________________ 115In(n,n')115mIn 4.95 1.64E-4 2.0 2.0 2.8 0.98 0.04 10.05 6.31E-5 1.0 2.6 2.8 0.93 0.04 20.15 1.20E-5 1.0 2.7 2.9 0.89 0.04 30.30 2.53E-6 1.0 2.7 2.9 0.88 0.04 40.50 5.36E-7 1.0 2.7 2.9 0.79 0.04 50.60 1.27E-7 2.0 2.2 3.0 0.81 0.04 60.90 2.78E-8 2.0 2.2 3.0 0.69 0.04 _______________________________________________________ ____________________________________________________ 197Au(n,g)197Au 5.0 8.18E-4 3.0 0.2 3.0 1.05 0.06 10.0 9.26E-4 3.0 0.2 3.0 1.04 0.06 20.0 9.97E-4 2.0 0.2 2.0 1.07 0.05 30.0 8.65E-4 2.0 0.2 2.0 1.11 0.05 40.0 5.66E-4 2.0 0.2 2.0 1.06 0.06 50.0 3.51E-4 2.0 0.2 2.0 1.13 0.07 60.0 1.44E-4 2.0 0.2 2.0 1.23 0.07 ______________________________________________________ 55Mn(n,g)56Mn 5.0 2.62E-5 3.0 11.4 11.8 0.98 0.13 10.0 3.33E-5 2.0 10.6 10.8 1.03 0.12 20.0 3.02E-5 2.0 9.8 10.0 0.91 0.11 30.0 2.52E-5 1.0 9.7 9.8 1.05 0.11 40.0 1.76E-5 1.0 9.1 9.2 1.11 0.10 50.0 9.92E-6 1.0 9.0 9.1 0.99 0.10 60.0 4.11E-6 1.0 9.3 9.4 1.06 0.11 ______________________________________________________ 58Ni(n,p)58Co 4.95 2.29E-4 2.0 9.8 10.0 1.06 0.10 10.05 6.64E-5 2.0 8.1 8.3 1.00 0.09 20.15 8.74E-6 2.0 6.6 6.9 0.95 0.07 30.30 1.42E-6 1.0 6.1 6.2 1.01 0.08 40.50 2.46E-7 1.0 5.9 6.0 1.00 0.08 50.70 4.20E-8 1.0 5.7 5.8 1.09 0.08 60.90 7.83E-9 1.0 5.5 5.6 1.10 0.09 _______________________________________________________ _______________________________________________________ 58Ni(n,2n)57Ni 4.95 1.77E-5 2.0 2.0 2.8 0.99 0.05 10.05 4.14E-6 2.0 2.0 2.8 0.96 0.05 20.15 4.06E-7 2.0 1.7 2.6 0.92 0.05 30.30 5.31E-8 2.0 1.7 2.6 0.89 0.06 _______________________________________________________________ CALC.DATA ENDS********************************************************