1 / 13

C6D6 Neutron sensitivity simulation with MCNP

C6D6 Neutron sensitivity simulation with MCNP. Vincenzo Variale, Nicola Colonna INFN-bari. Geometry used for simulations in MCNP. z =360 cm. C6D6. x=560 cm.

cooper-allison
Download Presentation

C6D6 Neutron sensitivity simulation with MCNP

An Image/Link below is provided (as is) to download presentation Download Policy: Content on the Website is provided to you AS IS for your information and personal use and may not be sold / licensed / shared on other websites without getting consent from its author. Content is provided to you AS IS for your information and personal use only. Download presentation by click this link. While downloading, if for some reason you are not able to download a presentation, the publisher may have deleted the file from their server. During download, if you can't get a presentation, the file might be deleted by the publisher.

E N D

Presentation Transcript

  1. C6D6 Neutron sensitivity simulation with MCNP Vincenzo Variale, Nicola Colonna INFN-bari

  2. Geometry used for simulations in MCNP z=360 cm C6D6 x=560 cm Transverse section of the experimental area. The violet boxes refer to C6D6 detectors very close to the Aluminum pipe; the green area is air and the red one is concrete surrounds the whole experimental area. The numbers refer to the surfaces as used in the code MCNPX Al pipe 40 cm

  3. c Created on: Wednesday, October 26, 2011 at 16:38 1 1 -1.1 -1 2 1 -1.1 -2 3 2 -0.001225 -4 1 2 (3 :-8) 4 3 -2.699 -3 8 5 228 -2.35 -5 4 6 0 5 1 rcc 0 0 2.7 0 0 6 6 $D1 2 rcc 0 0 -8.7 0 0 6 6 $D2 3 rcc -91.5 0 0 89.5 0 0 2.5 $ext Al tube 8 rcc -91.5 0 0 89.5 0 0 2.3 $int Al tube 4 rpp -280 280 -180 180 -180 180 5 rpp -320 320 -220 220 -220 220 mode n p phys:p 3j -1 imp:n 1 4r 0 $ 1, 6 imp:p 1 4r 0 $ 1, 6 imp:e 1 4r 0 $ 1, 6 m1 1002.70c -0.5 $MAT1 6000.70c -0.5 m2 7000.70c -0.78 $MAT2 8000.70c -0.22 m3 13027.70c -1 $MAT3 c m228 1000.02p -0.005558 $concrete (ordinary with ENDF-VI) c 8000.02p -0.4980765 11000.02p -0.017101 12000.02p -0.002565 c 13000.02p -0.045746 14000.02p -0.3150923 16000.02p -0.001283 c 19000.02p -0.019239 20000.02p -0.0829411 26000.02p -0.012398 m228 1001.70c -0.005558 8016.70c -0.498076 11023.70c -0.017101 12024.70c -0.001999 12025.70c -0.000264 12026.70c -0.000302 13027.70c -0.045746 14028.70c -0.289486 14029.70c -0.015181 14030.70c -0.010425 16032.70c -0.001216 16033.70c -1e-005 16034.70c -5.7e-005 19039.70c -0.01788 19040.70c -2e-006 19041.70c -0.001357 20040.70c -0.08019 20042.70c -0.000562 20043.70c -0.00012 20044.70c -0.00188 20046.70c -4e-006 20048.70c -0.000186 26054.70c -0.000707 26056.70c -0.01139 26057.70c -0.000265 26058.70c -3.6e-005 mt228 al27.10t Fe56.10t hortho.10t mt3 al27.10t sdef erg=0.025e-6 pos=7.2 0. 0. f8:p,e 1 $Energy distribution for Pulses created in cell 1 e8 0 0.2 140i 10. c t8 0 150i 10. ft8 geb 0 0.1 0.0225 f18:p,e 2 $Energy distribution for Pulses created in cell 1 e18 0 0.2 140i 10. ft18 geb 0 0.1 0.0225 nps 3000000 mplotfreq 1500000 loglog tally 8 cop tally 18 (Rcc: Vx , Vy, Vz ; Hx, Hy, Hz; R) Geometry data (Rpp: XminXmax ; YminYmax; ZminZmax) Material data Concrete elements Neutron source definition (here thermal neutrons. Further simulations with En= 1, 10, 100 eV 1, 10, 100 keV, 1 MeV; have been also considered)

  4. Simulation results for some neutron energies used Neutron of 1 eV Thermal neutrons Tally 8 : Pulse height, in pulses/Par. Num.

  5. Simulations for some other neutron energies Neutron energy , 100 eV Neutron energy , 100 keV

  6. From the simulation results shown before; by integrating the deposited energy in the C6D6 1 (for each source neutron energy); (a threshold energy of 200 keV has been used) C6D6+Al+Concrete Only C6D6 Neutron sensitivity (Energy deposited in C6D6 1 in MeV for neutron) vs the source neutron energies: the blue curve refers to the case in which Al pipe and concrete have been included in the calculation; the red curve refers to the case without Al pipe and concrete.

  7. Simulations of the interaction times in C6D6 for different source neutron energies have been considered: Thermal neutrons 100 eVneutrons Tally 6 : Energy Deposition (MeV/gm)

  8. 1 keVneutrons 100 keVneutrons

  9. The interaction times shown in the simulations, however, did not take into account the TOF of neutrons.To get the true interaction time, the neutron TOF (which of course will depend on the neutron energy), should be added to the times of the previous simulation results:TOF(Ek) = L/(2 Ek/mn)1/2 where: L = 185 mThen the ‘true’ time, t’(En) will be: t’(En)i = TOF(En)+ t(En)iwhere i refers to the time binning of the previous fig.sFrom the ‘true’ times can be determined the corresponding ‘true’ neutron energies, En’ :vn’= L/t’ and En’=1/2mnvn’2

  10. The deposited energy (Ed) in the C6D6 1 vs the ‘true’ neutron energies (En’)(for each source neutron energy case): the case of 1 eV The case of 100 eV

  11. For otherenergyneutronsources: 1 keVneutron source 10 keVneutron source

  12. PromptFraction Ed definitionfor eachneutronenergy: Ed : Depositedenergy in C6D6 1

  13. The prompt fraction of the deposited energy in detector C6D6 1 vs the source neutron energy (corrected):

More Related