MCNP TRANSPORT CODE SYSTEM & DETECTOR DESIGN

Transcription

1 MCNP TRANSPORT CODE SYSTEM & DETECTOR DESIGN Name: MAHMUT CÜNEYT KAHRAMAN Matr. Nr:

2 CONTENTS 1. Introduction of MCNP Code System 1.1. What is an input file? 1.2. What is an output file? 2. Detector Design 2.1. Creating the input file Surface definition Cell definition Source definition Material Definition Tallies Description 2.2. Creating the output file 2.3. Analysis result 2

3 1. Introduction of MCNP Code System MCNP Code System is a software package for simulating nuclear processes. It is used primarily for the simulation of nuclear processes, such as fission, but has the capability to simulate particle interactions involving neutrons, photons, and electrons. "Specific areas of application include, but are not limited to, radiation protection and dosimetry, radiation shielding, radiography, medical physics, nuclear criticality safety, detector design and analysis, nuclear oil well logging, accelerator target design, fission and fusion reactor design, decontamination and decommissioning." 1.1. What is an input file? The user creates an input file that is subsequently read by MCNP. This file contains information about the problem in areas such as: o the geometry specification o the description of materials and selection of cross-section evaluations, o the location and characteristics of the neutron, photon, or electron source, o the type of answers or tallies desired, and o any variance reduction techniques used to improve efficiency And units should be like shown below; lengths in centimeters, energies in MeV, times in shakes (10-8 sec), temperatures in MeV (kt), atomic densities in units of atoms/barn-cm, mass densities in g/cm^3, cross sections in barns (10-24 cm^2 ), heating numbers in MeV/collision, and atomic weight ratio based on a neutron mass of amu. In these units, Avogadro s number is x 10^24 3

4 1.2. What is an output file? After completing input file, mcnp.exe runs in CMD. Output file will be created automatically after completion of running. The output file contains tables of standard summary information to give the user a better idea of how the problem ran. This information can give insight into the physics of the problem and the adequacy of the Monte Carlo simulation. If errors occur during the running of a problem, detailed diagnostic prints for debugging are given. Analysis is made according to these data which are taken from output file. For instance, Decays, energies and uncertainties will be found in output file of detector design. 2. Detector Design 3I have designed Germanium detector. The measurements of surfaces and material data for design were taken from the picture shown below. Figure 1 Measurements of detector which was designed. 4

5 2.1. Creating the input file The input file for the Germanium detector was written on txt. format file. Input file is given below; > c cell definitions > imp:p=1 $Inner Copper detector > #1 imp:p=1 $Germanium Detector > #1 imp:p=1 $Dead Layer > imp:p=1 $Gap of Air > imp:p=1 $Plastic Covering > imp:p=1 $Water source > imp:p=1 $Plastic Covering > imp:p=1 $Universe > imp:p=0 $Rest of universe > > c Surfaces > 10 rcc > 20 rcc > 30 rcc > 40 rcc > 50 rcc > 60 rcc > 70 rcc > 80 so 80 > > sdef erg=d1 par=2 cel=6 pos axs rad=d2 ext=d3 > si1 L > sp1 D > si > si > f8:p 2 > e i 2.5 > phys:p > mode p > m $ Ger > m $ Al > m $ copper > m $ water > m $ plastic > m $Air nps

6 Surface definitions Surfaces are defined by: supplying coefficients to the analytic surface equations, for certain types of surfaces, known points on the surfaces. In design; 8 different surfaces were defined. Last one, 80 was defined as universe. The meanings of the each item is given in the example respectively; > 10 rcc A RCC Vx Vy Vz Hx Hy Hz R A Number of surface, RCC Right circular cylinder, Vx Vy Vz Center of base Hx Hy Hz Cylinder axis vector R Radius Cell definitions The cells are defined by the; intersections, unions, complements of the regions bounded by the surfaces. In design; 9 different cells were defined. The meanings of the each item is given in the example respectively; > #1 imp:p=1 Number of cell, number of the material, geometric data inside the 20 th surface, outside of 10 th surface, except the surface of cell 1, importance of cell Source definition Source of the simulation is given in this section, point or volume source. Independent probability distributions may be specified for the source variables of energy, time, position, and direction, and for other parameters such as starting cell(s) or surface(s). In design; volume source was defined. The meanings of the each item is given in the example respectively; > sdef erg=d1 par=2 cel=6 pos axs rad=d2 ext=d3 6

7 general source (sdef), energy distribution (erg=d1), 2=photons (par=2), the cell inside source (cell=6), reference point inside the cylinder (pos 0 0 6,1), direction of cylinder axis (axs 0 0 1),radial distribution(rad=d2) axial distribution (ext=d3) Material Definition This is usually specified as an integer value ZZZAAA where ZZZ is the atomic number and AAA is the atomic mass number or zero for a natural element. 6 different materials were defined; Germanium, Aluminum, Copper, Plastic, Water, Air. The meanings of the each item is given in the example respectively; > m $ water Number of material, atomic number of hydrogen, percentage of hydrogen in compound, atomic number of oxygen, percentage of oxygen in compound Tallies Description The tallies are identified by tally type and particle type as follows. Tallies are given the numbers 1, 2, 4, 5, 6, 7, 8, or increments of 10 thereof, and are given the particle designator :N, :P, or :E (or :N,P only in the case of tally type 6 or :P,E only in the case of tally type 8). F8 tally type was used for photon particles in detector design (pulse height distribution). > f8:p 2 > e i 2.5 The second line explains that value of 1 energy bin is where measurement starts from to 2.5 with a total of 4096 bins of energy. Finally when we run this input file in mcnp, we can see our detector and volume sources. 7

8 Figure 2 Detector and Source in MCNP 2.2. Creating the output file After completing the input file, CMD is run to create the output file. Command must be written as mcnp.exe name=nameoffile.txt on CMD. Then output file will be created in hard disc of C. Figure 3 CMD screen 8

9 2.3. Analysis result In output file, Data of decays and energies of particles can be found. Some decays and energies are given as example. With these data, we can create energy spectrum. cell 2 energy decays 1,00E+00 6,99E+02 6,10E+00 1,00E-01 1,22E+01 2,00E-01 1,83E+01 5,00E-01 2,44E+01 6,00E-01 3,05E+01 2,00E-01 3,66E+01 8,00E-01 4,27E+01 7,00E-01 4,88E+01 4,00E ,00E+02 7,00E+02 Energy Decays 6,00E+02 5,00E+02 4,00E+02 3,00E+02 Energy Decays 2,00E+02 1,00E+02 0,00E+00 0,00E+00 5,00E+03 1,00E+04 1,50E+04 2,00E+04 2,50E+04 3,00E+04 9

10 For more details and to be able to see small decay numbers, we can focus the energy spectrum. 5,00E+00 4,50E+00 Energy Decays 4,00E+00 3,50E+00 3,00E+00 2,50E+00 2,00E+00 1,50E+00 Energy Decays 1,00E+00 5,00E-01 0,00E+00 0,00E+00 5,00E+03 1,00E+04 1,50E+04 2,00E+04 2,50E+04 3,00E+04 Thank you. 10