Changes between Version 26 and Version 27 of GuarDyan_features


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Timestamp:
Feb 17, 2018, 11:02:54 PM (6 years ago)
Author:
dieda
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  • GuarDyan_features

    v26 v27  
    11= GUARDYAN Basic Features
    2 
    3 == Aims
    4 GUARDYAN simulates the time evolution of fission chains. The neutrons are followed from the beginning to the end of a time interval. This tracking philosophy differs from the usual source convergence calculations where neutrons are followed from generation to generation, i.e. from fission event until the next fission event. The source term for GUARDYAN is a "snapshot" of the neutron population at a time instance consisting of delayed neutron precursors waiting to "hatch" and prompt neutrons crossing the time boundary. The time dependent changes in the neutron population may originate from time dependent changes of the cross sections either as change of a material or of the temperature of a certain material. The current version of GUARDYAN is not capable of simulating thermal or any other feedback this feature is planned for future releases.
    5 
    6 
    72
    83== Software and Hardware Environment
    … …  
    149Hardware: 2 x Geforce GTX 690, 4 x  Geforce GTX 1080
    1510
    16 == Features
    17 === Geometry descriptors
    18 In order to run simulations on a real reactor geometry we need to create the tools for defining
    19 geometry. Similarly to other particle transport codes (Serpent, MCNP, OpenMC) in GUARDYAN geomety is  defined by cells containing isotopes. Cells are defined by bounding surfaces. These data can be given in ''xml'' files.
     11== Geometry descriptors
     12In order to run simulations on a real reactor geometry we need to create the tools for defining geometry. Similarly to other particle transport codes (Serpent, MCNP, OpenMC) in GUARDYAN geomety is  defined by cells containing isotopes. Cells are defined by bounding surfaces. These data can be given in ''xml'' files.
    2013
    21 ==== Surfaces
     14=== Surfaces
    2215In the current version general
    2316positioned sphere, plane, cone, torus and infinite cylinder are
    … …  
    8174Considering the most important isotopes in a e.g.  VVER-440 reactor with fresh fuel: 1H, 2H, 4He,10B, 11B 16O, 17O, 90Zr, 91Zr, 92Zr, 94Zr, 96Zr, 93Nb, 152Gd, 154Gd, 155Gd, 156Gd, 157Gd, 158Gd, 160Gd,174 Hf, 176Hf, 177Hf, 178Hf, 179Hf, 180Hf,235 U, 238U, with 8 different materials defined to store the cross section data for a given temperature 102,4 MB memory is needed on the GPU. For a realistic VVER-440 power plant with full detailed modelling the cross section date reaches 5Gb in memory need. The videocards available at our institute contain 2 (GTX 690) or 8 (GTX 1080) GB memory each for realistic simulations therefore modern videocards offer enough physical memory. This latter example is considered as extreme in memory need.
    8275
    83 == Neutron Transport
    84 
    85 The particle transport is based on conventional Woodcock method where we need to generate the majorant cross sections for a combined energy
    86 lattice in advance. For real scattering first we reduce the weight of the particle by the probability of survival using implicit capture then we draw a reaction by the cross sections. We consider reactions with the following MT numbers:
    87 
    88 MT=2, 5, 11, 16, 17, 18, 19, 20, 21, 22, 23, 24, 25, 28, 29, 30, 32, 33, 34, 35, 36, 37, 38, 41, 44, 45, 51-90, 101.
    89 
    90 After choosing the reaction, the determination of the new energy and angle of the neutron depends on the given ACE law (ACE Law 3, 4, 7,
    91 9, 11, 44, 61, 66 ). The angle distribution can be isotropic or given in table format. The coordinate system of the given data should also be considered.
    92 
    93 [[Image(UO_H2O_rods_material-map.png, 50%)]]
    94 
    95 ''' Fig. 3. x-y slice of the geometry of 60 uranium oxide rods in water cylinder'''
    96 
    97 
    98 Using the geometry seen in Figure 3 at t = 0s we place a 0.1MeV energy source at 0,0,0 point. Figure 4 shows how power density develops in
    99 time steps. In Figure 5 the distribution evolved after 1ms is plotted.
    100 
    101 [[Image(UO_H2O_rods_16timestep_z-sum.png​, 100%)]]
    102 
    103 ''' Fig. 4. Temporal and spatial evolution of released power '''
    104 
    105 
    106 [[Image(UO_H2O_rods_surf_z-sum_wiener.png​, 50%)]]
    107 
    108 ''' Fig. 5. Power density distribution after 0.001s '''
    109 
    110 There is a possibility of saving the trajectory of the neutrons. In Figure 6. for a 235U sphere at t = 0s we started neutrons at 1eV energy from 0,0,0 point. The first reactions of 500 neutrons is shown on the figure where we indicate elastic scattering with color green, fission with red, every other reaction with yellow.
    111 
    112 
    113 [[Image(traject.png​​, 50%)]]
    114 
    115 ''' Fig. 6. Green MT=2, red MT=18, every other reaction is indicated with yellow '''
    116 
    117 
    118 == Variance Reduction
    119 The time dependent tracking of the neutron population in a multiplying, near critical medium is very challenging in terms of Monte Carlo convergence. A naive analog game in most cases would statistically diverge, moreover  it will give  an underestimate of the power as the very low chance contributions of a high number of fission in certain chains see Fig. 7. Therefore the calculation is performed always keeping a single particle as a sample of the neutron population gaining or loosing weight at interactions. The neutron weight distribution must be kept around the mean for ensuring statistical convergence.
    120 The neutrons are followed from time interval to time interval and the population at the interval ends using splitting and Russian roulette while keeping the total population number constant.
    121 Having single, non-branching  calculations also supports the architecture of the GPU where threads can be set to single neutron chains.   
    122 
    123 [[Image(TDMCC_varP_analog_vs_nonanalog.png​​, 70%)]]
    124 
    125 ''' Fig. 7. Analog and non-analog simulation results for time dependent power evolution for a multiplying medium. Analog simulation produces an underestimate of the power '''
    126 
    127 Biased sampling schemes are applied at fission yield, delayed neutron, interaction type sampling with ongoing development regarding path length sampling and angular biasing.
    128 
    129 == Dynamic Capabilities
    130 GUARDYAN is meant to be a dynamics Monte Carlo code with thermohydraulic feedback. The current state of the code support time dependent cross section changes without feedback effects.
    131 
    132 
    133 == Validation
    134 
    135 GUARDYAN  is being validated against MCNP6, using ENDF-B.VII.I. Benchmarking geometries and tallies involve homogenously filled spheres with a point source in the middle and total leakage and flux energy-time spectra are calculated by both codes and results compared. Fluxes after certain interactions also compared for some isotopes.
    136 
    137 
    138 ||Element       ||R cm  ||Density g/cm^3        ||Energy  MeV   ||Nb || neutron Library||
    139 ||13 Al 27      ||10    ||2.6989        ||0,1   ||1,00E+06      ||70 c
    140 ||4 Be 9        ||5     ||2     ||0,1   ||1,00E+07      ||70 c
    141 ||26 Fe 56      ||10    ||7.874 ||0,1   ||1,00E+06      ||70 c
    142 ||1 H 1 ||   50 ||0,02  ||0,1   ||1,00E+06      ||70 c
    143 ||2 He 4        ||500   ||0,04  ||0,1   ||1,00E+06      ||70 c
    144 ||3 Li 7        ||50    ||0,5   ||0,1   ||1,00E+06      ||70 c
    145 ||8 O 16        ||500   ||0,004 ||0,1   ||1,00E+07      ||70 c
    146 ||92 U 238 (0,96)/239 (0,04)    ||10    ||10,8  ||0,1   ||1,00E+06      ||70 c
    147 ||40 Zr 90      ||5     ||6.52  ||0,1   ||1,00E+07      ||70 c
    148 ||Zircaloy      ||5     ||6,52  ||0,1   ||1,00E+07      ||70 c
    149 ||26 Fe 56 (0,95)/54 (0,05)     ||10    ||7.874 ||0,1   ||1,00E+06      ||70 c
    150 ||7 N 4 ||5     ||1.24982       ||0,1   ||1,00E+07      ||70 c
    151 ||9 F 19        ||5     ||1.696 ||0,1   ||1,00E+07      ||70 c
    152 ||11 Na 23      ||5     ||0.968e1       ||0,1   ||1,00E+06      ||70 c
    153 ||12 Mg 24      ||10    ||1.738 ||0,1   ||1,00E+06      ||70 c
    154 ||14 Si 28      ||10    ||2.33  ||0,1   ||1,00E+06      ||70 c
    155 ||94 Pu 239/40/41 (0,5/0,3/0,2) ||5     ||19.816        ||0,1   ||1,00E+06      ||70 c
    156 ||235 U || 10   ||19.1  ||0,1   ||1,00E+06      ||70 c
    157 
    158 '''Table 2. List of isotopes already validated'''
    159 
    160 Starting energies and sphere diameters are adopted to relevant interactions and neutronics.
    16176
    16277== Project Participants & Contacts