|Program name||Package id||Status||Status date|
|Package ID||Orig. computer||Test computer|
|NEA-1673/04||Linux-based PC,PC Windows,UNIX W.S.||Linux-based PC,PC Windows|
1- Problems to be solved
MVP/GMVP can solve eigenvalue and fixed-source problems. The multigroup code GMVP can solve forward and adjoint problems for neutron, photon and neutron-photon coupled transport. The continuous-energy code MVP can solve only the forward problems. Both codes can also perform time-dependent calculations.
2- Geometry description
MVP/GMVP employs combinatorial geometry to describe the calculation geometry. It describes spatial regions by the combination of the 3-dimensional objects (BODIes). Currently, the following objects (BODIes) can be used.
BODIes with linear surfaces: half space, parallelepiped, right parallelepiped, wedge, right hexagonal prism
BODIes with quadratic surface and linear surfaces: cylinder, sphere, truncated right cone, truncated elliptic cone, ellipsoid by rotation, general ellipsoid
Arbitrary quadratic surface and torus
The rectangular and hexagonal lattice geometry can be used to describe the repeated geometry. Furthermore, the statistical geometry model is available to treat coated fuel particles or pebbles for high temperature reactors.
3- Particle sources
Energy-, angle-, space- and time-dependent particle sources can be specified with various sampling functions.
4- Cross sections
The ANISN-type PL cross sections or the double-differential cross sections can be used in the multigroup code GMVP. On the other hand, the specific cross section libraries are used in the continuous-energy code MVP. The libraries are generated from the evaluated nuclear data (JENDL-4.0, ENDF/B-VII.1, JEFF-3.2 etc.) by using the LICEM code. The neutron cross sections in the unresolved resonance region are described with the probability table method. The neutron cross sections at arbitrary temperatures are available for MVP by just specifying the temperatures in the input data.
5- Boundary conditions
Vacuum, perfect reflective, isotropic reflective (white), periodic boundary conditions can be specified.
6- Variance reduction techniques
The basic variance reduction techniques Russian roulette kill and splitting are implemented. In addition, importance and weight window based on them are available. Path stretching and source biasing can be also used.
The track length, collision, point and surface crossing estimators are available. The eigenvalue is estimated by the track length, collision and analog estimators for neutron production and neutron balance methods. In the final estimation, the most probable value and its variance are calculated by the maximum likelihood method with the combination of the estimators.
GMVP calculates the eigenvalue, the particle flux and reaction rates in each spatial region, each energy group and each time bin for each material, each nuclide and each type of reactions, and their variances as the basic statistical parameters. In addition to these physical quantities, MVP calculates the effective microscopic and macroscopic cross sections and the corresponding reaction rates in the specified regions. These quantities are basically tallied for each spatial region but can be tallied for the arbitrary combination of the regions with options. Furthermore, the calculated quantities are output to files and can be then used for the input data of a drawing program mentioned later or a burnup calculation code MVP-BURN.
9- Drawing geometry
The CGVIEW code draws the cross-sectional view on an arbitrary plane and output it on a display or in the postscript or encapsulated postscript form. These functions are useful for checking the calculation geometry.
10- Burnup calculation
The auxiliary code MVP-BURN implemented in the MVP/GMVP system is available for burnup calculations.
Parallel calculations can be performed with standard libraries MPI and PVM.
MVP and GMVP are based on the continuous-energy and multigroup method, respectively. In the continuous-energy method, all reactions are treated explicitly as given in evaluated nuclear data. Pointwise cross sections and angular/energy distributions are basically used for particle tracking. For neutron thermal scattering, the free gas model is used to take into account the thermal motion of target nuclei or the scattering law data S(α,β) and elastic thermal scattering representation in the ENDF format are used to take into account the binding effect in liquids and solids. In the unresolved resonance region of neutron cross sections, the probability table method is used. For photon reactions, detailed and simple models are available. The detailed model includes the generation of fluorescent X-rays in the photoelectric effect and the correction factor of the Klein-Nishina differential cross section for the incoherent scattering but the simple model does not include them. In both models, Bremsstrahlung photons can be optionally taken into account with the thick target approximation. Energy ranges are from 0.00005 eV to 20 MeV for neutrons and from 1 keV to 100 MeV for photons. In the multigroup method, all reactions are treated according to multigroup cross section data given by users.
The following table lists CPU times for sample problems. All the calculations were performed on Linux with a single core of Intel Xeon CPU E5-2630, 2.30 GHz.
sample1 (Single assembly calculation): 2 min 14 sec
sample2 (Four assembly calculation): 57 sec
sample3 (Neutron-photon coupled problem): 3 min 20 sec
sample4 (Example of the statistical geometry model): 1 min 53 sec
sample5 (Example of the FREE-FRAME-LATTICE option): 37 sec
sample6 (Example of tally dimensions SOURCE-REGION and MARKER-REGION): 10 sec
sample7 (Example of a Feynman-alpha calculation): 29 sec
sample8 (Example of surface crossing and collision estimators): 14 sec
sample9 (Example of surface crossing and point estimators): 11 sec
sample1 (Takeda benchmark): 7 sec
sample2 (Calculation with the DDX-form cross sections): 1 sec
sample3 (Neutron-photon coupled problem): 16 sec
sample4 (Example of the statistical geometry model): 27 sec
sample5 (Example of the FREE-FRAME-LATTICE option): 40 sec
sample6 (Example of tally dimensions SOURCE-REGION and MARKER-REGION): 5 sec
sample7 (Example of a Feynman-alpha calculation): 6 sec
sample8 (Example of surface crossing and collision estimators): 11 sec
MVP/GMVP has the following unusual capabilities:
Burnup calculation with the MVP-BURN code.
Reactor noise analysis based on simulation of Feynman-alpha experiments.
Statistical geometry model is available to address the heterogeneity of coated fuel particles or pebbles for high temperature gas-cooled reactors.
MVP Version 3 implements the following new capabilities:
Perturbation calculation for effective multiplication factor.
Exact resonant elastic scattering model.
Calculation of reactor kinetics parameters (effective delayed neutron fraction and neutron generation time).
Photonuclear reaction model.
Simulation of delayed neutrons.
Generation of group constants.
CGVIEW: Program to draw cross-sectional views of MVP/GMVP calculation geometry.
MVPART: Program to generate cross section data at arbitrary temperatures.
MVPBURN: Program to perform burnup calculations with MVP.
NTXT2LB: Program to convert the text form of MVP libraries into the binary form.
NLB2TXT: Program to convert the binary form of MVP libraries into the text form.
GMVPLBCV: Program to convert the text form of multigroup cross section data into the binary form.
MVPFAT: Preprocessor for FORTRAN source codes.
|Package ID||Status date||Status|
|NEA-1673/04||21-FEB-2019||Tested at NEADB|
Yasunobu NAGAYA, Keisuke OKUMURA, Takeshi SAKURAI and Takamasa MORI, “MVP/GMVP Version 3: General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations Based on Continuous Energy and Multigroup Methods (Translated document),” JAEA-Data/Code 2016-019 (2017) [in Japanese].
MVP/GMVP runs on various platforms such as standard PCs, Mac or UNIX workstations. Disk space of about 80 Mbytes is required to make the executables from the program sources. A large amount of disk space is required for the cross section data of the MVP code. To install the cross section libraries for MVP (MVP libraries), disk space of about 1.1 Gbytes is required for JENDL-4.0. Doubled disk space is necessary for the 64-bit version of MVP (more than 2 Gbytes of memory is usable).
|Package ID||Computer language|
Yasunobu NAGAYA, Keisuke OKUMURA, Takeshi SAKURAI and Takamasa MORI
Research Group for Reactor Physics and Standard Nuclear Code System
Nuclear Data and Reactor Engineering Division
Nuclear Science and Engineering Center
Japan Atomic Energy Agency
Tokai-mura, Naka-gun, Ibaraki
Keywords: Monte Carlo method, continuous energy, criticality, multigroup, neutron transport, photon transport.