RSICC CODE PACKAGE CCC-739
1. NAME AND TITLE
MVP/GMVP II: General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations based on Continuous Energy and Multigroup Methods, Version 2.
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 text form of multigroup cross section data into binary.
MVPFAT : Preprocessor for FORTRAN source codes.
Reactor Physics Group, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki, Japan, through the OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France.
3. CODING LANGUAGE AND COMPUTER
Fortran 77, Windows PC (C00739MNYCP00). NEADB distributes as NEA-1673/01 (Windows) and NEA-1673/02 (unix/linux).
4. NATURE OF PROBLEM SOLVED
(1) Problems to be solved:
MVP/GMVP II 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:
The various forms of energy-, angle-, space- and time-dependent distribution functions can be specified.
(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-3.3, ENDF/B-VI.8, JEF3.0 etc.) by using the LICEM code. The neutron cross sections in the unresolved resonance region are described by 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.
(12) Other capabilities:
MVP/GMVP has a capability of reactor noise analysis based on simulation of Feynman-alpha experiments.
5. METHOD OF SOLUTION
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(alpha,beta), 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 in 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.
6. RESTRICTIONS OR LIMITATIONS
Dynamic memory allocation is available on all platforms except for Linux Fedora Core 3 and 4 with g77 and gcc compilers. The upper limit of the available memory size is 2GB for any platforms.
7. TYPICAL RUNNING TIME
Time's dependent on size of problem, ranging from a few minutes to days for the ICSBEP.
8. COMPUTER HARDWARE REQUIREMENTS
MVP/GMVP runs on various platforms: UNIX systems (Sun, HP, SGI, IBM, NEC, CRAY, Fujitsu, Hitachi, MIPS) and personal computers.
9. COMPUTER SOFTWARE REQUIREMENTS
The code system runs on Windows PC and Linux systems with Intel/Alpha processors. This package includes pre-compiled codes and binary-form MVP libraries for Windows. The developers compiled under Windows with a variety of compilers: Intel Visual Fortran, Compaq Visual Fortran, Fujitsu Fortran 90 and Lahey Fortran 95. GNU g77 0.5.26 and gcc 2.96 compilers can be used on linux.
a. included in package as PDF files in directory man_MVP:
Y. Nagaya, K. Okumura, T. Mori and M. Nakagawa, "MVP/GMVP II : General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations based on Continuous Energy and Multigroup Methods," [In Japanese and English] JAERI-1348 (2005).
K. Okumura, Y. Nagaya and T. Mori, “MVP-BURN: Burn-up Calculation Code Using a Continuous-energy Monte Carlo Code MVP,” [In Japanese and English] Draft report for JAEA-Data/Code (28 Jan. 2005).
b. background references (not included in package)
T. Mori, K. Okumura and Y. Nagaya, "Development of the MVP Monte Carlo Code at JAERI," Trans. Am. Nucl. Soc., 84, 45 (2001).
T. Mori, K. Okumura and Y. Nagaya, "Status of JAERI's Monte Carlo Code MVP for Neutron and Photon Transport Problems," Monte Carlo 2000 Conference, Lisbon, 23-26 October 2000, Proceeding p.625 (2000).
T. Mori and M. Nakagawa, "MVP/GMVP : General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations based on Continuous Energy and Multigroup Methods," JAERI-Data/Code 94-007 [in Japanese] (1994).
T. Mori, M. Nakagawa and M. Sasaki, "Vectorization of Continuous Energy Monte Carlo Method for Neutron Transport Calculation," J. Nucl. Sci. Technol., Vol. 29, No. 4, pp. 325-336 (1992).
M. Nakagawa, T. Mori and M. Sasaki, "Monte Carlo Calculations on Vector Supercomputers using GMVP," Prog. Nucl. Energy, 24, 183 (1990).
11. CONTENTS OF CODE PACKAGE
The code package includes the referenced documents listed above, Fortran source files, Windows PC executables, unix scripts, and sample cases transmitted on two DVD-ROM Discs.
12. DATE OF ABSTRACT
KEYWORDS: MONTE CARLO; COUPLED; CRITICALITY CALCULATIONS; CROSS SECTIONS; NEUTRON; GAMMA-RAY