RSICC Home Page RSICC CODE PACKAGE CCC-848

RSICC CODE PACKAGE CCC-855


****Individuals requesting access to the source version of VERA 4.1 must provide the specific code/module that they are developing, the computing systems upon which they are developing the code, and the manner in which they will control access to the VERA package in the end use statement of the request. The individual must provide the name of the person on the VERA development team with whom they are collaborating as well. If this information is not included with requests for access to the source version of VERA 4.1, then the request will be denied. ****


1.† NAME AND TITLE

VERA 4.1: Virtual Environment for Reactor Applications, Version 4.1


AUXILLARY PROGRAMS REQUIRED (Installed using omnus.sh)


CMAKE: https://cmake.org/files/v3.11/cmake-3.11.2.tar.gz

GCC: https://ftp.gnu.org/gnu/gcc/gcc-5.4.0/gcc-5.4.0.tar.gz

MPI: http://www.mpich.org/static/downloads/3.1.3/mpich-3.1.3.tar.gz


AUXILLARY LIBRARIES REQUIRED (Installed using vera_tpls/TPL_build/install_tpls.sh)


LAPACK/BLAS: https://github.com/CASL/vera_tpls.git

BOOST: https://github.com/CASL/vera_tpls.git

ZLIB: https://github.com/CASL/vera_tpls.git

HDF5: https://github.com/CASL/vera_tpls.git

PETSC: https://github.com/CASL/vera_tpls.git

SILO: https://github.com/CASL/vera_tpls.git

QT: https://github.com/CASL/vera_tpls.git


DATA LIBRARIES


mpact51g_71_4.3m2_03262018.fmt:     CASLís MPACT-formatted 51g neutron sub-group library

origen_library/origen_casl2.2.def.bof:     The ORIGEN reaction data from ENDF

origen.rev01.jeff56g:                              The ORIGEN JEFF data to supplement ENDF

origen_data/origen_casl2.2.yld.data:      The ORIGEN fission yield data

ENDF/B-VII.1 data for Shift:

ce_v7.1_endf(.xml)     Description of CE data in cekenolib_7.1 directory

cekenolib_7.1     Directory containing CE data for each nuclide


Additional non-standard cross section and depletion libraries for testing and backwards compatibility are also included.


 

2.† CONTRIBUTORS

The Consortium for the Advanced Simulation of Light Water Reactors (CASL), Oak Ridge, TN, USA


CASL is a U.S. DOE Energy Innovation Hub focused on the modeling and simulation of nuclear reactors. The core partners in CASL include:

Electric Power Research Institute

Idaho National Laboratory

Los Alamos National Laboratory

Massachusetts Institute of Technology

North Carolina State University

Oak Ridge National Laboratory

Sandia National Laboratories

Tennessee Valley Authority

University of Michigan

Westinghouse Electric Company


This release also includes contributions from the following contributing partners:

Core Physics, Inc.

Pennsylvania State University


For more information, see: http://www.casl.gov

 

3.† CODING LANGUAGE AND COMPUTER

FORTRAN, C/C++, Perl, PYTHON; LINUX (C00855PCX8600).

 

4.† NATURE OF PROBLEM SOLVED

Neutronics analysis can be performed for 2D lattices, 2D core and 3D core problems for pressurized water reactor geometries that can be used to calculate criticality and fission rate distributions by pin for input fuel compositions. MPACT uses the Method of Characteristics transport approach for 2D problems. For 3D problems, MPACT uses the 2D/1D method which uses 2D MOC in a radial plane and diffusion or Pn in the axial direction. MPACT includes integrated cross section capabilities that provide problem-specific cross sections generated using the subgroup methodology. The code can be executed both 2D and 3D problems in parallel to reduce overall run time and can be used for eigenvalue, fixed-source, and time-dependent problems.


The ORIGEN (Oak Ridge Isotope GENeration) capability, from SCALE, is used in MPACT to model the depletion, decay, and transmutation of hundreds to thousands of isotopes. It is integrated within MPACT to provide a complete neutronics capability.


A thermal-hydraulics capability is provided with CTF (an updated version of COBRA-TF) that allows thermal-hydraulics analyses for single and multiple assemblies using the simplified VERA common input. This distribution also includes coupled neutronics/thermal-hydraulics capabilities to allow calculations using MPACT coupled with CTF. The integrated MPACT, CTF, and ORIGEN capability is generally referred to as the VERA Core Simulator (VERA-CS). This capability can be used in both quasi-static and transient mode to model reactivity insertion accidents. A simplified fuel model is provided with VERA: an approximate fuel model in CTF that can model the transient fuel temperature distribution.


Shift is a general purpose radiation transport code that performs stochastic modeling of neutral particle physics using the Monte Carlo method. It can perform eigenvalue calculations as well as fixed source calculations in neutron, photon, or coupled neutron-photon mode. Shift is integrated into VERA for both in-core reactor analysis using eigenvalue mode and ex-core dosimetry using fixed-source mode. Shift is coupled to MPACT through VERA to enable source definitions for both fixed-source and eigenvalue problems.


The MAMBA code provides the ability for VERA to simulate the deposition of crud on the fuel rod surface. MAMBA solves the growth of the crud layer as well as the thermal solution, species transport, and chemical precipitation throughout the crud layer. MAMBA is tightly integrated into MPACT and CTF to provide direct feedback into the coupled simulation. MAMBA also includes a detailed mass balance capability which includes the generation of corrosion products from the steam generator and primary system piping, the deposition on core components, and removal from the coolant cleanup system.


Input/output capabilities include the VERA Common Input (VERAIn) script which converts the ASCII common input file to the intermediate XML used to drive all of the physics codes in the VERA. VERA component codes either read the VERA XML format directly or provide a preprocessor that converts the XML into native input for the component code. VERAView is an interactive graphical interface for the visualization and engineering analyses of output data from VERA. The python-based software is easy to install and intuitive to use, and provides instantaneous 2D and 3D images, 1D plots, and alpha-numeric data from VERA multi-physics simulations.


Testing within CASL has focused primarily on Westinghouse four-loop reactor geometries and conditions with example problems included in the distribution.


Physics components included in VERA 4.1:

MPACT: Neutron transport and cross-section physics.

CTF: Sub-channel resolved thermal-hydraulics with fuel rod fuel heat transfer model.

ORIGEN: Isotopic depletion and decay from a beta version of SCALE-6.2.3.

Shift: Monte Carlo neutron transport.

DAKOTA: Software library for Optimization, Uncertainty Quantification, and Sensitivity Analysis.


Infrastructure components included in VERA 4.1:

TriBITS Enhanced CMake based build system.

TRILINOS Software library for the solution of large scale complex numerical problems.

VERAIn VERA common input processor.

VeraShift Utility code for coupling of MPACT and CTF with Shift.


 

5.† METHOD OF SOLUTION

Details of the solution methods can be found in the included documentation and are briefly summarized below: .


MPACT is based on the Method of Characteristics transport approach for 2D problems with cross section weighting based on the subgroup methodology. The code can be executed in parallel to reduce overall run time. For 3D problems, MPACT uses the 2D/1D method which uses 2D MOC in a radial plane and diffusion or Pn in the axial direction. A 51-group library with subgroup parameters is provided. For simulating the time evolution of the reactor under operation, MPACT internally relies upon ORIGEN to solve the nuclide transmutation equations. ORIGEN provides isotopic information about the materials in the reactor as they undergo irradiation..


CTF (an updated version of the COBRA-TF code) is a subchannel thermal-hydraulics code that uses a two-fluid, three-field (i.e. fluid film, fluid drops, and vapor) modeling approach. Both sub-channel and three-dimensional (3D) Cartesian forms of nine conservation equations are available for LWR modeling. CTF includes a wide range of thermal-hydraulic models important to LWR safety analysis including flow-regime-dependent, two-phase wall heat transfer, inter-phase heat transfer and drag, droplet breakup, and quench-front tracking. Due to its 3D capabilities and extensive array of reactor thermal-hydraulic modeling capabilities, CTF has found much use in modeling of LWR rod-bundle transient analysis and Pressurized Water Reactor (PWR) whole-vessel, Loss-Of-Coolant Accident (LOCA) analysis.


MPACT has the ability to call CTF to obtain fuel temperatures and moderator density. This is done by directly calling the CTF solver every outer iteration and passing the power distribution. After CTF converges on a given power shape, the temperatures and densities are passed back to MPACT and applied to the cross-sections. A conditional check on the change in temperature and density is performed to determine if the subgroup calculation needs to be rerun to obtain new shielding parameters for the cross-section generation. This procedure continues until MPACT satisfies its internal convergence criteria on eigenvalue and fission source.


The MAMBA package simulates growth of crud, which refers to metal oxide corrosion products (primarily nickel ferrite, NiFe2O4) on fuel cladding and accumulation of boron in the porous crud. Precipitation of boron compounds, such as lithium tetraborate (Li2B4O7), can lead to a crud-induced power shift (CIPS) in the nuclear fuel. In addition, the crud itself can lead to crud-induced localized corrosion (CILC) due to reduced thermal transport and thus increased temperatures, which can cause mechanical failure of the fuel.


The role of MAMBA is to simulate the buildup of crud and the precipitation of boron rich compounds within the porous crud layer. Since the formation of crud is a fundamentally multiphysics problem, MAMBA is coupled to neutronic and thermal hydraulic solvers present in VERA to predict CIPS. MAMBA requires thermal hydraulics conditions as input, in particular the cladding surface heat flux, turbulent kinetic energy and the coolant temperature. Within VERA thermal hydraulic conditions are provided by CTF. Additionally, the crud source term is modeled in MAMBA, which originates from corrosion of steam generators and primary loop piping.


Shift is a general purpose radiation transport code that performs stochastic modeling of neutral particle physics using the Monte Carlo method. It can perform eigenvalue calculations as well as fixed source calculations in neutron, photon, or coupled neutron-photon mode. Shift is integrated into VERA for both in-core reactor analysis using eigenvalue mode and ex-core dosimetry using fixed-source mode. The main modules of Shift include physics, tallies, geometry, source definitions, parallel decomposition, and variance reduction. Shift is also coupled to internal deterministic discrete-ordinates and simplified PN solvers from the Denovo package. This enables the use of hybrid Monte Carlo methods for variance reduction. Shift is coupled to MPACT through VERA to enable source definitions for both fixed-source and eigenvalue problems.


6.† RESTRICTIONS OR LIMITATIONS

CASL reserves the right to pre-approve distribution of this release to non-CASL partners. Other limitations and known issues with the software are documented in the Release Notes.

 

7.† TYPICAL RUNNING TIME

 

8.† COMPUTER HARDWARE REQUIREMENTS

Linux platforms are supported. 32 cores or greater is recommended.

 

9.† COMPUTER SOFTWARE REQUIREMENTS

Linux based operating system with functioning gcc, g++ and gfortran compilers available, and X11 libraries. Detailed system software and third party library requirements are specified in the VERA Installation Guide. Specific OS versions tested are documented in the Release Notes.

 

10. REFERENCES

References and a list of supplied documentation are provided in the Release Notes.

 

11. CONTENTS OF CODE PACKAGE

The package will be transmitted on a DVD, which includes instructions for source and executable access, sample inputs, test problems, documentation and reference material.

 

12. DATE OF ABSTRACT

May 15, 2020

 

KEYWORDS:REACTOR PHYSICS, RADIATION TRANSPORT, THERMAL HYDRAULICS



RSICC CODE PACKAGE CCC-855



1.† NAME AND TITLE

VERA 4.1-EXE: Virtual Environment for Reactor Applications, Version 4.1 Executables only


AUXILLARY PROGRAMS REQUIRED (Installed using omnus.sh)


CMAKE: https://cmake.org/files/v3.11/cmake-3.11.2.tar.gz

GCC: https://ftp.gnu.org/gnu/gcc/gcc-5.4.0/gcc-5.4.0.tar.gz

MPI: http://www.mpich.org/static/downloads/3.1.3/mpich-3.1.3.tar.gz


AUXILLARY LIBRARIES REQUIRED (Installed using vera_tpls/TPL_build/install_tpls.sh)


LAPACK/BLAS: https://github.com/CASL/vera_tpls.git

BOOST: https://github.com/CASL/vera_tpls.git

ZLIB: https://github.com/CASL/vera_tpls.git

HDF5: https://github.com/CASL/vera_tpls.git

PETSC: https://github.com/CASL/vera_tpls.git

SILO: https://github.com/CASL/vera_tpls.git

QT: https://github.com/CASL/vera_tpls.git


DATA LIBRARIES


mpact51g_71_4.3m2_03262018.fmt:†††† CASLís MPACT-formatted 51g neutron sub-group library

origen_library/origen_casl2.2.def.bof:†††† The ORIGEN reaction data from ENDF

origen.rev01.jeff56g:†††††††† †††††††††††††††††††† The ORIGEN JEFF data to supplement ENDF

origen_data/origen_casl2.2.yld.data: †††† The ORIGEN fission yield data

ENDF/B-VII.1 data for Shift:

ce_v7.1_endf(.xml)†††† Description of CE data in cekenolib_7.1 directory

cekenolib_7.1†††† Directory containing CE data for each nuclide


Additional non-standard cross section and depletion libraries for testing and backwards compatibility are also included.


2.† CONTRIBUTORS

The Consortium for the Advanced Simulation of Light Water Reactors (CASL), Oak Ridge, TN, USA


CASL is a U.S. DOE Energy Innovation Hub focused on the modeling and simulation of nuclear reactors. The core partners in CASL include:

Electric Power Research Institute

Idaho National Laboratory

Los Alamos National Laboratory

Massachusetts Institute of Technology

North Carolina State University

Oak Ridge National Laboratory

Sandia National Laboratories

Tennessee Valley Authority

University of Michigan

Westinghouse Electric Company


This release also includes contributions from the following contributing partners:

Core Physics, Inc.

Pennsylvania State University


For more information, see: http://www.casl.gov

3.† CODING LANGUAGE AND COMPUTER

FORTRAN, C/C++, Perl, PYTHON; LINUX (C00855PCX8601).

4.† NATURE OF PROBLEM SOLVED

Neutronics analysis can be performed for 2D lattices, 2D core and 3D core problems for pressurized water reactor geometries that can be used to calculate criticality and fission rate distributions by pin for input fuel compositions. MPACT uses the Method of Characteristics transport approach for 2D problems. For 3D problems, MPACT uses the 2D/1D method which uses 2D MOC in a radial plane and diffusion or Pn in the axial direction. MPACT includes integrated cross section capabilities that provide problem-specific cross sections generated using the subgroup methodology. The code can be executed both 2D and 3D problems in parallel to reduce overall run time and can be used for eigenvalue, fixed-source, and time-dependent problems.


The ORIGEN (Oak Ridge Isotope GENeration) capability, from SCALE, is used in MPACT to model the depletion, decay, and transmutation of hundreds to thousands of isotopes. It is integrated within MPACT to provide a complete neutronics capability.


A thermal-hydraulics capability is provided with CTF (an updated version of COBRA-TF) that allows thermal-hydraulics analyses for single and multiple assemblies using the simplified VERA common input. This distribution also includes coupled neutronics/thermal-hydraulics capabilities to allow calculations using MPACT coupled with CTF. The integrated MPACT, CTF, and ORIGEN capability is generally referred to as the VERA Core Simulator (VERA-CS). This capability can be used in both quasi-static and transient mode to model reactivity insertion accidents. A simplified fuel model is provided with VERA: an approximate fuel model in CTF that can model the transient fuel temperature distribution.


Shift is a general purpose radiation transport code that performs stochastic modeling of neutral particle physics using the Monte Carlo method. It can perform eigenvalue calculations as well as fixed source calculations in neutron, photon, or coupled neutron-photon mode. Shift is integrated into VERA for both in-core reactor analysis using eigenvalue mode and ex-core dosimetry using fixed-source mode. Shift is coupled to MPACT through VERA to enable source definitions for both fixed-source and eigenvalue problems.


The MAMBA code provides the ability for VERA to simulate the deposition of crud on the fuel rod surface. MAMBA solves the growth of the crud layer as well as the thermal solution, species transport, and chemical precipitation throughout the crud layer. MAMBA is tightly integrated into MPACT and CTF to provide direct feedback into the coupled simulation. MAMBA also includes a detailed mass balance capability which includes the generation of corrosion products from the steam generator and primary system piping, the deposition on core components, and removal from the coolant cleanup system.


Input/output capabilities include the VERA Common Input (VERAIn) script which converts the ASCII common input file to the intermediate XML used to drive all of the physics codes in the VERA. VERA component codes either read the VERA XML format directly or provide a preprocessor that converts the XML into native input for the component code. VERAView is an interactive graphical interface for the visualization and engineering analyses of output data from VERA. The python-based software is easy to install and intuitive to use, and provides instantaneous 2D and 3D images, 1D plots, and alpha-numeric data from VERA multi-physics simulations.


Testing within CASL has focused primarily on Westinghouse four-loop reactor geometries and conditions with example problems included in the distribution.


Physics components included in VERA 4.1-EXE:

MPACT: Neutron transport and cross-section physics.

CTF: Sub-channel resolved thermal-hydraulics with fuel rod fuel heat transfer model.

ORIGEN: Isotopic depletion and decay from a beta version of SCALE-6.2.3.

Shift: Monte Carlo neutron transport.

DAKOTA: Software library for Optimization, Uncertainty Quantification, and Sensitivity Analysis.


Infrastructure components included in VERA 4.1-EXE:

TriBITS Enhanced CMake based build system.

TRILINOS Software library for the solution of large scale complex numerical problems.

VERAIn VERA common input processor.

VeraShift Utility code for coupling of MPACT and CTF with Shift.


5.† METHOD OF SOLUTION

Details of the solution methods can be found in the included documentation and are briefly summarized below: .


MPACT is based on the Method of Characteristics transport approach for 2D problems with cross section weighting based on the subgroup methodology. The code can be executed in parallel to reduce overall run time. For 3D problems, MPACT uses the 2D/1D method which uses 2D MOC in a radial plane and diffusion or Pn in the axial direction. A 51-group library with subgroup parameters is provided. For simulating the time evolution of the reactor under operation, MPACT internally relies upon ORIGEN to solve the nuclide transmutation equations. ORIGEN provides isotopic information about the materials in the reactor as they undergo irradiation..


CTF (an updated version of the COBRA-TF code) is a subchannel thermal-hydraulics code that uses a two-fluid, three-field (i.e. fluid film, fluid drops, and vapor) modeling approach. Both sub-channel and three-dimensional (3D) Cartesian forms of nine conservation equations are available for LWR modeling. CTF includes a wide range of thermal-hydraulic models important to LWR safety analysis including flow-regime-dependent, two-phase wall heat transfer, inter-phase heat transfer and drag, droplet breakup, and quench-front tracking. Due to its 3D capabilities and extensive array of reactor thermal-hydraulic modeling capabilities, CTF has found much use in modeling of LWR rod-bundle transient analysis and Pressurized Water Reactor (PWR) whole-vessel, Loss-Of-Coolant Accident (LOCA) analysis.


MPACT has the ability to call CTF to obtain fuel temperatures and moderator density. This is done by directly calling the CTF solver every outer iteration and passing the power distribution. After CTF converges on a given power shape, the temperatures and densities are passed back to MPACT and applied to the cross-sections. A conditional check on the change in temperature and density is performed to determine if the subgroup calculation needs to be rerun to obtain new shielding parameters for the cross-section generation. This procedure continues until MPACT satisfies its internal convergence criteria on eigenvalue and fission source.


The MAMBA package simulates growth of crud, which refers to metal oxide corrosion products (primarily nickel ferrite, NiFe2O4) on fuel cladding and accumulation of boron in the porous crud. Precipitation of boron compounds, such as lithium tetraborate (Li2B4O7), can lead to a crud-induced power shift (CIPS) in the nuclear fuel. In addition, the crud itself can lead to crud-induced localized corrosion (CILC) due to reduced thermal transport and thus increased temperatures, which can cause mechanical failure of the fuel.


The role of MAMBA is to simulate the buildup of crud and the precipitation of boron rich compounds within the porous crud layer. Since the formation of crud is a fundamentally multiphysics problem, MAMBA is coupled to neutronic and thermal hydraulic solvers present in VERA to predict CIPS. MAMBA requires thermal hydraulics conditions as input, in particular the cladding surface heat flux, turbulent kinetic energy and the coolant temperature. Within VERA thermal hydraulic conditions are provided by CTF. Additionally, the crud source term is modeled in MAMBA, which originates from corrosion of steam generators and primary loop piping.


Shift is a general purpose radiation transport code that performs stochastic modeling of neutral particle physics using the Monte Carlo method. It can perform eigenvalue calculations as well as fixed source calculations in neutron, photon, or coupled neutron-photon mode. Shift is integrated into VERA for both in-core reactor analysis using eigenvalue mode and ex-core dosimetry using fixed-source mode. The main modules of Shift include physics, tallies, geometry, source definitions, parallel decomposition, and variance reduction. Shift is also coupled to internal deterministic discrete-ordinates and simplified PN solvers from the Denovo package. This enables the use of hybrid Monte Carlo methods for variance reduction. Shift is coupled to MPACT through VERA to enable source definitions for both fixed-source and eigenvalue problems.


6.† RESTRICTIONS OR LIMITATIONS

CASL reserves the right to pre-approve distribution of this release to non-CASL partners. Other limitations and known issues with the software are documented in the Release Notes.

7.† TYPICAL RUNNING TIME

8.† COMPUTER HARDWARE REQUIREMENTS

Linux platforms are supported. 32 cores or greater is recommended.

9.† COMPUTER SOFTWARE REQUIREMENTS

Linux based operating system with functioning gcc, g++ and gfortran compilers available, and X11 libraries. Detailed system software and third party library requirements are specified in the VERA Installation Guide. Specific OS versions tested are documented in the Release Notes.

10. REFERENCES

References and a list of supplied documentation are provided in the Release Notes.

11. CONTENTS OF CODE PACKAGE

The package will be transmitted on a DVD, which includes instructions for source and executable access, sample inputs, test problems, documentation and reference material.

12. DATE OF ABSTRACT

May 15, 2020

KEYWORDS:REACTOR PHYSICS, RADIATION TRANSPORT, THERMAL HYDRAULICS