An arbitrary volume V with surface area S 16 2.2 Cartesian geometry coordinate system 18 2.3 Doma... more An arbitrary volume V with surface area S 16 2.2 Cartesian geometry coordinate system 18 2.3 Domain discretization with one-cell and two-cell block inversion (slab geometry, LI)) 25 2.4 Eigenvalues as a function of X for the analytic SI and DSA 0 in slab geometry (c1 .0) 35 2.5 Two paths leading from analytic transport to discrete diffusion 41 2.6 High-frequency functions on fine and coarse grids 44 2.7 Schedule of grids 47 2.8 Spatial coordinates for cell V 49 2.9 Cell indices and unknowns for tm >0 and flm >0 in x-y geometry SCB and UCB schemes 55 3.1 Fourier analysis for LD, LLD, SCB and UCB M4S DSA in slab geometry (c1 .0, S16) 68 3.2 Fourier analysis for LD M4S DSA in slab geometry (c1.0, S16) 68 3.3 Fourier analysis for LLD and SCB M4S DSA in slab geometry (c=l.0, S16) 69 3.4 Fourier analysis for UCB M4S DSA in slab geometry (c1.0, S16) 69 3.5 Fourier analysis for asymptotic continuous equation (c1 .0, LxiXy) 76 82 3.6. Cell indices in SCB scheme 3.7 Flow diagram for the multi-level technique 90
The embedded self-shielding method (ESSM) was previously developed to provide an integrated appro... more The embedded self-shielding method (ESSM) was previously developed to provide an integrated approach for lattice calculations. ESSM consists of three interrelated components: (1) processing multigroup cross sections, (2) computing resonance–shielded multigroup cross sections, and (3) performing neutron transport calculations in lattice physics calculations. The self-shielding computation in ESSM is embedded within the transport solution, which provides information for treating heterogeneous self-shielding effects. The resulting shielded cross sections provide data for the transport calculation. This approach integrates the self-shielding and transport components of the lattice physics calculations. Earlier publications have described this aspect of ESSM; however, another equally important aspect of ESSM is that the processing of the multigroup cross section library is integrated with the self-shielding procedure in a consistent manner. This paper describes the procedures that have b...
In this paper, we demonstrate a two step procedure for a pebble bed reactor core analysis. In the... more In this paper, we demonstrate a two step procedure for a pebble bed reactor core analysis. In the first step, we generate equivalent cross sections from a 1-D slab spectral geometry model with the help of the equivalence theory. In the second step, we perform a diffusion calculation by using the equivalent cross sections generated in the first step. A simple 2-D benchmark problem derived from the PMBR-400 reactor was introduced to verify the two step procedure. We compared the two step solutions with the Monte Carlo solutions for the problem and found that the two step solutions agreed well with the Monte Carlo solutions within an acceptable error range. A strong spectral interaction between the core and the reflector has been one of the main concerns in the analysis of pebble bed reactor cores. To resolve this problem, VSOP adopted an iteration between the local spectrum calculation in a spectral zone and the global core calculation (1). In VSOP, the whole problem domain is divided...
The latest research associated with the very high temperature gas-cooled reactor (VHTR) is focuse... more The latest research associated with the very high temperature gas-cooled reactor (VHTR) is focused on the verification of a system performance and safety under operating conditions for the VHTRs. As a part of those, an international gas-cooled reactor program initiated by IAEA is going on. The key objectives of this program are the validation of analytical computer codes and the evaluation of benchmark models for the projected and actual VHTRs. [1] New reactor physics analysis procedure for the prismatic VHTR is under development by adopting the conventional two-step procedure. [2] In this procedure, a few group constants are generated through the transport lattice calculations using the HELIOS [3] code, and the core physics analysis is performed by the 3-dimensional nodal diffusion code MASTER [4]. We evaluated the performance of the HELIOS/MASTER code package through the benchmark calculations related to the GT-MHR (Gas Turbine-Modular Helium Reactor) to dispose weapon plutonium. ...
In fiscal year 2018, the US Nuclear Regulatory Commission (NRC) expressed an interest in using th... more In fiscal year 2018, the US Nuclear Regulatory Commission (NRC) expressed an interest in using the US Department of Energy (DOE) Office of Nuclear Energy (NE) advanced modeling and simulation tools to evaluate advanced fuel concepts such as accident-tolerant fuel (ATF). This interest evolved into formal cooperation between DOE and NRC to ensure that advanced modeling and simulation (M&S) capability is available to the NRC for the analyses of ATF concepts. The Consortium for Advanced Simulation of Light Water Reactors (CASL) has developed, applied, and deployed advanced modeling and simulation capabilities to enhance the operational performance, efficiency, and safety of light water reactors (LWRs). As a result of this cooperative effort between DOE and NRC, potential ATF concepts of interest to the industry were identified and simulated with CASL's Virtual Environment for Reactor Applications (VERA) for 2D pressurized water reactor (PWR) 17 × 17 lattices. Four ATF concepts were identified by CASL's Westinghouse collaborators, and the benchmark specifications for these ATF concepts were generated as presented in this report. These benchmark problems were set up to test the ability of VERA to simulate these ATF concepts, and the results generated by VERA were compared against two Monte Carlo codes: (1) Serpent, which was developed at VTT Technical Research Centre of Finland Ltd, and (2) Shift, which was developed at Oak Ridge National Laboratory (ORNL). The depletion parameters and flags used to run these models differ between all three codes, and there were differences seen between VERA, Shift, and Serpent, but overall, the agreement shown was considered sufficient to progress to core modeling and evaluation of these ATF concepts using VERA. The differences identified in this document require further investigation by modeling single fuel pin depletion to compare the isotopic evolution of the fuel during a depletion calculation. Prior depletion benchmarking efforts between VERA and Shift have shown closer agreement for UO2, so recent changes to the code and the data should be investigated to identify the cause of these differences.
The evaluated nuclear data file (ENDF)/B-7.0 v4.1m3 MPACT 47-group library has been used as a mai... more The evaluated nuclear data file (ENDF)/B-7.0 v4.1m3 MPACT 47-group library has been used as a main library for the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronics simulator in simulating pressurized water reactor (PWR) problems. Recent analysis for the high void boiling water reactor (BWR) fuels and burnt fuels indicates that the 47-group library introduces relatively large reactivity bias. Since the 47group structure does not match with the SCALE 6.2 252-group boundaries, the CASL Virtual Environment for Reactor Applications Core Simulator (VERA-CS) MPACT library must be maintained independently, which causes quality assurance concerns. In order to address this issue, a new 51-group structure has been proposed based on the MPACT 47g and SCALE 252-g structures. In addition, the new CASL library will include a 19-group structure for gamma production and interaction cross section data based on the SCALE 19group structure. New AMPX and MPACT 51-group libraries have been developed with the ENDF/B-7.0 and 7.1 evaluated nuclear data. The 19-group gamma data also have been generated for future use, but they are only available on the AMPX 51-g library. In addition, ENDF/B-7.0 and 7.1 MPACT 252-g libraries have been generated for verification purposes. Various benchmark calculations have been performed to verify and validate the newly developed libraries. Generation of the V4.2m5 AMPX and MPACT 51 and 252-Group Libraries Consortium for Advanced Simulation of LWRs iv CASL-U-2016-1177-000 Rev.0 TABLE OF CONTENTS REVISION LOG ii EXECUTIVE SUMMARY iii TABLE OF CONTENTS iv ACRONYMS vii 1. INTRODUCTION 2. PROGRAMS AND DATA 2.1 Programs, data and computer to generate the AMPX MG library 2.2 Programs and computer to generate the MPACT MG library 3. GENERATION OF THE 51(n)/252(n)/19()-GROUP AMPX LIBRARY 3.1 Procedure 3.2 The 51(n)/252(n)/19()-group AMPX master library 3.2.1 Generation of 51-g and 252-g cross sections and Bondarenko data 3.2.2 Generation of the first AMPX 51-g and 252-g libraries 3.2.3 Generation of IR parameters and homogeneous F-factors 3.2.4 Generation of the second AMPX 51-g and 252-g library 3.2.5 Generation of heterogeneous F-factors 3.2.6 Generation of the third AMPX 51-g and 252-g library 4. GENERATION OF THE MPACT 51-G AND 252-G LIBRARY 4.1 Subgroup data generation 4.2 Generation of transport correction factors for 1 H 4.3 Generation of 238 U resonance data with and without epithermal upscattering 4.4 Subgroup data generation with the 238 U SPH factors 4.5 Generation of the 51-group MPACT libraries 5. BENCHMARK CALCULATION 5.1 The MPACT 51-g library results 5.2 The MPACT 252-g library results 6. DISCUSSION AND CONCLUSION REFERENCE APPENDIX A.1 Input files to generate MG XSs, Bondarenko data and Gamma XSs for 235 U APPENDIX A.2 Input file to generate the first AMPX MG library APPENDIX A.3 Input files to generate IR parameters and homogeneous F-factors for 235 U APPENDIX A.4 Input file to generate the second AMPX MG library APPENDIX A.5 Input files to generate heterogeneous F-factors for 235 U APPENDIX A.6 Input file to generate the third AMPX 51-g library APPENDIX B.1 The FF2RI and SUBGR input file and standard subgroup level file APPENDIX B.2 The CE-KENO and MERIT input files and editing program APPENDIX B.3 The DECLIB input file to generate the ENDF/B-7.0 MPACT 51-g library APPENDIX B.4 The DECLIB input file to generate the ENDF/B-7.1 MPACT 51-g library APPENDIX C.1 Benchmark problems for the given atomic number densities at various burnup points Generation of the V4.2m5 AMPX and MPACT 51 and 252-Group Libraries CASL-U-2016-1177-000 Rev.0 v Consortium for Advanced Simulation of LWRs Generation of the V4.2m5 AMPX and MPACT 51 and 252-Group Libraries Consortium for Advanced Simulation of LWRs vi CASL-U-2016-1177-000 Rev.0 Generation of the V4.2m5 AMPX and MPACT 51 and 252-Group Libraries CASL-U-2016-1177-000 Rev.0 1 Consortium for Advanced Simulation of LWRs This document includes a detailed procedure and information to generate the AMPX and MPACT 51-group libraries with ENDF/B-7.0 and 7.1. Generation of the V4.2m5 AMPX and MPACT 51 and 252-Group Libraries Consortium for Advanced Simulation of LWRs 2 CASL-U-2016-1177-000 Rev.0 Generation of the V4.2m5 AMPX and MPACT 51 and 252-Group Libraries Consortium for Advanced Simulation of LWRs 4 CASL-U-2016-1177-000 Rev.0 Generation of the V4.2m5 AMPX and MPACT 51 and 252-Group Libraries Consortium for Advanced Simulation of LWRs 6 CASL-U-2016-1177-000 Rev.0 Table 3.1 Neutron 51-Group and Gamma 19-Group Structures Neutron 47-g Neutron 51-g Gamma 19-g Group Upper bound Group Upper bound Group Upper bound 33 9
The default energy deposition model in the CASL neutronics code MPACT assumes all fission energy ... more The default energy deposition model in the CASL neutronics code MPACT assumes all fission energy is deposited locally in fuel rods. Furthermore, equilibrium delayed energy release is assumed for both steady-state and transient calculations. These approximations limit the accurate representation of the heat generation distribution in space and its variations over time, which are essential for power distribution and thermal-hydraulic coupling calculations. In this paper, an improved energy deposition model is presented in both the spatial and time domains. Spatially, the energy deposition through fission, neutron capture, and slowing-down reactions are explicitly modeled to account for the heat generation from all regions of a reactor core, and a gamma smearing scheme is developed that utilizes the gamma sources from neutron fission and capture. In the time domain, the delayed energy release is modeled by solving an additional equation of delayed heat emitters, similar to the equation...
Notice: This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 w... more Notice: This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the US Department of Energy (DOE). The US government retains and the publisher, by accepting the article for publication, acknowledges that the US government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for US government purposes. DOE will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (http://energy.gov/downloads/doe-public-access-plan). An In-scattering Transport Correction by the Neutron Leakage Conservation Method and Its Application to the DeCART Multigroup Library
New nuclear design procedure is under development for the reactor physics analysis of the very hi... more New nuclear design procedure is under development for the reactor physics analysis of the very high temperature gas-cooled reactor (VHTR). The conventional two-step procedure developed for the commercial pressurized water reactors (PWR) is adopted as a standard procedure for the prismatic and pebble type VHTR reactor physics analysis. We employed HELIOS [1] code for the transport lattice calculation to generate few group constants, and MASTER [2] code for the 3-D core calculation to perform the reactor physics analysis. The neutronic characteristics of VHTR is quite different from the PWR one in many aspects. VHTR employs a graphite moderator which results in long neutron diffusion length. A particulate fuel with multicoating layers, called TRISO, is employed to achieve a high fuel performance and fission gas confinement, which is randomly dispersed in a graphite matrix. This causes a so-called double heterogeneity problem in the lattice calculation requiring a special treatment. Th...
As an effort to resolve the whole-core flat flux assumption introduced in the formulation of the ... more As an effort to resolve the whole-core flat flux assumption introduced in the formulation of the slowing down fixed source problem appearing in the application of the subgroup method for resonance treatment in heterogeneous systems, a new definition of the equivalence cross section is introduced which establishes equivalence between a heterogeneous and a homogeneous system in terms of equal sensitivity of reaction rate on the perturbation in the resonance cross section. The derivation to obtain the heterogeneous sensitivity coefficient is carried out through the use of the sensitivity theory to yield a fixed source problem which is different only in the right hand side source term from that of the conventional formulation. This derivation guarantees positive equivalence cross sections unlike the conventional formulation. The new approach is evaluated by employing an analytic Pn solver for a simple one dimensional two-region problem consisting of fuel and moderator. The results indic...
Energy. The United States Government retains and the publisher, by accepting the article for publ... more Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. Preliminary Assessment of Resonance Interference Consideration by Using 0-D Slowing Down Calculation in the Embedded Self-Shielding Method*
In VHTRs, a particulate fuel with multi-coating layers, which is called TRISO, is employed to ach... more In VHTRs, a particulate fuel with multi-coating layers, which is called TRISO, is employed to achieve a high fuel performance. The TRISO fuels are utilized in two fuel types, either a cylindrical compact or a spherical pebble. In both fuel types, TRISO particles are randomly dispersed in a graphite matrix with a relatively low volume fraction. This special fuel configuration leads to the so-called double-heterogeneity problem, which requires special computational methodology. Currently, most conventional lattice codes cannot handle the doubleheterogeneity media and only a few computer code systems such as DRAGON[1] are applicable to the problem, with a limited accuracy. Furthermore, a fullscale application of the Monte Carlo method to the double-heterogeneous problem is also very challenging due to the huge number of TRISO fuels randomly distributed. In this paper, a novel method is proposed to eliminate the double-heterogeneity, which makes the conventional lattice codes applicable...
The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CA... more The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3-D whole core transport code being developed for the CASL toolset, Virtual Environment for Reactor Analysis (VERA). The v4.1m3 MPACT 47-group library with ENDF/BVII.0, which was developed previously by using Oak Ridge National Laboratory AMPX/SCALE code packages, includes deficiencies, especially for burnt fuel cases. New v4.2m5 MPACT 51-group libraries with ENDF/B-VII.0 and VII.1 have been developed to have better accuracy by improving the library generation methodology. This study discusses a detailed procedure to generate the MPACT 51-group libraries and results for various benchmark problems.
This research focuses on the development of a new fast and robust lattice physics methodology for... more This research focuses on the development of a new fast and robust lattice physics methodology for modeling and simulation of doubly heterogeneous (DH) nuclear fuels containing tristructural isotropic (TRISO) coated fuel particles and fully ceramic microencapsulated (FCM) fuel. The specific aims of this study are to determine if existing complex methods for DH calculations can be integrated with the Embedded Self-Shielding Method (ESSM) developed at Oak Ridge National Laboratory ORNL. New ESSM methodologies for the DH resonance self-shielding calculation have been developed using the Sanchez and Pomraning’s method. These methodologies have been incorporated into a transport lattice code based on method of characteristics (MOC) for spatial discretization. Benchmark problems including reference solutions have been developed to validate new DH lattice physics methodology. Benchmark results show that new DH capability for resonance self-shielding and eigenvalue calculations is working re...
A new algorithm has been developed to automatically optimize the coarse energy group structure fo... more A new algorithm has been developed to automatically optimize the coarse energy group structure for the SCALE multigroup procedures based on either a pointwise slowing down calculation or intermediate resonance approximation for resonance self-shielding. A coarse group structure will be determined to minimize reactivity differences between the fine and coarse group calculations for several variations of states. A new 56-group structure has been developed for the pressurized water reactor and boiling water reactor fuels by using the algorithm with the SCALE 252-group structure, which could be used in the regular SCALE and intermediate resonance approaches. The computational results for benchmark problems show that the new 56-group structure developed by using coarse group optimization is reasonable.
In general the two dimensional discrete ordinates transport code DORT has been used for an evalua... more In general the two dimensional discrete ordinates transport code DORT has been used for an evaluation of neutron and gamma fluxes during a shielding design of nuclear reactors. It is very complicated and it takes too much time for shielding designers to prepare input data such as a geometrical modeling and a source distribution and to process an output of the results from the shielding analysis. The GEOSHIELD code was developed to save the time spent preparing a geometrical model and an output processing. The GEOSHIELD code is composed of a module for a geometrical modeling by using a combinatorial geometry, a module for a fixed source redistribution, a module for a DORT processing, and a module for a graphical processing of the output activities. The evaluation of an irradiation of a fast neutron which has an energy of higher than 1.0 MeV is very important to verify the integrity of an internal structure including a pressure vessel. The GEOSHIELD code was applied to evaluate a fast neutron fluence distribution on the internal structures inside the reactor pressure vessel of the SMART reactor and the MCNP was used for verification of the result from the GEOSHIELD calculation. Result of the GEOSHIELD and MCNP showed good agreement each other.
An arbitrary volume V with surface area S 16 2.2 Cartesian geometry coordinate system 18 2.3 Doma... more An arbitrary volume V with surface area S 16 2.2 Cartesian geometry coordinate system 18 2.3 Domain discretization with one-cell and two-cell block inversion (slab geometry, LI)) 25 2.4 Eigenvalues as a function of X for the analytic SI and DSA 0 in slab geometry (c1 .0) 35 2.5 Two paths leading from analytic transport to discrete diffusion 41 2.6 High-frequency functions on fine and coarse grids 44 2.7 Schedule of grids 47 2.8 Spatial coordinates for cell V 49 2.9 Cell indices and unknowns for tm >0 and flm >0 in x-y geometry SCB and UCB schemes 55 3.1 Fourier analysis for LD, LLD, SCB and UCB M4S DSA in slab geometry (c1 .0, S16) 68 3.2 Fourier analysis for LD M4S DSA in slab geometry (c1.0, S16) 68 3.3 Fourier analysis for LLD and SCB M4S DSA in slab geometry (c=l.0, S16) 69 3.4 Fourier analysis for UCB M4S DSA in slab geometry (c1.0, S16) 69 3.5 Fourier analysis for asymptotic continuous equation (c1 .0, LxiXy) 76 82 3.6. Cell indices in SCB scheme 3.7 Flow diagram for the multi-level technique 90
The embedded self-shielding method (ESSM) was previously developed to provide an integrated appro... more The embedded self-shielding method (ESSM) was previously developed to provide an integrated approach for lattice calculations. ESSM consists of three interrelated components: (1) processing multigroup cross sections, (2) computing resonance–shielded multigroup cross sections, and (3) performing neutron transport calculations in lattice physics calculations. The self-shielding computation in ESSM is embedded within the transport solution, which provides information for treating heterogeneous self-shielding effects. The resulting shielded cross sections provide data for the transport calculation. This approach integrates the self-shielding and transport components of the lattice physics calculations. Earlier publications have described this aspect of ESSM; however, another equally important aspect of ESSM is that the processing of the multigroup cross section library is integrated with the self-shielding procedure in a consistent manner. This paper describes the procedures that have b...
In this paper, we demonstrate a two step procedure for a pebble bed reactor core analysis. In the... more In this paper, we demonstrate a two step procedure for a pebble bed reactor core analysis. In the first step, we generate equivalent cross sections from a 1-D slab spectral geometry model with the help of the equivalence theory. In the second step, we perform a diffusion calculation by using the equivalent cross sections generated in the first step. A simple 2-D benchmark problem derived from the PMBR-400 reactor was introduced to verify the two step procedure. We compared the two step solutions with the Monte Carlo solutions for the problem and found that the two step solutions agreed well with the Monte Carlo solutions within an acceptable error range. A strong spectral interaction between the core and the reflector has been one of the main concerns in the analysis of pebble bed reactor cores. To resolve this problem, VSOP adopted an iteration between the local spectrum calculation in a spectral zone and the global core calculation (1). In VSOP, the whole problem domain is divided...
The latest research associated with the very high temperature gas-cooled reactor (VHTR) is focuse... more The latest research associated with the very high temperature gas-cooled reactor (VHTR) is focused on the verification of a system performance and safety under operating conditions for the VHTRs. As a part of those, an international gas-cooled reactor program initiated by IAEA is going on. The key objectives of this program are the validation of analytical computer codes and the evaluation of benchmark models for the projected and actual VHTRs. [1] New reactor physics analysis procedure for the prismatic VHTR is under development by adopting the conventional two-step procedure. [2] In this procedure, a few group constants are generated through the transport lattice calculations using the HELIOS [3] code, and the core physics analysis is performed by the 3-dimensional nodal diffusion code MASTER [4]. We evaluated the performance of the HELIOS/MASTER code package through the benchmark calculations related to the GT-MHR (Gas Turbine-Modular Helium Reactor) to dispose weapon plutonium. ...
In fiscal year 2018, the US Nuclear Regulatory Commission (NRC) expressed an interest in using th... more In fiscal year 2018, the US Nuclear Regulatory Commission (NRC) expressed an interest in using the US Department of Energy (DOE) Office of Nuclear Energy (NE) advanced modeling and simulation tools to evaluate advanced fuel concepts such as accident-tolerant fuel (ATF). This interest evolved into formal cooperation between DOE and NRC to ensure that advanced modeling and simulation (M&S) capability is available to the NRC for the analyses of ATF concepts. The Consortium for Advanced Simulation of Light Water Reactors (CASL) has developed, applied, and deployed advanced modeling and simulation capabilities to enhance the operational performance, efficiency, and safety of light water reactors (LWRs). As a result of this cooperative effort between DOE and NRC, potential ATF concepts of interest to the industry were identified and simulated with CASL's Virtual Environment for Reactor Applications (VERA) for 2D pressurized water reactor (PWR) 17 × 17 lattices. Four ATF concepts were identified by CASL's Westinghouse collaborators, and the benchmark specifications for these ATF concepts were generated as presented in this report. These benchmark problems were set up to test the ability of VERA to simulate these ATF concepts, and the results generated by VERA were compared against two Monte Carlo codes: (1) Serpent, which was developed at VTT Technical Research Centre of Finland Ltd, and (2) Shift, which was developed at Oak Ridge National Laboratory (ORNL). The depletion parameters and flags used to run these models differ between all three codes, and there were differences seen between VERA, Shift, and Serpent, but overall, the agreement shown was considered sufficient to progress to core modeling and evaluation of these ATF concepts using VERA. The differences identified in this document require further investigation by modeling single fuel pin depletion to compare the isotopic evolution of the fuel during a depletion calculation. Prior depletion benchmarking efforts between VERA and Shift have shown closer agreement for UO2, so recent changes to the code and the data should be investigated to identify the cause of these differences.
The evaluated nuclear data file (ENDF)/B-7.0 v4.1m3 MPACT 47-group library has been used as a mai... more The evaluated nuclear data file (ENDF)/B-7.0 v4.1m3 MPACT 47-group library has been used as a main library for the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronics simulator in simulating pressurized water reactor (PWR) problems. Recent analysis for the high void boiling water reactor (BWR) fuels and burnt fuels indicates that the 47-group library introduces relatively large reactivity bias. Since the 47group structure does not match with the SCALE 6.2 252-group boundaries, the CASL Virtual Environment for Reactor Applications Core Simulator (VERA-CS) MPACT library must be maintained independently, which causes quality assurance concerns. In order to address this issue, a new 51-group structure has been proposed based on the MPACT 47g and SCALE 252-g structures. In addition, the new CASL library will include a 19-group structure for gamma production and interaction cross section data based on the SCALE 19group structure. New AMPX and MPACT 51-group libraries have been developed with the ENDF/B-7.0 and 7.1 evaluated nuclear data. The 19-group gamma data also have been generated for future use, but they are only available on the AMPX 51-g library. In addition, ENDF/B-7.0 and 7.1 MPACT 252-g libraries have been generated for verification purposes. Various benchmark calculations have been performed to verify and validate the newly developed libraries. Generation of the V4.2m5 AMPX and MPACT 51 and 252-Group Libraries Consortium for Advanced Simulation of LWRs iv CASL-U-2016-1177-000 Rev.0 TABLE OF CONTENTS REVISION LOG ii EXECUTIVE SUMMARY iii TABLE OF CONTENTS iv ACRONYMS vii 1. INTRODUCTION 2. PROGRAMS AND DATA 2.1 Programs, data and computer to generate the AMPX MG library 2.2 Programs and computer to generate the MPACT MG library 3. GENERATION OF THE 51(n)/252(n)/19()-GROUP AMPX LIBRARY 3.1 Procedure 3.2 The 51(n)/252(n)/19()-group AMPX master library 3.2.1 Generation of 51-g and 252-g cross sections and Bondarenko data 3.2.2 Generation of the first AMPX 51-g and 252-g libraries 3.2.3 Generation of IR parameters and homogeneous F-factors 3.2.4 Generation of the second AMPX 51-g and 252-g library 3.2.5 Generation of heterogeneous F-factors 3.2.6 Generation of the third AMPX 51-g and 252-g library 4. GENERATION OF THE MPACT 51-G AND 252-G LIBRARY 4.1 Subgroup data generation 4.2 Generation of transport correction factors for 1 H 4.3 Generation of 238 U resonance data with and without epithermal upscattering 4.4 Subgroup data generation with the 238 U SPH factors 4.5 Generation of the 51-group MPACT libraries 5. BENCHMARK CALCULATION 5.1 The MPACT 51-g library results 5.2 The MPACT 252-g library results 6. DISCUSSION AND CONCLUSION REFERENCE APPENDIX A.1 Input files to generate MG XSs, Bondarenko data and Gamma XSs for 235 U APPENDIX A.2 Input file to generate the first AMPX MG library APPENDIX A.3 Input files to generate IR parameters and homogeneous F-factors for 235 U APPENDIX A.4 Input file to generate the second AMPX MG library APPENDIX A.5 Input files to generate heterogeneous F-factors for 235 U APPENDIX A.6 Input file to generate the third AMPX 51-g library APPENDIX B.1 The FF2RI and SUBGR input file and standard subgroup level file APPENDIX B.2 The CE-KENO and MERIT input files and editing program APPENDIX B.3 The DECLIB input file to generate the ENDF/B-7.0 MPACT 51-g library APPENDIX B.4 The DECLIB input file to generate the ENDF/B-7.1 MPACT 51-g library APPENDIX C.1 Benchmark problems for the given atomic number densities at various burnup points Generation of the V4.2m5 AMPX and MPACT 51 and 252-Group Libraries CASL-U-2016-1177-000 Rev.0 v Consortium for Advanced Simulation of LWRs Generation of the V4.2m5 AMPX and MPACT 51 and 252-Group Libraries Consortium for Advanced Simulation of LWRs vi CASL-U-2016-1177-000 Rev.0 Generation of the V4.2m5 AMPX and MPACT 51 and 252-Group Libraries CASL-U-2016-1177-000 Rev.0 1 Consortium for Advanced Simulation of LWRs This document includes a detailed procedure and information to generate the AMPX and MPACT 51-group libraries with ENDF/B-7.0 and 7.1. Generation of the V4.2m5 AMPX and MPACT 51 and 252-Group Libraries Consortium for Advanced Simulation of LWRs 2 CASL-U-2016-1177-000 Rev.0 Generation of the V4.2m5 AMPX and MPACT 51 and 252-Group Libraries Consortium for Advanced Simulation of LWRs 4 CASL-U-2016-1177-000 Rev.0 Generation of the V4.2m5 AMPX and MPACT 51 and 252-Group Libraries Consortium for Advanced Simulation of LWRs 6 CASL-U-2016-1177-000 Rev.0 Table 3.1 Neutron 51-Group and Gamma 19-Group Structures Neutron 47-g Neutron 51-g Gamma 19-g Group Upper bound Group Upper bound Group Upper bound 33 9
The default energy deposition model in the CASL neutronics code MPACT assumes all fission energy ... more The default energy deposition model in the CASL neutronics code MPACT assumes all fission energy is deposited locally in fuel rods. Furthermore, equilibrium delayed energy release is assumed for both steady-state and transient calculations. These approximations limit the accurate representation of the heat generation distribution in space and its variations over time, which are essential for power distribution and thermal-hydraulic coupling calculations. In this paper, an improved energy deposition model is presented in both the spatial and time domains. Spatially, the energy deposition through fission, neutron capture, and slowing-down reactions are explicitly modeled to account for the heat generation from all regions of a reactor core, and a gamma smearing scheme is developed that utilizes the gamma sources from neutron fission and capture. In the time domain, the delayed energy release is modeled by solving an additional equation of delayed heat emitters, similar to the equation...
Notice: This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 w... more Notice: This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the US Department of Energy (DOE). The US government retains and the publisher, by accepting the article for publication, acknowledges that the US government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for US government purposes. DOE will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (http://energy.gov/downloads/doe-public-access-plan). An In-scattering Transport Correction by the Neutron Leakage Conservation Method and Its Application to the DeCART Multigroup Library
New nuclear design procedure is under development for the reactor physics analysis of the very hi... more New nuclear design procedure is under development for the reactor physics analysis of the very high temperature gas-cooled reactor (VHTR). The conventional two-step procedure developed for the commercial pressurized water reactors (PWR) is adopted as a standard procedure for the prismatic and pebble type VHTR reactor physics analysis. We employed HELIOS [1] code for the transport lattice calculation to generate few group constants, and MASTER [2] code for the 3-D core calculation to perform the reactor physics analysis. The neutronic characteristics of VHTR is quite different from the PWR one in many aspects. VHTR employs a graphite moderator which results in long neutron diffusion length. A particulate fuel with multicoating layers, called TRISO, is employed to achieve a high fuel performance and fission gas confinement, which is randomly dispersed in a graphite matrix. This causes a so-called double heterogeneity problem in the lattice calculation requiring a special treatment. Th...
As an effort to resolve the whole-core flat flux assumption introduced in the formulation of the ... more As an effort to resolve the whole-core flat flux assumption introduced in the formulation of the slowing down fixed source problem appearing in the application of the subgroup method for resonance treatment in heterogeneous systems, a new definition of the equivalence cross section is introduced which establishes equivalence between a heterogeneous and a homogeneous system in terms of equal sensitivity of reaction rate on the perturbation in the resonance cross section. The derivation to obtain the heterogeneous sensitivity coefficient is carried out through the use of the sensitivity theory to yield a fixed source problem which is different only in the right hand side source term from that of the conventional formulation. This derivation guarantees positive equivalence cross sections unlike the conventional formulation. The new approach is evaluated by employing an analytic Pn solver for a simple one dimensional two-region problem consisting of fuel and moderator. The results indic...
Energy. The United States Government retains and the publisher, by accepting the article for publ... more Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. Preliminary Assessment of Resonance Interference Consideration by Using 0-D Slowing Down Calculation in the Embedded Self-Shielding Method*
In VHTRs, a particulate fuel with multi-coating layers, which is called TRISO, is employed to ach... more In VHTRs, a particulate fuel with multi-coating layers, which is called TRISO, is employed to achieve a high fuel performance. The TRISO fuels are utilized in two fuel types, either a cylindrical compact or a spherical pebble. In both fuel types, TRISO particles are randomly dispersed in a graphite matrix with a relatively low volume fraction. This special fuel configuration leads to the so-called double-heterogeneity problem, which requires special computational methodology. Currently, most conventional lattice codes cannot handle the doubleheterogeneity media and only a few computer code systems such as DRAGON[1] are applicable to the problem, with a limited accuracy. Furthermore, a fullscale application of the Monte Carlo method to the double-heterogeneous problem is also very challenging due to the huge number of TRISO fuels randomly distributed. In this paper, a novel method is proposed to eliminate the double-heterogeneity, which makes the conventional lattice codes applicable...
The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CA... more The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3-D whole core transport code being developed for the CASL toolset, Virtual Environment for Reactor Analysis (VERA). The v4.1m3 MPACT 47-group library with ENDF/BVII.0, which was developed previously by using Oak Ridge National Laboratory AMPX/SCALE code packages, includes deficiencies, especially for burnt fuel cases. New v4.2m5 MPACT 51-group libraries with ENDF/B-VII.0 and VII.1 have been developed to have better accuracy by improving the library generation methodology. This study discusses a detailed procedure to generate the MPACT 51-group libraries and results for various benchmark problems.
This research focuses on the development of a new fast and robust lattice physics methodology for... more This research focuses on the development of a new fast and robust lattice physics methodology for modeling and simulation of doubly heterogeneous (DH) nuclear fuels containing tristructural isotropic (TRISO) coated fuel particles and fully ceramic microencapsulated (FCM) fuel. The specific aims of this study are to determine if existing complex methods for DH calculations can be integrated with the Embedded Self-Shielding Method (ESSM) developed at Oak Ridge National Laboratory ORNL. New ESSM methodologies for the DH resonance self-shielding calculation have been developed using the Sanchez and Pomraning’s method. These methodologies have been incorporated into a transport lattice code based on method of characteristics (MOC) for spatial discretization. Benchmark problems including reference solutions have been developed to validate new DH lattice physics methodology. Benchmark results show that new DH capability for resonance self-shielding and eigenvalue calculations is working re...
A new algorithm has been developed to automatically optimize the coarse energy group structure fo... more A new algorithm has been developed to automatically optimize the coarse energy group structure for the SCALE multigroup procedures based on either a pointwise slowing down calculation or intermediate resonance approximation for resonance self-shielding. A coarse group structure will be determined to minimize reactivity differences between the fine and coarse group calculations for several variations of states. A new 56-group structure has been developed for the pressurized water reactor and boiling water reactor fuels by using the algorithm with the SCALE 252-group structure, which could be used in the regular SCALE and intermediate resonance approaches. The computational results for benchmark problems show that the new 56-group structure developed by using coarse group optimization is reasonable.
In general the two dimensional discrete ordinates transport code DORT has been used for an evalua... more In general the two dimensional discrete ordinates transport code DORT has been used for an evaluation of neutron and gamma fluxes during a shielding design of nuclear reactors. It is very complicated and it takes too much time for shielding designers to prepare input data such as a geometrical modeling and a source distribution and to process an output of the results from the shielding analysis. The GEOSHIELD code was developed to save the time spent preparing a geometrical model and an output processing. The GEOSHIELD code is composed of a module for a geometrical modeling by using a combinatorial geometry, a module for a fixed source redistribution, a module for a DORT processing, and a module for a graphical processing of the output activities. The evaluation of an irradiation of a fast neutron which has an energy of higher than 1.0 MeV is very important to verify the integrity of an internal structure including a pressure vessel. The GEOSHIELD code was applied to evaluate a fast neutron fluence distribution on the internal structures inside the reactor pressure vessel of the SMART reactor and the MCNP was used for verification of the result from the GEOSHIELD calculation. Result of the GEOSHIELD and MCNP showed good agreement each other.
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