Loading EBR-II/README.md +20 −21 Original line number Diff line number Diff line Loading @@ -6,35 +6,34 @@ The SCALE full core model of the EBR-II was developed based on the benchmark spe Experiments of the Organisation for Economic Co-operation and Development (OECD), Nuclear Energy Agency (NEA): > E. S. Lum, C. K. Pope, R. Stewart, B. Byambadorj, Q. Beaulieu (2018), > _Evaluation of Run 138B at Experimental Breeder Reactor II, a prototypic > liquid metal reactor,_ EBR2-LMFR-RESR-001, NEA/NSC/DOC[2006]1, Rev. 0, OECD/NEA. * E. S. Lum, C. K. Pope, R. Stewart, B. Byambadorj, Q. Beaulieu (2018), _Evaluation of Run 138B at Experimental Breeder Reactor II, a prototypic liquid metal reactor,_ EBR2-LMFR-RESR-001, NEA/NSC/DOC[2006]1, Rev. 0, OECD/NEA. ## Model citation Cite the following references when using this model: > [1] F. Bostelmann, G. Ilas, W. Wieselquist (2021), _Nuclear Data Sensitivity > Study for the EBR-II Fast Reactor Benchmark Using SCALE with ENDF/B-VII.1 > and ENDF/B-VIII.0,_ Journal of Nuclear Engineering, 2, 345-367, > doi:[10.3390/jne2040028](https://www.mdpi.com/2673-4362/2/4/28). > > [2] F. Bostelmann, G. Ilas, C. Celik, A. M. Holcomb, and W. Wieselquist (2021), > _Nuclear Data Assessment for Advanced Reactors,_ > [NUREG/CR-7289, ORNL/TM-2021/2002](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7289/index.html), > Oak Ridge National Laboratory, Oak Ridge, TN. * F. Bostelmann, G. Ilas, W. Wieselquist (2021), _Nuclear Data Sensitivity Study for the EBR-II Fast Reactor Benchmark Using SCALE with ENDF/B-VII.1 and ENDF/B-VIII.0,_ Journal of Nuclear Engineering, 2, 345-367, doi:[10.3390/jne2040028](https://www.mdpi.com/2673-4362/2/4/28). * F. Bostelmann, G. Ilas, C. Celik, A. M. Holcomb, and W. Wieselquist (2021), _Nuclear Data Assessment for Advanced Reactors,_ [NUREG/CR-7289, ORNL/TM-2021/2002](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7289/index.html), Oak Ridge National Laboratory, Oak Ridge, TN. ## Content Input and output files for EBR-II full core sensitivity and uncertainty calculations with the TSUNAMI-Shift sequence are provided. Calculation settings: - SCALE version: 6.3.0 - Neutron transport: Shift - Sensitivity analysis: TSUNAMI/IFP - Neutron histories: 50,000 x (9,900 + 100) - Latent generations: 20 - Group structure for sensitivity tallies: 302 fast group structure * SCALE version: 6.3.0 * Neutron transport: Shift * Sensitivity analysis: TSUNAMI/IFP * Neutron histories: 50,000 x (9,900 + 100) * Latent generations: 20 * Group structure for sensitivity tallies: 302 fast group structure ``` EBR-II Loading @@ -44,7 +43,7 @@ EBR-II └── ebr2.tsu_ce.ce_v7.1_endf.out ``` - inp: text input file - out: text output file - sdf: sensitivity data file * inp: text input file * out: text output file * sdf: sensitivity data file HTR-10/README.md +20 −21 Original line number Diff line number Diff line Loading @@ -6,9 +6,9 @@ The SCALE full core model of the HTR-10 was developed based on the benchmark spe the International Handbook of Reactor Physics Experiments of the Organisation for Economic Co-operation and Development (OECD), Nuclear Energy Agency (NEA): > W. K. Terry, L. M. Montierth, S. S. Kim, J. J. Cogliati, A. M. Ougouag (2007), > _Evaluation of the initial critical configuration of the HTR-10 pebble-bed reactor,_ > HTR10-GCR-RESR-001, NEA/NSC/DOC[2006]1, Rev. 0, OECD/NEA. * W. K. Terry, L. M. Montierth, S. S. Kim, J. J. Cogliati, A. M. Ougouag (2007), _Evaluation of the initial critical configuration of the HTR-10 pebble-bed reactor,_ HTR10-GCR-RESR-001, NEA/NSC/DOC[2006]1, Rev. 0, OECD/NEA. A description of the basic design features can be found in [NUREG/CR-7107](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7107/index.html). Loading @@ -17,27 +17,26 @@ A description of the basic design features can be found in Cite the following references when using this model: > [1] F. Bostelmann, G. Ilas, C. Celik, A. M. Holcomb, and W. Wieselquist (2021), > _Nuclear Data Assessment for Advanced Reactors,_ > [NUREG/CR-7289, ORNL/TM-2021/2002](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7289/index.html), > Oak Ridge National Laboratory, Oak Ridge, TN. > > [2] G. Ilas, D. Ilas, R. P. Kelly, E. E. Sunny (2012), _Validation of SCALE > for High Temperature Gas-Cooled Reactor Analysis,_ > [NUREG/CR-7107, ORNL/TM-2011/161](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7107/index.html), > Oak Ridge National Laboratory, Oak Ridge, TN. * F. Bostelmann, G. Ilas, C. Celik, A. M. Holcomb, and W. Wieselquist (2021), _Nuclear Data Assessment for Advanced Reactors,_ [NUREG/CR-7289, ORNL/TM-2021/2002](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7289/index.html), Oak Ridge National Laboratory, Oak Ridge, TN. * G. Ilas, D. Ilas, R. P. Kelly, E. E. Sunny (2012), _Validation of SCALE for High Temperature Gas-Cooled Reactor Analysis,_ [NUREG/CR-7107, ORNL/TM-2011/161](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7107/index.html), Oak Ridge National Laboratory, Oak Ridge, TN. ## Content Input and output files for HTR-10 full core sensitivity and uncertainty calculations with the TSUNAMI-Shift sequence are provided. Calculation settings: - SCALE version: 6.3.0 - Neutron transport: Shift - Sensitivity analysis: TSUNAMI/IFP - Neutron histories: 50,000 x (9,900 + 100) - Latent generations: 20 - Group structure for sensitivity tallies: 252 group structure * SCALE version: 6.3.0 * Neutron transport: Shift * Sensitivity analysis: TSUNAMI/IFP * Neutron histories: 50,000 x (9,900 + 100) * Latent generations: 20 * Group structure for sensitivity tallies: 252 group structure ``` HTR-10 Loading @@ -47,6 +46,6 @@ HTR-10 └── htr10.tsu_ce.ce_v7.1_endf.out ``` - inp: text input file - out: text output file - sdf: sensitivity data file * inp: text input file * out: text output file * sdf: sensitivity data file INL-A-MET/README.md +22 −23 Original line number Diff line number Diff line Loading @@ -8,37 +8,36 @@ Laboratory. The INL Design A concept is an adaptation of the Megapower HPR developed by Los Alamos National Laboratory. The original design contains oxide fuel. However, for this project, metal fuel was assumed: > J. W. Sterbentz Sterbentz, J. W. Werner, A. J. Hummel, J. C. Kennedy, > R. C. O. Brien, A. M. Dion, R. N. Wright, K. P. Ananth, Krishnan P (2018), > _Preliminary Assessment of Two Alternative Core Design Concepts for the Special > Purpose Reactor,_ INL/EXT-17-43212, Rev. 1, Idaho National Laboratory, > Idaho Falls, ID. * J. W. Sterbentz Sterbentz, J. W. Werner, A. J. Hummel, J. C. Kennedy, R. C. O. Brien, A. M. Dion, R. N. Wright, K. P. Ananth, Krishnan P (2018), _Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor,_ INL/EXT-17-43212, Rev. 1, Idaho National Laboratory, Idaho Falls, ID. ## Model citation Cite the following references when using this model: > [1] F. Bostelmann, G. Ilas, C. Celik, A. M. Holcomb, and W. Wieselquist (2021), > _Nuclear Data Assessment for Advanced Reactors,_ > [NUREG/CR-7289, ORNL/TM-2021/2002](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7289/index.html), > Oak Ridge National Laboratory, Oak Ridge, TN. > > [2] E. Walker, S. Skutnik, W. Wieselquist, A. Shaw, and F. Bostelmann (2021), > _SCALE Modeling of the Fast-Spectrum Heat Pipe Reactor,_ > [ORNL/TM-2021/2021](https://www.osti.gov/biblio/1871124-scale-modeling-fast-spectrum-heat-pipe-reactor), > Oak Ridge National Laboratory, Oak Ridge, TN, 2021. * F. Bostelmann, G. Ilas, C. Celik, A. M. Holcomb, and W. Wieselquist (2021), _Nuclear Data Assessment for Advanced Reactors,_ [NUREG/CR-7289, ORNL/TM-2021/2002](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7289/index.html), Oak Ridge National Laboratory, Oak Ridge, TN. * E. Walker, S. Skutnik, W. Wieselquist, A. Shaw, and F. Bostelmann (2021), _SCALE Modeling of the Fast-Spectrum Heat Pipe Reactor,_ [ORNL/TM-2021/2021](https://www.osti.gov/biblio/1871124-scale-modeling-fast-spectrum-heat-pipe-reactor), Oak Ridge National Laboratory, Oak Ridge, TN, 2021. ## Content Input and output files for INL-A-MET full core sensitivity and uncertainty calculations with the TSUNAMI-Shift sequence are provided. Calculation settings: - SCALE version: 6.3.0 - Neutron transport: Shift - Sensitivity analysis: TSUNAMI/IFP - Neutron histories: 50,000 x (9,900 + 100) - Latent generations: 20 - Group structure for sensitivity tallies: 302 fast group structure * SCALE version: 6.3.0 * Neutron transport: Shift * Sensitivity analysis: TSUNAMI/IFP * Neutron histories: 50,000 x (9,900 + 100) * Latent generations: 20 * Group structure for sensitivity tallies: 302 fast group structure ``` INL-A-MET Loading @@ -48,6 +47,6 @@ INL-A-MET └── hpr_inla_met.tsu_ce.ce_v7.1_endf.out ``` - inp: text input file - out: text output file - sdf: sensitivity data file * inp: text input file * out: text output file * sdf: sensitivity data file MET1000/README.md +10 −9 Original line number Diff line number Diff line Loading @@ -26,12 +26,12 @@ Cite the following references when using this model: Input and output files for MET1000 full core sensitivity and uncertainty calculations with the TSUNAMI-Shift sequence are provided. Calculation settings: - SCALE version: 6.3.0 - Neutron transport: Shift - Sensitivity analysis: TSUNAMI/IFP - Neutron histories: 50,000 x (9,900 + 100) - Latent generations: 20 - Group structure for sensitivity tallies: 302 fast group structure * SCALE version: 6.3.0 * Neutron transport: Shift * Sensitivity analysis: TSUNAMI/IFP * Neutron histories: 50,000 x (9,900 + 100) * Latent generations: 20 * Group structure for sensitivity tallies: 302 fast group structure ``` MET1000 Loading @@ -41,6 +41,7 @@ MET1000 └── met1000.tsuce.ce_v7.1_endf.out ``` - inp: text input file - out: text output file - sdf: sensitivity data file * inp: text input file * out: text output file * sdf: sensitivity data file * No newline at end of file MSRE/README.md +22 −25 Original line number Diff line number Diff line Loading @@ -10,39 +10,36 @@ and Development (OECD), Nuclear Energy Agency (NEA). The major references used to develop the provided models are the following: > [1] D. Shen et al. (2019), _Molten-Salt Reactor Experiment (MSRE) Zero-Power First > Critical Experiment with U-235,_ MSRE-MSR-EXP-001, International Handbook > of Reactor Physics Experiments, OECD/NEA. > > [2] M. Fratoni, D. Shen, G. Ilas, and J. Powers (2020), _Molten Salt Reactor Experiment > Benchmark Evaluation,_ [DOE-UCB-8542](https://www.osti.gov/servlets/purl/1617123), 16-10240, University of California, Berkeley, CA. > > [3] R. C. Robertson, _MSRE design and operations report part I: Description of reactor design,_ ORNL-TM-0728, 1965. * D. Shen et al. (2019), _Molten-Salt Reactor Experiment (MSRE) Zero-Power First Critical Experiment with U-235,_ MSRE-MSR-EXP-001, International Handbook of Reactor Physics Experiments, OECD/NEA. * M. Fratoni, D. Shen, G. Ilas, and J. Powers (2020), _Molten Salt Reactor Experiment Benchmark Evaluation,_ [DOE-UCB-8542](https://www.osti.gov/servlets/purl/1617123), 16-10240, University of California, Berkeley, CA. * R. C. Robertson, _MSRE design and operations report part I: Description of reactor design,_ ORNL-TM-0728, 1965. ## Model citation Cite the following references when using this model: > [1] F. Bostelmann, S. E. Skutnik, E. Walker, G. Ilas, and W. A. Wieselquist (2021), > _Modeling of the Molten Salt Reactor Experiment with SCALE,_ Nuclear Technology, > doi:[10.1080/00295450.2021.1943122](https://doi.org/10.1080/00295450.2021.1943122) > > [2] F. Bostelmann, G. Ilas, C. Celik, A. M. Holcomb, and W. Wieselquist (2021), > _Nuclear Data Assessment for Advanced Reactors,_ > [NUREG/CR-7289, ORNL/TM-2021/2002](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7289/index.html), > Oak Ridge National Laboratory, Oak Ridge, TN. * F. Bostelmann, S. E. Skutnik, E. Walker, G. Ilas, and W. A. Wieselquist (2021), _Modeling of the Molten Salt Reactor Experiment with SCALE,_ Nuclear Technology, doi:[10.1080/00295450.2021.1943122](https://doi.org/10.1080/00295450.2021.1943122) * F. Bostelmann, G. Ilas, C. Celik, A. M. Holcomb, and W. Wieselquist (2021), _Nuclear Data Assessment for Advanced Reactors,_ [NUREG/CR-7289, ORNL/TM-2021/2002](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7289/index.html), Oak Ridge National Laboratory, Oak Ridge, TN. ## Content Input and output files for MSRE full core sensitivity and uncertainty calculations with the TSUNAMI-Shift sequence are provided. Calculation settings: - SCALE version: 6.3.0 - Neutron transport: Shift - Sensitivity analysis: TSUNAMI/IFP - Neutron histories: 25,000 x (19,900 + 100) - Latent generations: 20 - Group structure for sensitivity tallies: 252 group structure * SCALE version: 6.3.0 * Neutron transport: Shift * Sensitivity analysis: TSUNAMI/IFP * Neutron histories: 25,000 x (19,900 + 100) * Latent generations: 20 * Group structure for sensitivity tallies: 252 group structure :warning: The models provided here might differ from the models used for the studies summarized in the report since models are updated based upon feedback. :warning: Loading @@ -55,6 +52,6 @@ MSRE └── msre.tsu_ce.ce_v7.1_endf.out ``` - inp: text input file - out: text output file - sdf: sensitivity data file * inp: text input file * out: text output file * sdf: sensitivity data file Loading
EBR-II/README.md +20 −21 Original line number Diff line number Diff line Loading @@ -6,35 +6,34 @@ The SCALE full core model of the EBR-II was developed based on the benchmark spe Experiments of the Organisation for Economic Co-operation and Development (OECD), Nuclear Energy Agency (NEA): > E. S. Lum, C. K. Pope, R. Stewart, B. Byambadorj, Q. Beaulieu (2018), > _Evaluation of Run 138B at Experimental Breeder Reactor II, a prototypic > liquid metal reactor,_ EBR2-LMFR-RESR-001, NEA/NSC/DOC[2006]1, Rev. 0, OECD/NEA. * E. S. Lum, C. K. Pope, R. Stewart, B. Byambadorj, Q. Beaulieu (2018), _Evaluation of Run 138B at Experimental Breeder Reactor II, a prototypic liquid metal reactor,_ EBR2-LMFR-RESR-001, NEA/NSC/DOC[2006]1, Rev. 0, OECD/NEA. ## Model citation Cite the following references when using this model: > [1] F. Bostelmann, G. Ilas, W. Wieselquist (2021), _Nuclear Data Sensitivity > Study for the EBR-II Fast Reactor Benchmark Using SCALE with ENDF/B-VII.1 > and ENDF/B-VIII.0,_ Journal of Nuclear Engineering, 2, 345-367, > doi:[10.3390/jne2040028](https://www.mdpi.com/2673-4362/2/4/28). > > [2] F. Bostelmann, G. Ilas, C. Celik, A. M. Holcomb, and W. Wieselquist (2021), > _Nuclear Data Assessment for Advanced Reactors,_ > [NUREG/CR-7289, ORNL/TM-2021/2002](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7289/index.html), > Oak Ridge National Laboratory, Oak Ridge, TN. * F. Bostelmann, G. Ilas, W. Wieselquist (2021), _Nuclear Data Sensitivity Study for the EBR-II Fast Reactor Benchmark Using SCALE with ENDF/B-VII.1 and ENDF/B-VIII.0,_ Journal of Nuclear Engineering, 2, 345-367, doi:[10.3390/jne2040028](https://www.mdpi.com/2673-4362/2/4/28). * F. Bostelmann, G. Ilas, C. Celik, A. M. Holcomb, and W. Wieselquist (2021), _Nuclear Data Assessment for Advanced Reactors,_ [NUREG/CR-7289, ORNL/TM-2021/2002](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7289/index.html), Oak Ridge National Laboratory, Oak Ridge, TN. ## Content Input and output files for EBR-II full core sensitivity and uncertainty calculations with the TSUNAMI-Shift sequence are provided. Calculation settings: - SCALE version: 6.3.0 - Neutron transport: Shift - Sensitivity analysis: TSUNAMI/IFP - Neutron histories: 50,000 x (9,900 + 100) - Latent generations: 20 - Group structure for sensitivity tallies: 302 fast group structure * SCALE version: 6.3.0 * Neutron transport: Shift * Sensitivity analysis: TSUNAMI/IFP * Neutron histories: 50,000 x (9,900 + 100) * Latent generations: 20 * Group structure for sensitivity tallies: 302 fast group structure ``` EBR-II Loading @@ -44,7 +43,7 @@ EBR-II └── ebr2.tsu_ce.ce_v7.1_endf.out ``` - inp: text input file - out: text output file - sdf: sensitivity data file * inp: text input file * out: text output file * sdf: sensitivity data file
HTR-10/README.md +20 −21 Original line number Diff line number Diff line Loading @@ -6,9 +6,9 @@ The SCALE full core model of the HTR-10 was developed based on the benchmark spe the International Handbook of Reactor Physics Experiments of the Organisation for Economic Co-operation and Development (OECD), Nuclear Energy Agency (NEA): > W. K. Terry, L. M. Montierth, S. S. Kim, J. J. Cogliati, A. M. Ougouag (2007), > _Evaluation of the initial critical configuration of the HTR-10 pebble-bed reactor,_ > HTR10-GCR-RESR-001, NEA/NSC/DOC[2006]1, Rev. 0, OECD/NEA. * W. K. Terry, L. M. Montierth, S. S. Kim, J. J. Cogliati, A. M. Ougouag (2007), _Evaluation of the initial critical configuration of the HTR-10 pebble-bed reactor,_ HTR10-GCR-RESR-001, NEA/NSC/DOC[2006]1, Rev. 0, OECD/NEA. A description of the basic design features can be found in [NUREG/CR-7107](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7107/index.html). Loading @@ -17,27 +17,26 @@ A description of the basic design features can be found in Cite the following references when using this model: > [1] F. Bostelmann, G. Ilas, C. Celik, A. M. Holcomb, and W. Wieselquist (2021), > _Nuclear Data Assessment for Advanced Reactors,_ > [NUREG/CR-7289, ORNL/TM-2021/2002](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7289/index.html), > Oak Ridge National Laboratory, Oak Ridge, TN. > > [2] G. Ilas, D. Ilas, R. P. Kelly, E. E. Sunny (2012), _Validation of SCALE > for High Temperature Gas-Cooled Reactor Analysis,_ > [NUREG/CR-7107, ORNL/TM-2011/161](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7107/index.html), > Oak Ridge National Laboratory, Oak Ridge, TN. * F. Bostelmann, G. Ilas, C. Celik, A. M. Holcomb, and W. Wieselquist (2021), _Nuclear Data Assessment for Advanced Reactors,_ [NUREG/CR-7289, ORNL/TM-2021/2002](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7289/index.html), Oak Ridge National Laboratory, Oak Ridge, TN. * G. Ilas, D. Ilas, R. P. Kelly, E. E. Sunny (2012), _Validation of SCALE for High Temperature Gas-Cooled Reactor Analysis,_ [NUREG/CR-7107, ORNL/TM-2011/161](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7107/index.html), Oak Ridge National Laboratory, Oak Ridge, TN. ## Content Input and output files for HTR-10 full core sensitivity and uncertainty calculations with the TSUNAMI-Shift sequence are provided. Calculation settings: - SCALE version: 6.3.0 - Neutron transport: Shift - Sensitivity analysis: TSUNAMI/IFP - Neutron histories: 50,000 x (9,900 + 100) - Latent generations: 20 - Group structure for sensitivity tallies: 252 group structure * SCALE version: 6.3.0 * Neutron transport: Shift * Sensitivity analysis: TSUNAMI/IFP * Neutron histories: 50,000 x (9,900 + 100) * Latent generations: 20 * Group structure for sensitivity tallies: 252 group structure ``` HTR-10 Loading @@ -47,6 +46,6 @@ HTR-10 └── htr10.tsu_ce.ce_v7.1_endf.out ``` - inp: text input file - out: text output file - sdf: sensitivity data file * inp: text input file * out: text output file * sdf: sensitivity data file
INL-A-MET/README.md +22 −23 Original line number Diff line number Diff line Loading @@ -8,37 +8,36 @@ Laboratory. The INL Design A concept is an adaptation of the Megapower HPR developed by Los Alamos National Laboratory. The original design contains oxide fuel. However, for this project, metal fuel was assumed: > J. W. Sterbentz Sterbentz, J. W. Werner, A. J. Hummel, J. C. Kennedy, > R. C. O. Brien, A. M. Dion, R. N. Wright, K. P. Ananth, Krishnan P (2018), > _Preliminary Assessment of Two Alternative Core Design Concepts for the Special > Purpose Reactor,_ INL/EXT-17-43212, Rev. 1, Idaho National Laboratory, > Idaho Falls, ID. * J. W. Sterbentz Sterbentz, J. W. Werner, A. J. Hummel, J. C. Kennedy, R. C. O. Brien, A. M. Dion, R. N. Wright, K. P. Ananth, Krishnan P (2018), _Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor,_ INL/EXT-17-43212, Rev. 1, Idaho National Laboratory, Idaho Falls, ID. ## Model citation Cite the following references when using this model: > [1] F. Bostelmann, G. Ilas, C. Celik, A. M. Holcomb, and W. Wieselquist (2021), > _Nuclear Data Assessment for Advanced Reactors,_ > [NUREG/CR-7289, ORNL/TM-2021/2002](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7289/index.html), > Oak Ridge National Laboratory, Oak Ridge, TN. > > [2] E. Walker, S. Skutnik, W. Wieselquist, A. Shaw, and F. Bostelmann (2021), > _SCALE Modeling of the Fast-Spectrum Heat Pipe Reactor,_ > [ORNL/TM-2021/2021](https://www.osti.gov/biblio/1871124-scale-modeling-fast-spectrum-heat-pipe-reactor), > Oak Ridge National Laboratory, Oak Ridge, TN, 2021. * F. Bostelmann, G. Ilas, C. Celik, A. M. Holcomb, and W. Wieselquist (2021), _Nuclear Data Assessment for Advanced Reactors,_ [NUREG/CR-7289, ORNL/TM-2021/2002](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7289/index.html), Oak Ridge National Laboratory, Oak Ridge, TN. * E. Walker, S. Skutnik, W. Wieselquist, A. Shaw, and F. Bostelmann (2021), _SCALE Modeling of the Fast-Spectrum Heat Pipe Reactor,_ [ORNL/TM-2021/2021](https://www.osti.gov/biblio/1871124-scale-modeling-fast-spectrum-heat-pipe-reactor), Oak Ridge National Laboratory, Oak Ridge, TN, 2021. ## Content Input and output files for INL-A-MET full core sensitivity and uncertainty calculations with the TSUNAMI-Shift sequence are provided. Calculation settings: - SCALE version: 6.3.0 - Neutron transport: Shift - Sensitivity analysis: TSUNAMI/IFP - Neutron histories: 50,000 x (9,900 + 100) - Latent generations: 20 - Group structure for sensitivity tallies: 302 fast group structure * SCALE version: 6.3.0 * Neutron transport: Shift * Sensitivity analysis: TSUNAMI/IFP * Neutron histories: 50,000 x (9,900 + 100) * Latent generations: 20 * Group structure for sensitivity tallies: 302 fast group structure ``` INL-A-MET Loading @@ -48,6 +47,6 @@ INL-A-MET └── hpr_inla_met.tsu_ce.ce_v7.1_endf.out ``` - inp: text input file - out: text output file - sdf: sensitivity data file * inp: text input file * out: text output file * sdf: sensitivity data file
MET1000/README.md +10 −9 Original line number Diff line number Diff line Loading @@ -26,12 +26,12 @@ Cite the following references when using this model: Input and output files for MET1000 full core sensitivity and uncertainty calculations with the TSUNAMI-Shift sequence are provided. Calculation settings: - SCALE version: 6.3.0 - Neutron transport: Shift - Sensitivity analysis: TSUNAMI/IFP - Neutron histories: 50,000 x (9,900 + 100) - Latent generations: 20 - Group structure for sensitivity tallies: 302 fast group structure * SCALE version: 6.3.0 * Neutron transport: Shift * Sensitivity analysis: TSUNAMI/IFP * Neutron histories: 50,000 x (9,900 + 100) * Latent generations: 20 * Group structure for sensitivity tallies: 302 fast group structure ``` MET1000 Loading @@ -41,6 +41,7 @@ MET1000 └── met1000.tsuce.ce_v7.1_endf.out ``` - inp: text input file - out: text output file - sdf: sensitivity data file * inp: text input file * out: text output file * sdf: sensitivity data file * No newline at end of file
MSRE/README.md +22 −25 Original line number Diff line number Diff line Loading @@ -10,39 +10,36 @@ and Development (OECD), Nuclear Energy Agency (NEA). The major references used to develop the provided models are the following: > [1] D. Shen et al. (2019), _Molten-Salt Reactor Experiment (MSRE) Zero-Power First > Critical Experiment with U-235,_ MSRE-MSR-EXP-001, International Handbook > of Reactor Physics Experiments, OECD/NEA. > > [2] M. Fratoni, D. Shen, G. Ilas, and J. Powers (2020), _Molten Salt Reactor Experiment > Benchmark Evaluation,_ [DOE-UCB-8542](https://www.osti.gov/servlets/purl/1617123), 16-10240, University of California, Berkeley, CA. > > [3] R. C. Robertson, _MSRE design and operations report part I: Description of reactor design,_ ORNL-TM-0728, 1965. * D. Shen et al. (2019), _Molten-Salt Reactor Experiment (MSRE) Zero-Power First Critical Experiment with U-235,_ MSRE-MSR-EXP-001, International Handbook of Reactor Physics Experiments, OECD/NEA. * M. Fratoni, D. Shen, G. Ilas, and J. Powers (2020), _Molten Salt Reactor Experiment Benchmark Evaluation,_ [DOE-UCB-8542](https://www.osti.gov/servlets/purl/1617123), 16-10240, University of California, Berkeley, CA. * R. C. Robertson, _MSRE design and operations report part I: Description of reactor design,_ ORNL-TM-0728, 1965. ## Model citation Cite the following references when using this model: > [1] F. Bostelmann, S. E. Skutnik, E. Walker, G. Ilas, and W. A. Wieselquist (2021), > _Modeling of the Molten Salt Reactor Experiment with SCALE,_ Nuclear Technology, > doi:[10.1080/00295450.2021.1943122](https://doi.org/10.1080/00295450.2021.1943122) > > [2] F. Bostelmann, G. Ilas, C. Celik, A. M. Holcomb, and W. Wieselquist (2021), > _Nuclear Data Assessment for Advanced Reactors,_ > [NUREG/CR-7289, ORNL/TM-2021/2002](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7289/index.html), > Oak Ridge National Laboratory, Oak Ridge, TN. * F. Bostelmann, S. E. Skutnik, E. Walker, G. Ilas, and W. A. Wieselquist (2021), _Modeling of the Molten Salt Reactor Experiment with SCALE,_ Nuclear Technology, doi:[10.1080/00295450.2021.1943122](https://doi.org/10.1080/00295450.2021.1943122) * F. Bostelmann, G. Ilas, C. Celik, A. M. Holcomb, and W. Wieselquist (2021), _Nuclear Data Assessment for Advanced Reactors,_ [NUREG/CR-7289, ORNL/TM-2021/2002](https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7289/index.html), Oak Ridge National Laboratory, Oak Ridge, TN. ## Content Input and output files for MSRE full core sensitivity and uncertainty calculations with the TSUNAMI-Shift sequence are provided. Calculation settings: - SCALE version: 6.3.0 - Neutron transport: Shift - Sensitivity analysis: TSUNAMI/IFP - Neutron histories: 25,000 x (19,900 + 100) - Latent generations: 20 - Group structure for sensitivity tallies: 252 group structure * SCALE version: 6.3.0 * Neutron transport: Shift * Sensitivity analysis: TSUNAMI/IFP * Neutron histories: 25,000 x (19,900 + 100) * Latent generations: 20 * Group structure for sensitivity tallies: 252 group structure :warning: The models provided here might differ from the models used for the studies summarized in the report since models are updated based upon feedback. :warning: Loading @@ -55,6 +52,6 @@ MSRE └── msre.tsu_ce.ce_v7.1_endf.out ``` - inp: text input file - out: text output file - sdf: sensitivity data file * inp: text input file * out: text output file * sdf: sensitivity data file
