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 Acceleratordriven systems; Monte Carlo NParticle eXtended (MCNPX); Nuclear reactions – Computer simulation; Radioactive wastes — Transmutation; Spallation (Nuclear physics) (4)
 Monte Carlo NParticle eXtended (MCNPX); Neutron flux; Neutrons; Nuclear reactions – Computer simulation; Radiative transfer; Radioactive wastes — Transmutation; Spallation (Nuclear physics); Spent reactor fuels (2)
 Monte Carlo method; Nuclear reactors; Radiation — Dosage; Radiation dosimetry; Shielding (Radiation); Statistics (2)
 Nuclear reactors; Monte Carlo method; Spent reactor fuels; Stochastic processes (2)
 Laser beams; Monte Carlo NParticle eXtended (MCNPX); Nuclear reactions – Computer simulation; Radioactive wastes — Transmutation; Spallation (Nuclear physics) (1)

 Neutron flux; Neutrons; Particles (Nuclear physics); Radioactive wastes—Transmutation; Spallation (Nuclear physics); Spent reactor fuels (1)
 Monte Carlo NParticle eXtended (MCNPX); Neutron flux; Neutrons; Nuclear reactors; Particles (Nuclear physics); Radioactive wastes — Transmutation; Spallation (Nuclear physics); Spent reactor fuels; Transmutation (Chemistry) (1)
 Neutron flux; Neutrons; Nuclear reactors; Particles (Nuclear physics); Radioactive wastes—Transmutation; Spallation (Nuclear physics); Spent reactor fuels; Transmutation (Chemistry) (1)
 Monte Carlo NParticle eXtended (MCNPX); Neutron flux; Neutrons; Radiative transfer; Radioactive wastes — Transmutation; Spallation (Nuclear physics); Spent reactor fuels (1)
Articles 1  15 of 15
FullText Articles in Physics
Implementation Of Uncertainty Propagation In Triton/Keno, Charlotta Sanders, Denis Beller
Implementation Of Uncertainty Propagation In Triton/Keno, Charlotta Sanders, Denis Beller
Reactor Campaign (TRP)
Monte Carlo methods are beginning to be used for three dimensional fuel depletion analyses to compute various quantities of interest, including isotopic compositions of used nuclear fuel. The TRITON control module, available in the SCALE 5.1 code system, can perform threedimensional (3D) depletion calculations using either the KENO V.a or KENOVI Monte Carlo transport codes, as well as the twodimensional (2D) NEWT discrete ordinates code. To overcome problems such as spatially nonuniform neutron flux and nonuniform statistical uncertainties in computed reaction rates and to improve the fidelity of calculations using Monte Carlo methods, uncertainty propagation is needed for ...
Monaco/Mavric Evaluation For Facility Shielding And Dose Rate Analysis, Charlotta Sanders, Denis Beller
Monaco/Mavric Evaluation For Facility Shielding And Dose Rate Analysis, Charlotta Sanders, Denis Beller
Reactor Campaign (TRP)
The dimensions and the large amount of shielding required for Global Nuclear Energy Partnership (GNEP) facilities, advanced radiation shielding, and dose computation techniques are beyond today’s capabilities and will certainly be required. With the Generation IV Nuclear Energy System Initiative, it will become increasingly important to be able to accurately model advanced Boiling Water Reactor and Pressurized Water Reactor facilities, and to calculate dose rates at all locations within a containment (e.g., resulting from radiations from the reactor as well as the from the primary coolant loop) and adjoining structures (e.g., from the spent fuel pool).
The ...
Implementation Of Uncertainty Propagation In Triton/Keno: To Support The Global Nuclear Energy Partnership, Charlotta Sanders, Denis Beller
Implementation Of Uncertainty Propagation In Triton/Keno: To Support The Global Nuclear Energy Partnership, Charlotta Sanders, Denis Beller
Reactor Campaign (TRP)
Monte Carlo methods are beginning to be used for threedimensional fuel depletion analyses to compute various quantities of interest, including isotopic compositions of used fuel.1 The TRITON control module, available in the SCALE 5.1 code system, can perform three dimensional (3D) depletion calculations using either the KENO V.a or KENOVI Monte Carlo transport codes, as well as the twodimensional (2 D) NEWT discrete ordinates code. For typical reactor systems, the neutron flux is not spatially uniform. For Monte Carlo simulations, this results in nonuniform statistical uncertainties in the computed reaction rates. For spatial regions where the flux ...
Monaco/Mavric Evaluation For Facility Shielding And Dose Rate Analysis: To Support The Global Nuclear Energy Partnership, Charlotta Sanders, Denis Beller
Monaco/Mavric Evaluation For Facility Shielding And Dose Rate Analysis: To Support The Global Nuclear Energy Partnership, Charlotta Sanders, Denis Beller
Reactor Campaign (TRP)
Monte Carlo methods are used to compute fluxes or dose rates over large areas using mesh tallies. For problems that demand that the uncertainty in each mesh cell be less than some set maximum, computation time is controlled by the cell with the largest uncertainty. This issue becomes quite troublesome in deeppenetration problems, and advanced variance reduction techniques are required to obtain reasonable uncertainties over large areas.
In this project the MAVRIC sequence will be evaluated along with the Monte Carlo engine Monaco to investigate its effectiveness and usefulness in facility shielding and dose rate analyses. A previously MCNPevaluated cask ...
Radiation Transport Modeling Using Parallel Computational Techniques, William Culbreth, Denis Beller
Radiation Transport Modeling Using Parallel Computational Techniques, William Culbreth, Denis Beller
Reactor Campaign (TRP)
One of the most significant tools available for the design and analysis of acceleratordriven systems, such as the systems proposed for transmutation, is the highenergy particle transport code MCNPX. The MCNPX code suite, developed by the national laboratories, allows researchers and engineers to model the complex interactions of highenergy particles with the target and related systems, including the spallation reaction and subsequent neutron multiplication expected in the accelerator targets.
The next stage in the development of the MCNPX code suite is to validate the code by comparing the theoretical predictions from the models with experimental observations. Additionally, the nuclear database ...
Radiation Transport Modeling Using Parallel Computational Techniques, William Culbreth, Denis Beller
Radiation Transport Modeling Using Parallel Computational Techniques, William Culbreth, Denis Beller
Reactor Campaign (TRP)
The second year of this project involved modeling several aspects of the LANCSE beam experiments:
 Modeling targets of varying diameter in air, in a vacuum, and in the presence of humid air;
 Modeling various proton beam profiles;
 Modeling the effects of offaxis proton beam impingement on the target;
 Modeling the asymmetry introduced by the steel table below the target;
 Modeling the effect of varying ratios of Pb to Bi and the effect of impurities; and
 Modeling the system, including other structures within the test room.
With the experience gained through modeling these systems, the UNLV researchers plan, with the assistance ...
Radiation Transport Modeling Using Parallel Computational Techniques, William Culbreth
Radiation Transport Modeling Using Parallel Computational Techniques, William Culbreth
Reactor Campaign (TRP)
The Advanced Fuel Cycle Initiative (AFCI) program will rely on the use of accurate calculations and simulations of criticality and shielding for the separation process of the longlived isotopes that present a significant safety hazard in commercial spent fuel. To help design and verify the safety of the separation process, the neutronics code MCNPX will be used to model the distribution of neutron flux within the fuel blanket and to determine the neutron multiplication, k_{eff}. However, the cross section libraries and computational methods used by MCNPX at these neutron energies still have some uncertainty and will require validation.
Currently ...
Radiation Transport Modeling Using Parallel Computational Techniques, William Culbreth, Denis Beller
Radiation Transport Modeling Using Parallel Computational Techniques, William Culbreth, Denis Beller
Reactor Campaign (TRP)
The second year of this project involved modeling several aspects of the LANCSE beam experiments:
 Modeling targets of varying diameter in air, in a vacuum, and in the presence of humid air;
 Modeling various proton beam profiles;
 Modeling the effects of offaxis proton beam impingement on the target;
 Modeling the asymmetry introduced by the steel table below the target;
 Modeling the effect of varying ratios of Pb to Bi and the effect of impurities; and
 Modeling the system, including other structures within the test room.
With the experience gained through modeling these systems, the UNLV researchers plan, with the assistance ...
Monte Carlo Verification And Modeling Of LeadBismuth Spallation Targets, Daniel R. Lowe
Monte Carlo Verification And Modeling Of LeadBismuth Spallation Targets, Daniel R. Lowe
Reactor Campaign (TRP)
No abstract provided.
Radiation Transport Modeling Of BeamTarget Experiments For The Aaa Project: Quaterly Report, June 01 August 31, 2002, William Culbreth
Radiation Transport Modeling Of BeamTarget Experiments For The Aaa Project: Quaterly Report, June 01 August 31, 2002, William Culbreth
Reactor Campaign (TRP)
The national development of technology to transmute nuclear waste depends upon the generation of highenergy neutrons produced by proton spallation. Proton accelerators, such as LANSCE at the Los Alamos National Laboratory, are capable of producing 800 MeV protons. By bombarding a lead /bismuth target, each proton may generate up to 25 neutrons that can activate fission of transuranic isotopes. Students at UNLV have been involved in radiation transport calculations in collaboration with researchers at the Los Alamos National Laboratory and at the Argonne National Laboratory.
Radiation Transport Modeling Of BeamTarget Experiments For The Aaa Project: Quaterly Report, William Culbreth
Radiation Transport Modeling Of BeamTarget Experiments For The Aaa Project: Quaterly Report, William Culbreth
Reactor Campaign (TRP)
The national development of technology to transmute nuclear waste depends upon the generation of high energy neutrons produced by proton spallation. Proton accelerators, such as LANSCE at the Los Alamos National Laboratory, are capable of producing 800 MeV protons. By bombarding a lead/bismuth target, each proton may generate 500 or more neutrons that can activate fission products or induce the fission of transuranic isotopes.
The Monte Carlo radiation transport code MCNPX developed at LANL is an important tool in the design of transmuter technology. It must be validated, however, for the neutron energy that will be employed. Experiments are ...
Project Continuation Proposal: Radiation Transport Modeling Of BeamTarget Experiments For The Aaa Project, William Culbreth
Project Continuation Proposal: Radiation Transport Modeling Of BeamTarget Experiments For The Aaa Project, William Culbreth
Reactor Campaign (TRP)
The AAA program will rely on the use of an acceleratorbased transmuter1 to expose spent nuclear fuel to highenergy neutrons. The neutron flux will be sufficient to activate or fission the longlived isotopes of Tc, I, Pu, Am, Cm, and Np that present a significant radiological hazard in commercial spent fuel. Transmuter fuel will be subcritical and a highenergy proton accelerator is needed to maintain the necessary neutron flux through the use of a neutron spallation target. The maximum neutron energy produced by spallation (~ 800 MeV) is significantly higher than that produced by a commercial light water reactor (~ 2 MeV ...
Radiation Transport Modeling Of BeamTarget Experiments, William Culbreth, Denis Beller
Radiation Transport Modeling Of BeamTarget Experiments, William Culbreth, Denis Beller
Reactor Campaign (TRP)
In the first year of the UNLV effort, researchers planned to develop the models of the experimental systems to predict the neutron flux and leakage from the experimental targets using the MCNPX code suite in order to help determine these missing parameters. To support these models, the researchers project, or estimate, values for the unknown parameters describing various events and phenomena occurring within the beamtarget experiment. The results of these simulations will then be compared against the observed neutron leakage rates and energies. The estimates for the unknown parameters are then revised to correlate with the observed values (these parameters ...
Radiation Transport Modeling Of BeamTarget Experiments For The Aaa Project: Quaterly Report, William Culbreth
Radiation Transport Modeling Of BeamTarget Experiments For The Aaa Project: Quaterly Report, William Culbreth
Reactor Campaign (TRP)
The national development of technology to transmute nuclear waste depends upon the generation of high energy neutrons produced by proton spallation. Proton accelerators, such as LANSCE at the Los Alamos National Laboratory, are capable of producing 800 MeV protons. By bombarding a lead/bismuth target, each proton may generate 500 or more neutrons that can activate fission products or induce the fission of transuranic isotopes.
The Monte Carlo radiation transport code MCNPX developed at LANL is an important tool in the design of transmuter technology. It must be validated, however, for the neutron energy that will be employed. Experiments are ...
Radiation Transport Modeling Of BeamTarget Experiments For The Aaa Project, William Culbreth
Radiation Transport Modeling Of BeamTarget Experiments For The Aaa Project, William Culbreth
Reactor Campaign (TRP)
The AAA program will rely on the use of an acceleratorbased transmuter to expose spent nuclear fuel to highenergy neutrons. The neutron flux will be sufficient to activate or fission the longlived isotopes of Tc, I, Pu, Am, Cm, and Np that present a significant safety hazard in commercial spent fuel. Transmuter fuel will be subcritical and a highenergy proton accelerator is needed to maintain the necessary neutron flux through the use of a neutron spallation target. The maximum neutron energy produced by spallation (~ 600 MeV) is significantly higher than that produced by a commercial light water reactor (~ 2 MeV ...