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At least 757 records · Page 42
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Verification & Validation of High-Order Short-Characteristics-Based Deterministic Transport Methodology on Unstructured Grids

The research team has developed a practical, high-order, discrete-ordinates, short characteristics neutron transport code for three-dimensional configurations represented on unstructured tetrahedral grids that can be used for realistic reactor physics applications at both the assembly and core levels. This project will perform a comprehensive verification and validation of this new computational tool against both a continuous-energy Monte Carlo simulation (e.g. MCNP) and experimentally measured data, an essential prerequisite for its deployment in reactor core modeling. Verification is divided into three phases. The team will first conduct spatial mesh and expansion order refinement studies to monitor convergence of the numerical solution to reference solutions. This is quantified by convergence rates that are based on integral error norms computed from the cell-by-cell difference between the code’s numerical solution and its reference counterpart. The latter is either analytic or very fine- mesh numerical solutions from independent computational tools. For the second phase, the team will create a suite of code-independent benchmark configurations to enable testing the theoretical order of accuracy of any particular discretization of the discrete ordinates approximation of the transport equation. For each tested case (i.e. mesh and spatial approximation order), researchers will execute the code and compare the resulting numerical solution to the exact solution on a per cell basis to determine the distribution of the numerical error. The final activity comprises a comparison to continuous-energy Monte Carlo solutions for zero-power critical configuration measurements at Idaho National Laboratory’s Advanced Test Reactor (ATR). Results of this comparison will allow the investigators to distinguish between modeling errors and the above-listed discretization errors introduced by the deterministic method, and to separate the sources of uncertainty.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Boron-10 ABUNCL Models of Fuel Testing

The Department of Energy Office of Nuclear Safeguards and Security (NA-241) is supporting the project Coincidence Counting With Boron-Based Alternative Neutron Detection Technology at Pacific Northwest National Laboratory (PNNL) for the development of a 3He proportional counter alternative neutron coincidence counter. The goal of this project is to design, build and demonstrate a system based upon 10B-lined proportional tubes in a configuration typical for 3He-based coincidence counter applications. This report provides results from MCNP simulations of the General Electric Reuter-Stokes Alternative Boron-Based Uranium Neutron Coincidence Collar (ABUNCL) active configuration model with fuel pins previously measured at Los Alamos National Laboratory. A comparison of the GE-ABUNCL simulations and simulations of 3He based UNCL-II active counter (the system for which the GE-ABUNCL was targeted to replace) with the same fuel pin assemblies is also provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Release of Continuous Representation for S(α,β) ACE Data

For low energy neutrons, the default free gas model for scattering cross sections is not always appropriate. Molecular effects or crystalline structure effects can affect the neutron scattering cross sections. These effects are included in the S(α; β) thermal neutron scattering data and are tabulated in file 7 of the ENDF6 format files. S stands for scattering. α is a momentum transfer variable and is an energy transfer variable. The S(α; β) cross sections can include coherent elastic scattering (no E change for the neutron, but specific scattering angles), incoherent elastic scattering (no E change for the neutron, but continuous scattering angles), and inelastic scattering (E change for the neutron, and change in angle as well). Every S(α; β) material will have inelastic scattering and may have either coherent or incoherent elastic scattering (but not both). Coherent elastic scattering cross sections have distinctive jagged-looking Bragg edges, whereas the other cross sections are much smoother. The evaluated files from the NNDC are processed locally in the THERMR module of NJOY. Data can be produced either for continuous energy Monte Carlo codes (using ACER) or embedded in multi-group cross sections for deterministic (or even multi-group Monte Carlo) codes (using GROUPR). Currently, the S(α; β) files available for MCNP use discrete energy changes for inelastic scattering. That is, the scattered neutrons can only be emitted at specific energies— rather than across a continuous spectrum of energies. The discrete energies are chosen to preserve the average secondary neutron energy, i.e., in an integral sense, but the discrete treatment does not preserve any differential quantities in energy or angle.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Verification of MCNP6.1 and MCNP6.1.1 for Criticality Safety Applications

MCNP6.1 is the new production version of MCNP [1] released in June 2013. An updated beta version, MCNP6.1.1, is targeted for release in summer 2014 to enable the use of several new features for homeland security and nonproliferation applications. The beta version does not include new capabilities for the criticality safety community, but does run significantly faster than the production version. From 2003 – 2013, the production version was MCNP5, with MCNP5-1.60 the final release in 2010.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Spent Fuel Assembly Modeling for Unattended Gamma Emission Tomography (UGET) Project [Slides]

NGSI-SF project provided good basis for spent fuel assemblies. UGET assemblies have both DU pins and water replacing fuel in the center of the bundle. Both MCNP geometry files and CINDER90 isotopic compositions files were created and used by LLNL and/or others. Four different spent fuel assembly types were represented: PWR, BWR, VVER-440, and the VVER-1000. Not enough information was known about irradiation of assemblies in Forsmark or Olkiluoto (JNT 1510) benchmarks for detailed calculations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Final report on accident tolerant fuel performance analysis of APMT-Steel Clad/UO₂ fuel and APMT-Steel Clad/UN-U₃Si₅ fuel concepts

In FY2014 our group completed and documented analysis of new Accident Tolerant Fuel (ATF) concepts using BISON. We have modeled the viability of moving from Zircaloy to stainless steel cladding in traditional light water reactors (LWRs). We have explored the reactivity penalty of this change using the MCNP-based burnup code Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify changes to structural integrity resulting from thinner stainless steel claddings. We account for thermal and irradiation creep, fission gas swelling, thermal swelling and fuel relocation in the models for both Zircaloy and stainless steel claddings. Additional models that account for the lower oxidation stainless steel APMT are also invoked where available. Irradiation data for HT9 is used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied towards cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided. In the first section (section 1-3), we present results accepted for publication in the 2014 TOPFUEL conference regarding the APMT/UO₂ ATF concept (J. Galloway & C. Unal, Accident Tolerant and Neutronically Favorable LWR Cladding, Proceedings of WRFPM 2014, Sendai, Japan, Paper No.1000050). Next we discuss our preliminary findings from the thermo-mechanical analysis of UN-U₃Si₅ fuel with APMT clad. In this analysis we used models developed from limited data that need to be updated when the irradiation data from ATF-1 test is available. Initial results indicate a swelling rate less than 1.5% is needed to prevent excessive clad stress.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Monte Carlo Simulations of Neutron Resonance Transmission Analysis with the Dense Plasma Focus

Simulations were performed using the Monte Carlo radiation transport code MCNP in order to analyze the feasibility of using a dense plasma focus to perform neutron resonance transmission analysis. It was found that the dense plasma focus is well-suited for neutron resonance transmission analysis because of the relatively short pulse duration and large neutron yield. This makes lowenergy neutron spectroscopy using time of flight much more accurate, allowing for more detailed assays of smaller quantities of special nuclear material.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Simulations of Lithium-Based Neutron Coincidence Counter for Gd-Loaded Fuel

The Department of Energy Office of Nuclear Safeguards and Security (NA-241) is supporting the project Lithium-Based Alternative Neutron Detection Technology Coincidence Counting for Gd-loaded Fuels at Pacific Northwest National Laboratory for the development of a lithium-based neutron coincidence counter for nondestructively assaying Gd loaded nuclear fuel. This report provides results from MCNP simulations of a lithium-based coincidence counter for the possible measurement of Gd-loaded nuclear fuel. A comparison of lithium-based simulations and UNCL-II simulations with and without Gd loaded fuel is provided. A lithium-based model, referred to as PLNS3A-R1, showed strong promise for assaying Gd loaded fuel.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2015

Results for Phase I of the IAEA Coordinated Research Program on HTGR Uncertainties

The quantification of uncertainties in design and safety analysis of reactors is today not only broadly accepted, but in many cases became the preferred way to replace traditional conservative analysis for safety and licensing analysis. The use of a more fundamental methodology is also consistent with the reliable high fidelity physics models and robust, efficient, and accurate codes available today. To facilitate uncertainty analysis applications a comprehensive approach and methodology must be developed and applied. High Temperature Gas-cooled Reactors (HTGR) has its own peculiarities, coated particle design, large graphite quantities, different materials and high temperatures that also require other simulation requirements. The IAEA has therefore launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modeling (UAM) in 2013 to study uncertainty propagation specifically in the HTGR analysis chain. Two benchmark problems are defined, with the prismatic design represented by the General Atomics (GA) MHTGR-350 and a 250 MW modular pebble bed design similar to the HTR-PM (INET, China). This report summarizes the contributions of the HTGR Methods Simulation group at Idaho National Laboratory (INL) up to this point of the CRP. The activities at INL have been focused so far on creating the problem specifications for the prismatic design, as well as providing reference solutions for the exercises defined for Phase I. An overview is provided of the HTGR UAM objectives and scope, and the detailed specifications for Exercises I-1, I-2, I-3 and I-4 are also included here for completeness. The main focus of the report is the compilation and discussion of reference results for Phase I (i.e. for input parameters at their nominal or best-estimate values), which is defined as the first step of the uncertainty quantification process. These reference results can be used by other CRP participants for comparison with other codes or their own reference results. The status on the Monte Carlo modeling of the experimental VHTRC facility is also discussed. Reference results were obtained for the neutronics stand-alone cases (Ex. I-1 and Ex. I-2) using the (relatively new) Monte Carlo code Serpent, and comparisons were performed with the more established Monte Carlo codes MCNP and KENO-VI. For the thermal-fluids stand-alone cases (Ex. I-3 and I-4) the commercial CFD code CFX was utilized to obtain reference results that can be compared with lower fidelity tools.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Impact of ASTM Standard E722 update on radiation damage metrics

The impact of recent changes to the ASTM Standard E722 is investigated. The methodological changes in the production of the displacement kerma factors for silicon has significant impact for some energy regions of the 1-MeV(Si) equivalent fluence response function. When evaluating the integral over all neutrons energies in various spectra important to the SNL electronics testing community, the change in the response results in an increase in the total 1-MeV(Si) equivalent fluence of 2 7%. Response functions have been produced and are available for users of both the NuGET and MCNP codes.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2015

A Methodology for Calculating Radiation Signatures

A rigorous formalism is presented for calculating radiation signatures from both Special Nuclear Material (SNM) as well as radiological sources. The use of MCNP6 in conjunction with CINDER/ORIGEN is described to allow for the determination of both neutron and photon leakages from objects of interest. In addition, a description of the use of MCNP6 to properly model the background neutron and photon sources is also presented. Examinations of the physics issues encountered in the modeling are investigated so as to allow for guidance in the user discerning the relevant physics to incorporate into general radiation signature calculations. Furthermore, examples are provided to assist in delineating the pertinent physics that must be accounted for. Finally, examples of detector modeling utilizing MCNP are provided along with a discussion on the generation of Receiver Operating Curves, which are the suggested means by which to determine detectability radiation signatures emanating from objects.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Radiation Characterization Summary: ACRR Polyethylene-Lead-Graphite (PLG) Bucket Located in the Central Cavity on the 32-Inch Pedestal at the Core Centerline (ACRR-PLG-CC-32-cl).

This document presents the facility-recommended characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the Annular Core Research Reactor (ACRR) for the polyethylene-lead-graphite (PLG) bucket in the central cavity on the 32-inch pedestal at the core centerline. The designation for this environment is ACRR-PLG-CC-32-cl. The neutron, prompt gamma-ray, and delayed gamma-ray energy spectra, uncertainties, and covariance matrices are presented as well as radial and axial neutron and gamma-ray fluence profiles within the experiment area of the bucket. Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse operations are presented with conversion examples. Acknowledgements The authors wish to thank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work. Also thanks to David Ames for his assistance in running MCNP on the Sandia parallel machines.

Vehar, David W.↗
DOE OSTI2014

In-Situ Gamma-Ray Assay of the West Cell Line in the 235-F Plutonium Fuel Form Facility

On August 29 th , 2013, scientists from SRNL took a series of in-situ gamma-ray measurements in the maintenance trench beneath Cells 6-9 on the west line of the PuFF facility using an uncollimated, high-purity germanium detector. The detector efficiency was estimated using a combination of MCNP simulations and empirical measurements. Data analysis was performed using three gamma-rays emitted by Pu-238 (99.85 keV, 152.7 keV, and 766.4 keV) providing three independent estimates of the mass of Pu-238 holdup in each of the cells. The weighted mean of these three results was used as the best estimate of Pu-238 holdup in the West Cell Line of PuFF. The results of the assay measurements are found in the table below along with the results from the scoping assay performed in 2006. All uncertainties in this table (as well as the rest of the report) are given as 1σ. The total holdup in the West Cell Line was 2.4 ± 0.7 grams. This result is 0.6 g higher than the previous estimate, a 0.4σ difference.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

In-Situ Gamma-Ray Assay of the East Cell Line in the 235-F Plutonium Fuel Form Facility

On September 17 th -19 th , 2013, scientists from SRNL took a series of in-situ gamma-ray measurements in the maintenance trench beneath Cells 1-5 on the east line of the PuFF facility using a well-collimated, high-purity germanium detector. The cell interiors were assayed along with the furnaces and storage coolers that protrude beneath the cells. The detector efficiency was estimated using a combination of MCNP simulations and empirical measurements. Data analysis was performed using three gamma-rays emitted by Pu-238 (99.85 keV, 152.7 keV, and 766.4 keV) providing three independent estimates of the mass of Pu-238 holdup in each of the cells. The weighted mean of these three results was used as the best estimate of Pu-238 holdup in the East Cell Line of PuFF. The results of the assay measurements are found in the table on the following page along with the results from the scoping assay performed in 2006. All uncertainties in this table (as well as the rest of the report) are reported at 1σ. Summing the assay results and treating MDAs as M 238Pu = 0 ± MDA, the total holdup in the East Cell Line was 240 ± 40 grams. This result is 100 grams lower than the previous estimate, a 0.55σ difference. The uncertainty in the Pu-238 holdup is also reduced substantially relative to the 2006 scoping assay. However, the current assay results are in agreement with the 2006 scoping assay results due to the large uncertainty associated with the 2006 scoping assays. The current assay results support the conclusion that the 2006 results bound the Pu-238 mass in Cells 1-5. These results should be considered preliminary since additional measurements of the East Cell line are scheduled for 2017 and 2018. Those measurements will provide detailed information about the distribution of Pu-238 in the cells to be used to refine the results of the current assay.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO 2 particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the power coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.

Connaway, H. M.↗
DOE OSTI2015

Aqueous Solution Vessel Thermal Model Development II

The work presented in this report is a continuation of the work described in the May 2015 report, “Aqueous Solution Vessel Thermal Model Development”. This computational fluid dynamics (CFD) model aims to predict the temperature and bubble volume fraction in an aqueous solution of uranium. These values affect the reactivity of the fissile solution, so it is important to be able to calculate them and determine their effects on the reaction. Part A of this report describes some of the parameter comparisons performed on the CFD model using Fluent. Part B describes the coupling of the Fluent model with a Monte-Carlo N-Particle (MCNP) neutron transport model. The fuel tank geometry is the same as it was in the May 2015 report, annular with a thickness-to-height ratio of 0.16. An accelerator-driven neutron source provides the excitation for the reaction, and internal and external water cooling channels remove the heat. The model used in this work incorporates the Eulerian multiphase model with lift, wall lubrication, turbulent dispersion and turbulence interaction. The buoyancy-driven flow is modeled using the Boussinesq approximation, and the flow turbulence is determined using the k-ω Shear-Stress-Transport (SST) model. The dispersed turbulence multiphase model is employed to capture the multiphase turbulence effects.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Neutronics Conversion Analyses of the Laue-Langevin Institute (ILL) High Flux Reactor (RHF)

The following report describes the neutronics results obtained with the MCNP model of the RHF U7Mo LEU reference design that has been established in 2010 during the feasibility analysis. This work constitutes a complete and detailed neutronics analysis of that LEU design using models that have been significantly improved since 2010 and the release of the feasibility report. When possible, the credibility of the neutronics model is tested by comparing the HEU model results with experimental data or other codes calculations results. The results obtained with the LEU model are systematically compared to the HEU model. The changes applied to the neutronics model lead to better comparisons with experimental data or improved the calculation efficiency but do not challenge the conclusion of the feasibility analysis. If the U7Mo fuel is commercially available, not cost prohibitive, a back-end solution is established and if it is possible to manufacture the proposed element, neutronics analyses show that the performance of the reactor would not be challenged by the conversion to LEU fuel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗