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At least 919 records · Page 51
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Analysis of Approximations in Modeling of BWR Bundle Void Distributions

In boiling water reactors, complex heterogeneous bundle designs, control blades adjacent to the corner of bundles, and the presence of boiling can lead to complex internal void distributions. A few approximations exist to model these void distributions. They could be modeled using a 1D axial solver in which each axial node is assumed to be at an average void, or each pin cell could be modeled with its own void concentration. In the latter case, the void could be discretized in pin-centered or coolant-centered channels. The goal of this project was to quantify the effect of using the different approximations for modeling internal void distributions on neutronics calculations. Using 3D void distributions calculated with CTF, Monte Carlo Neutral Particle (MCNP) transport code models were created for GE-9 and GE-14 lattices. For each model, the internal void distribution from CTF at a given axial node was selected, and a lattice calculation was carried out with MCNP. Comparisons between models using a lattice-averaged void, or using a void distribution in coolant-centered channels, showed large differences in reactivity which in some cases were well above 1,000 pcm, and it also showed differences in normalized fission rates greater than 20%. It was also found that using a lattice average void can lead to a significant difference in the estimation of the worth of a control blade. The differences found when comparing results from models using pin-centered and coolant-centered channels were up to 200 pcm in reactivity and up to 1.4% in the normalized fission rates. In addition to these two sets of comparisons, MCNP models were set up so that each subchannel had a saturated liquid component around the fuel pins and a saturated vapor component in the center to approximate annular flow. In comparison to the models using coolant-centered subchannels, up to 1–3% differences in normalized fission rates could be found.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Verification of Boundary Conditions in MCNP

MCNP (Monte Carlo N-Particle) is code that is used to model the transport of different particles. To verify in MCNP6.1 that arrays formed with reflective and periodic boundary conditions produce a neutron multiplication factor (keff) that is equal to or higher than an array of the same size with no boundary conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Radiation Damage Calculation Methodology

This report outlines the consensus radiation damage calculation methodology used by the Neutronics Analysis Group at Idaho National Laboratory. This methodology includes calculation of radiation damage rates via the Monte Carlo N-Particle (MCNP) code, and calculation of radiation damage in mixed materials, such as alloys. A consistent method for calculating the average displacement threshold energy of a mixed material has been established by the Neutronics Analysis Group. No standard method currently exists to account for radiation damage caused by neutron-induced transmutations. Only one documented case addresses transmutation radiation damage in materials containing nickel, and that method was incorporated by the Neutronics Analysis Group into their radiation damage calculation methodology.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Hand Calculation Methods for Nuclear Criticality Safety

This primer provides an overview of the most common hand calculation methods used for criticality safety calculations. The most widely used tools available to a nuclear criticality safety (NCS) practitioner are probably the common Monte Carlo or deterministic criticality safety codes, which can be used to model very complex systems. However, use of these codes can obscure the parameters to which a particular fissile system may be sensitive, whereas the hand calculation methods can be used to delve into the ways each parameter may affect the reactivity of a fissile material system. Furthermore, practitioners must avoid using computer codes as devices that take inputs and simply provide outputs (i.e., a “black box”). Many years ago, pioneers such as Joe Thomas, David Smith, and Hugh Paxton, among others in the field of nuclear criticality safety, took the time before the advent of high-speed desktop computers to create simple hand methods for criticality safety analyses. Some of the methods can be used for single fissile units; others are applicable to fissile units arranged into simple array configurations. This primer discusses the applicability of the various methods, illustrates how they are used, and provides an interpretation of the various results. The NCS practitioner will need to spend time to master the methods that could be most useful; however, they can provide the practitioner with fast and accurate answers to criticality safety problems if they are used correctly and if critical data exist for the problem at hand. Hand calculation methods can be used as a starting point for more advanced calculations, and in many circumstances, they can provide sensitivity and perturbation information more quickly than using a criticality code.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Niowave Neutron Source Converter: Lead-Bismuth-Eutectic (LBE) Windowless Target Design and Evaluation

Los Alamos National Laboratory (LANL) is working with Niowave on the design and evaluation of their lead-bismuth-eutectic (LBE) windowless target (i.e., neutron source converter). Niowave plans to use 200 kW electron beam at 40 MeV beam energy to produce neutrons by photonuclear reaction with LBE. Then, the neutrons undergo fission at the surrounding uranium target assembly (UTA) to produce Molybdenum 99 (Mo-99) as a fission product, which eventually decays to Technetium-99 (Tc-99m). Tc-99m is one of important radioisotopes that is used for medical diagnostics. LANL conducted 3D multiphysics analysis for the Niowave neutron converter design and provided design assessment in thermohydraulic aspects. LANL conducted radiation transport calculations using Monte-Carlo N-Particle (MCNP) code with unstructured meshing scheme. The 3D volumetric heating profiles in the LBE and Stainless-Steel (SS) housing were imported into multiphase computational fluid dynamics (CFD) to obtain 3D temperature profiles of LBE and SS through conjugate heat transfer (CHT) analysis. The key findings are: LBE film thickness at the center of the beam is approximately 1.6 cm with a maximum LBE velocity of approximately 1.8 m/s, which is below a 2 m/s limit to avoid erosion issues on supporting structures; Heat deposition in the LBE peaks at ~1 cm depth from the LBE free-surface because of the forward interactions of electron, photon, and neutron with LBE; LBE maximum temperature is ~360 °C, which is below LBE evaporation initiative temperature, ~450 °C; LBE-SS interface temperature is ~350 °C, which is below the safety thermal limit to prevent severe corrosion on SS. The results indicate that Niowave’s neutron converter design satisfies both hydraulic and thermal criteria for safe operation. By virtue of such computational analysis, Niowave can move toward establishing an experimental setup to experimentally test their LBE neutron converter. The following sections describe the detailed work done by LANL.

43 PARTICLE ACCELERATORS↗
DOE OSTI2025

Characterization of Gas Bremsstrahlung sources at the Electron-Ion Collider

This report estimates the potential impact of Gas Bremsstrahlung, generated as the electron beam traverses the straight sections of the storage ring and impinges onto selected locations of the Electron-Ion Collider. The source terms were calculated for routine operation scenarios at three different energies: 5, 10 and 18 GeV, using the FLUKA Monte Carlo particle transport and interaction code. The source term’s power distributions were validated and compared with the Gas Bremsstrahlung produced in the long straight sections of the National Synchrotron Light Source II storage ring.

43 PARTICLE ACCELERATORS↗
DOE OSTI2026

Rapid Gamma Simulations of TRISO Fuel Elements

As energy demand rises, nuclear energy, particularly from reactors that use tristructural isotropic (TRISO) fuels, has gained attention due to the fuel’s enhanced resistance to radiation damage and high temperatures. This report investigates the modeling capabilities of the Gamma Detector Response and Analysis Software (GADRAS) for TRISO fuels, focusing on the gamma signatures of TRISO particles, which have not been extensively explored. Using the Monte Carlo N-Particle (MCNP) code as a benchmark, we developed both homogeneous and heterogeneous models of TRISO pebbles to analyze gamma spectra. Our findings reveal that the homogeneous and heterogeneous models produced different gamma signatures. Additionally, the GADRAS heterogeneous model significantly reduces computation times compared to MCNP, enabling effective modeling of gamma signatures for safeguards applications. This advancement is essential for the International Atomic Energy Agency (IAEA) in detecting anomalies and potential smuggling attempts in TRISO reactor fuel elements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Estimation of radiation fields generated by injected beam losses at the EIC's RCS

This technical note provides a general estimate of radiation fields generated by injection fault events at the electron-ion collider´s (EIC) Rapid Cycling Synchrotron (RCS), calculated with the Monte Carlo particle transport and interaction code FLUKA. Calculations were performed for two major injection loss scenarios that involve iron targets and featured different electron beam energy and current values. The results presented here constitute a first order assessment of several radiological quantities associated with these electromagnetic showers and their potential effect on environmental safety and health (ESH) systems in the vicinity of injection areas.

43 PARTICLE ACCELERATORS↗
DOE OSTI2022

Efficiency Studies of Fast Neutron Tracking Using MCNP

Fast neutron identification and spectroscopy is of great interest to nuclear physics experiments. Using the neutron elastic scattering, the fast neutron momentum can be measured. Wang and Morris introduced the theoretical concept that the initial fast neutron momentum can be derived from up to three consecutive elastic collisions between the neutron and the target, including the information of two consecutive recoil ion tracks and the vertex position of the third collision or two consecutive elastic collisions with the timing information. Here, we also include the additional possibility of measuring the deposited energies from the recoil ions. In this paper, we simulate the neutron elastic scattering using the Monte Carlo N-Particle Transport Code (MCNP) and study the corresponding neutron detection and tracking efficiency. The corresponding efficiency and the scattering distances are simulated with different target materials, especially natural silicon (92.23%28Si, 4.67%29Si, and 3.1%30Si) and helium-4 (4He). The timing of collision and the recoil ion energy are also investigated, which are important characters for the detector design. We also calculate the ion traveling range for different energies using the software, “The Stopping and Range of Ions in Matter (SRIM)”, showing that the ion track can be most conveniently observed in 4He unless sub-micron spatial resolution can be obtained in silicon.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1975

Solar-wind and solar-flare maturation of the lunar regolith

Detailed predictions concerning the depth and time-dependent accumulation of solar-wind effects and solar-flare tracks in lunar dust grains were obtained by use of an adaptation of a Monte Carlo soil-mixing computer code described by Duraud et al. (1975). The predictions are compared to experimental measurements obtained by analyzing lunar dust grains as well as artificially irradiated minerals by a variety of techniques. A study of amorphous coatings of solar-wind radiation-damaged material on certain lunar grains sets limits on the integrated residence time of these grains in the ancient solar wind. Other topics discussed include solar wind maturation, the peculiar shape of the experimental distribution of central track densities in 50-micron grains, and the interpretation of both track gradients in 50-micron feldspars and the relatively 'low' concentration of solar-wind species implanted in ilmenite grains.

Bibring, J. P.↗
NASA NTRS1991

Monte Carlo analysis of dissociation and recombination behind strong shock waves in nitrogen

Computations are presented for the relaxation zone behind strong, 1D shock waves in nitrogen. The analysis is performed with the direct simulation Monte Carlo method (DSMC). The DSMC code is vectorized for efficient use on a supercomputer. The code simulates translational, rotational and vibrational energy exchange and dissociative and recombinative chemical reactions. A model is proposed for the treatment of three body-recombination collisions in the DSMC technique which usually simulates binary collision events. The model improves previous models because it can be employed with a large range of chemical-rate data, does not introduce into the flow field troublesome pairs of atoms which may recombine upon further collision (pseudoparticles) and is compatible with the vectorized code. The computational results are compared with existing experimental data. It is shown that the derivation of chemical-rate coefficients must account for the degree of vibrational nonequilibrium in the flow. A nonequilibrium-chemistry model is employed together with equilibrium-rate data to compute the flow in several different nitrogen shock waves.

Boyd, I. D.↗
NASA NTRS1998

Ly(alpha) Photolysis in the Primitive Solar Nebula

This is the final report for the third year of work on this project. Our proposal was to quantitatively investigate the importance of photochemistry in the solar nebula. In the generally accepted theory for the chemical evolution of the primitive solar nebula, Prinn and Fegley argued that photochemistry is unimportant, and that thermochemistry controls the relative abundances of molecular species throughout the planet-forming region. They provided useful estimates of the chemical energy available to the solar nebula from a variety of sources, and established that even the small photolysis rate due to starlight is more important than the photolysis rate from direct sunlight (although small, the UV flux from starlight could have processed a non-negligible fraction of the solar nebula. The reason for this is that the opacity of the disk was so large that direct sunlight could only penetrate to 0.1 AU or so, despite the expectation that the protosun, if comparable to a T-Tauri star, would be emitting up to 10(exp 4) more H I Ly(alpha) photons than the current sun. We developed a Monte Carlo resonance fine radiative transfer code, capable of accurately calculating the radiation field of H I Ly(alpha), He I 584 A, and He II 304 A emissions throughout the nebula and the nearby interstellar medium in which it is embedded. We applied the code to two appropriate models of the primitive solar nebula. Our model provided the photolysis rates of various species over the entire surface layer of the nebula, and from this we evaluated the importance of UV photochemistry due to backscattered solar UV resonance line emissions on different parts of the nebula. The results discussed below were presented.

Gladstone, G. Randall↗
NASA NTRS2002

System Level Uncertainty Assessment for Collaborative RLV Design

A collaborative design process utilizing Probabilistic Data Assessment (PDA) is showcased. Given the limitation of financial resources by both the government and industry, strategic decision makers need more than just traditional point designs, they need to be aware of the likelihood of these future designs to meet their objectives. This uncertainty, an ever-present character in the design process, can be embraced through a probabilistic design environment. A conceptual design process is presented that encapsulates the major engineering disciplines for a Third Generation Reusable Launch Vehicle (RLV). Toolsets consist of aerospace industry standard tools in disciplines such as trajectory, propulsion, mass properties, cost, operations, safety, and economics. Variations of the design process are presented that use different fidelities of tools. The disciplinary engineering models are used in a collaborative engineering framework utilizing Phoenix Integration's ModelCenter and AnalysisServer environment. These tools allow the designer to join disparate models and simulations together in a unified environment wherein each discipline can interact with any other discipline. The design process also uses probabilistic methods to generate the system level output metrics of interest for a RLV conceptual design. The specific system being examined is the Advanced Concept Rocket Engine 92 (ACRE-92) RLV. Previous experience and knowledge (in terms of input uncertainty distributions from experts and modeling and simulation codes) can be coupled with Monte Carlo processes to best predict the chances of program success.

Charania, A. C.↗
NASA NTRS2013

Modeling Analysis for NASA GRC Vacuum Facility 5 Upgrade

A model of the VF5 test facility at NASA Glenn Research Center was developed using the direct simulation Monte Carlo Hypersonic Aerothermodynamics Particle (HAP) code. The model results were compared to several cold flow and thruster hot fire cases. The main uncertainty in the model is the determination of the effective sticking coefficient -- which sets the pumping effectiveness of the cryopanels and oil diffusion pumps including baffle transmission. An effective sticking coefficient of 0.25 was found to provide generally good agreement with the experimental chamber pressure data. The model, which assumes a cold diffuse inflow, also fared satisfactorily in predicting the pressure distribution during thruster operation. The model was used to assess other chamber configurations to improve the local effective pumping speed near the thruster. A new configuration of the existing cryopumps is found to show more than 2x improvement over the current baseline configuration.

cryopumping↗
DOE OSTI2017

A Versatile Coupled Test Reactor Concept

A Versatile Coupled Test Reactor (VCTR) is a highly reconfigurable, sodium-cooled, coupled fast and thermal spectrum test reactor, which provides (1) a fast flux irradiation environment prototypical of potential fast reactor designs, (2) thermal and epithermal flux irradiation environments complementary to those of ATR and HFIR, and (3) other possibilities, including beam tubes for scientific experiments, irradiation vehicles for isotope production, etc. The ongoing design of a new versatile coupled thermal-fast test reactor at INL that can accommodate both fast and thermal irradiation, technology and safety features tests is described in this paper. A coupled reactor is defined as a reactor with two distinct spectral zones (Fast and Thermal), which are neutronically coupled to each other: some neutrons born in zone F cause fission in zone T and vice-versa. Only fast neutrons are allowed to go from one zone to the other, which is assured by a thermal neutron filter. A distinction of coupling can be made depending on the main purpose of using fuel assemblies with some level of neutron moderation (thermal zone) together with fuel assemblies with as little moderation as possible (fast zone). It can be used mainly to minimize the core power and fissile inventory, or it can be used mainly to provide additional control of the irradiation conditions in the thermal zone. The steady-state neutronics calculations were performed using the three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code Serpent2. Although, a few changes were necessary to support the Serpent code in calculating the coupling coefficients. Steady state and transient thermal analyses were performed with RELAP5-3D. RELAP5-3D is the latest in the RELAP5 code series developed at INL. One of the core configurations using only low enriched uranium fuel could provide both a high fast neutron flux and a high thermal neutron flux—respectively 3.5E15 n/s.cm2 and 1E15 n/s.cm2—in large volumes while maintaining core power below 300 MW. The use of plutonium in the fast zone provides more flexibility in fuel assembly design than when low enriched uranium is used, the designer has more degrees of freedom to work with; in particular it could provide additional control of the irradiation conditions in the thermal zone. The results showed that the current design fulfills most R&D needs requirements providing high thermal and fast fluxes and wide possibilities for diversified future reactor features test in appropriate loops, both in steady state and during transients.

coupled spectrum↗
DOE OSTI2019

General Handling Calculations for the Modified Hospital Can

Containers are evaluated using a Monte-Carlo N-Particle transport code (MCNP). The code calculated keff is compared to critical experiment benchmarks to determine a bias. Additional margin of safety is added to the bias to establish an upper subcritical limit (USL). Modeled cases are considered subcritical if the neutron multiplication factor plus two sigma remains below the USL. Net weight limits allow for any amount of fissile/moderator mixture to exist so long as it meets the weight requirement, therefore all mixture types must be evaluated. Cases that exceed the USL must either be controlled against or have justification why that configuration cannot reasonably exist. Current and legacy loadings and container dimensions must be taken into consideration.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

As-Run Physics Analysis for the AGC-4 Experiment Irradiated in the ATR

This Engineering Calculations Analysis Report (ECAR) documents the results of the Advanced Test Reactor (ATR) detailed physics analyses performed to calculate the displacements per atom (DPA) and the fast neutron fluence (E > 0.1 MeV) of the Advanced Graphite Creep (AGC) experiment, AGC-4, irradiated in the ATR East Flux Trap (EFT) (see Figure 1) during ATR Cycle 157D, 158A, 162A, 162B, 164A, 164B, 166A, and Cycle 166B. This ECAR also reports the neutron and photon heat rates for the materials of the AGC-4 experiment for ATR Cycle 158A (timestep 19), which provides to the maximum heating. The results for these evaluations and analysis are reported herein. The AGC-4 as-run specimen neutron fast fluence (E > 0.1 MeV), DPA, and material heat rate calculations were performed using a general-purpose Monte Carlo N-Particle (MCNP) code. All calculated results are tabulated herein.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Investigating Fission Reaction Rate Ratio Sensitivities [Abstract]

Reaction rate ratios are a measurable parameter for reactor and criticality applications. A number of foil irradiations and fission chamber measurements have been performed for critical assemblies at Los Alamos National Laboratory starting in the 1950’s including (i) Godiva, a bare HEU spherical assembly; (ii) Flattop-25, a spherical assembly consisting of an HEU core and a natural uranium reflector; (iii) Jezebel, a bare 239 Pu assembly; and (iv) Flattop-Pu, a spherical assembly consisting of a 239Pu core and a natural uranium reflector. Fission ratio data for 238 U(n,f)/ 235 U(n,f), 237 Np(n,f)/ 235 U(n,f), 233 U(n,f)/ 235 U(n,f) and 239 Pu(n,f)/ 235 U(n,f) were obtained and reported. The EUCLID (Experiments Underpinned by Computational Learning for Improvements in nuclear Data) project at Los Alamos National Laboratory (LANL) aims to constrain nuclear data by using a suite of measurement types beyond k-effective. Recent investigations include the use of pulsed spheres for nuclear data validation and other measurement methods of interest. One focus of the work is to determine if other methods are complimentary to the critical experiments utilized for nuclear data validation. It is anticipated the investigations will help inform methods that may be utilized in machine learning algorithms for nuclear validation. In order to use a measurement type for nuclear validation, it is necessary to obtain cross-section sensitivities for parameters. This work looks at reaction rate ratio sensitivities with SENSMG and Monte Carlo N-Particle R Code Version 6.21.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗