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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 1,801 records · Page 100
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DOE OSTI2017

Flux in a 1D-Reactor Cell by Response Matrix

Legacy computational reactor physics software tools and protocols currently used in support of the Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice. Thus, they are difficult, if not impossible, to verify and validate (V and V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of their application and protocols from the 1960's and 1970's is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity transport theory based computational software and models to replace the legacy methods, with appropriate (V and V) protocols according to applicable national standards. The new computational tools include: - The 2-dimensional HELIOS deterministic transport code for integrated ATR core fuel and experiment management and Core Safety Analysis Package (CSAP) physics data generation. The MCNP5 3-dimensional stochastic simulation (Monte Carlo) code for verification and validation applications and for specialized ATR core analysis along with its longstanding traditional role in experiment support - The 2-dimensional deterministic NEWT transport code and the 3-dimensional KENO6 stochastic simulation code, with their SCALE-based support infrastructure for use in V and V applications and as supporting modules for sensitivity-uncertainty analysis via the TSUNAMI system. The ATTILA 3-dimensional discrete-ordinates transport code for specialized experiment support and possibly additional applications. Regardless of application, every neutron transport code, new or legacy, requires verification. The aim of this presentation is to provide a neutron transport benchmark for a simple 1D-reactor cell in plane geometry. The benchmark finds use as a standard of comparison for the codes listed above intended for the ATR upgrade. The method developed here is in contrast to the finite Fourier transform solution found in Ref. 1, which is limited to isotropic scattering and particularly difficult to numerically implement unless the total cross section is constant over the cell.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Sensitivity Coefficients for Diffusion Coefficients and Other Reactor Physics Parameters using CE TSUNAMI-3D

The TSUNAMI (Tools for Sensitivity and Uncertainty Analysis Methodology Implementation) capabilities within the SCALE code system leverage sensitivity coefficients for many criticality safety applications, including quantifying the data-induced uncertainty in the eigenvalue of critical systems, quantifying computational biases, and selecting applicable benchmark experiments. TSUNAMI sensitivity and uncertainty analysis tools have traditionally been limited to criticality safety application spaces, but the introduction of Generalized Perturbation Theory (GPT) sensitivity methods by Jessee in 2009 and Perfetti in 2013 has extended the scope of SCALE adjoint-weighted sensitivity and uncertainty analysis to include reactor physics applications. The continuous-energy (CE) TSUNAMI-3D GPT sensitivity analysis methods were originally developed to calculate sensitivity coefficients for a limited number of reaction-rate responses using CE Monte Carlo methods. These tools were recently applied in an Oak Ridge National Laboratory (ORNL)-directed research and development project to improve the efficiency of {sup 252}Cf production in the High Flux Isotope Reactor (HFIR) by understanding the sensitivity of key heavy actinide capture-to- fission ratios to potential system design changes. This work further extends the applicability of the CE TSUNAMI- 3D GPT methods to reactor physics applications by enabling the calculation of sensitivity coefficients for an expanded set of reaction rate responses-including a first-ever capability for conducting sensitivity analysis of neutron diffusion coefficients.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Distinguishing Research Reactor Fuels Using Neutron Multiplicity - 25615

Neutron measurements are one of the reliable means of interrogation of fissile material. Neutron measurements are used for total counting (pure metal fissile samples) or counting coincidence/multiplicity (for impure fissile samples). Various flavors of neutron measurement are used routinely by the IAEA to determine the fissile mass of an unknown sample. The objective of this work is to design an experiment to distinguish the two different fuel types using neutron multiplicity method. The design of experiment is performed using Monte Carlo N- Particle (MCNP) transport code. General Electric (GE) designed 1-MWt pool-type research reactor at UMass Lowell (UMLRR) uses Materials Testing Reactor (MTR) plate-type fuel consisting of 19 full and 2 partial fuel assemblies, controlled by four boron control blades and one regulating control rod. Two different fuel types are available at the UMLRR facility. In 2008, UMass Lowell obtained slightly burnt fuel assemblies from the decommissioned WPI research reactor. The simulation results discussed in this paper show that within the limits of statistical error, the single and double moments for the WPI and UMLRR fuels can be measured. The two fuels have very similar composition and enrichments resulting in very similar neutron multiplicity moments. This implies that neutron multiplicity measurements alone would not be sufficient to distinguish between the two fuel types at low burnups. The change in the observed single-double ratio between each fuel type is less than 6 % which falls below the uncertainty in the efficiency of the JCC71 slab detectors. The neutron signature from low burnup and low enriched uranium fuel are not strong, however, these can be measured under low background and long measurement times. This would be an important consideration in designing the experiments and parallel gamma measurements will also be planned. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Neutronic design and fuel cycle analysis of a fluoride salt-cooled High Temperature Reactor (FHR)

The Fluoride salt-cooled High temperature Reactor (FHR) is a new Gen IV reactor concept that can operate under near atmospheric pressure circumstances and further enhance the inherent safety. In this study, an FHR core design with 165 megawatts of thermal output (MWth) is proposed. The reactor core employs tri-structural-isotropic (TRISO) particle fuel within prismatic graphite blocks as the basic fuel form, FLiBe (2 {sup 7}LiF-BeF{sub 2}) as the primary coolant, and a three-batch fuel cycle scheme. Sensitivity analyses on various parameters were performed to optimize the cycle length and neutronic parameters. The fuel cycle of this core design was evaluated in detail from 4 aspects: cycle length, power peaking factor (PPF), discharge burnup, and temperature coefficient. It was found that larger fuel channel pitch (FCP) would have a relatively harder neutron spectrum, and yield relatively longer cycle length, lower, better fuel temperature coefficient (FTC), and moderator temperature coefficient (MTC). In addition, burnable poison (BP), Er{sub 2}O{sub 3}, can effectively reduce PPF, hold down the multiplication factor, and more importantly it can improve the MTC. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Approach to nuclear criticality in IAN-R1 research reactor

The Republic of Colombia is operating a small 100 kW(t) research reactor. The reactor was provided to Colombia under the U.S. Atoms for Peace Program, and which was fueled with MTR HEU fuel enriched nominally to 93% U{sup 235}. With the cooperation of the International Atomic Energy Agency IAEA, it was prepared a new safety analysis report for performing an HEU to LEU conversion of the IAN-R1 reactor and manufacture TRIGA type LEU (19.7%, enriched) fuel to replace the original MTR-HEU fuel plate assemblies. This paper describes the activities which were achieved during the approach to nuclear criticality. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Development of Data Reporting Standards for High-Temperature Gas-cooled Reactor (HTGR) Nuclear Energy University Program (NEUP) Thermal-Fluid Experiments

Since 2009, the U.S. Department of Energy (DOE) Office of Nuclear Energy's Nuclear Energy University Program (NEUP) has been at the forefront of nuclear research, specifically concentrating on advancing high-temperature gas-cooled reactor (HTGR) technologies. By Fiscal Year 2023, NEUP has authorized 35 projects dedicated to HTGR research, each contributing significantly to the enhancement of our understanding of this technology. The outcomes of these diverse projects have been disseminated through final NEUP reports, peer-reviewed journal articles, and presentations at academic conferences, forming a comprehensive tapestry of knowledge. Despite the substantial value of these findings, their dissemination has been fragmented, posing challenges for accessibility to researchers and policymakers and leading to underutilization of DOE investments. Recognizing this critical gap and its potential consequences for the future of nuclear research, the Advanced Reactor Technologies (ART) Gas-Cooled Reactor (GCR) program conducted an extensive survey of completed and ongoing HTGR NEUP projects. This survey enabled the compilation of crucial data, resulting in the development of a specialized public-access database tailored for computational fluid dynamics and system code validation, specifically designed for HTGR applications. However, the data collection process revealed a significant challenge in central data organization due to individual researchers from different institutes employing varying logics and preferences for recording and documenting experimental data. Consequently, an urgent need has been identified to establish a standardized reporting format for HTGR experimental projects. Addressing this issue is essential for enhancing collaboration, maximizing the impact of DOE investments, and ensuring the seamless advancement of HTGR technologies in nuclear research.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Experimental Studies of Coolant Temperature and Velocity on an Integrated Water Model of a Fast Reactor in Different Operating Regimes

The results of experimental studies of a next-generation fast reactor performed using an integrated water model (scale 1:10) are presented: local velocity over height and radius of the top chamber in a plane in a direction from core center to the intermediate heat exchangers and coolant temperature in the top chamber and other elements of the circulation loop with forced circulation and emergency cool-down by natural convection. The data were obtained using a specially designed system for performing measurements which is incorporated in the bench and provides high accuracy and rate of acquisition. Stable temperature stratification of the coolant is observed in the top, cold, and pressure chambers of the reactor, elevator enclosure, and reactor vessel cooling system as well as at the exit from the intermediate and autonomous heat exchangers in different operating regimes.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Emergency Heat Removal System for the Secondary Loop of a Fast Sodium Reactor and its Efficiency Evaluation

An emergency heat removal system (SAOT) that removes heat from the secondary loop of a fast sodium reactor by air cooling the surfaces of the piping and equipment of the heat-sink loops of the secondary loop is described and ways to optimize it are indicated. A method of evaluating the efficiency of such an SAOT is described, and the results of an evaluation for a fast sodium reactor are presented as an example showing that the system can be used in fast sodium reactors of any power capacity.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Dynamics of Power Pulses in a Research Reactor with Neptunium Nuclear Fuel

The current trend in neutron-aided scientific research presupposes the development of new high-intensity pulsed sources. One variant of such an installation could be a pulsed reactor with a fissile-isotope core – {sup 237}Np. The present work is devoted to a study of the dynamics of power pulses of a neptunium reactor taking account of fast temperature feedback during pulse buildup. The problem is reduced to a simple model of a single frequency oscillator combined with the equation of single point kinetics of a reactor neglecting delayed neutrons. It is shown that on the basis of such an approximation, owing to feedback on the temperature of a fuel element the estimated energy of accidental power pulses decreases by almost a factor of 10 compared with a calculation neglecting fast feedback.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Problems of Radioactive Graphite Management During Decommissioning of Nuclear Reactors

The results of an inspection of samples of the graphite masonry from the RFT research reactor at the National Research Center Kurchatov Institute are presented. These studies made it possible to determine methods and procedures for extracting and packaging the radioactive graphite from the reactor pit during decommissioning. It is shown that management of radioactive graphite requires the development of methods and procedures for determining the radiation characteristics of graphite blocks, first and foremost, the development of highly efficient methods of measuring the specific activity of {sup 14}C. These methods will make it possible to separate radioactive graphite according to the level of contamination and to determine subsequent technologies for handling it during decommissioning of power reactors. It is noted that {sup 14}C can be used as a source of heat.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Coupled reactors analysis: New needs and advances using Monte Carlo methodology

Coupled reactors and the coupling features of large or heterogeneous core reactors can be investigated with the Avery theory that allows a physics understanding of the main features of these systems. However, the complex geometries that are often encountered in association with coupled reactors, require a detailed geometry description that can be easily provided by modern Monte Carlo (MC) codes. This implies a MC calculation of the coupling parameters defined by Avery and of the sensitivity coefficients that allow further detailed physics analysis. The results presented in this paper show that the MC code SERPENT has been successfully modifed to meet the required capabilities.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Physics analysis and design of heavy water reflected thermal test reactor

Here, this work investigates the option of modifying the Advanced Test Reactor by replacing the current beryllium reflector with heavy water. Such a change may provide some potential benefits for not only increasing the thermal irradiation capabilities but also resolving other problems such as reflector integrity issues due to fast fluence damage, which is always a limiting factor in the lifetime of the current beryllium reflector. This paper presents the analysis and estimation of the ATR core physics parameters by replacing the current beryllium reflector with heavy water (D 2 O). The paper first describes the details of two selected conceptual designs, which are partially reflected with either beryllium or graphite, and how they are derived from the baseline beryllium reflector concept. Then, reactor physics performance parameters for the two new concepts are assessed by comparing with those of the baseline concept. The performance parameters considered in this paper include in-pile tube neutron and gamma fluxes and heating rates, maximum loop voiding reactivity, core power behavior with different power splits, predicted cycle length with a given fuel loading, and thermal hydraulic analysis with a higher lobe power split. It is important to note that this study focuses on the reactor physics aspects and does not delve into the engineering challenges associated with such a design modification.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

At-power subcritical multiplication in the advanced test reactor

The degree to which a given physical subdivision of a critical reactor acts as a neutron source for other lobes is not analytically characterized for general application. ATR and ATRC both operate at power, and individual power-producing lobes rely on each other as neutron sources in order to maintain constant power, which in general requires either exactly critical multiplication within a reactor or an external neutron source. Here, this work shows that lobe powers in ATR or ATRC can be related with subcritical multiplication theory, demonstrating that the lobes are not individually critical. This explanation is significant for ATR due to the desire to irradiate a large variety of experiments simultaneously, each having its impact on the core neutron population. This work also establishes the lack of relationship between experiment reactivity worth and cross-sections calculated from fundamental nuclear data. For any physical subdivision of any other critical reactor, it is likewise true that the subdivision undergoes only subcritical multiplication.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Conceptual design of a new homogeneous reactor for medical radioisotope Mo-99/Tc-99m production

To partly solve the global and regional shortages of Mo-99 supply, a conceptual design of a nitrate-fuel-solution based homogeneous reactor dedicated for Mo-99/Tc-99m medical radioisotope production is proposed. The modified LEU Cintichem process for Mo-99 extraction which has been licensed and demonstrated commercially for decades by BATAN is taken into account as a key design consideration. The design characteristics and main parameters are identified and the advantageous aspects are shown by comparing with the BATAN's existing Mo-99 supply chain which uses a heterogeneous reactor (RSG GAS multipurpose reactor)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Thermal-Hydraulic and Neutronic Phenomena Important in Modeling and Simulation of Liquid-Fuel Molten Salt Reactors

In this paper, we discuss liquid-fuel molten salt (cooled) reactors (MSRs); how they will operate under normal, transient, and accident conditions; and the results of an expert elicitation to determine the corresponding thermal-hydraulic and neutronic phenomena important to understanding their behavior. Identifying these phenomena will enable the U.S. Nuclear Regulatory Commission (NRC), U.S. Department of Energy, and industry to develop or identify modeling functionalities and tools required to carry out confirmatory and licensing analyses that examine the validity and accuracy of an applicant's calculations and help determine the margin of safety in plant design. The NRC frequently does an expert elicitation using a Phenomena Identification and Ranking Table (PIRT) to identify and evaluate the state of knowledge of important modeling phenomena. However, few details about the design of these reactors and the sequence of events during accidents are known, so the process used was considered a preliminary PIRT. A panel comprising a group of subject matter experts met to define phenomena that would need to be modeled and considered the impact/importance of each phenomenon with respect to specific figures of merit (FoMs) (e.g., salt temperature, velocity, and composition). Each FoM reflected a potential impact on radionuclide release or loss of a barrier to release. The panel considered what the path forward might be with respect to being able to model the phenomenon in a simulation code. Results are explained for both thermal and fast spectrum designs, with an emphasis on the thermal-hydraulic takeaways. It was concluded that compared to light water reactors, the lack of high-pressure operation, energetic break flow, depressurization, and quench front tracking may simplify some aspects of an MSR analysis. However, MSRs have new phenomena both for a license applicant and NRC confirmatory analysis. There is a need for enhanced understanding of physical properties for MSRs that encompass several individual thermophysical properties, including thermal conductivity, viscosity, specific heat, density, optical properties, thermodynamic properties, volatilities, solubilities, etc. Salt composition is closely linked to both these properties and the neutronics of the system. Additionally, the large number of MSR concepts and system designs means that there is wide variation in the potential modeling needs for these systems.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Reactor Performance and Safety Characteristics of Beryllium-Based Composite Moderators as Replacements for Graphite in mHTGRs

Here, this study evaluates beryllium-based two-phase composite moderators as an alternative to graphite in an evaluation of reactor performance and safety characteristics. Historically, modular high-temperature gas-cooled reactors (mHTGRs) use graphite as a moderator because of its high moderating ratio and reasonable thermal properties; however, graphite has unfavorable properties under irradiation, which can require component replacement and a significant radioactive waste burden. In this assessment, we explore advanced moderators comprised of magnesium oxide (MgO) as the host matrix and beryllium metal and/or beryllium oxide (Be and/or BeO) as the entrained moderating phase. For the reactor performance and thermal-hydraulic safety analysis, the core design model of the General Atomics mHTGR-350 was used to demonstrate the feasibility of a “drop-in” replacement of graphite using the beryllium-based moderators. We employed the neutronics code Serpent to analyze the moderating behavior of the composite moderators with comparisons drawn to graphite. We performed a scoping analysis of accidents for mHTGRs using RELAP to show that these moderators do not present impediments to safety and are expected to stay within temperature limits. Measured thermophysical properties of the composite moderators are used in the thermal-hydraulic assessments. Our analysis reveals that the two-phase composite MgO-matrix beryllium-based moderators are a suitable replacement for graphite.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Application of MELCOR for Simulating Molten Salt Reactor Accident Source Terms

Molten Salt Reactor (MSR) systems can be divided into two basic categories: liquid-fueled MSRs in which the fuel is dissolved in the salt, and solid-fueled systems such as the Fluoride-salt-cooled High-temperature Reactor (FHR). The molten salt provides an impediment to fission product release as actinides and many fission products are soluble in molten salt. Nonetheless, under accident conditions, some radionuclides may escape the salt by vaporization and aerosol formation, which may lead to release into the environment. Here, we present recent enhancements to MELCOR to represent the transport of radionuclides in the salt and releases from the salt. Some soluble but volatile radionuclides may vaporize and subsequently condense to aerosol. Insoluble fission products can deposit on structures. Thermochimica, an open-source Gibbs Energy Minimization (GEM) code, has been integrated into MELCOR. With the appropriate thermochemical database, Thermochimica provides the solubility and vapor pressure of species as a function of temperature, pressure, and composition, which are needed to characterize the vaporization rate and the state of the salt with fission products. Since thermochemical databases are still under active development for molten salt systems, thermodynamic data for fission product solubility and vapor pressure may be user specified. This enables preliminary assessments of fission product transport in molten salt systems. In this paper, we discuss modeling of soluble and insoluble fission product releases in a MSR with Thermochimica incorporated into MELCOR. Separate-effects experiments performed as part of the Molten Salt Reactor Experiment in which radioactive aerosol was released are discussed as needed for determining the source term.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing {sup 233}U from {sup 232}Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

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