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At least 1,621 records · Page 90

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DOE OSTI2017

Conceptual Model of Iodine Behavior in the Subsurface at the Hanford Site

Isotopes of iodine were generated during plutonium production within the nine production reactors at the U.S. Department of Energy Hanford Site. The short half-life 131I that was released from the fuel into the atmosphere during the dissolution process (when the fuel was dissolved) in the Hanford Site 200 Area is no longer present at concentrations of concern in the environment. The long half-life 129I generated at the Hanford Site during reactor operations was (1) stored in single-shell and double-shell tanks, (2) discharged to liquid disposal sites (e.g., cribs and trenches), (3) released to the atmosphere during fuel reprocessing operations, or (4) captured by off-gas absorbent devices (silver reactors) at chemical separations plants (PUREX, B-Plant, T-Plant, and REDOX). Releases of 129I to the subsurface have resulted in several large, though dilute, plumes in the groundwater. There is also 129I remaining in the vadose zone beneath disposal or leak locations. The fate and transport of 129I in the environment and potential remediation technologies are currently being studied as part of environmental remediation activities at the Hanford Site. A conceptual model describing the nature and extent of subsurface contamination, factors that control plume behavior, and factors relevant to potential remediation processes is needed to support environmental remedy decisions. Because 129I is an uncommon contaminant, relevant remediation experience and scientific literature are limited. In addition, its behavior in subsurface is different from that of other more common and important contaminants (e.g., U, Cr and Tc) in terms of sorption (adsorption and precipitation), and aqueous phase species transformation via redox reactions. Thus, the conceptual model also needs to both describe known contaminant and biogeochemical process information and identify aspects about which additional information is needed to effectively support remedy decisions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Evaluation of Iodine Remediation Technologies in Subsurface Sediments (Interim Status Report)

Isotopes of iodine were generated during plutonium production from nine production reactors at the U.S. Department of Energy Hanford Site. The long half-life 129 I generated at the Hanford Site during reactor operations was 1) stored in single-shell and double-shell tanks, 2) discharged to liquid disposal sites (e.g., cribs and trenches), 3) released to the atmosphere during fuel reprocessing operations, or 4) captured by off-gas absorbent devices (silver reactors) at chemical separations plants (PUREX, B-Plant, T-Plant, and REDOX). Releases of 129 I to the subsurface have resulted in several large, though dilute, plumes in the groundwater, including the plume in the 200-UP-1 operable unit. There is also 129 I remaining in the vadose zone beneath disposal or leak locations. Because 129 I is an uncommon contaminant, relevant remediation experience and scientific literature are limited. Iodine cannot be degraded/transformed to a less toxic form. Therefore, the primary metric for remediation is changing the fate and transport properties in a way that works to decrease radioiodine concentrations in groundwater. To this end, targeted laboratory experiments were conducted to provide proof-of-principle screening of remedial options and to evaluate scale-up and performance elements. The evaluation of different iodine remediation technologies will occur over multiple years and fall into three different categories: 1) geochemical/biological sequestration, 2) microbial volatilization, and 3) enhanced pump-and-treat. This report provides status on two in situ sequestration methods (apatite and iron oxides) and two methods examining enhanced pump-and-treat.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NASA NTRS2022

Gas-Liquid Flows Through Porous Media in Microgravity: Packed Bed Reactor Experiment-2

Modifications were made to the Packed Bed Reactor Experiment (PBRE) and flown on the ISS as PBRE-2 to eliminate external pressure oscillations at higher liquid flow rates and the packing diameter was reduced to increase the pressure gradient for lower flows. It is found that gas hold-up is a function of bed history at low liquid and gas flow rates whereas higher gas hold-up and pressure gradients are observed for the test conditions following a liquid only pre-flow compared to the test conditions following a gas only pre-flow period. Over the range of flow rates tested, the capillary force is the dominant contributor to the pressure gradient, which is found to be linear with the superficial liquid velocity but is a much weaker function of the superficial gas velocity, and varying inversely with the particle diameter.

Mahsa Taghavi↗

NASA NTRS2010

A Computer Model for Analyzing Volatile Removal Assembly

A computer model simulates reactional gas/liquid two-phase flow processes in porous media. A typical process is the oxygen/wastewater flow in the Volatile Removal Assembly (VRA) in the Closed Environment Life Support System (CELSS) installed in the International Space Station (ISS). The volatile organics in the wastewater are combusted by oxygen gas to form clean water and carbon dioxide, which is solved in the water phase. The model predicts the oxygen gas concentration profile in the reactor, which is an indicator of reactor performance. In this innovation, a mathematical model is included in the computer model for calculating the mass transfer from the gas phase to the liquid phase. The amount of mass transfer depends on several factors, including gas-phase concentration, distribution, and reaction rate. For a given reactor dimension, these factors depend on pressure and temperature in the reactor and composition and flow rate of the influent.

Guo, Boyun↗

DOE OSTI2021

Demonstration of MELCOR and MACCS Capabilities for Molten Salt Reactor Decay Heat Removal During both Normal Operations and Salt Spill Scenarios

This report provides a demonstration of MELCOR and MELCOR Accident Consequence Code System (MACCS) capabilities to perform a dose assessment for a Molten Salt Reactor (MSR) off-gas system. A primary containment system salt spill is used as the off-normal scenario, along with a normal operation dose assessment for comparison. This report discusses the tools, methods, and information used in this assessment so that it may be utilized as a starting point for future advanced reactor consequence analyses. This report also highlights several gaps, to include the need for reactor inventory information specific to advanced reactors, and the need for specific atmospheric transport models that take into account the unique deposition behaviors of tritium and carbon-14, and makes recommendations for closing these gaps. This report satisfies the DOE NE Milestone M4RD-21SN0601062.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

The HTR-Proteus Benchmark: Analysis and Use as a Verification and Validation Case

This presentation outlines the evaluation and application of the HTR-Proteus benchmark as a verification and validation (V&V) case for advanced reactor modeling tools. The work supports the U.S. Department of Energy’s HALEU Availability Program (HAP) and the joint DOE/NRC DNCSH project, which aims to reduce criticality safety uncertainties in commercial-scale HALEU fuel cycle and transportation systems. The HTR-Proteus experiments, conducted at the Paul Scherrer Institute, provide high-fidelity data for TRISO-fueled, graphite-moderated pebble bed reactors with high neutron leakage—conditions relevant to HALEU transport scenarios. This study focuses on Cores 4.2 and 4.3 of the HTR-Proteus benchmark, analyzing key sources of uncertainty including pebble packing, TRISO particle positioning, and core height. Using Project Chrono for realistic pebble geometries and Serpent, SHIFT, and MCNP for neutronics simulations, the study quantifies the impact of these uncertainties on the effective multiplication factor (keff). Results show that a sample size of 110 pebble configurations is sufficient to converge keff, with ±30 pcm uncertainty due to packing randomness. TRISO positioning and core height variations also significantly influence keff, highlighting the importance of detailed modeling in V&V efforts. The benchmark serves as a valuable test case for validating the Griffin reactor physics code and improving confidence in HALEU system simulations.

73 - NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2015

SENSITIVITY EVALUATION OF THE DAILY THERMAL PREDICTIONS OF THE AGR-2 EXPERIMENT IN THE ADVANCED TEST REACTOR

A temperature sensitivity evaluation has been performed for the AGR-2 fuel experiment on an individual capsule. A series of cases were compared to a base case by varying different input parameters into the ABAQUS finite element thermal model. These input parameters were varied by ±10% to show the temperature sensitivity to each parameter. The most sensitive parameters are the outer control gap distance, heat rate in the fuel compacts, and neon gas fraction. Thermal conductivity of the fuel compacts and thermal conductivity of the graphite holder were in the middle of the list for sensitivity. The smallest effects were for the emissivities of the stainless steel and graphite, along with gamma heat rate in the non-fueled components. Sensitivity calculations were also performed varying with fast neutron fluence, showing a general temperature rise with an increase in fast neutron fluence. This is a result of the control gas gap becoming larger due to the graphite holder shrinkage with neutron damage. A smaller sensitivity is due to the thermal conductivity of the fuel compacts and graphite holder decreasing with fast neutron fluence.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Cryogenic system with GM cryocooler for krypton, xenon separation from hydrogen-helium purge gas

In the thorium molten salt reactor (TMSR), fission products such as krypton, xenon and tritium will be produced continuously in the process of nuclear fission reaction. A cryogenic system with a two stage GM cryocooler was designed to separate Kr, Xe, and H{sub 2} from helium purge gas. The temperatures of two stage heat exchanger condensation tanks were maintained at about 38 K and 4.5 K, respectively. The main fluid parameters of heat transfer were confirmed, and the structural heat exchanger equipment and cold box were designed. Designed concentrations after cryogenic separation of Kr, Xe and H{sub 2} in helium recycle gas are less than 1 ppb.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

Development of an In Situ Fission Gas Release Instrument for Fuel Sample Irradiations in the High Flux Isotope Reactor

Experimental measurement of gaseous fission product release with respect to temperature and burnup is a critical aspect of understanding nuclear fuel performance, validating predictive models, and qualifying new fuels. To measure this phenomenon in real-time, Oak Ridge National Laboratory has developed an instrument for measuring in situ fission gas release from small-scale fuel samples irradiated in the High Flux Isotope Reactor (HFIR). The instrument uses a continuous flow of Heover the fuel samples to sweep gaseous fission products from a sealed capsule in the HFIR Be reflector to an instrument station adjacent to the reactor. The instrument station houses two high-purity germanium (HPGe) detectors that measure decay gamma rays from fission products passing through a room temperature dwell chamber placed over the detector crystal. The sealed capsules in the reactor are designed to modulate fuel sample temperatures between 700 and 1,100°C by changing the Ar/He gas mixture surrounding the capsules during irradiation. N-type thermocouples are incorporated into the capsule housing to record real-time fuel temperatures. The capsules are heated primarily by prompt gamma rays emitted from the HFIR core with minimal heat contributions from fission in the fuel samples to minimize temperature gradients in the specimens for separate-effects characterization of the material. This paper describes modeling of time-dependent nuclear heating and fission product formation in fuel samples, thermal characteristics of the in-core capsules, and expected gaseous fission product gamma spectra at the HPGe instrument station.

Mulligan, Padhraic L [ORNL] (ORCID:000000025826540↗

NASA NTRS1991

Nuclear Thermal Propulsion: A Joint NASA/DOE/DOD Workshop

Papers presented at the joint NASA/DOE/DOD workshop on nuclear thermal propulsion are compiled. The following subject areas are covered: nuclear thermal propulsion programs; Rover/NERVA and NERVA systems; Low Pressure Nuclear Thermal Rocket (LPNTR); particle bed reactor nuclear rocket; hybrid propulsion systems; wire core reactor; pellet bed reactor; foil reactor; Droplet Core Nuclear Rocket (DCNR); open cycle gas core nuclear rockets; vapor core propulsion reactors; nuclear light bulb; Nuclear rocket using Indigenous Martian Fuel (NIMF); mission analysis; propulsion and reactor technology; development plans; and safety issues.

Clark, John S.↗

DOE OSTI2025

Bayesian calibration of bubble size dynamics applied to CO 2 gas fermenters

To accelerate the scale-up of gaseous CO 2 fermentation reactors, computational models need to predict gas-to-liquid mass transfer which requires capturing the bubble size dynamics, i.e. bubble breakup and coalescence. However, the applicability of existing models beyond air–water mixtures remains to be established. Here, an inverse modeling approach, accelerated with a neural network surrogate, calibrates the breakup and coalescence closure models, that are used in class methods for population balance modeling (PBM). The calibration is performed based on experimental results obtained in a CO 2 -air–water-coflowing bubble column reactor. Bayesian inference is used to account for noise in the experimental dataset and bias in the simulation results. To accurately capture gas holdup and interphase mass transfer, the results show that the breakage rate needs to be increased by one order of magnitude. In conclusion, the inferred model parameters are then used on a separate configuration and shown to also improve bubble size distribution predictions.

09 BIOMASS FUELS↗

DOE OSTI2022

Development of an MC&A toolbox for liquid-fueled molten salt reactors with online reprocessing (Final Report)

A critical barrier to the deployment of MSRs is the absence of a well-defined nuclear material control and accounting (MC&A) approach, a vital prerequisite to meet NRC licensing requirements as well as facilitating future international exports. Liquid-fueled MSR variants (especially those incorporating online refueling or reprocessing) present a unique set of challenges to traditional nuclear material control and accountancy. Unlike solid-fueled light-water reactors, traditional item-counting methods cannot be applied as an accountancy strategy. Rather, MC&A approaches to MSR variants (including both uranium and thorium fueled designs) are more analogous to bulk material handling facilities (e.g., enrichment and reprocessing); yet further complicating matters is the fact that fuel medium is also highly radioactive. Meanwhile, the space of MSRs covers a broad range of design parameters, including thermal and fast spectra designs; operation in actinide breeder or burner modes, choice of actinide fuel used (e.g., 235 U, 232 Th / 233 U, denatured 233 U, etc.), pool or loop-type configuration, and even different salt chemistry. Each of these design choices introduce significant challenges to MC&A approaches within MSR facilities. We propose to bridge this gap for liquid-fueled MSRs by developing a modular, component-based test framework for evaluating viable process monitoring and MC&A techniques specifically suited to liquid-fueled MSR system variants employing online refueling or reprocessing. This test platform will consist of a toolbox of independent process modules representing discrete physical units (such as the reactor core, off-gas processing, decay tanks, and actinide separation units), each with its own self-contained physics responsive to the input mass flow, along with appropriate measurement models that can be coupled to key flow points. These dynamic physical signatures thus afford the ability to test the viability and efficacy of potential accountancy techniques under the full range of reactor operating conditions. As process modules are connected via mass flows, the result is a reconfigurable, generic MSR mass flow model capable of serving as an MC&A test platform for a broad spectrum of possible MSR configurations. The resulting MSR MC&A toolbox will enable robust assessment of accountancy strategies for this unique facility type, including analysis of physical feedbacks arising both from depletion of the fuel over time as well as from potential off-normal events, including those introduced by equipment failures (e.g., a pump failure) as well as by malicious action (i.e., attempts to divert material). The proposed toolbox both addresses a critical needs area for the MPACT analysis toolkit while leveraging existing MPACT-sponsored tools, especially with respect to simulation of measurement and accountancy techniques for advanced fuel cycle facilities. Beyond enabling new analysis capabilities for MSR systems, the design of this toolbox will be to such to maximize compatibility with existing MPACT tools, such to enhance existing facility MC&A analysis capabilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors (Final Project Report)

Advanced (i.e., non-light-water) reactor technologies offer developers new opportunities to enhance the reliability, efficiency, and safety of nuclear power reactors through changes in fundamental design and operation. A variety of advanced design concepts are being developed that generally trend toward increased reliance on innovative, inherent, and passive safety features. Many of these features are substantially different from components now found in large light water reactors (LWRs) that dominate the commercial operating fleet. Similarly, non-light water reactor (non-LWR) suppliers are also pursuing market niches quite different from the regional baseload power generation sites typically associated with large LWR facilities.

42 ENGINEERING↗

NASA NTRSDate not supplied

Flow Regime Transitions and Pressure Drop in Gas-Liquid Flows through Packed Beds in Microgravity

The Packed Bed Reactor Experiment (PBRE) focused on the hydrodynamics of gas-liquid flow in porous media, specifically on flow regime transition and pressure drop correlation. The reactor included a cylindrical test section of 60 cm in length, 5 cm in diameter and filled with 2 mm spherical glass beads. Liquid water and Nitrogen gas were flowed into a centrifugal phase separator after exiting the test section. Five absolute pressure transducers and two high-speed video cameras spanned the test section to measure the pressure drop and capture flow regime behavior. Video observations, analysis of pressure traces, and pressure drop data were used to identify flow patterns. It was found that the slope of pressure gradient data when plotted versus the gas or liquid Reynolds numbers changed when the flow regime changed. This observation was used to identify the flow regime boundaries. At high liquid flow rates (high interaction regime), dispersed bubble and pulse flow regimes were detected. At low liquid flow rates (low interaction regime), large bubble and gas channeling regimes were observed. The transition from the big bubble regime to the gas channeling regime occurred at a critical gas flow rate independent of the liquid flow rate. Similarly, the transition from the low interaction to the high interaction regime occurred at a critical liquid flow rate. For these different flow regimes, different two-phase flow friction factors and pressure gradient correlations were fitted as a function of the gas and liquid modified Reynolds numbers and Suratman number. It was found that in all the flow regimes the capillary contribution to the pressure drop is significant and can be as high as 90% of the total pressure drop.

Brian Motil↗

DOE OSTI2021

Enabling Load Following Capability in the Transatomic Power MSR

This project is dedicated towards designing a fuel processing system that enables liquid-fueled molten salt reactors (MSR) to load follow by removing the dissolved xenon in the fuel salt. As one of the Gen-IV nuclear reactor concepts, the molten salt reactor receives increasing development interests in the recent years. One distinguishing feature of the liquid-fueled molten salt reactor is its improved ability to operate in a load-following mode by including the unique online fission product removal system. Load-following means that the reactor changes its power output based on the demand on the grid. Most of the current operating nuclear reactors have limited load-following ability and operate as the base load on the grid. Due to the rapid increase of solar energy, the requirement on load-following capacity is significantly increased because of the varying power output of the solar panels, yet the traditional load-following capacity is expected to decrease as the decarbonization of the grid continues. Therefore, the ability to perform load-following operation for the nuclear reactors will greatly enhance the resilience of the grid and make nuclear energy more economically competitive. This load-following feature is included in many commercial molten salt reactor designs, such as the designs by Transatomic Power, Terrestrial Energy, and Flibe Energy. Unfortunately, detailed analysis of the fuel processing system for commercial scale MSRs is still lacking, as well as how the fuel processing quantitively impacts the load-following operation. Moreover, experimental data for many of the underlying physics of fuel processing is limited. This project aims to pave the way for the fuel processing technology to advance to the commercial stage by performing combined experimental and simulation research. During the project period, four interconnected aspects of the development of the fuel processing system in liquid-fueled molten salt reactors are investigated. These aspects are the simulation and analysis of the fission product removal system, the fuel cycle simulation, the coupled reactor neutronics and thermal hydraulics transient simulation, and the gaseous fission product removal experiment. Multiphase CFD simulations are performed for components of the processing systems, and simplified air-water experiments are carried out to provide validation data. It is concluded that the CFD simulation can satisfactorily predict the system level performance of the components, and engineering models are constructed based on this success. Fuel cycle analysis is performed for two representative MSR design, the MSBR and the Transatomic Power MSR. Open-source code SaltProc is developed to incorporate the unique fuel processing system of the MSRs. It is concluded that the removal of xenon is essential for load-following operation in thermal spectrum MSR and Molten Salt Breeder Reactor. For the Transatomic Power MSR, the xenon poisoning effect is negligible due to its relatively fast neutron spectrum, though the overall fuel cycle economics still benefits from the removal of xenon. Coupled reactor neutronics and thermal hydraulics transient simulation is performed specifically for the Transatomic Power MSR. It is concluded that the reactor core design could perform power ramping fast enough to satisfy load-following operation. Combining the findings from each aspect, it is concluded that the load-following operation of a thermal neutron MSR is dependent upon the removal of xenon, which could be achieved for a commercial sized reactor using continuous inert gas sparging in a separate system with reasonable dimensions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRSDate not supplied

Analysis of Volatile Compounds from CO2 Removal Systems

One of the primary concerns when designing CO2 scrubber systems that will be integrated with a Sabatier reactor to produce water and methane is the amount of water released from the scrubber. Because the gas stream entering a Sabatier reactor must be compressed, water entering the reactor can condense and compromise the integrity of the system, thus rendering its valuable conversion capability useless. When the Johnson Space Center Environmental Chemistry Laboratory was tasked to develop an assay to quantify the water concentration in air samples from CO2 scrubbers, additional testing was also performed to see if any other compounds were being concentrated on the scrubbers. It was thought that the efficiency of the scrubber systems could be quantified by comparing the differences in samples from the ambient air on the International Space Station (ISS) to the exit gas of the scrubber. As this analysis was carried out, it became evident that the concentrations of certain volatile compounds were higher in the samples from the scrubbers than they were in nominal environmental samples. This meant these compounds were being retained and concentrated on the scrubber beds. Based on this finding, concerns were raised about their potential for these compounds to poison the Sabatier reactor. Further investigation was required to identify these compounds due to their high concentrations and unique matrix of the CO2 scrubber exhaust. This paper describes these events as well as the process that was developed to identify the volatile compounds that increased. An examination of how much the certain compounds can be concentrated by the scrubber systems is also included.

Volatiles↗

NASA NTRSDate not supplied

Analysis of Volatile Compounds From CO2 Removal Systems

Volatiles↗

DOE OSTI2015

Zirconium Recycle Test Equipment for Hot Cell Operations

The equipment components and assembly support work were modified for optimized, remote hot cell operations to complete this milestone. The modifications include installation of a charging door, Swagelok connector for the off-gas line between the reactor and condenser, and slide valve installation to permit attachment/replacement of the product salt collector bottle.

42 ENGINEERING↗