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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Regulatory Review of Micro-Reactors: Initial Considerations

The U.S. Nuclear Regulatory Commission (NRC) is working to have an effective and efficient mission readiness for reactors that differ considerably from those currently licensed. Micro-reactors, that is, reactors that have a power of roughly no more than 10 MW, are one class of these advanced reactors. This report is one of the first steps taken to articulate the many technical and regulatory issues that will need to be addressed for NRC to have the ability to review license applications for micro-reactors. Many of the issues center around the fact that a) these reactors may be operated remotely or semi-autonomously and b) it will be difficult to analyze risk from new, unique, technologies. Initial thoughts are given on how probabilistic methods could be used to determine risk and how the current approach for reviewing non-power reactors could be useful for micro-reactors.

42 ENGINEERING↗
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 OSTI2022

Multiscale Experiments and Multiphysics Simulation of Multiphase Flow for Transportable Small Modular Reactors

A new type of safe, small, transportable nuclear reactor would address the intense and ever-growing global demand for energy produced via a resilient, carbon-free energy source. In this regard, transportable small modular reactors (SMRs) are being designed and developed for electricity generation within small/micro-grid/off-grid isolated systems, as well as for heat generation in industrial/residential applications. These reactors feature the capability to be fully factory fabricated and then directly transported to utilities’ sites as “plug-and-play” systems. Research and development (R&D) programs are underway at Idaho National Laboratory (INL) to successfully design, develop, and demonstrate such safe-by-design mobile reactor technologies, in collaboration with partner organizations. Multiscale experimental facilities and multiphysics simulation tools are required for reactor design verification and validation (V&V), and licensing. These advanced reactors are intended to feature passive safety systems such as passive containment cooling systems (PCCS), which consist of multiphase flows and multispecies distributions. This seminar talk will focus on designing and analyzing transportable SMR PCCS by using multiscale experiments and multiphysics computational fluid dynamics (CFD) simulations to support reactor licensing and safety. The corresponding research challenges are addressed via supportive verification and validation results generated by the models and simulation tools in combination with selective parametric and uncertainty analysis. This solution approach could blaze the trail for commercial adoption of such technologies. The facilities, simulation capabilities, and research opportunities available at INL in regard to such reactors and the integrated energy systems with which they go hand in hand are also discussed briefly, and may spark interest in deeper research as well as new collaborative projects.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

The Deimos Experiment: Advanced Reactor Testbed

Advanced reactor initiatives are growing significantly through programs nationwide. This research area includes small modular reactors, microreactors, and space reactors. Many of the reactors being designed are untested concepts. They include unique moderators, varying fuel types, high temperatures, and compact configurations. The shift in fuel type, from highly enriched uranium (HEU) to high assay, low enriched uranium (HALEU), is particularly important as it has driven many of the other changes. For example, lower enrichment requires advanced moderators, which in turn require different reflectors to make the systems compact. The change in materials including the transition from HEU to HALEU affects the temperature feedback of the systems. Additionally, these advanced reactor concepts generally have a thermal neutron spectrum in contrast to earlier fast spectrum advanced reactor. With the extensive changes from previous reactor designs, validation experiments are needed. The National Criticality Experiments Research Center (NCERC) is uniquely equipped to perform such experiments. The Deimos experiment, designed for execution at NCERC, will serve as a testbed for advanced reactor concepts. It will use HALEU fuel in a graphite matrix, provide the ability to use advanced moderators, and allow measurements of temperature reactivity coefficients (TRCs).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Removing of Core Constructions of the MR and RFT Reactors in NRC 'Kurchatov Institute' - 16054

In frame of the FTP 'Providing nuclear and radiation safety of Russia in 2008 and for the period till 2015' works of a decommissioning of the MR and RFT research reactors are carried out in National Research Center 'Kurchatov Institute'. Within these works, the equipment of 9 loop-back facilities of the MR reactor was dismantled. In the conditions, which are most approached to natural, one researched a heat, hydrodynamic and strength characteristics of elements of cores and the capital equipment of the majority of power reactors at these facilities. In 2014-2015 dismantle of core constructions of the MR reactor was carried out and dismantle of a graphite laying of the RFT reactor is begun. The beryllium blocks, which were constructional elements of the core, and the graphite blocks playing a reactor reflector role were also taken, packed into cases or containers and removed: part on an accumulative platform, and part are placed in storage of high level radioactive waste of the Center for decay and further sending for long storage. Works were performed by remote controlled mechanisms, control of their performance was exercised by means of remote diagnostics: Gammavizor, the scanning spectrometer systems. The following methods and the organizational principles were applied to providing of radiation safety of these works: - the constant radiation control including individual control of a dose of external radiation, control of volume activity of radionuclides in air, annual control of internal radiation doses of the personnel; - dust suppression and fixing of radioactive contamination by dispersion of polymeric structures; - carrying out decontamination on a shift basis rooms with use of detergents; - use of the remote controlled mechanisms equipped with hydraulic scissors, nippers, application of the diamond are sharp that allowed to perform the majority of technological operations remotely, using 'cold' methods of fragmentation; - development on each type of works of the operation project and schedules. As a result, of works of 2011-2014 of the decommissioning of the MR and RFT research reactors more than 630 t of the equipment were dismantled. From them more than 455 t of radioactive waste was. The total activity of radwaste exceeded 300.0 GBq. From the pool of the reactor, about 160 channels of loop-back installations and near 230 beryllium and graphite are taken, fragmented and packed. The total activity on radionuclides {sup 60}Co and {sup 137}Cs of them was near 55 TBq. From the territory of the center about 2550 m{sup 3} of firm and 4600 m{sup 3} of liquid radwaste are removed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Thermalhydraulic calculation for boiling water reactor and its natural circulation component

Thermalhydraulic of reactor core is the thermal study on fluids within the core reactor, i.e. analysis of the thermal energy transfer process produced by fission reaction from fuel to the reactor coolant. This study include of coolant temperature and reactor power density distribution. The purposes of this analysis in the design of nuclear power plant are to calculate the coolant temperature distribution and the chimney height so natural circulation could be occurred. This study was used boiling water reactor (BWR) with cylinder type reactor core. Several reactor core properties such as linear power density, mass flow rate, coolant density and inlet temperature has been took into account to obtain distribution of coolant density, flow rate and pressure drop. The results of calculation are as follows. Thermal hydraulic calculations provide the uniform pressure drop of 1.1 bar for each channels. The optimum mass flow rate to obtain the uniform pressure drop is 217g/s. Furthermore, from the calculation it could be known that outlet temperature is 288°C which is the saturated fluid’s temperature within the system. The optimum chimney height for natural circulation within the system is 14.88 m.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Design Studies for a Multiple Application Thermal Reactor for Irradiation Experiments (MATRIX)

The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Should unforeseen circumstances lead to the decommissioning of ATR, the U.S. Government would be left without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. A survey was conducted in order to catalogue the anticipated needs of potential customers. Then, concepts were evaluated to fill the role for this reactor, dubbed the Multi-Application Thermal Reactor Irradiation eXperiments (MATRIX). The baseline MATRIX design is expected to be capable of longer cycle lengths than ATR given a particular batch scheme. The volume of test space in In-Pile-Tubes (IPTs) is larger in MATRIX than in ATR with comparable magnitude of neutron flux. Furthermore, MATRIX has more locations of greater volume having high fast neutron flux than ATR. From the analyses performed in this work, it appears that the lead MATRIX design can be designed to meet the anticipated needs of the ATR replacement reactor. However, this design is quite immature, and therefore any requirements currently met must be re-evaluated as the design is developed further.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Successes and Lessons Learned in the Design, Construction and Use of a Tool for Sampling the WR-1 Reactor Core - 1900z

Whiteshell Laboratories, located in Manitoba, Canada, provided research facilities for the Canadian nuclear industry since the early 1960's. The facility was centered on an organically cooled, heavy water moderated nuclear reactor in the world, Whiteshell Reactor 1 (WR-1). WR-1 was safely shut down in 1985, de-fueled and drained, and has been in a safe Storage with Surveillance (Deferred Decommissioning) state until present day. In support of an Environmental Assessment for proposed In Situ Decommissioning of WR-1, detailed characterization information regarding the reactor core components (calandria, calandria tubes, pressure tubes) was necessary to verify inventory estimates and neutron activation models produced circa 1992. The Fuel Channels of WR-1 (pressure tubes) were expected to be the most activated components of the reactor core, and to account for about 60% of the reactor core inventory. The interior surface of the fuel channel is also the most readily accessible surface within the reactor core. As such the decision was made to collect a sample of one of each of 3 Fuel Channel material types (Stainless Steel, Zr-2.5Nb, Ozhenite) for comparison to modelled estimates of neutron activation of various core materials. The expected dose rates from the samples were between 5 and 100 mSv (500-10,000 mrem) on contact based on a 200 mg sample. Sampling the Fuel Channels required verification the Fuel Channel was drained of organic coolant, removal of residual organic coolant from the Fuel Channel wall, removal of material from the interior surface of the Fuel Channel at the centerline of flux (∼25' below the operator position), and extraction of that material from the core. Several options for existing tooling and technologies, including hydraulic scrapers, and horizontal drill technologies, were looked at for their applicability to WR-1. While suitable approaches to higher radiation field sample collection, they were overly complex, costly and time consuming for the configuration at WR-1. In an effort to reduce costs, and to simplify the sampling process, a specialized and simple tool was developed to mechanically scrape a small sample of the interior wall of the Fuel Channels, collect the sample into a vial, and remove the sample from the core for analysis. The tool was designed, constructed, tested and implemented in the field entirely by Canadian Nuclear Laboratories Staff. The tool required no powered components, or electronic guidance or measurements; and was designed and constructed for a fraction of the cost of other commercially available technologies for sample collection within reactor cores. The samples were successfully collected without any injuries, contamination events or equipment failures. The tool remains operational in the event additional sampling is required. This paper provides additional details of the design, construction, testing and use of the tool and the successes and lessons learned during the various stages of design and implementation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Minimization of Nuclear Waste using AHWR Reactor Technology

Nuclear sciences have grown from creating reactors to a greater understanding of how to dispose off the waste exhibited by the nuclear activities. Various methodologies have been proposed and are being used in order to do the same. This poster poses great attention to 'Using a fuel cycle dedicated to the production of the lower quantity of minor actinides, is one of the key targets in reducing nuclear waste in india'. A significant fraction of thorium is utilized as a fertile host in the Advanced Heavy Water Reactor (AHWR300-LEU) fuel. Being lower in the periodic table, it leads to a reduction of minor actinides. It produces only half of the minor actinides as a waste as compared to modern LWRs. As thorium oxide has inert nature, it benefits in its prolonged storage. Therefore, safe disposal of spent fuel is achieved due to the inert matrix, on completion of one fuel cycle mode. AHWR is a 300 MWe, vertical, pressure-tube type, boiling light water-cooled, and heavy water-moderated reactor. Employs natural circulation for removal of heat from the reactor core. In-situ conversion of {sup 232}Th into {sup 233}U. The reactor design permits flexible fuel cycle options. There are two types of AHWR depending upon the fuel cycle options. Uses thorium-based oxide fuels: two fuel cycle options. With the adoption of thorium fuel cycle, AHWR300-LEU produces less minor actinides. As compared to modern LWRSs, it produces almost 37% less amount of actinides for the same amount of energy production, which in turn minimizes the burden on waste disposal requirements. But, it is associated with radionuclides such as {sup 231}Pa (produced through absorption of one neutron and emission of two neutrons). Since, the reaction rate of (n,2n) reactions is very low in thermal neutron reactors, thus making the fuel cycle the favorable one. Thorium oxide (ThO{sub 2}) is chemically stable and is the highest oxidation state of thorium, therefore the inert nature of the matrix is advantageous for safe disposal of spent fuel in case of adoption of open fuel cycle mode and offers prolonged storage of spent fuel. The AHWR is thorium fueled Indian nuclear reactor that assimilates several new features with diminished environmental concern due to reduced waste generation. Since thorium fuel cycle produces less long lived nuclear waste, thus it can possibly bring the world cleaner and more sustainable nuclear energy. Currently, India being the only country with large scale research program of thorium nuclear power serves a platform for the development and demonstration of all technologies and advanced safety systems.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Atmospheric {sup 41}Ar Plume Tracking of the UF Training Reactor Using a Large-Volume NaI Detector

The ability to track and monitor gaseous radioactive materials is important in the event of a radioactive release such as from a nuclear power plant accident, weapons testing, or other unintended release scenarios. The International Monitoring System (IMS) is currently in place to monitor for weapons testing and has been used to monitor nuclear power plant accidents. However, only 80 radionuclide stations for the IMS exist in the world and are stationary, limiting the resolution of any radioactive plumes detected. Furthermore, the IMS radionuclide stations are designed to be sensitive specifically to atmospheric radio-xenon and no other radionuclides. Historically, monitoring radioactive releases has been limited to the detection of radio-xenon and radio-krypton, both arising as fission products from nuclear fuel. After the accident at Three Mile Island Unit 2, scientists were interested in finding the dose the public received from released radioactive material. This was achieved by analyzing radiation monitors in the exhaust stack and filters within the reactor building. furthermore, a series of 20 thermoluminescent dosimeters (TLD's), 8 air sampling station positioned around Three Mile Island, and a helicopter survey which used a NaI(Tl) detector array provided data on radionuclides, specifically {sup 137}Cs, that escaped the confines of the reactor building. The TLD's and air sampling stations were stationary, and therefore limited the resolution of the radioactive release plume. The helicopter survey using the NaI detectors was limited by the technology available at the time. The NaI detector array output was compressed from 1024 channels to 256 channels, limiting the energy resolution of the spectral data. The helicopter used a microwave ranging system to establish position. Since this survey was taken in October of 1982, advancements in technology allow spectral data to be saved in its full 1024 channel format and the advent of GPS can give a more accurate account of the detector position. An updated mobile system for tracking a radioactive release would help to alleviate these issues as well as gain the ability to track a radioactive release in real-time. We investigate the feasibility of mobile tracking through study and measurement of the {sup 41}Ar plume released from the University of Florida Training Reactor (UFTR) research reactor. We use a large-volume NaI(Tl) detector placed on the roof near the reactor. This detector may also be placed and moved on a cart to survey the area surrounding the reactor. We compare the spectral data taken when the reactor is running at full power to the background to analyze the effectiveness of the detector at identifying atmospheric {sup 41}Ar. Finally we discuss the implications of these measurements with regards to validation of plume models and the capability to track releases over a localized area from the source. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

HFIR Core Modeling and Analysis for the PROSPECT Reactor Antineutrino Experiment

Nuclear reactors have been employed as a copious source of antineutrinos for over six decades. Research initially directed at the study of antineutrino physics has led to applications of these highly penetrating particles to reactor safeguards because the antineutrino flux and spectrum can be related to the state of the reactor core, including burnup evolution. One of the main challenges of antineutrino-based safeguards is operation of the detectors near surface (within the reactor containment or in the close proximity). Lack of overburden results in increased contribution to the detector noise from cosmic radiation. Similarly to safeguards, experiments to definitively confirm or exclude the existence of sterile neutrino oscillations require detector technologies capable of operating at short baselines (tens of meters) at the surface. The Precision Reactor Oscillation and Spectrum Experiment (PROSPECT) at the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is an experiment with a main objective to make a precision measurement of the flux and energy spectrum of antineutrinos emitted from nuclear reactors in order to search for the oscillation signature of sterile neutrinos. Fissions of highly-enriched uranium of HFIR will serve as a source of antineutrinos. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Comparing Sensitivity/Uncertainty Analysis Results for LR-0 Salt Experiments with Salt Reactor Models

Recent reactor physics measurements conducted at Research Centre Rez (RC Rez) used the LR-0 low power nuclear reactor to investigate neutronic impacts and uncertainties of FLiNa (60 mol% LiF and 40 mol% NaF) and FLiBe (2 LiF + BeF{sub 2}) salt targets. This included using FLiBe originally from the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL), which contained enriched lithium (99.994 mol% {sup 7}Li). In collaboration with RC Rez, ORNL performed sensitivity/uncertainty (S/U) analyses of these LR-0 experiments with salt targets. Molten salt reactor technologies can be categorized into two general types: solid-fueled or liquid-fueled. Solid-fueled molten salt technologies are perhaps best represented by fluoride-salt-cooled high-temperature reactors (FHRs), which are solid-fueled and salt-cooled. They are potentially promising because they offer high-temperature outlet heat using a low-pressure, high-power-density system (relative to high temperature gas-cooled reactors). However, FHRs have never been built; therefore, no direct neutronic benchmarks exist for code or cross section validation work. Liquid-fueled molten salt reactors (MSRs) use a liquid salt as both fuel and coolant. MSRs can be designed to use a variety of fertile or fissile species as nuclear fuel, different neutron energy spectra, and a multitude of approaches for adding or removing material from the salt. The MSRE generated some operational data for fluoride salt MSRs, but no formal benchmarks currently exist based on the MSRE data, and the data available do not cover all operational conditions. Given these data limitations, ORNL performed S/U analyses using application models for FHR and MSR concepts to (1) identify potential sources of bias in FHR and MSR simulations resulting from cross section uncertainties and to (2) compare the sensitivity of neutron multiplication in FHR and MSR concepts to cross section data on an energy-dependent basis for specific nuclides against similar sensitivity results from LR-0 with a FLiBe salt target. In the future, these S/U analyses could inform the design of additional salt experiments or provide part of the basis for follow-up work to refine nuclear data. Please see the final project report for additional details beyond this paper.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Multicomponent gas mixture parametric CFD study of condensation heat transfer in small modular reactor system safety

Safety is always the primary concern for designing and analyzing nuclear reactor systems. The requirements for the safety margin for advanced small modular reactor (SMR) systems are targeted even higher than the conventional commercial large-scale nuclear reactors incorporating the passive and inherent safety systems. The SMR systems are designed with the condensation passive containment cooling system (PCCS), which plays a critical role in removing reactor heat during a steam release accident case. However, the presence of non-condensable gas (NCG), like air, reduces the heat transfer performance. This physics phenomenon becomes multifactorial for nuclear reactor containment during a fuel failure accident case that releases hydrogen gas. Besides, the mixture component of steam-air-hydrogen varies in reactor accident cases, which need simulation and validation keeping parameters of importance. Reviews showed that previous studies for SMR’s PCCCS did not cover the condensation heat transfer (CHT) in the presence of multicomponent NCG mixture parametric computational fluid dynamics (CFD) simulation and validation, making a research gap in the SMR design safety. A comprehensive CHT parametric CFD study was performed for SMR PCCS to fill this research gap. This study used experimental data as simulation 3D physics domain inlet and outlet boundary conditions. However, the wall boundary conditions were constant temperature, curve-fit, and annular coolant for verifying the turbulence models. Parametric simulations were performed, verified, and optimized for steam-NCGs mixtures. The multicomponent gases, multiphase mixtures, and fluid film condensation models were applied with associated turbulence models. The results of the parametric study were evaluated for realistic reactor conditions. Results showed that parametric study provided critical insight about the dependency of multicomponent gas mixture parameters that supports reactor safety design, analysis, and licensing.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Fuel cycle and neutronic performance of a spectral shift molten salt reactor design

Here, the fuel cycle performance and core design of the Transatomic Power liquid-fueled molten salt reactor concept is analyzed. This advanced reactor concept uses configurable zirconium hydride moderator rod assemblies to shift the neutron spectrum in the core from intermediate at beginning of life to thermal at end of life. With a harder spectrum during the early years of reactor operation, this spectral shift design drives captures in fertile 238 U. The converted fissile plutonium makes up over 50% of the fissile material in the fuel salt over the last half (~15 years) of reactor operation. A softer spectrum late in reactor life helps drive the fuel to a burnup of 90 GWd/MTU. Continuously changing physics necessitates time-dependent analyses resolved over long timescales (i.e., months to years), as this concept does not meet an equilibrium condition. The spectral shift and molten salt reactor material feeds and removals enable this concept to perform better in fuel cycle metrics, increasing resource utilization by more than 50% compared with a typical light water reactor (i.e., from ~0.6% to ~1%). Finally, these metrics are compared to similar fuel cycles using alternate technologies. Additional core design and analysis challenges associated with the spectral shift and use of molten salt reactor technology are identified and discussed.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2024

Nuclear fuels for transient test reactors

Transient test reactors with the ability to test fissile specimens under extreme conditions have been crucial tools in the development of nuclear technologies. Less than 10 unique facility designs have ever been constructed, most of which remain operational today and still use the original nuclear fuel constructed for them more than 40 years ago. Historic fuel systems for transient test reactors vary in significant ways which have marked influences on reactor capabilities. Eventually, new fuel will be needed to support the longevity of transient test reactor missions. This paper reviews precedent transient reactor fuel systems in the context of their unique requirements. A few key conclusions are illustrated by comparing and contrasting these transient test reactors. Fuel composites which are mostly graphite can enable transient reactors with very high neutron fluence capability (>2E16 n/cm 2 ) and are amenable to longer “shaped” transients but cannot achieve pulses <10 ms in duration. Reducing the graphite-to-uranium ratio can yield a very narrow pulse capability but delivers less fluence and requires cores with considerably more fissile material. Designs based on uranium dioxide (UO 2 ) make use of readily available materials to create compact cores with narrow pulse width capabilities but with moderate neutron fluence capabilities (~2E15 n/cm 2 ). Uranium zirconium hydride (U-ZrHx) is a well-established fuel system for pulsing reactors which has been intermittently manufactured throughout the decades. U-ZrHx offers similar capabilities to UO 2 designs in terms of nuclear kinetics, but with about half the fluence capability (~1E15 n/cm 2 ). An evolution of the UO 2 system, termed “ternary ceramic” fuel, shows that dispersing UO 2 in zirconium oxide and calcium oxide can increase fluence capability greatly (~7E15 n/cm 2 ), but is not presently a commonly available fuel form. A unique composite of UO 2 and beryllium oxide (UO 2 -BeO) can be used to create a core with similar kinetics and compact core geometry as U-ZrHx designs, but with significantly higher fluence capability (~6E15 n/cm 2 ). Like ternary ceramic fuel, newly fabricated UO 2 -BeO would require reestablishing its historic manufacturing process which would be further complicated by the health hazards associated with beryllium. In conclusion, like most engineering problems, there is no perfect solution, but this paper outlines the advantages and disadvantages of candidate fuel options to help guide detailed evaluations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Experimental demonstration of high-temperature (>1000 °C) heat extraction from a moving-bed oxidation reactor for thermochemical energy storage

Previously developed reduction-oxidation (redox) thermochemical energy storage technologies must store their products at high temperatures, complicating handling and transportation. This work describes a countercurrent, tubular, moving bed oxidation reactor at laboratory scale that produces high grade heat and allows solids to enter and exit the system at ambient temperatures. The particles implemented in the system consist of a novel magnesium manganese-oxide material well-suited for thermochemical energy storage. Output heat is obtained via a separate extraction gas flow, which exits from the middle portion of the main reactor tube. With this design, reactor temperatures in excess of 1000°C and extraction temperatures above 950°C were achieved. Deviation between the two measurements is a result of extraction thermocouple placement and losses in the reactor extraction arm; improvements to these parameters would bring the extraction temperature closer to the bed temperature. The reactor produces enough energy via oxidation to sustain both heat extraction and continued chemical reaction. During one representative steady state experiment at a particle flow rate of 1.5 g/s, an average of 447 W was extracted from the reactor out of an estimated 1083 W of released chemical energy for a duration of 70 minutes. Among the four experiments, the maximum bench-scale oxidation reactor energy efficiency of 36.2% and corresponding round-trip efficiency of 13.7% considering both redox reactors were demonstrated. Characteristics of an ideal system are considered and future improvements are proposed.

25 ENERGY STORAGE↗
DOE OSTI2022

Monitoring Operational States of a Nuclear Reactor Using Seismoacoustic Signatures and Machine Learning

Monitoring nuclear reactors is an important safety and security task with growing requirements. We explore the possibility of using seismic and acoustic data for inferring the power level of an operating reactor. Continuous data recorded at a single seismoacoustic station that is located about 50 m away from a research reactor was visualized and analyzed. The data show a clear correlation between seismoacoustic features and reactor main operational states. We designed a workflow that includes two machine learning (ML) models to classify the reactor operational states (OFF, transition, and ON) and estimate reactor power levels (10%, 30%, 50%, 70%, and 90%). We applied and compared five ML algorithms for the reactor OFF-transition-ON and four approaches for the power level classification. We also compared the performance of ML models trained with seismic-only, acoustic-only, and both types of data. Five-fold cross validations were implemented to assure a thorough evaluation of the model performances. Additionally, the results show the extreme boosting gradient algorithm worked best for the first model, whereas random forests performed best for the second model. Combining seismic and acoustic data leads to better performance than using a single type of data. Seismic data contributed more than acoustic data for both models. We reached an accuracy of 0.98 for reactor OFF and ON. The accuracies for the transition state and power levels are less optimal with a minimum accuracy of 0.66. However, our results suggest seismic and acoustic data contain useful information about the transition state as well as power levels. Seismic and acoustic data could be integrated with other observations to improve monitoring performance.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗