Uranium Zirconium Hydride Research Articles

Optimizing Hydride Stability in U-ZrHx Nuclear Fuel: The “Goldilocks Radius”

Nuclear-powered microreactors show great promise for opening new nuclear energy markets due to the flexibility offered by their rapid/streamlined in-factory fabrication, transportability, and self-regulating nature. The economic benefits of any commercialized nuclear reactor, however, rely on the system’s ability to produce large amounts of heat and efficiently convert that heat into electrical power reliably for long periods of time. Uranium-zirconium hydride (U-ZrHx) is currently being considered for compact reactor designs because it is a well-known nuclear fuel system that is self-moderating, but this fuel, which has historically been used for research reactors, has not been optimized for commercial power production. This paper analyzes the hydride stability of standard 304 stainless steel–clad U-ZrHx fuel under commercially relevant conditions. Fuel element design parameters, including physical dimensions, as-fabricated hydrogen content, burnup, peak fuel temperature, temperature gradient, operational fuel cycle duration, and volumetric heat generation rate, are discussed with a focus on hydrogen distribution and phase stability within the fuel element. Hydride stability declines more rapidly as the coolant temperature, burnup, and fuel cycle duration increase. Using a fuel-cladding gap material with heat transfer properties superior to air, such as helium or sodium, is essential to prolonging fuel hydride stability. The fuel’s physical dimensions are also important. At very small fuel diameters, the H/Zr ratio in the fuel meat decreases too rapidly due to the hydrogen content’s dependence on fuel meat volume. Conversely, the fuel meat temperature and temperature gradient exacerbate hydrogen loss at very large fuel diameters. We find that the most important parameter to consider when optimizing the hydride stability of U-ZrHx fuel is the relationship between the fuel meat radius and the power density in the fuel. A simple equation is empirically determined that relates the “Goldilocks radius,” that is, the fuel radius for which the H/Zr ratio is most stable, to the power density in the fuel.

Uranium-zirconium hydride nuclear fuel performance in the NaK-cooled MARVEL microreactor

The Microreactor Applications Research Validation and EvaLuation (MARVEL) Project is producing a high temperature NaK-cooled nuclear reactor test bed at the Idaho National Laboratory to ultimately improve integration of microreactors to end-user applications. This ambitious effort seeks to design, authorize, construct, test, and operate the reactor within five years. The computational software package chosen for high fidelity fuel performance modeling, BISON, has been upgraded to model the performance of the MARVEL reactor's selected fuel system: 304 stainless steel-clad U-ZrHx. In this manuscript, we begin by describing the MARVEL fuel element. The known properties and relationships of the fuel element constituents are recapitulated from data available in the literature. These historical relationships assume a particular U-ZrH microstructure which previously was not well-understood; therefore, the fuel's microstructure is experimentally confirmed using electron microscopy and X-ray diffraction. This information is then used to analyze the thermomechanical performance of the MARVEL fuel element during a hypothetical extreme accident scenario (a loss of flow accident during a loss of coolant pressure accident) using the nuclear fuel performance code BISON. The analysis shows that the primary phenomenon which could compromise the structural integrity of the MARVEL fuel element is the same as for TRIGA reactors: high temperature internal gas overpressurization-induced rupture of the cladding arising from excessive hoop stresses. Despite modeling the most severe MARVEL accident scenario, the highest anticipated temperature in the fuel, about 718 °C, results in hoop stresses of less than 10 MPa. Based on this analysis, a conservative steady state fuel meat temperature limit of 900 °C precludes cladding damage. The newly incorporated fuel performance simulation capabilities developed in this work may also be extended to model this fuel system under other conditions or reactors.

Optimized core design and shield analysis of a medium temperature heat pipe cooled reactor

Medium Temperature heat pipe cooled Reactor (MTR) is a new reactor concept proposed by Li et al. in (2022). Low enriched uranium zirconium hydride fuel and mercury heat pipe are introduced in MTR. The unique advantages of MTR have been verified (Li et al., 2022) and it is considered as a potential candidate for energy supply in decentralized markets. In this work, an optimized MTR core design is proposed with detailed neutronic analysis and evaluation on mercury heat pipe, reflector and control rod arrangement, further increasing the economic efficiency and mobility of MTR. Heat transfer limit of the improved heat pipe model is further increased, contributing to a great reduction on core volume. Reflector material and thickness are analyzed in detail for mass minimization. Core physics characteristics are analyzed for four control rod arrangement schemes. The control mechanism is further simplified with only 9 control rods involved in the normal regulation of the core. Core lifetime is far more than 10 years. High inherent safety of the core is ensured. Besides, preliminary design of the shielding system is also carried out. A shieling layer of 60 cm LiH and 20 cm Pb can reduce the total dose to less than 40 % of the design limits. The total mass of the optimized core is greatly reduced and its specific power is increased by 60 % compared with that of the previous design.

A medium temperature heat pipe cooled reactor

Heat pipe cooled reactor is a potential candidate for special purpose energy supply. The use of high temperature heat pipe results in a high thermal efficiency and a high passive safety. However, its development has been limited by several shortcomings in engineering, such as the restriction of low nuclear power and the material choosing in high temperature, etc. In this work, a Medium temperature Heat pipe cooled Reactor (MHR) with very high inherent safety is proposed. Mercury heat pipes are used to cool the solid reactor with uranium zirconium hydride fuel pins, which results in a limited temperature in reactor core and a low requirement to high temperature materials. The fission heat is transferred to electricity by a superheat steam/water cycle with an efficiency of approximately 33%. The unique merits of the MHR include low requirements to materials, low enrichment, transient negative temperature coefficient, excellent startup characteristics and so on. The control rod worth, neutron energy spectrum, neutron flux distribution, temperature coefficient and core lifetime were analyzed. It was found that the MHR includes a large transient negative temperature coefficient, which means that a sharp reactivity increase in the core could be controlled immediately. Finally, a very high inherent safety is achieved.

Neutronic feasibility study of ([formula omitted] fuel in a modelled TRIGA research reactor

This paper presents a comparative analysis when transitioning from fuel meat based on uranium zirconium hydride (U-ZrH1.65) to one based on thorium-uranium-233 zirconium hydride ((Th-233U)-ZrH1.65) as an alternative option for TRIGA reactors. The aim of this study was to investigate the neutronic characteristics of (Th-233U)-ZrH1.65 fuels with two different Th-233U weight contents (i.e., 8.5 wt% and 12 wt%) and various U-233 enrichment levels. Using MCNP6.2 computer code based on the Monte Carlo method, a three-dimensional (3D) hypothetical TRIGA model was developed to investigate various neutronic parameters, such as the effects of U-233 enrichment and different (Th-233U) weight contents on the effective multiplication factor, burnup behavior, spent fuel composition, neutron flux, and fission product poison concentration. Since fuel temperature coefficients of reactivity and effective delayed neutron fractions play important roles in the reactor’s kinetics, these were also investigated, and the results were compared to the model with U-ZrH1.65 fuel. The obtained results revealed that the presence of Th-232 in the fuels did not drastically reduce the fuel temperature coefficients due to the zirconium hydride within the fuels providing most of the feedback in a TRIGA reactor. In contrast, the presence of U-233 in the (Th-233U)-ZrH1.65 fuels did reduce the effective delayed neutron fraction. Despite this, the lower transuranic generation in cores fueled by (Th-233U)-ZrH1.65 make this fuel advantageous in terms of achieving longer cycle lengths and less-hazardous nuclear waste for disposal.

Effect of irradiation of silicon photodiode arrays for ITER radial x-ray camera investigated by measuring response and current-voltage characteristics.

The response and current-voltage (I-V) characteristics of irradiated and non-irradiated silicon photodiode arrays (SPDAs) for use in the International Thermonuclear Experimental Reactor camera are measured and compared. Irradiation experiments are carried out using a uranium-zirconium hydride pulsed reactor. The total equivalent 1MeV neutron fluence with energy above 0.01MeV is ∼9.89 × 1013ncm-2. The output signal of the irradiated SPDA (XD2) shows a nonlinear trend during the irradiation experiment. The final signal is about 5.6% of the original one in the visible light region. Tests on the Experimental Advanced Superconducting Tokamak (EAST) show that the XD2 signal is 70%-80% of that of a non-irradiated SPDA (XD3). This indicates that irradiated SPDAs can still observe plasma radiation after exposure to 9.89 × 1013ncm-2 neutron fluence. However, because the neutron fluence of external camera detectors will reach 1.4 × 1016ncm-2 in D-T phase, the SPDAs might become unusable at some point. The responsivity ratio of irradiated and non-irradiated SPDAs is about 4%-20% from 7 to 13 keV. The degradation of responsivity is related to the energy level. After irradiation, the reversed dark current rises from 0.1 to 10 nA to a level of around 1 µA. In terms of tests of XD2 on EAST, zero bias is a good working condition for irradiated SPDAs.

Neutronic investigation of the thorium-based mixed-oxide as an alternative fuel in the TRIGA Mark-II research reactor – Part I: A beginning of life calculations

This article presents a neutronic investigation of the effect of replacing some uranium zirconium hydride (U-ZrH1.6) fuels with thorium-based mixed-oxide ((Th,U)O2) fuels in the TRIGA Mark-II research reactor without changing the core geometry, while the only variable is the fuel material.The possibility of utilizing thorium mixed oxide fuel in the TRIGA Mark-II reactor is assessed using the famous seed-blanket configuration or also known as Radkowsky Thorium Fuel (RTF) concept. Using the RTF concept, a long core lifetime can be achieved by enhancing the internal conversion rate of fertile to fissile materials, thanks to the thermal capture cross-section of 232Th that is around three times that of 238U. Additionally, on the fuel cycle aspect of using the thorium-based fuel is that the thorium cycle produces much less plutonium and other transuranic radioactive elements than uranium fuel cycle.Actually, neutronic and kinetic calculations, at the beginning of life conditions, for nine different seed-blanket configurations have been investigated using the MCNPX2.6 code. These calculations for all proposed configurations in comparison with the conventional U-ZrH1.6 fueled core were performed to pick out an optimal core(s) design ensuring the most efficient use of (Th,U)O2 fuel in it and with minimum changes in the main safety parameters. The parameters consist of effective multiplication factor keff, delayed neutron fraction βeff, neutron flux distribution, power distribution and radial hot rod power peaking factor (FHR), and shutdown margin (SDM).The calculation results exhibit that the highest initial criticality achieved is 1.07716. The results of the other neutronic investigations of thorium-based mixed-oxide fuel elements and their positions, including their numbers in the core are summarized and discussed in the following paper.

Initial Criticality Analysis of Malaysia’s TRIGA Research Reactor Using TRIGLAV Computer Code

Malaysian Nuclear Agency hosts the 1MW Thermal TRIGA MARK II research reactor since 1982. It first initial criticality was achieved on 28th June 1982 with loading of solid fuel elements like Uranium Zirconium Hydride. TRIGA MARK II started its operation of the research reactor on the same year with a 1MW power generation. Training, Research, Isotope Production and General Atomic (TRIGA) is designed to successfully actualize the variety fields of fundamental nuclear research, the manpower training and the production of radioisotopes. This study deals with the initial criticality analysis of the TRIGA research reactor using TRIGLAV reactor physics computer program. For this purpose, a model of its initial core will be developed and simulated using the software and the results will be validated against the experimental result as mentioned in the final safety analysis report (FSAR). The TRIGLAV computer code solves the neutron diffusion equation by using a finite differences method with iteration of fission density. TRIGLAV is based on four group time independent diffusion equation in two dimensional cylindrical (r, θ) geometry. TRIGLAV can also be applied for criticality of reactor, fuel burn-up calculations, and power distribution and flux distributions calculations of the core and also the reactivity predictions of the reactor.

An investigation into the feasibility of thorium fuels utilization in seed-blanket configurations for TRIGA PUSPATI Reactor (RTP)

Thorium is one of the elements that needs to be explored for nuclear fuel research and development. One of the popular core configurations of thorium fuel is seed-blanket configuration or also known as Radkowsky Thorium Fuel concept. The seed will act as a supplier of neutrons, which will be placed inside of the core. The blanket, on the other hand, is the consumer of neutrons that is located at outermost of the core. In this work, a neutronic analysis of seed-blanket configuration for the TRIGA PUSPATI Reactor (RTP) is carried out using Monte Carlo method. The reactor, which has been operated since 1982 use uranium zirconium hydride (U-ZrH1.6) as the fuel and have multiple uranium weight which are 8.5, 12 and 20 wt.%. The pool type reactor is one and only research reactor that located in Malaysia. The design of core included the Uranium Zirconium Hydride located at the centre of the core that will act as the seed to supply neutron. The thorium oxide that will act as blanket situated outside of seed region will receive neutron to transmute 232Th to 233U. The neutronmultiplication factor or criticality of each configuration is estimated. Results show that the highest initial criticality achieved is 1.30153.

Utilization of thorium and U-ZrH1.6 fuels in various heterogeneous cores for TRIGA PUSPATI Reactor (RTP)

The use of thorium as nuclear fuel has been an appealing prospect for many years and will be great significance to nuclear power generation. There is an increasing need for more research on thorium as Malaysian government is currently active in the national Thorium Flagship Project, which was launched in 2014. The thorium project, which is still in phase 1, focuses on the research and development of the thorium extraction from mineral processing ore. Thus, the aim of the study is to investigate other alternative TRIGA PUSPATI Reactor (RTP) core designs that can fully utilize thorium. Currently, the RTP reactor has an average neutron flux of 2.797 x 1012 cm-2/s-1 and an effective multiplication factor, keff, of 1.001. The RTP core has a circular array core configuration with six circular rings. Each ring consists of 6, 12, 18, 24, 30 or 36 U-ZrH1.6 fuel rods. There are three main type of uranium weight, namely 8.5, 12 and 20 wt.%. For this research, uranium zirconium hydride (U-ZrH1.6) fuel rods in the RTP core were replaced by thorium (ThO2) fuel rods. Seven core configurations with different thorium fuel rods placements were modelled in a 2D structure and simulated using Monte Carlo n-particle (MCNPX) code. Results show that the highest initial criticality obtained is around 1.35101. Additionally there is a significant discrepancy between results from previous study and the work because of the large estimated leakage probability of approximately 21.7% and 2D model simplification.

Evaluation of a uranium zirconium hydride fuel rod option for conversion of the MIT research reactor (MITR) from highly-enriched uranium to low-enriched uranium

The conversion of the Massachusetts Institute of Technology Reactor (MITR) from the use of highly-enriched uranium (HEU) fuel-plate assemblies to low-enriched uranium (LEU) by replacing the HEU fuel plates with specially designed General Atomics (GA) uranium zirconium hydride (UZrH) LEU fuel rods is evaluated. The margin to critical heat flux (CHF) in the core, which is cooled by light water at low pressure, is evaluated analytically for steady-state operation. A form of the Groeneveld CHF lookup table method is used and described in detail. A CHF ratio of 1.41 was found in the present analysis at 10MW with engineering hot channel factors included. Thus, the nominal reactor core power, and neutron flux performance, would need to be reduced by at least 25% in order to meet the regulatory requirement of a minimum CHF ratio of 2.0.

Irradiation effects on thermal properties of LWR hydride fuel

Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH1.6) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

Improving the Neutronic Characteristics of a Boiling Water Reactor by Using Uranium Zirconium Hydride Fuel Instead of Uranium Dioxide Fuel

The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide (UO₂) and uranium zirconium hydride (UZrH 1.6 ) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with UO₂ contains 8 × 8 fuel rods while that fueled with UZrH 1.6 contains 9 × 9 fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. UZrH 1.6 fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

Reactivity temperature coefficient evaluation of uranium zirconium hydride fuel element in power reactor

An in-core fuel management code package for uranium zirconium hydride power reactor, which is developed on the basis of the assembly lattice code TPFAP and the core calculating code BMFGD for LWR, is firstly introduced in this paper. The inherent safety of the U–ZrHx element which is mainly caused by the high prompt negative temperature coefficient is then evaluated, because the weight percentage of uranium, fuel rod radius and fuel temperature of U–ZrHx element will be different in power reactor from those in research reactor, and these changes may make obvious effect on the prompt negative temperature coefficient. The influence of weight percentage of uranium, fuel rod radius, fuel temperature, content of hydrogen and additional poison on prompt negative temperature coefficient for uranium zirconium hydride element are calculated respectively in this paper, and then the results are analyzed theoretically. The study shows that the absolute value of prompt negative temperature coefficient reduces observably along with the increasing of Uranium weight percentage from 10wt% in research reactor to maximum 45wt% in power reactor. Smaller radius, higher operating temperature and longer core life make little effect on the prompt negative temperature coefficient in the condition of high weight percentage of U. Additional poison erbium in fuel makes prompt negative temperature coefficient much more negative. Anyway, high prompt negative temperature coefficient can be achieved for U–ZrHx power reactor with higher weight percentage of U, smaller radius and Er poison, in other words, the characteristics of inherent safety of U–ZrHx core can be retained at power reactors.

Simulation of Loss of Coolant Accident in an Integral Pressurized Water Reactor (IPWR)

In this paper research has been carried out on the Loss of Coolant Accident (LOCA) in an Integral Pressurized Water Reactor(IPWR) by using thermal hydraulic system code Relap5/Mod3.4.The designing of Integrated Pressurized Water Reactor (IPWR) incorporates the safety and reliability of the reactor to withstand under accidental vulnerabilities. In this study, the reactor under consideration is Uranium Zirconium Hydride Nuclear Power Reactor INSURE-100 with the power output of 100MW.In the current research, the reactor has been described in detail according to the requirement for the simulation of LOCA using Relap5 code with the possibility of occurrence of the time sequence of events. The graphs obtained shows good agreement for the safe operation of IPWR under LOCA.

On the use of a moderation layer to improve the safety behavior in sodium cooled fast reactors

This work shows the effect of the use of moderating layers on the sodium void effect in sodium cooled fast breeder reactors. The moderating layers consisting of either boron carbide B4C or uranium–zirconium hydride UZrH cause a strong reduction of the sodium void effect. Additionally these layers improve the fuel temperature effect and the coolant effect of the system. The use of the UZrH is significantly more effective for the reduction of the sodium void effect as well as for the improvement of the fuel temperature and the coolant effect. All changes cause by the insertion of the UZrH layer cause a significantly increased stability of the fast reactor system against transients. The moderating layers have only a small influence on the breeding effect and on the production of minor actinides.

Normal Operational State Results of Integral Pressurized Water Reactor by Using Relap5/Mod3.4

In this paper, a research has been carried out on the normal operational state of IPWR by using the thermal hydraulic system code Relap5/Mod3.4.In this study the conceptual study analysis of the reactor named as Inherent Safe Uranium Zirconium Hydride Nuclear Power Reactor INSURE-100 is considered but is only based on the conceptual study so the current research focuses on the normal operational of the reactor by using the Relap5/Mod3.4 code. For this purpose, two passive safety methods have been included for the safe operation as well as for transient analysis of the reactor. In the first one, Passive Residual Heat Removal System (PRHRS) has been modeled by taking the heat exchanger out of the Reactor Pressure Vessel (RPV) in the water tank so that it can absorb core decay heat in case of transient conditions and in the second one heat exchanger is placed both in inside and outside the RPV so that there can be another way to absorb the core residual heat. Considering these concepts figured out the normal operational state of the reactor by using Relap5/Mod3.4 in comparison with the conceptual design study of the reactor under consideration and the results extracted can be a good agreement for the transient analysis of the reactor.

Transient hydride fuel behavior in LWRs

The transient response of a uranium zirconium hydride (U 0.31ZrH 1.6) fuel element to conditions typical of a light water reactor reactivity insertion accident has been studied. Hydrogen diffusion within the fuel is treated as a function of the time-dependent temperature profile. Temperature and hydrogen concentration dependence of thermal properties of the fuel is also considered. The set of coupled equations describing the transient conduction of heat and diffusion of hydrogen are solved with a Crank–Nicolson discretization scheme for heat diffusion and an explicit Euler scheme for hydrogen diffusion.

Fission Gas Release from Uranium-Zirconium Hydride Fuel During Heating to Melting

Fission gas release and fuel expansion of irradiated uranium-zirconium hydride fuel were measured during fuel heating to melting. The first indication of melting occurred at approximately the uranium-cladding eutectic temperature, whereas complete melting took place at the U metal melting point. Fission gas release began with the onset of uranium-cladding melting and gradually increased with rising temperature. Two sharp peaks in the fission gas release were observed corresponding to the events of uranium-zirconium eutectic melting and uranium melting. The fractional release was much higher than that measured previously for fuel without cladding at comparable temperatures.

Research Reactor Benchmarks

A criticality benchmark experiment performed at the Jozef Stefan Institute TRIGA Mark II research reactor is described. This experiment and its evaluation are given as examples of benchmark experiments at research reactors. For this reason the differences and possible problems compared to other benchmark experiments are particularly emphasized. General guidelines for performing criticality benchmarks in research reactors are given. The criticality benchmark experiment was performed in a normal operating reactor core using commercially available fresh 20% enriched fuel elements containing 12 wt% uranium in uranium-zirconium hydride fuel material. Experimental conditions to minimize experimental errors and to enhance computer modeling accuracy are described. Uncertainties in multiplication factor due to fuel composition and geometry data are analyzed by sensitivity analysis. The simplifications in the benchmark model compared to the actual geometry are evaluated. Sample benchmark calculations with the MCNP and KENO Monte Carlo codes are given.