High Power Research Reactors Research Articles

Investigation of Buddhist and Bon votive statues at the Very low power reactor VR-1

The nuclear analytical techniques are excellent tools for the investigation in various research disciplines, i.e. cultural heritage preservation, archaeology or anthropology. Most of these radioanalytical techniques require sufficient sources of radiation. So far, neutron imaging has mostly been used at high-power neutron sources, i.e. high-power research reactors. On the other hand, small nuclear research reactors can provide more options for investigations in multidisciplinary research. Since the development of the NIFFLER neutron imaging facility in 2020, neutron imaging has been one of the nuclear analytical methods used at the very low-power research reactor VR-1 of the Czech Technical University in Prague. This paper deals with the investigation of the internal structures of Buddhist and Bon votive statues from Central Asia at the Training reactor VR-1. The experimental investigation was conducted to explore the possibilities of using neutron imaging for cultural heritage research at very low-power research reactors such as the VR-1 reactor. The results of these unique experiments clearly showed the potential of the VR-1 reactor or possibly other very low-power research reactors in neutron imaging for multidisciplinary research, which can provide important data for anthropologists, archaeologists or experts in cultural heritage preservation.

Integrated simulation of U-10Mo monolithic fuel swelling behavior

A separate computational branch has been implemented within the DART (Dispersion Analysis Research Tool) computational code to simulate the swelling behavior of U-10Mo monolithic fuel under the operating conditions of high-power research and test reactors (RTRs). The monolithic branch of the DART code implements a mechanistic rate-theory-based fission-gas-behavior model for the calculation of fission gas swelling, as well as a suite of thermal, physical, and mechanical models to consider various processes occurring in RTR fuels during irradiation. To accurately simulate and eventually predict U-10Mo monolithic fuel irradiation behavior, the code uses materials properties calculated with lower length-scale computational methods, such as gas atom diffusivity and U-Mo surface energy from atomic simulations and grain-morphology-specific recrystallization kinetics (recrystallized fuel volume fractions versus fission density) predicted using the phase-field method. The remainder of the fission gas behavior parameters used in the model were calibrated with measured intergranular bubble size distributions. With this integrated simulation approach, the swelling behavior of U-10Mo monolithic fuel was simulated for various initial grain sizes at different operating conditions and compared with measured data. Furthermore, because limited experimental data exist for parameter calibration, detailed sensitivity studies for the important parameters used in the fission gas behavior model were performed to examine their impact on both intergranular gas bubble morphology at low fission density, and on total porosity at high fission density.

Study of cooling strategy for pressure vessel in pool research reactor at accident situation

Abstract Pressure vessel in pool layout is usually used in high power and middle pressure research reactor, Flow inversion and residual heat removal is important problem. The heat removal strategy at accident situation for a pressure vessel in pool layout research reactor was studied in the paper, many cases was researched for the ability of heat removal and safety of core.

Hydraulic design of an advanced second shutdown system for a research reactor

There has been a significant increase in the demand for nuclear safety around the world. Safety concerns are less major for research reactors than for commercial reactors owing to their low operating temperature and power. Because most research reactors do not have a second shutdown system for the shutdown of the core, damage to the core may occur upon failure of the shutdown system. Recent research on aspects, such as utilization of radiation isotopes and neutron utilization fields, has increased the demand for high-power research reactors. This has led to the development and extensive application of the second shutdown system for research reactors. This study analyzes the hydraulic design of the heavy water dump system (HWDS) using the instantaneous dump of the heavy water reflector for the research reactor. It also investigates the hydraulic phenomena in the heavy water system that can be generated by the HWDS. The analysis was conducted on the several parameters namely, three design candidates of HWDS, piping size, valve type, dumping height, and valve running time, using the MARS-KS system analysis code. Results were used to assess the major factors affecting the system performance. An optimal design was developed to minimize the amount of stagnant heavy water in the dumping pipe and ensure that the dumping time was within 15 s. The findings of this study are applicable to research reactors that do not have a second shutdown system and to the next-generation research reactors, and are expected to contribute in enhancing the safety of research reactors.

Study of cooling strategy for pressure vessel in pool research reactor at accident situation

Abstract Pressure vessel in pool layout is usually used in high power and middle pressure research reactor, Flow inversion and residual heat removal is important problem. The heat removal strategy at accident situation for a pressure vessel in pool layout research reactor was studied in the paper, many cases was researched for the ability of heat removal and safety of core.

The use of U3Si2/Al dispersion fuel for high power research reactors

In order to investigate the irradiation performance of U3Si2/Al dispersion fuel at relatively aggressive reactor conditions, compared to how this fuel is typically used in a research or test reactor, post irradiation examination was performed on two fuel plates that were irradiated at ∼270 W/cm2 average surface heat flux and to a maximum local burnup greater than 8 × 1021 fissions/cm3 in the Advanced Test Reactor as part of the RERTR-8 experiment. As part of the non-destructive examinations that were performed on both fuel plates, detailed thickness measurements using a high-fidelity measurement bench were performed on one fuel plate and a segment of the second fuel plate. This paper provides the first reported results of such high-resolution thickness measurements for U3Si2/Al dispersion fuel plates, containing primarily the U3Si2 fuel phase, irradiated at such power levels to high burnup. For destructive examination using optical metallography, samples were generated from both irradiated plates. Localized regions of high swelling (characteristic of internal blistering) were observed on the surface of both fuel plates, suggesting that the performance limits had been reached for each plate. Destructive examinations showed that the fuel particles that are primarily U3Si2, which become amorphous during irradiation, in both plates contained relatively large fission gas bubbles with evidence of fission gas bubble interconnection.

Exploring the electronic structure and thermal properties of UAl3 using density functional theory calculations

Among the intermetallic compounds of uranium, U-Al alloy is found to be promising nuclear fuel for high-power research reactors due to its high thermal conductivity and structural stability. U-Al system can form three different intermetallic compounds namely, UAl2, UAl3 and UAl4. Here, we systematically explored the electronic structure and various thermal properties of the cubic UAl3 system. Pseudo-potential plane wave based DFT calculations are used along with the spin-orbit coupling. The structural parameter calculated using the generalised gradient approximation is comparable to the experimental results. The phonon dispersion plot indicates the dynamic stability of the structure. Thermophysical properties including free energy, molar specific heat, coefficient of thermal expansion, bulk modulus, etc. have been evaluated within the framework of quasi-harmonic approximation. Both the electronic and lattice contribution to thermal conductivity has been evaluated through the Boltzmann transport theory.

Advanced passive residual heat removal system for high power pool-type research reactor

A novel passive residual heat removal system is proposed to enhance the safety and cost effectiveness of open pool-type research reactor with high power. As the pool-type research reactors typically have forced downward flow through the reactor core, a stagnant coolant in the middle of flow transition from forced to natural circulation can cause damage to fuel bundles. The main strategy of the proposed system is to delay the flow inversion time by prolonging the core downward flow until the decay heat is significantly calmed down. A gravity core cooling tank is devised to drag the downward flow on by gravity. To demonstrate the new system with the gravity core cooling tank, experimental results from a small-scale facility were compared with results from theoretical analysis and CFD simulation. The theoretical prediction of a 15 MWt real scale system shows that the advanced passive residual heat removal system can perform as an intermediate heat removal system to natural convection by providing sufficient mass flow to remove decay heat from a high power research reactor.

Characterization of surface layers formed on DU10Mo ingots after processing steps and high humidity exposure

The design of monolithic UMo fuel elements fabricated from low-enriched uranium for use in high-power research reactors requires bonding of the fuel foil to either Al cladding or a Zr barrier layer. Processing of the UMo ingot to final foil form has the potential to generate surface layers on the foil that differ from the bulk, metallic UMo. The interfacial properties between the UMo and Zr or Al cladding layers will then be determined by these surface layers. We use x-ray photoelectron spectroscopy, cross-sectional scanning electron microscopy, and atomic force microscopy to characterize the composition, oxidation state, and morphology of the surface layers that form after hot rolling and cold rolling depleted U–10 wt% Mo alloy (DU10Mo). A thick uranium nitride layer is observed after hot rolling, although its origin is likely from a previous processing step. The efficacy of acid etching in HNO3 is compared to that of electropolishing in H2SO4 to remove surface nitride and oxide layers, and both methods are found to be similarly effective. Both laboratory (low humidity) air exposure and longer rinse times in water are shown to promote the formation of surface oxide layers. Exposure of both acid-etched and electropolished DU10Mo foils to humid air (97% relative humidity) for six weeks results in formation of a thick oxide layer due to corrosion. The oxide layer on the acid-etched foil is thicker and more highly oxidized than the oxide layer that forms on the electropolished foil, and these differences in oxidation behavior are attributed to higher surface roughness on the acid-etched foil. In general, Mo is found to play a role as a sacrificial element, typically exhibiting a larger ratio of Mo6+/Mo4+ than U6+/U4+. This is unexpected, given the greater thermodynamic driving force to form U oxides than Mo oxides.

Interaction between Al and atomic layer deposited (ALD) ZrN under high-energy heavy ion irradiation

Uranium-molybdenum (U-Mo) particles dispersed in an aluminum matrix is the most promising candidate fuel to convert high-power research and test reactors in Europe from using high-enriched to using low-enriched fuel. However, chemical interaction between the U-Mo and the Al matrix leads to undesirable fuel behavior. Zirconium nitride (ZrN) is used as a diffusion barrier between the U-Mo fuel particles and the Al matrix. To understand the potential microstructural evolution of ZrN during irradiation, a high-energy heavy ion (84 MeV Xe) irradiation experiment was performed on atomic layer deposited (ALD) nanocrystalline ZrN deposited on an Al plate. A fluence of 1.86×1017 ions/cm2, or 90.3 dpa was reached during this experiment. Both analytic transmission electron microscopy (TEM) and synchrotron microbeam X-ray diffraction (μXRD) techniques were utilized to investigate the kinetics of radiation-induced grain growth of ZrN at various radiation doses based on the Williamson-Hall analyses. The grain growth kinetics can be described by a power law expression, Dn−D0n=Kϕ, with n = 5.1. The Al-ZrN interaction products (Al3Zr and AlN) created by radiation-induced ballistic mixing/radiation-enhanced diffusion and their corresponding formation mechanism were determined from electron diffraction and elemental composition analysis. These experimental results were confirmed by first principle thermodynamic density functional theory (DFT) calculations. The results from this ion irradiation study were also compared to in-pile irradiation data from physical vapor deposited (PVD) ZrN samples for a comprehensive evaluation of the interaction between Al and ZrN and its influence on diffusion barrier performance.

Elevated temperature tensile tests on DU–10Mo rolled foils

Studies were completed to obtain tensile mechanical properties for uranium-10 wt.% molybdenum (U10Mo) foils which were subjected to four different thermomechanical processing conditions. UMo alloy foils are being investigated to support fuel conversion of high power research reactors from their current high enriched fuel form to a low enriched fuel form. Mechanical properties of the fuel foil have an effect on irradiation performance and fuel fabrication and therefore are required to support modeling and qualification of new low-enriched uranium monolithic fuel plate designs. The data contained in this document contributes to fuel qualification by fulfilling the requirement that physical properties related to fuel meat be established. It is expected that depleted uranium-10 wt% Mo (DU–10Mo) mechanical behavior is representative of the low-enriched U10Mo to be used in actual fuel plates; therefore DU–10Mo was studied to simplify material processing, handling, and testing requirements. In this report, the different thermomechanical treatments included variations of wrought hot and cold rolling reduction and post rolling annealing. Each of the four foils was hot rolled. After hot rolling reduction, three of the four foils were further reduced by cold rolling. One of the three was reduced a further 20% by cold rolling, and the remaining two were reduced 50% by cold rolling. Following cold rolling reduction, one of the two foils which had been reduced 50% by cold rolling was annealed at 650 °C. Performing this analysis allows assessment of the impact of foil fabrication history on the resultant tensile properties DU–10Mo fuel foils.Tensile properties of DU–10Mo at room temperature through approximately 400 °C determined from the tests conducted herein suggest the material is stronger and has lower ductility than what has been reported previously in the literature. The explanation for these differences has yet to be determined, but is likely related to differences in grain size and/or impurity content, and variation in fabrication history. At the highest temperatures tested (550 °C) better agreement between the values reported here and available literature was found. As expected, yield and ultimate tensile strength decreased with increasing test temperature. Generally, the yield stress for all foil processing conditions was found to be in the range of 1100 MPa for room temperature tests, and in the range of 200 MPa for tests conducted at 550 °C. Ultimate tensile stress was in the range of 1175 MPa at room temperature, decreasing to approximately 225 MPa at 550 °C. Elongation increased significantly, from 0 to 2% at room temperature to 50% or more for the tests at 550 °C.

Analysis and comparison of focused ion beam milling and vibratory polishing sample surface preparation methods for porosity study of U-Mo plate fuel for research and test reactors

Uranium-Molybdenum (U-Mo) low enriched uranium (LEU) fuels are a promising candidate for the replacement of high enriched uranium (HEU) fuels currently in use in a high power research and test reactors around the world. Contemporary U-Mo fuel sample preparation uses focused ion beam (FIB) methods for analysis of fission gas porosity. However, FIB possess several drawbacks, including reduced area of analysis, curtaining effects, and increased FIB operation time and cost. Vibratory polishing is a well understood method for preparing large sample surfaces with very high surface quality. In this research, fission gas porosity image analysis results are compared between samples prepared using vibratory polishing and FIB milling to assess the effectiveness of vibratory polishing for irradiated fuel sample preparation. Scanning electron microscopy (SEM) imaging was performed on sections of irradiated U-Mo fuel plates and the micrographs were analyzed using a fission gas pore identification and measurement script written in MatLab. Results showed that the vibratory polishing method is preferentially removing material around the edges of the pores, causing the pores to become larger and more rounded, leading to overestimation of the fission gas porosity size. Whereas, FIB preparation tends to underestimate due to poor micrograph quality and surface damage leading to inaccurate segmentations. Despite the aforementioned drawbacks, vibratory polishing remains a valid method for porosity analysis sample preparation, however, improvements should be made to reduce the preferential removal of material surrounding pores in order to minimize the error in the porosity measurements.

Microstructure of as atomized and annealed U-Mo7 particles: A SEM/EBSD study of grain growth

Significant progresses in the performances under in-pile irradiation of particular U-Mo based fuels have been observed over the last fifteen years. One of the remaining issues has still to be tackled for use as a LEU fuel in the high power research reactors: the U-Mo recrystallization and its associated swelling have to be controlled or delayed. One way to mitigate this problem would be to optimize the initial microstructure of U-Mo atomized particle, by homogenizing Mo concentration and increasing grain size. This paper mainly focuses on U-Mo grain growth.Based on samples prepared in the framework of KOMO-5 and EMPIrE tests, a methodological work based on the use of EBSD is presented. In particular, surface preparation procedures are proposed for powders and rods, this last one being most likely readily applicable for plate analysis.As-atomized microstructures are analyzed in detail and subsequently compared to those obtained on particles annealed at 1000 °C under various conditions. It is found that 1 h annealing under vacuum is a good compromise of temperature and time to meet the development goals, provided that few impurity precipitates are present within U-Mo particles, since these can impact grain growth.

THERMAL-HYDRAULIC ANALYSIS OF IRT-4M IN REACTIVITY INSERTION ACCIDENT AT VR-1 REACTOR

The paper deals with thermal-hydraulic analysis during reactivity insertion accident, i.e. a step increase of nuclear system reactivity by 0.7 eff, at VR-1 Reactor. The reactor utilizes IRT-4M type of fuel assemblies, and even though these fuel assemblies are designed for an operation at the high-power research reactors, they might be also used for zero-power reactors. The thermal-hydraulic analyses must take into account several specific assumptions that are derived from VR-1 reactor specifications. The reactor does not require a forced water flow for a fuel cooling, the core is placed in an open vessel with atmospheric pressure, and amount of coolant water in the vessel is sufficient for providing the inlet water at room temperature for the whole event. Coolant circulation is expected to be formed only by natural convection.

Background Radiation Measurements at High Power Research Reactors

Background Radiation Measurements at High Power Research Reactors

Background radiation measurements at high power research reactors

Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. The general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

Improved Irradiation Performance of Uranium-Molybdenum/Aluminum Dispersion Fuel by Silicon Addition in Aluminum

Uranium-molybdenum fuel particle dispersion in aluminum is a form of fuel under development for conversion of high-power research and test reactors from highly enriched to low-enriched uranium in the U.S. Global Threat Reduction Initiative program (also known as the Reduced Enrichment for Research and Test Reactors program). Extensive irradiation tests have been conducted to find a solution for problems caused by interaction layer growth and pore formation between U-Mo and Al. Adding a small amount of Si (up to ~5 wt%) in the Al matrix was one of the proposed remedies. The effect of silicon addition in the Al matrix was examined using irradiation test results by comparing side-by-side samples with different Si additions. Interaction layer growth was progressively reduced with increasing Si addition to the matrix Al, up to 4.8 wt%. The Si addition also appeared to delay pore formation and growth between the U-Mo and Al.

Reduced interaction layer growth of U–Mo dispersion in Al–Si

Development of high U-density U–Mo fuel particle dispersion in Al is needed to convert high power research and test reactors from HEU to LEU. Interaction layer growth between U–Mo and Al poses a challenge to this goal. The KOMO-4 test was designed at KAERI and irradiated in the HANARO reactor to ∼50% burnup of initial 19.75% U-235 enrichment at ∼200°C. The main objective of the test was to examine the effect of the Si content in the matrix up to 8wt.%. U–Mo/Al–Si dispersion samples with a Si addition in the range 0–8wt.% in the matrix were tested. A sample with pre-irradiation Si-containing interaction layers (ILs) was also tested. As the Si content in the matrix increases, the IL growth was progressively reduced. Contrary to the thermodynamics prediction and out-of-pile observations, however, Si accumulation in the ILs occurred near the IL–matrix interface with only a slight increase in concentration. The effect of the pre-formed ILs was insignificant in reducing IL growth.

Properties of DU–10 wt% Mo alloys subjected to various post-rolling heat treatments

Studies were completed to obtain mechanical properties of depleted uranium–molybdenum (U–Mo) alloys subjected to different post-processing treatments using microhardness, quasi-static tensile tests, and scanning electron microscopy failure analysis. U–Mo alloy foils are currently under investigation for potential fuel conversion of high power research reactors to low enriched uranium fuel. Although mechanical properties take on a secondary effect during irradiation, an understanding of the alloy behavior during fabrication and the effects of irradiation on the integrity of the fuel are essential. In general, the microhardness was insensitive to annealing temperature but decreased with annealing duration. Yield strength, Young's modulus, and ultimate tensile strength were affected in varying manners with both increasing annealing temperature and duration, and subjecting the alloy to rolling. The failure mode was insensitive to annealing conditions, but was significantly controlled by the impurity concentration of the alloy, especially carbon. Values obtained from literature are also provided with reasonable agreement based on extrapolation of annealing duration, even though processing conditions and applications were quite different in some instances.

Mechanical Properties of DU-xMo Alloys with x = 7 to 12 Weight Percent

Mechanical properties of six depleted uranium-molybdenum (U-Mo) alloys have been obtained using microhardness, quasistatic tensile tests, and scanning electron microscopy (SEM) failure analysis. U-Mo alloy foils are currently under investigation for potential conversion of high power research reactors to low enriched uranium fuel. Although mechanical properties take on a secondary effect during irradiation, an understanding of the alloy behavior during fabrication and the effects of irradiation on the integrity of the fuel is essential. In general, the microhardness, yield strength, Young’s modulus, and ultimate tensile strength improved with increasing Mo content. Microhardness measurements were very sensitive to local composition, while the failure mode was significantly controlled by the impurity concentration of the alloy, especially carbon. Values obtained from literature are also provided with reasonable agreement, even though processing conditions and applications were quite different in some instances.