High Flux Research Reactor Research Articles

Neutron detection in the frame of spatial magnetic spin resonance

This work is related to neutron detection in the context of the polarised neutron optics technique of spatial magnetic spin resonance. By this technique neutron beams may be tailored in their spectral distribution and temporal structure. We have performed experiments with very cold neutrons (VCN) at the high-flux research reactor of the Institut Laue Langevin (ILL) in Grenoble to demonstrate the potential of this method. A combination of spatially and temporally resolving neutron detection allowed us to characterize a prototype neutron resonator. With this detector we were able to record neutron time-of-flight spectra, assess and minimise neutron background and provide for normalisation of the spectra owing to variations in reactor power and ambient conditions at the same time.

The Characteristics and Irradiation Capabilities of MARIA Research Reactor in NCBJ Świerk

The high flux research reactor MARIA is a water and beryllium moderated reactor of a pool type with graphite reflector and pressurised channels containing concentric six-tube assemblies of fuel elements. It has been designed to provide high degree of flexibility. The fuel channels are situated in a matrix containing beryllium blocks and enclosed by lateral reflector made of graphite blocks in aluminium cans. The MARIA reactor is equipped with vertical channels for irradiation of target materials, a rabbit system for short irradiations and six horizontal neutron beam channels. The MARIA reactor reached its first criticality in December 1974. The reactor was in operation until 1985 when it was shut down for modernization. The modernization encompassed refurbishment and upgrading of technological systems such as: enlargement of beryllium matrix, inspection of the graphite bocks, upgrading of ventilation and temperature systems.

Investigation of Experimental Dispersion Fuel Elements with an Aluminum-Based Matrix for High-Flux Research Reactors

A variant of experimental dispersion fuel elements with fuel based on uranium dioxide as fissile material in a silumin matrix, which have been proposed for modernizing the core of the SM reactor and are irradiated in a wide spectrum of neutron-physical and thermophysical parameters in the reflector channel of the reactor, is investigated. A nonuniformity with fuel density two times higher than the average over a fuel element, viz., a cluster of uranium dioxide particles in the central zone of the kernel, has been found in the distribution of the fuel phase in the fuel elements. The thermophysical calculations show that for such a distribution of fuel particles the temperature of the fuel composition with heat flux density from the fuel-element surface >6 MW/m2 can reach the melting temperature of the matrix. This is confirmed by investigations of fuel elements with appreciably larger volume and shape change. To increase the serviceability of the fuel elements, ways of improving their design and the material composition of the fuel mix are proposed.

Beijing Isotope-Separation-On-Line Neutron-Rich Beam Facility and its application prospects

Radioactive-ion beam (RIB) experiments open up new exciting fields and frontiers in nuclear physics and nuclear astrophysics. The experiments can provide unprecedented opportunities to explore new phenomena such as new magic numbers and probe the isospin mechanisms occurring in shell evolution. Other experiments probe the limits of nuclear stability on the neutron-rich side, provide insights into the synthesis of superheavy elements and nuclear giant halo structures, and offer the possibility to investigate new nuclear decay modes as well as the nucleosynthesis of nuclides beyond Iron. To achieve this goal, many nuclear physics laboratories all over the world are building large-scale powerful new RIB facilities. The “Beijing Isotope-Separation-On-Line Neutron-Rich Beam Facility (BISOL)” is a joint project proposed by the China Institute of Atomic Energy (CIAE) and Peking University. This facility will combine two techniques: Projectile fragmentation (PF) and isotope separation on line (ISOL). The BISOL facility will operate based on a dual driver system, including reactor driving and intense deuteron-beam driving modes. The existing high flux research reactor, the China Advanced Research Reactor at CIAE provides reactor driving. At the BISOL facility ISOL separation of fission fragments, post acceleration, and fragmentation of neutron-rich fission fragment beams will be available. By combining ISOL and PF techniques in one facility the technical advantages of both can be merged. The BISOL will deliver intensive rare ion beams of very neutron-rich isotopes; the intensity of these beams will be of the order of at least 1-2 magnitudes higher than currently available ones. Intense deuteron-beam driving is an established technology, which provides an intense neutron beam source and can be operated independently in the facility. Characterizing material performance and performing lifetime experiments are the challenges facing researchers working on the current and next-generation nuclear energy systems. Intense radiation sources, especially neutron sources, are indispensable for these experiments. The intense deuteron accelerator proposed in this project will be able to generate intense neutron beam. These beams will be able to probe nuclear materials and will provide vital results to push forward the boundaries of nuclear physics.

Heat Transfer Calculation on Plate-Type Fuel Assembly of High Flux Research Reactor

Heat transfer characteristics of fuel assemblies for a high flux research reactor with a neutron trap are numerically investigated in this study. Single-phase turbulence flow is calculated by a commercial code, FLUENT, where the computational objective covers standard and control fuel assemblies. The simulation is carried out with an inlet coolant velocity varying from 4.5 m/s to 7.5 m/s in hot assemblies. The results indicate that the cladding temperature is always lower than the saturation temperature in the calculated ranges. The temperature rise in the control fuel assembly is smaller than that of the standard fuel assembly. Additionally, the assembly with a hot spot is specially studied, and the safety of the research reactor is also approved.

Light and heavy water dynamic structure factor for neutron transport codes

In this study, we report on recent neutron inelastic scattering experiments performed at the Institut Laue-Langevin (ILL) for H2O and D2O. The measured dynamic structure factors S(q, ω) have been reduced, normalised and transformed into the S(α, β) formalism, where α and β stand for the unit-less momentum and energy transfers, respectively. The measurements were complemented with molecular dynamics simulations. After processing with NJOY, new water neutron scattering cross-sections have been generated for use with e.g. the Monte Carlo N-Particle (MCNP) software in view to improve the accuracy of the nuclear facility models. As an example, we present improved accuracy calculations for the safety rod insertion impact on the criticality factor keff for the ILL high flux research reactor.

Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel

U–Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800°C of in-pile irradiated U–Mo/Al–Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500°C and 670°C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500°C FGs are released from IDL/matrix interfaces. The second peak at 670°C representing the main release of FGs originates from the interaction between U–Mo and matrix in the vicinity of the cladding.

Evidence of amorphous interdiffusion layer in heavy ion irradiated U–8wt%Mo/Al interfaces

Abstract U–Mo/Al based nuclear fuels are worldwide considered as the most promising high density fuel for the conversion of high flux research and test reactors from highly enriched uranium to lower enrichment. However in-pile growth of an amorphous interdiffusion layer at the U–Mo/Al interfaces strongly limits the performances of this fuel. Several in-pile tests have been performed to optimize the composition. In this paper, a breakthrough in simulating the U–8wt%Mo/Al behavior under out-of-pile irradiation is reported. It is shown that an amorphous U–8wt%Mo/Al interdiffusion layer (IDL) is obtained by heavy ion irradiation ( 127 I) in a U–Mo/Al diffusion couple under controlled temperature conditions. The properties of this IDL coincide with the results obtained from in-pile tests. This methodological work clearly indicates that heavy ion irradiations could be routinely applied for optimizing composition of U–Mo/Al nuclear fuels. In other words these out-of-pile tests using ion beams could become a representative, efficient and economic step before in-pile irradiation.

A multi-detector, digitizer based neutron depth profiling device for characterizing thin film materials

Neutron depth profiling (NDP) is a mature, nondestructive technique used to characterize the concentration of certain light isotopes in a material as a function of depth by measuring the residual energy of charged particles in neutron induced reactions. Historically, NDP has been performed using a single detector, resulting in low intrinsic detection efficiency, and limiting the technique largely to high flux research reactors. In this work, we describe a new NDP instrument design with higher detection efficiency by way of spectrum summing across multiple detectors. Such a design is capable of acquiring a statistically significant charged particle spectrum at facilities limited in neutron flux and operation time.

Neutron-Use Optimization with Virtual Experiments to Facilitate Research-Reactor Conversion to Low-Enriched Fuel

Converting research reactors from highly enriched uranium (HEU) fuel to more proliferation-resistant low-enriched fuel is critical for achieving the objective of ending the use of directly weapon-usable materials in the civilian nuclear fuel cycle. The most challenging type of reactors to convert are high-flux research reactors, which, along with upcoming strong spallation sources, are the most important neutron sources for sophisticated neutron scattering experiments. Advanced Monte-Carlo computer codes are now available that make it possible to track neutrons from the neutron source, through neutron guides, to the detector of a neutronic experimental setup, including realistic samples. These “virtual experiments” allow optimizing the performance of complete beamlines, where in many cases a large unused potential exists for increasing the neutron flux at the sample or detector position. The Monte-Carlo codes VITESS and McStas are used to compare results for typical neutron scattering setups using typical versus state-of-the-art technologies. The analysis shows that performance gains due to instrument upgrades or neutron guide renewals can dwarf potential neutron flux losses due to conversion to low-enriched fuel. Combined convert-and-upgrade strategies therefore offer unique opportunities for reactor operators and neutron scientists to significantly improve the overall performance of research facilities, and turn them into centers of excellence, while supporting the objective of phasing out the use of highly enriched uranium in the civilian nuclear fuel cycle as soon as possible.

Quantitative crystallographic analysis of as-fabricated full size U–Mo/Al(Si) nuclear fuel plates

The optimisation of high density metallic U–Mo/Al(Si) nuclear fuels, developed for use in high neutron flux research reactor without proliferation issues, requires an excellent control of the manufacturing conditions for getting a full benefit of Si powder additions. Eleven full size plates (from the IRIS and E-FUTURE experimental programs) have been analysed by diffraction using synchrotron high energy X-rays. For the first time, both the levels of the γU–Mo phase destabilisation in the U–Mo particle and characteristics (size and composition) of the Si rich protective layers around these particles have been quantified at a macroscopic scale. These results enable to evaluate the influence of the different parameters implied in the manufacturing process and could also provide solutions for producing enhanced fuels.

Technetium-99m production issues in the United Kingdom

Nuclear Medicine developed when it was realised that a radioisotopic substitution of Iodine-131 for the stable Iodine-127 would follow the same metabolic pathway in the body enabling the thyroid to be imaged and the thyroid uptake measured. The Iodine could be complexed with pharmaceutical substrates to enable other organs to be imaged, but its use was limited and high gamma energy and beta emission restricted the activity of each radiopharmaceutical used, leading to long acquisition times and degraded images. As a pure gamma emitter of 140 keV and with a 6-h half-life, Technetium-99m is a better radionuclide and images a wider range of bodily organs. However, its short half-life also requires it to be eluted from its mother radionuclide, Mo-99, in a generator, delivered weekly from radiopharmaceutical companies who obtain the Mo-99 in liquid form from high-flux research reactors. All went well till around 2007, when the NRU Reactor in Canada was closed and all other reactors went down for various periods for unrelated problems, leading to widespread Mo-99 shortages. Although the reactors have since recovered, they are 48 to 57 years old, and it seems that few governments have made any future provision such as building replacement reactors.

Neutron Visual Sensing Techniques Using the High Intensity Neutron Sources

In this manuscript, the fundamentals and merits of neutron visual sensing techniques, high-speed consecutive visual sensing technique, high-speed-scanning dynamic 3D visualization technique, multi-beam 3D velocimetry technique and energy-selective neutron imaging technique are summarized. These techniques have been developed mainly for the R&D of the nuclear energy system using the high-flux research reactors, JRR-3 and JRR-4, and a high-intensity pulsed neutron source, J-PARC. Very unique visualization and measurement, e.g. dynamic 3D velosimetry in JRR-4 were successfully carried out by these neutron visual sensing techniques. An energy-selective neutron imaging, e.g. the neutron resonance absorption imaging in J-PARC is extremely unique and newly developed visualization and measurement technique. Analytical investigation for the feasibility of multiphase flow experiment in J-PARC was also reported using a Monte Carlo neutron simulation code. It was confirmed from the analytical work that the measurement error of the void fraction of two-phase flow can be reduced by the energy-selective neutron imaging technique.

From High-Enriched to Low-Enriched Uranium Fuel in Research Reactors

Since the 1970's, global efforts have been going on to replace the high-enriched (>90% 235 U), low-density UAI x research reactor fuel with high-density, low-enriched (<20% 235 U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched material because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative (GTRI) and the Reduced Enrichment for Research and Test Reactors (RERTR) program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high-enriched ones without loss of performance. Most success has presently been obtained with U 3 Si 2 dispersion fuel, which is currently used in many research reactors in the world. However, efforts to search for a replacement with even higher density, which will also allow the conversion of some high flux research reactors that currently cannot change to U 3 Si 2 (eg. BR2 in Belgium), have continued and are for the moment mainly directed towards the U(Mo) alloy fuel (7-10 w% Mo). This paper provides an overview of the past efforts and presents the current status of the U(Mo) development.

Computational-experimental studies of the neutronics of the SM reactor with different variants of the neutron-trap arrangement

The results of experimental studies of the neutronics of the high-flux SM reactor with different arrangements of the neutron trap are presented. The MCU series of high-precision computer programs implementing the Monte Carlo method is used for computations. Experimental data on reactivity effects, the effectiveness of safety and control rods, and the coefficients of nonuniformity of energy release in the core have been obtained in experiments on a critical assembly ‐ a physical model of the SM reactor ‐ and directly in experiments in the reactor. The error is 4.2‐10% in determining the reactivity parameters and 5‐10% for the relative energy release in the fuel elements. Information on the neutron field formed in the volume of the neutron trap has been obtained for two arrangements of the beryllium and water moderators. The differential and integral energy spectra of the neutrons in the energy interval from 0.5 eV to 20 MeV are obtained for three points inside the trap (external, central series, center). The flux density of thermal, superthemal, and fast neutrons are determined. The SM high-flux research reactor was designed to produce radionuclides and test reactor materials, fuel compositions, and fuel elements [1]. The results of the studies performed with this reactor are used to substantiate the serviceability and safety of nuclear reactors; the radionuclides produced find applications in industry and medicine. The SM reactor has been operated with three different arrangements of the neutron trap. In 1961‐1990, an experimental channel which was connected to an autonomous secondary loop with light water as the coolant was located at the center of the reactor. The next phase of the construction of the reactor, the purpose of which was to increase operational safety, was completed in 1991‐1992 [2]. To increase 252 Cf production, the radionuclide in greatest demand at the time, the water neutron trap was replaced with a centralized beryllium block for holding transuranium targets. The coolant in the primary loop of the reactor started to remove heat from the central block. The cross section of the rebuilt SM reactor is shown in Fig. 1. In 2002, the central beryllium block in the central moderating cavity of the reactor was replaced with a separator structure consisting of 27 14-mm in diameter and 0.5-mm thick zirconium tubes with water in the space between the tubes. The purpose of the change in the construction was to increase production efficiency of 63 Ni, 75 Se, and other isotopes which were in demand at the time and to accumulate which a low-energy neutron spectrum with high flux density is required. The cross section of the central moderating cavity of the SM reactor with different arrangements of the neutron trap is shown in Fig. 2.

Development of a fuel element with low harmful neutron absorption for the SM high-flux research reactor

Information about the SM research reactor and its characteristic features and advantages over other research reactors is presented. The reasons for updating the reactor and the optimal method of solving the problem are indicated. The upgrade program preserves the essential structural features of the rector and allows for the insertion of additional irradiation channels in the fuel part of the core by removing some fuel elements. The reactivity loss arising in so doing is compensated by increasing the uranium content in the remaining fuel elements. A new type of fuel element based on materials with reduced harmful absorption of neutrons is being developed to improve further the technical and economic performance of the reactor. The design and the technology of the fuel element have been developed for three implementations, and experimental fuel elements for reactor tests have been fabricated. The fuel elements have been checked for adherence to the requirements. It has been shown that normal operation of the fuel elements is possible with heat flux density at the surface 9-12 MW/m 2 , which meets the initial requirements.

Transmutation of Technetium into stable Ruthenium in high flux conceptual research reactor

The effectiveness of transmutation for the long-lived fission product Technetium-99 in a high flux research reactor is evaluated, taking into account its large capture cross section in the thermal and epithermal regions. Calculation of the evolution of the Ruthenium concentration under irradiation was performed using the ChainSolver 2.20 code. The results on Technetium transmutation suggest an effective use of this kind of research reactors and describes a new concept of multi-recycle Technetium transmutation using HFRTRAN (high flux research reactor for transmutation).

Radiation resistance of high-density uranium—molybdenum dispersion fuel for nuclear research reactors

The results of radiation tests are discussed and the character of the failure of fuel compositions and the operability of fuel elements is analyzed as a function of the type of fuel and the irradiation conditions. The intense interaction of the fuel with the matrix material is considered as the main factor limiting the operability of fuel elements in power-dense high-flux nuclear reasearch reactors. It is concluded that low-enrichment high-density uranium—molybdenym fuel can provide reliable operation of dispersion fuel elements in low-and medium-power research reactors. Such fuel can be used in power-dense high-flux research reactors if the fuel load is decreased below the maximum admissible amount, the compatibility of the uranium—molybdenum alloy with an aluminum matrix is radically improved, or fuel elements with a different construction, for example, monolithic, are used.

Coupling of Time-Dependent Neutron Transport Theory with the Thermal Hydraulics Code ATHLET and Application to the Research Reactor FRM-II

We introduce a new coupled neutronics/thermal hydraulics code system for analyzing transients of nuclear power plants and research reactors, based on a neutron transport theory approach. For the neutron kinetics, we have developed the code DORT-TD, a time-dependent extension of the well-known discrete ordinates code DORT. DORT-TD uses a fully implicit time integration scheme and is coupled via a general interface to the thermal hydraulics system code ATHLET, a generally applicable code for the analyses of LWR accident scenarios. Feedback is accounted for by interpolating multigroup cross sections from precalculated libraries, which are generated in advance for user-specified, discrete sets of thermal hydraulic parameters, e.g., fuel and coolant temperature. The coupled code system is applied to the high-flux research reactor FRM-II (Germany). Several design basis accidents are considered, namely the unintended control rod withdrawal, the loss of offsite power, and the loss of the secondary heat sink as well as a hypothetical transient with large reactivity insertion.

The SSC RIAR high-flux research reactors: experience and possibilities of testing materials and mock-ups for fusion

A significant part of the operation parameters of the high power loaded elements of a fusion reactor can be reproduced in a high-flux fission reactors with considerable neutron spectrum hardness. This paper presents basic characteristics of the SM and BOR-60 high-flux reactors that are being currently under operation at RIAR as well as some examples of their usage for scientific programs performed for validation of radiation resistance of structural materials and mock-ups of some fusion reactor units. In particular, at present the program on the examination of developed copper alloys, welded and brazed joints of copper//steel type after irradiation in the SM reactor is performed for validation of the ITER divertor plate. Beryllium of various grades is examined for DEMO reactor blanket. A unique experiment was performed on simulation of expected operation conditions of the ITER divertor plate using mock-ups under fission reactor conditions.