Uranium Plates Research Articles

Characterization and verification of low-enriched uranium plate using a non-destructive methodology

Non-destructive methods play a crucial role in evaluating the integrity of nuclear materials. During the production of low-enriched mini-uranium plates, some destructive tests were initially employed to assess parameters such as uranium meat density, homogeneity, 235U enrichment, and meat distribution within the clad. While these destructive methods provide precise measurements during the manufacturing process, they become impractical and costly post-fabrication. Consequently, non-destructive tests are employed for quality control, verification, and inspection of nuclear materials. Laboratories and factories are developing non-destructive methods that yield results closely aligned with destructive measurements. The present research focuses on creating non-destructive measurements for verifying 235U enrichment and meat distribution in mini-uranium plates. This innovative approach involves utilizing 235U gamma emission as input to calculate enrichment and meat distribution. High purity germanium (HPGe) and sodium iodide (NaI) detectors alongside MCNPX code for efficiency correction were used in the proposed non-destructive measurements, respectively. The obtained results were validated and verified against certified measurements and x-ray images of mini-uranium plate. The enrichment measurement exhibited consistency with certified measurements, with about 0.8% difference. Similarly, the meat distribution measurement aligned with the x-ray image of the plate. This research proposes a practical application for nuclear material control during mini-uranium plate fabrication and for ensuring accountability and verification of inventory in storage facilities.

Measurement of 233U-HU Substitution Reactivity Worth in KUCA for Integral Validation of 233U Nuclear Data

In order to perform an integral evaluation of 233U nuclear data, measurements of substitution reactivity worth between 233U sample and high enriched uranium (HU) sample in the Kyoto University Critical Assembly (KUCA) were carried out. In this study, the experimental core was consisted of 24 fuel elements and 1 special fuel element with the 233U or HU sample. The fuel element was consisted of 31 unit cells and sandwiched by a upper and a lower polyethylene reflector. The unit has one enriched uranium plate of 1/16” thickness, three polyethylene plates of 1/8” thickness. The special fuel element was loaded at the center of the core. The substitution reactivity worth was defined as difference between the excess reactivity of the core in which the 233U sample or the HU sample inserted the core. The excess reactivities ware measured by the positive period method. As the result, the substation reactivity was obtained 0.0146±0.0006 %∆k/k. A numerical result by MVP3 with JENDL-4.0 was 0.0141±0.0002 %∆k/k and a C/E was 0.964 ± 0.041. On the other hand, the calculated substitution reactivity worth with JENDL-5 was 0.0168±0.0002 %∆k/k and a C/E was 1.151±0.047.

Multimodal imaging of muon based on scattering and secondary induced neutrons

Muon scattering imaging technology can be used to detect nuclear material and is of considerable significance in nuclear safety. However, it is difficult to distinguish special nuclear materials from high-Z objects effectively by using the existing muon scattering imaging technologies. Muon-induced neutrons emitted from special nuclear materials can help to identify the existence of special nuclear materials. However, this method has long imaging time and low imaging quality. Multimodal imaging of muon uses both the information about scattering muons penetrating the material and the information about muons stopped by material and generating secondary induced neutrons, which can overcome the shortcomings of single imaging method effectively. The detection model is set up based on Geant4. The simulation programs of muon imaging in coincidence with muon induced neutrons, scattering imaging of muon, and multimodal imaging of muon are developed by using Cosmic-ray Shower Library as particle source, and the imaging algorithms are implemented respectively on the basis of the simulated data. Two imaging models are designed for muon scattering imaging. The first one is a single <sup>235</sup>U cube, and the second one is composed of four cubes, namely <sup>235</sup>U cube, <sup>239</sup>Pu cube, lead cube and aluminum cube. This simulation has completed muon scattering imaging of single cube and four cubes. In the part of muon imaging in coincidence with muon induced neutrons, the neutronic gain of the HEU (90% <sup>235</sup>U) plate, LEU (20% <sup>235</sup>U) plate, and DU (0.2% <sup>235</sup>U) plate, as well as the relationship between the neutronic gain of these three uranium plates and the energy and charged properties of the muon are obtained by simulation, and then two imaging models are set up. The first one is composed of four cubes, namely <sup>235</sup>U cube, <sup>239</sup>Pu cube, lead cube, and aluminum cube, and the other is comprised of multilayer nuclear components. The 2D and 3D reconstruction results of multi-objects and multilayer nuclear components are obtained through muon imaging in coincidence with muon induced neutrons. Then the multimodal imaging of muon for three cubes is realized in the presence or absence of iron shielding shell. The imaging capabilities are compared with the muon scattering imaging capacities and muon imaging capacities in coincidence with muon induced neutrons. Simulation studies indicate that multimodal imaging of muon based on scattering and secondary induced neutrons can effectively combine the advantages of every single imaging method. The multimodal imaging of muon can take advantage of available information more efficiently, which is helpful in improving the imaging quality. Multimodal imaging of muon not only has the advantages of short imaging time and high imaging quality, but also can distinguish special nuclear material from other high-Z materials clearly, which is vital for detecting special nuclear materials.

REACTOR PHYSICS EXPERIMENT IN A GRAPHITE-MODERATION SYSTEM FOR HTGR

The Japan Atomic Energy Agency (JAEA) started the Research and Development (R&D) to improve nuclear prediction techniques for High Temperature Gas-cooled Reactors (HTGRs). The objectives are to introduce a generalized bias factor method to avoid full mock-up experiment for the first commercial HTGR and to introduce reactor noise analysis to High Temperature Engineering Test Reactor (HTTR) experiment to observe sub-criticality. To achieve the objectives, the reactor core of graphite-moderation system named B7/4”G2/8”p8EUNU+3/8”p38EU(1) was newly composed in the B-rack of Kyoto University Critical Assembly (KUCA). The core is composed of the fuel assembly, driver fuel assembly, graphite reflector, and polyethylene reflector. The fuel assembly is composed of enriched uranium plate, natural uranium plate and graphite plates to realize the average fuel enrichment of HTTR and it’s spectrum. However, driver fuel assembly is necessary to achieve the criticality with the small-sized core. The core plays a role of the reference core of the bias factor method, and the reactor noise was measured to develop the noise analysis scheme. In this study, the overview of the criticality experiments is reported. The reactor configuration with graphite moderation system is rare case in the KUCA experiments, and this experiment is expected to contribute not only for an HTGR development but also for other types of a reactor in the graphite moderation system such as a molten salt reactor development.

Benchmarks of Criticality in Solid-Moderated and Solid-Reflected Cores at Kyoto University Critical Assembly

To elucidate the accuracy of benchmarks of criticality at the Kyoto University Critical Assembly (KUCA), uncertainty analysis is conducted for manufacturing tolerances in highly enriched uranium (HEU) plates and modeling of core configurations in addition to nuclear data. For evaluation of eigenvalue bias, eigenvalue calculations are conducted using MCNP6.1 and SCALE6.2/KENO-VI together with ENDF/B-VII.1. The modeling of reference core configurations and material properties with average values is validated through a comparison between calculated and measured results. The uncertainty induced by nuclear data is evaluated with SCALE6.2/TSUNAMI-3D together with ENDF/B-VII.1 for sensitivity calculations and 56groupcov.7.1 for the covariance matrix. In the breakdown of the uncertainty induced by nuclear data, the impact of 235U shows significant dominance, about 900 pcm in hard and soft spectrum cores. Furthermore, uncertainty evaluation by manufacturing tolerances in HEU plates and reproducibility of control rod positions demonstrates that the impact of variation on measured reactivity is minor. Through experimental analyses, the index of accuracy in benchmark experiments of criticality is conducive to the reliability of benchmarks at KUCA.

Neutron imaging with fission and thermal neutrons at NECTAR at MLZ

The instrument NECTAR is located at beam port SR10 of the neutron source FRM II at the Heinz Maier-Leibnitz Zentrum (MLZ). With a pair of moveable uranium plates placed in front of the entrance window of the beam tube, a fission neutron spectrum with a mean energy of 1.9 MeV can be used for neutron imaging applications. Via remote control these plates can be removed and a thermal neutron spectrum (mean energy at 28 meV) gets available for experiments. While the fission neutron spectrum is regularly used, some upgrades of the instrument are necessary to make the thermal neutron spectrum routinely available for user experiments. This includes additional equipment like a new sample stage and a second detector system foreseen to extend the capabilities of NECTAR. The current state of the instrumentation and necessary changes for the future thermal beam option and its usage for standard user experiments will be presented. First measurements were carried out with a temporary flight tube installed and a compact detector (510 mm × 180 mm x 180 mm) for thermal neutrons with a spatial resolution in the range of 100 μm. The feasibility of the thermal beam option could already be verified at an L/D ratio of 240 and a thermal neutron flux of 7.92·106 cm−2 s−1. The thermal neutron beam option adds a pure thermal neutron spectrum – Maxwell spectrum originating from the moderator without alteration by a secondary source or converter – to the energy ranges available for neutron imaging at MLZ instruments. It also offers a unique possibility to combine two quite different neutron energy ranges at a single instrument including their respective advantages. The thermal neutron beam option is funded by BMBF in the frame of research project 05K16VK3.

Poly- Versus Mono-Energetic Dual-Spectrum Non-Intrusive Inspection of Cargo Containers

In this paper, based on an invited talk at SORMA (May 2016), we present an overview of x-ray sources, detectors and system configurations for non-intrusive inspection (NII) of cargo containers. Our emphasis is on dual-energy x-ray NII for detecting high-atomic-number ( $\text {Z}\geq 72$ ) materials such as tungsten shielding and special nuclear materials (SNM). Standard single-energy (MeV and above) x-rays needed to penetrate and image cargo provide little SNM contrast, whereas dual-energy x-ray NII is demonstrated as a way to improve the selectivity of materials with Z $\text {Z}\geq 72$ . For two possible dual-energy x-ray source technologies—polyenergetic dual-energy (PDE) and quasi-monoenergetic x-ray sources (QMXS)—we investigate their trade-offs and future prospects using experimental and simulated results. The reduced scatter and larger separation of low- and high-energy photons provided by QMXS offers improved high-Z material contrast, but practical considerations such as flux and pulse rate need to be solved before making a deployable system. Straight-ray simulations show factor of four increases in contrast for QMXS over PDE scans of tin (Z=50) and iron (Z=26) relative to a uranium plate (Z=92) behind 20 cm of iron simulated cargo.

The Thermal Neutron Beam Option for NECTAR at MLZ

The beam port SR10 at the neutron source FRM II of Heinz Maier-Leibnitz Zentrum (MLZ) is equipped with a moveable assembly of two uranium plates, which can be placed in front of the entrance window of the beam tube via remote control. With these plates placed in their operating position the thermal neutron spectrum produced by the neutron source FRM II is converted to fission neutrons with 1.9 MeV of mean energy. This fission neutron spectrum is routinely used for medical applications at the irradiation facility MEDAPP, for neutron radiography and tomography experiments at the facility NECTAR and for materials testing. If, however, the uranium plates are in their stand-by position far off the tip of the beam tube and the so-called permanent filter for thermal neutrons is removed, thermal neutrons originating from the moderator tank enter the beam tube and a thermal spectrum becomes available for irradiation or activation of samples. By installing a temporary flight tube the beam may be used for thermal neutron radiography and tomography experiments at NECTAR. The thermal neutron beam option not only adds a pure thermal neutron spectrum to the energy ranges available for neutron imaging at MLZ instruments but it also is an unique possibility to combine two quite different neutron energy ranges at a single instrument including their respective advantages.The thermal neutron beam option for NECTAR is funded by BMBF in frame of research project 05K16VK3.

Adjustment method of deterministic control rods worth computation based on measurements and auxiliary Monte Carlo runs

The control rods (CRs) worth is key parameter for the research reactors (RRs) operation and utilization. Control rods worth computation is a challenge for the full deterministic calculation methodology, including the few group cross section generation, and the core analysis. The purpose of this work is to interpret our codes system, and their applicability of obtaining reliable CRs worth by some engineering adjustments. Cross sections collapsing in three energy groups is made by WIMS and SN2 codes, while the core analysis is performed by CITATION. We use these codes for the design, construction, and operation of our research reactor CMRR (China Mianyang Research Reactor). However, due to the intrinsic deficiency of the diffusion theory and homogenizing approximation, the directly obtained results, such as CRs worth and neutron flux distributions are not satisfactory. So two points of simple adjustments are made to generate the few group cross-sections with the assistance of measurements and auxiliary Monte Carlo runs. The first step is to adjust the fuel cross sections by changing properly the mass of a non-fissile material, such as the mass of the two 0.4mm Cd wires existing at both sides of each uranium plate, so that the core model of CITATION can get good eigenvalue when all CRs are completely extracted. The second step is to revise the shim absorber cross section of CRs by adjusting the hafnium mass, so that the CITATION model can get correct critical rods position. In this manuscript, the JRR-3M (Japan Research Reactor No. 3 Modified) reactor is employed as a demonstration. Final revised results are validated with the stochastic simulation and experimental measurement values, including the critical rods position, the differential/integral CR curves, and the thermal neutron flux distributions.

Production of Fission Product 99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis

Abstract The main objective of 235U irradiation is to obtain the 99mTc isotope, which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short lifetime, is a reaction of radioactive decay of 99Mo into 99mTc. One of the possible sources of molybdenum can be achieved in course of the 235U fission reaction. The paper presents activities and the calculation results obtained upon the feasibility study on irradiation of 235U targets for production of 99Mo in the MARIA research reactor. Neutronic calculations and analyses were performed to estimate the fission products activity for uranium plates irradiated in the reactor. Results of dummy targets irradiation as well as irradiation uranium plates have been presented. The new technology obtaining 99Mo is based on irradiation of high-enriched uranium plates in standard reactor fuel channel and calculation of the current fission power generation. Measurements of temperatures and the coolant flow in the molybdenum installation carried out in reactor SAREMA system give online information about the current fission power generated in uranium targets. The corrective factors were taken into account as the heat generation from gamma radiation from neighbouring fuel elements as well as heat exchange between channels and the reactor pool. The factors were determined by calibration measurements conducted with aluminium mock-up of uranium plates. Calculations of fuel channel by means of REBUS code with fine mesh structure and libraries calculated by means of WIMS-ANL code were performed.

Modeling and analysis of high-explosive driven perturbed plate experiments at Los Alamos

We have carried out several experiments on the Los Alamos proton radiography (pRad) facility to explore the growth of perturbations subjected to shockless acceleration. These experiments have involved both Tantalum and depleted Uranium plates with various initial amplitudes. The experimental platform is based on the one first developed by Barnes et al. [1] and further advanced by Raevsky [2]. This paper presents both the data for these experiments and an initial attempt to model the experiments using the simulation code FLAG [3].

Natural convection cooling for LEU irradiated fuel plates

The important thermal hydraulic limiting phenomena were studied during the natural convection cooling of a proposed irradiated Low Enriched Uranium (LEU) plate’s target. The irradiated LEU targets are used for the production of the medical isotope Molybdenum-99 (Mo-99), whose decay product, Technetium-99 (Tc-99), is used in the majority of medical diagnostic imaging procedures. After finishing the stage of irradiation for the LEU targets, they reach a high level of radioactivity, so the direct handling and maneuvering for these targets at that time becomes quite difficult. The targets should stay for a period of time for decay and cooling either at the reactor core situation or at a special basket inside the reactor main pool. This parametric study was performed to verify the thermal–hydraulic limiting phenomena (Onset of Nucleate Boiling, Pulsed Boiling Power ration, Onset of Net Vapor Generation and Burn-out Power ratio) during the natural convection cooling against the different power ratio of the irradiated plate (plate decay power at the establishment of natural convection regime to the steady state plate operational full power during the irradiation). The study was performed at two different positions, namely inside the reactor core with a long chimney, and in a storage basket. Half load and full load were considered, and both of these possibilities were discussed and analyzed with considering different power ratios. All the limiting phenomena satisfy the safety margins except the ONBR. The ONBR requires certain conditions to satisfy the safety margin.

Strain and stress tensors of rolled uranium plate by Rietveld refinement of TOF neutron-diffraction data

We report the complete macroscopic average strain and stress tensors for a cold-rolled uranium plate, based on the neutron TOF measurements. Both tensors were determined by the least-squares refinement of the interplanar spacings of 19 Bragg reflections. Based on the pole figures, as determined by GSAS, a triclinic sample symmetry of the uranium plate was assumed. Strain and stress are tensile in both the transverse and rolling directions and very small in the normal direction (through the thickness of the plate). Shear strain and stress components are compressive and of significant magnitude.

Long-term studies of the optical components in the ZEUS calorimeter using a moving 60Co source

The response of the ZEUS sampling calorimeter has been studied over a period of 17 years using a moving 60Co source. The aim was to monitor the stability of the calorimeter with respect to radiation damage, ageing and mechanical damage. The measurements started in 1990 during the installation of the first calorimeter modules and ended in 2007 after the last HERA run. Within the experimental uncertainties, the longitudinal uniformity of the polymethylmethacrylate (PMMA) based wavelength shifters (Y7) did not change. There is also no evidence for radiation damage in the polystyrene (PS) based scintillators (SCSN-38) due to particle losses of the electron and proton beams. On the other hand, the investigations clearly indicate that the attenuation length of the SCSN-38 scintillators decreased continuously over the years. A comparison with data obtained from a test calorimeter showed that the same effect observed in the ZEUS calorimeter after 17 years would require a uniform exposure of all scintillators by a dose of ( 2.8 ± 1 ) kGy of γ rays. However, it can be excluded that the degradation observed in the ZEUS calorimeter can be attributed to the radioactivity of the uranium plates of the calorimeter. Thus we expect that the decrease of the attenuation length is due to ageing of the scintillators. The ageing effects observed in SCSN-38 are compared to data taken from the literature.

Critical Mass and Subcritical Experiments Interlaced with Nb-1Zr, Re, Mo, Ta-2.5W Fueled with Highly Enriched Uranium in Support of the Prometheus Project

The National Aeronautics and Space Administration is considering nuclear power sources for space exploration. A series of critical mass experiments was designed to address the development, performance, and design of a space nuclear reactor being considered to support the Prometheus project. These experiments consisted of interlacing the refractory metals rhenium (Re), molybdenum (Mo), tantalum-2.5 wt% tungsten (Ta-2.5W), and niobium-1 wt% zirconium (Nb-1Zr) with moderating materials (graphite or polyethylene) and were fueled by highly enriched uranium plates. These experiments are designed to assess the adequacy of and uncertainty in refractory metal neutron cross-section evaluations for use in Prometheus nuclear reactor design work. The critical experiments were designed in the energy spectrum closely resembling or bracketing that in the proposed space reactor. For each material (Re, Mo, Ta-2.5W and Nb-1Zr), four critical configurations were designed and performed to measure the sensitivity of keff to the material under four different and progressively softer neutron spectra (core center spectrum, harder than core average spectrum, softer than core average spectrum, and accident flooded spectrum). The thicknesses of the graphite or polyethylene moderator and reflector plates were adjusted to achieve the desired neutron spectrum. One critical and 18 subcritical experiments provided for measurements of material neutronic behavior in a simple cylindrical geometry configuration that was modeled in MCNP with ENDF/B-VI.6 cross-section data and compared to the extrapolated or predicted critical mass for all the experiments. These experiments were performed at the Los Alamos National Laboratory using the Planet vertical lift critical assembly at the Los Alamos Critical Experiment Facility.

Neutron diffraction texture study of deformed uranium plates

The texture of two depleted uranium (DU) samples, labelled DUWR and DUWR2, were studied by neutron diffraction. DUWR was prepared by warm rolling of a cast ingot, and DUWR2 was prepared by adding 20% tensile strain to the warm-rolled DUWR. Complete three-dimensional orientation distribution functions were determined using four neutron pole figures for the DUWR, and using six neutron pole figures for the DUWR2 sample, by the WIMV method of the program popLA. The textures of the two samples were essentially identical to each other. They could be described by a twisted helical density tube spiralling continuously along the ψ-axis of the Euler space. The projection of the backbone of the density tube along the θ-axis cast a linear shadow running parallel to the diagonal of the φ-ψ plane, which could be defined by a φ=ψ+90° (and φ=ψ+270°) relation. The helical tube was confined within narrow θ-angle limits, from 14° to 30° with the peak orientation at (103) 〈0 10〉. The diffraction patterns of the DUWR2 sample were measured from the normal direction to the rolling surface of the sample, up to the scattering angle of 108° using a 0.15 nm neutron beam. The Rietveld profile refinement using the textured diffraction pattern was quite satisfactory when the texture effect to the entire diffraction profile was corrected for by the corresponding pole density from the inverse pole figure.

Measurement of Fuel Importance Distribution in Non-uniformly Distributed Fuel Systems.

A reactivity effect due to a spatial variation of nuclear fuel concentration is an important problem for nuclear criticality safety in a reprocessing plant. As a theory estimating this reactivity effect, the Goertzel and fuel importance theor~es are well known. It has been shown that the Goertzel's theory is valid in the range of our experiments based on measurements of reactivity effect and thermal neutron flux in non-uniformly distributed fuel systems. On the other hand, there have been no reports concerning systematic experimental studies on the flatness of fuel importance which is a more general index than the Goertzel's theory. It is derived from the perturbation theory that the fuel importance is proportional to the reactivity change resulting from a change of small amount of fuel mass. Using a uniform and three kinds of nonuniform fuel systems consisting of 93.2% enriched uranium plates and polyethylene plates, the fuel importance distributions were measured. As a result, it was found experimentall...

共分散測定法に基づく実効遅発中性子割合の測定

The effective delayed neutron fraction βeff is an important reactor physics parameter, because it plays a role of a conversion factor between the calculated reactivity in Δk/k unit and the measured one in dollar unit. In 1970's, central-sample-worth measurements were done to accumulate the basic data for fast reactor design. Calculated values to these measurements were 30% overestimate. It was pointed out that the 10% of this discrepancy was due to the βeff. Recently, as an origin of systematic discrepancy between the calculated and measured reactivity is due to the βeff, needs for experimental method obtaining the βeff with 1-2% accuracy have been stressed in order to re-examine delayed neutron data and calculational method. In this report, we describe the βeff measurements by using the method proposed by E.F. Bennett at the Kyoto University Critical Assembly (KUCA) of the Kyoto University Research Reactor Institute. This method is based on the measurements of covariance between neutron counts and the absolute fission rate. Measurements were performed with the 3/8″ P36EU thermal neutron system which was composed of 93.3% enriched uranium plates of 1/16″ thickness and 3/8″ polyethylene moderator plates, and the 1/8″ P80EU system which was composed of 1/8″ polyethylene plates and had slightly hard neutron spectrum. As the results of several measurements, the βeff of the 3/8″ P36EU system ranged from 7.60 ±0.18 to 7.96±0.19, and from 8.07±0.18 to 8.14±0.18 for the 1/8″ P80EU system. The latter is slightly large because of hard neutron spectrum. The experimental accuracy is ±2.2%- ±2.9% which did not clear the target accuracy of 1-2%. But this may be improved by means of an improvement of fission rate measurement. Comparing with these measurements and the calculated values based on the JENDL-2 nuclear data file, they agreed within the range form -4% to 1.5%. Then we concluded that the covariance method was usable to evaluate the βeff value of thermal neutron system.

Construction and beam test of the ZEUS forward and rear calorimeter

The forward and rear calorimeters of the ZEUS experiment are made of 48 modules with maximum active dimensions of 4.6 m height, 0.2 m width, 7 λ depth and maximum weight of 12 t. It consists of 1 X 0 uranium plates interleaved with plastic scintillator tiles read out via wavelength shifters and photomultipliers. The mechanical construction, the achieved tolerances as well as the optical and electronics readout are described. Ten of these modules have been tested with electrons, hadrons and muons in the momentum range 15–100 GeV/ c. Results on resolution, uniformity and calibration are presented. Our main result is the achieved calibration accuracy of about 1% obtained by using the signal from the uranium radioactivity.

Measurements of longitudinal and transverse profiles for hadron showers in the range 10–100 GeV and comparisons with Monte Carlo simulations

Data on the longitudinal and transverse shower profiles initiated by 10, 20, 30, 50, 75, and 100 GeV positive hadron beams are presented. They have been measured with a 6λ deep sampling hadronic calorimeter using 3.2 mm thick depleted uranium plates as absorber and 3.0 mm scintillator layers as active material. The scintillator is read out with wavelength shifter plates coupled to photomultipliers. The data are compared to Monte Carlo calculations, simulating both the development of hadronic cascades as well as detector effects.