Investigation of Al-B4C Metal Matrix Composites Produced by Friction Stir Additive Processing

Aluminium—boron carbide metal matrix composites (Al-B4C MMCs) belong to the class of materials extensively used in the nuclear industry as a thermal neutron absorber in spent fuel casks. This article investigates a novel production method of Al-B4C MMCs—Friction Stir Additive Processing (FSAP)—as an alternative production method to casting or sintering. FSAP is derived from friction stir welding, which can be used to local modifications of microstructure, or it can be used to incorporate the second phase into the processed material. During this study, a variant of FSAP for MMC production was proposed, and its mechanical and thermal neutron absorbing properties have been investigated. Further, the influence of neutron irradiation on mechanical properties has been studied. Results show that FSAP can successfully produce Al-B4C MMCs with 7 mm thickness. Neutron irradiation causes only a slight increase in hardness, while its effect on tensile properties remains inconclusive.

Features of a BWR Neutron Absorber Melt Relocation in an Oxidative Environment During the CLADS-MADE-02 Test

The work on the understanding of the accident progression at the units of the Fukushima Dai-Ichi Nuclear Power Plant (1F) is ongoing. This contribution gives a part of detailed investigations of the control blade melt propagation downwards through the prototypic BWR bundle assembly during the CLADS-MADE-02 test, where the conditions of the 1F Unit 3 was simulated. Interesting features emerged in an oxidative environment.

Improvement of Conversion Ratio of Thorium Fuel in LWR by Adding Neutron Absorber

The reproduction factor of Th232 is high in the thermal energy range and there is a possibility to achieve the breeding in LWRs. However, it is necessary to improve the conversion ratio since the breeding is difficult in LWRs. The conversion ratio can be improved by suppressing capture rate of Pa233 and by promoting capture rate of Th232. In addition, these capture rates can be modified by adding neutron absorber. Therefore, the neutron absorber is focused for improving the conversion ratio in this study. The high resonance peaks of Pa233 capture cross section exist around 1∼100 eV. The resonance peaks of Th232 are higher than 10 eV. Thus, when the 1∼10 eV neutrons are suppressed in the fuel, the Pa233 resonance capture reaction is suppressed and the Th232 resonance capture reaction is promoted by neutron spectrum hardening. Therefore, six neutron absorbers that have high capture cross section peaks at 1∼10 eV were selected. The PWR pin cell calculations were carried out by Monte Carlo code MVP. The fuels are composed of a base material and an absorber. The base material is an oxidized fuel composed of U233(10 wt%), Th232(89.95 wt%), and Pa233(0.05 wt%). The amount of neutron absorber was adjusted so that the infinite multiplication factor becomes 1.33. The impact of adding neutron absorber on the reaction rate was evaluated. As the result, the hardening of the neutron spectrum leads increase of the capture rate of Pa233, and the capture rate of Th232 in the epithermal energy range is increased. The change of capture rate of Th232 is greater than that of Pa232. Therefore, the conversion ratio is found to be improved by adding neutron absorber.

ANALYSES OF NEUTRON ABSORBER MATERIALS ON THE SAFETY PARAMETERS IN THE RSG-GAS REACTOR

Shutdown system in RSG-GAS reactor is using neutron absorber. There are 3 kinds of absorber material in research reactors including Ag-In-Cd alloy, B4C, and Hf. In this works, analyses of different neutron absorbers on the main safety core parameters in the RSG-GAS research reactor are selected for analyses. Their integral effects on the main neutronic core parameters important to safety issues are investigated. These parameters are core excess reactivity, shutdown margin, total reactivity worth of control rods, PPF and neutron flux . The RSG-GAS core silicide fuel is selected as the case study to verify calculations. A three-dimensional, four-group diffusion model is selected for core calculations. The well-known WIMSD-5B and Batan-3DIFF reactor codes are used to carry out these calculations. It is found that the largest shutdown margin is gained using the B4C; also the lowest PPF is gained using the Hf material. The maximum point power densities belong to the inside fuel regions surrounding the CIP (centre irradiation position), surrounded by control fuel elements, and the peripheral fuel elements beside the berrylium reflector. The greatest and least fluctuation of the point power densities are gained by using B4C and Ag-In-Cd alloy, respectively.

Leaching behavior of cementitious material immobilizing Cs-containing B-10 enriched boric acid waste

<p>Recently, boric acid enriched in B-10 has received attention over natural boric acid in nuclear industry, because the elevated content of B-10 is a prospective neutron absorber. Advantages connected to the use of B-10 enriched boric acid are the increased controllability of reactor core which results in use of reduced amount of boric acid and, subsequently, the reduction in the amount of the radioactive boric acid waste produced during reactor operation. In the other hand, consequent radioactive boric acid waste requires an adequate stabilization technology as it contains fission products of health concerns, importantly C<sub>s</sub>-137. Cementation is one of the proven, commercially viable, durable, widely used, simple and flexible technology for immobilization of low-level radioactive wastes (Hyatt and Ojovan, 2019). General integrity and durability of the cementitious waste form containing boric acid is B-leachability dependent (Rostamiparsa et al, 2020). The B-10 enriched boric acid leaching is expected to control also the C<sub>s</sub>-leaching. However, no study is found in which this is proven and the different geochemical behavior and phase distribution of the B and C<sub>s</sub> might cause deviations. This calls for the investigation of the connection between B- and C<sub>s</sub>-leaching behaviors in cementitious materials, in this case, especially focusing on B-10 enriched boric acid waste form. In this ongoing experimental work the B- and C<sub>s</sub>-leaching behavior of cementitious materials are studied, which are made of Portland cement, boric acid enriched in B-10 isotope and C<sub>s</sub>Cl. Boron- and Cs-leachability from the cementitious matrix are investigated in parallel by a standardized reference leaching test (ASTM, 2017). The tests are carried out by immersing the 28 days cured cement paste samples in deionized water in a glass bottle. Leachant renewal and solution sampling are done on a daily basis for the whole leaching test period of 11 days. Analysis of leached fractions are quantitatively measured by ICP-OES. Characterization of solid samples are conducted by XRD, SEM-EDX and Raman micro-spectroscopy methods. This is the first study to shed light on the connection between B-leaching and C<sub>s</sub>-leaching in cementitious materials containing B-10 enriched boric acid.</p> <p>Acknowledgements</p> <p>Our special thanks goes to Környezettudományi Centrum, Eötvös Loránd University and the Prémium_2017-13 research grant.</p> <p>References</p> <p>ASTM (2017). Standard Test Method for Accelerated Leach Test for Diffusive Releases From        Solidified Waste and a Computer Program to Model Diffusive, Fractional Leaching from Cylindrical Waste Forms. ASTM Standard C1308-08(2017), West Conshohocken, PA.</p> <p>Hyatt, N.C & Ojovan, M.I. (2019). Special Issue: Materials for Nuclear Waste Immobilization. Materials, 12(21), 3611.</p> <p>Rostamiparsa, M., Szabó-Krausz, Z., Fábián, M., Falus, G., Szabó, C., & Völgyesi, P. (2020). Experimental assessment of interaction between boric acid enriched in boron-10 and cementitious matrix. In EGU General Assembly Conference Abstracts (p. 19441).</p>

CRITICALITY SAFETY ANALYSIS OF GBC-32 SPENT FUEL CASK WITH IMPROVED NEUTRON ABSORBER CONCEPT

Higher enrichment of nuclear fuel along the manufacturing limit of boron content in steel and aluminum alloys represents a significant challenge in designing spent fuel transport and storage facilities. One possible solution for spent fuel pools and casks is the burnup credit method that allows for decreasing very high safety margins associated with fresh fuel assumption in spent fuel facilities. An alternative solution based on placing neutron absorber material directly into the fuel assembly is proposed here. A neutron absorber permanently fixed in guide tubes decreases system reactivity more efficiently than absorber sheets between the assemblies. The efficiency of the newly proposed concept is demonstrated on the criticality safety analysis of the GBC-32 spent fuel cask. Absorber rods from 8 different elements are placed within Westinghouse OFA 17x17 guide tubes. Currently used boron is a good option because of high absorption cross section, low atomic mass and chemical compatibility with various alloys. Alternative options (e.g., Sm, Eu, Gd, Dy, Hf, Re, Ir) are based on very good absorbers that do not require alloy compatibility since the absorbers can be placed inside zirconium or steel cladding. Because of high efficiency of the newly proposed absorber concept, boron content in BORAL sheets can be decreased to more competitive economics. Moreover, fuel assembly pitch is investigated in order to change cask wall inner diameter that will result in lower material consumption for the cask wall with the same shielding thickness.

Some Considerations on the Energy Deposition During a RIA Transient Based On Monte Carlo Simulations

Specific research reactors are capable of reproducing reactivity injection accidents in order to study the behavior of the nuclear fuel pins in accidental situations. In the CABRI research reactor, the fuel pin to be examined (test pin) is placed in the center of the core in a dedicated test loop. It is then subjected to a power transient, obtained by the fast depressurization of the 3He neutron absorber gas from the transient rods located in the core. One of the central parameters of the experiment is the energy deposition in the test pin, which is currently not measured during a transient. Instead, it is assumed that the relative energy distribution between the core and the test pin is constant regardless the operational state of the reactor. Currently, this correlation is measured in steady state. As such, the impact of the variations in the neutron flux, fuel and moderator temperatures during the transient is assumed equivalent on the energy deposition in the core and in the test pin. The goal of this work is to improve our knowledge on the mechanisms involved in the transient energy deposition. The aim of this paper is to present a methodological approach for the energy deposition estimation during a CABRI transient, based on static Monte Carlo calculations. The results suggest that the transient energy deposition rate is mainly dependent on the helium pressure and the Doppler feedback, and the relative energy distribution between the core and test pin changes during the transient.

Control Rod Blades Size Reduction Using Underwater Plasma Cutting and its Effects on Boron Carbide Powder Scattering

Control rod blades are comprised of a stainless steel sheath, which contains neutron absorber tubes (filled with boron carbide powder). During decommissioning, the first stage of size reduction consists of cutting the connector (bottom portion) of the control rod, while the second stage consists of separating the blades of the control rod by cutting through the tie rod. The last stage consists of segmenting the control rod blades by cutting through absorber tubes. In this study, the control rod blades segmentation (last stage of size reduction) is investigated using an actual control rod (unused). During the experiments, we used a forming press on the cut locations followed by a plasma arc cutting underwater. The purpose of this cutting technique is to minimize the scattering of boron carbides into water by using the stainless sheath melt to seal the absorber tubes. After the segmentation, we confirmed the sealing of the absorber tubes by visually examining the cut cross-sections. The water analysis showed that the boron carbide scattering was relatively low (only 0.07% of the total boron carbides was scattered). Finally, we confirmed that the off-gas emission is considerably reduced by using Argon plasma instead of Argon-Hydrogen plasma.