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The parameters of the coolant moving in the channel between the fuel elements in the WWER-1200 core with a design steam content of 0.1 at the outlet are determined. The corrosion of the outer surface of these elements is modeled taking into account the heat and mass exchange between the near-wall layer of the coolant flow (with a thickness of the order of the thickness of the evaporating microlayer under the steam bubble) and its core. It is found that the concentration of ammonia in the near-wall layer of the flow is 0.35 of its concentration in the flow core, and this concentration level is typical of the upper half of the core. A formula is proposed for calculating the thickness of the oxidized layer of the outer surface of the fuel elements taking into account the concentration of oxidizers in the near-wall layer of the coolant flow. An increase in the thickness of the oxidized layer due to the effect of boiling is estimated at 30% compared to the thickness observed upon completion of operation during four annual cycles without boiling. [ABSTRACT FROM AUTHOR]

The results of experiments are presented on testing the separation of uranium from U-Be nuclear fuel solutions by extraction. It was shown that the degree of uranium extraction using a solution of 10 vol% TBP in RED-3 from a nitric acid solution of this fuel is 98.7%. Transfer of uranium into the re-extract is 97.9%. More than 99.99 % of beryllium goes into the extraction raffinate intended for further vitrification. The results of the study were used for the preparation of technological regulations on reprocessing of U-Be SNF at the RT-1 plant of the Mayak Production Association. The industrial-scale reprocessing of this fuel confirmed the values determined in the study. [ABSTRACT FROM AUTHOR]

The results of experiments on development of dissolution modes of the uranium–beryllium nuclear fuel produced on an enlarged batch of unirradiated fuel elements, are reported. The feasibility of a two-stage fuel dissolution mode was confirmed, which makes it possible to reduce the total time of the process and reduce the rate of gas evolution. The composition of the formed gas phase was determined. The weight of undissolved solid residue was estimated. The results of the study were used in the preparation of technological regulations for the reprocessing of uranium–beryllium SNF at the RT-1 plant of the Mayak Production Association. [ABSTRACT FROM AUTHOR]

This paper presents the results of work on the application of plasma-chemical technology for producing uranium and plutonium oxides and their homogeneous mixtures from nitrate solutions of a given composition for the purpose of manufacturing fuel element cores. The technology is based on the process of high-temperature decomposition of aqueous solutions of salts by spraying them into a flow of the heat-transfer gas (air) heated to 5000–6000 K in a plasma torch. It is a highly productive short-stage process. No reagents are required for salt precipitation; there are no filtration and high-temperature treatment precipitation operations; the resulting powders are chemically active. It was found that the transformation time of salt solution drops to oxides is 0.05–0.10 s for the length of 0.5–0.7 m. The temperature in the reactor varies in the range of 1200–1800 K; no sintering and fusion of particles occurs. The apparatus-technological scheme of the plant has been developed. The technological equipment is made in a nuclear-safe design and installed in the exhaust boxes. The scheme includes a high-frequency induction plasmatron (plasma torch) with coaxial gas inlet into the discharge chamber. The high-frequency induction discharge is excited by inserting a grounded ignition electrode into the plasmatron discharge chamber. The plasmatron design provides the possibility of long continuous operation at high plasma temperatures with the creation of intense plasma-jet processes for processing radioactive solutions. The reactor provides ways to prevent product accumulation on the walls (gas curtain method, electric pulse cleaning of cylindrical surfaces, vortex tangential gas supply at the level of the outlet holes of the injectors). Equipment for the separation of nanodispersed oxide powders from high-temperature steam- and powder-gas flows (vortex precipitators, filters with a metal-woven partition) has been developed and applied. The degree of oxide recovery in vortex precipitators is 88–92% per stage; in filters with a metal-cloth partition, the degree of purification reaches 105–108. Plasma-chemical oxides are nanocrystalline and nanoscale materials, represented by crystalline and amorphous structures. Mixed powders are a homogeneous mixture of oxides in the form of solid solutions. The particle size does not exceed 3 μm; the crystallite size is about 10 nm. Fuel elements were manufactured from mixed powders at FSUE Mayak, which passed successful tests in the Bor-60 reactor. Seven industrial fuel elements were manufactured for the BN-600 reactor, six of which were loaded into the reactor. [ABSTRACT FROM AUTHOR]

Reliable transportation of fuel elements is vital for High Temperature Reactor (HTR) power plant. Therefore, the safety performance of the containment boundary in HTR-PM600 fresh fuel transport package under impact loading should be assessed. Specifically, factors such as the structural integrity, bolt safety, and sealing performance of the flange are considered. However, it is not convenient to assess the sealing performance using the helium leakage test without a prototypical experiment. Therefore, to accurately assess the safety performance in the design stage, we propose a detection technique combing with the finite element method to measure the leak channel in the flange for HTR-PM600 transport package. The results show that the boundary containment has no failure, and there is no complete leak channel in the flange of HTR-PM600 package. Hence, this package has a great safety performance under impact loading. This work also provides references for design of fresh fuel transport packages in pebble-bed reactors.

One of the factors capable of resulting in deconfinement of fuel element cladding upon long term storage of spent nuclear fuel (SNF) is reorientation of hydrides. This work estimates the extent of reorientation of hydrates in unirradiated gas filled specimens of fuel element claddings of VVER-1000 reactor (Zr‒1% Nb) under the conditions simulating conditions of dry storage. The specimens with the hydrogen content of 100 and 200 ppm were exposed to thermomechanical tests according to various scenarios at hoop stresses in cladding up to 100 MPa. It has been established that the extent of reorientation of hydrides and related decrease in plasticity of cladding material are insignificant. The most significant factor influencing the extent of reorientation is the level of hoop stresses. [ABSTRACT FROM AUTHOR]

• Burn-up is an essential parameter for nuclear fuel elements to ensure reactor safety. • Non-destructive analysis is an essential technique for precisely determining burn-up. • Istanbul Technical University Triga Mark II research reactor fuel elements were investigated. • The gamma spectroscopy of certain fuels in the ITU research reactor core was conducted. In this study, burn-up values ​​of spent fuels were determined using Non-Destructive Analysis (NDA) with the help of 137Cs burn-up indicator. Four spent fuel elements (F30, D18, B5 and E14) that burned in the core of the Istanbul Technical University Triga Mark II research reactor over the years were investigated with two different methods. In addition, the individual activity of the fuel elements along the axial fuel length was determined by the High Purity Germanium (HPGe) detector. As expected, it was determined that the fuel element B5, located in the center of the reactor core, exhibited the highest burn-up value with both methods. The calculations conducted using method 1 revealed a burn-up rate of 5.40% for the B5 fuel element, while method 2 yielded a burn-up value of 205.59 MWd/tHM. Experimental results were compared with previous simulation burn-up studies. [ABSTRACT FROM AUTHOR]

Background The solid fuel thorium element molten salt reactors (MSR) have attracted more attention recent years. A3-3 graphite is chosen as the fuel matrix for MSR, thus its irradiation behavior and mechanical property is very important before the application. Purpose The study aims to observe the irradiation defects and hardness of A3-3 matrix graphite after ion irradiation by slow positron beam and nano-indentation, respectively. Methods The matrix graphite of fuel elements was irradiated with 1 MeV Xe ions to fluence of 5.8×1014 ions·cm-2 and 2.9×1015 ions·cm-2 respectively at room temperature. The slow positron beam and nano-indentation were employed to investigate the effect of Xe ions irradiation on vacancy defects and hardness of matrix graphite. The changes in irradiation induced defects distribution with depth and fluence were analyzed according to the obtained positron annihilation S parameters versus positron incidence energy or depth curves, compared to SRIM (Stopping and Range of Ions in Matter) calculation. Results Results from slow positron beam measurement show that 1 MeV Xe ions irradiation in matrix graphite introduces a damage layer with depth of about 600 nm, and the damage peak locates at about 250~350 nm in depth, consisted with SRIM simulation. The S parameters in irradiation samples increase significantly compared to virgin sample, which suggests that a high concentration of vacancy-type defects appeared within irradiation damage layer. In addition, the S parameters increase with the irradiation fluence, which shows that the concentration or size of vacancy-type defects increases. The nano-indentation results show that the hardness of irradiated graphite matrix is enhanced. Conclusions The enhanced hardness of A3-3 matrix graphite after ion irradiation is ascribed to the pinning of basal plane dislocation by the high concentration of vacancy type defects introduced by irradiation, consisted with the slow positron beam analysis. Slow positron beam is a very sensitive tool to study the irradiation defects.