Experimental study on elemental behaviors in fluorination of nuclear spent fuel with flame reactor

Our proposed spent nuclear fuel reprocessing technology named FLUOREX is a hybrid system based on reprocessing technologies of fluorination and solvent extraction for light water reactor fuel. In the current research, we experimentally clarified solid–gas transfer behaviours of the fluorides in the FLUOREX process and identified the volatile and non-volatile compounds in the fluorination. We carried out a fluorination experiment for simulated spent nuclear fuel and solid separation from the UF₆ gas stream. The distribution ratios of fission product elements in the experimental apparatus were evaluated. Molybdenum, Te, Nb, and Ru were volatilized by fluorination and they accompanied the UF₆ gas. However, 22.9% of the Ru and 3.4% of the Nb were retained as solids in the experimental apparatus, contrary to the fact that their partial pressures in the experiment were lower than their vapor pressures. Rubidium, Sr, Zr, Ce, and Nd were completely recovered as solid fluorides, and these results agreed with the prediction based on boiling points of their fluorides. Antimony was completely recovered as a solid; nevertheless, the boiling point of antimony pentafluoride was lower than the process temperature, and that was attributed to the formation of a non-volatile antimony oxyfluoride.     

Adsorption of Molybdenum Hexafluoride on Magnesium Difluoride for Uranium Purification in FLUOREX Reprocessing

The nuclear fuel reprocessing method FLUOREX is a hybrid system based on fluoride volatility and using solvent extraction. Spent nuclear fuel is fluorinated, and most of the uranium is recovered as UF₆ gas. UF₆ contains some volatile fission product (FP) fluorides, so we considered their elimination from UF₆ by adsorbing them on fluoride adsorbents. We experimentally examined the adsorption of MoF₆ on MgF₂ adsorbent; MoF₆ is present as a volatile FP fluoride in UF₆ produced by the fluorination of spent nuclear fuel. The adsorption isotherm of MoF₆ adsorption on MgF₂ was obtained at MoF₆ pressures from 10^?4 to 50 kPa. The saturated adsorption amount was 1:3 ± 0:4 mg/m² at MoF₆ pressures from 10^?4 to 1 kPa. At MoF₆ pressure of about 10^?3 kPa, the saturated adsorption amount had no dependence on adsorption temperatures from 398 to 463 K. We deduced that MoF₆ was adsorbed as a monomolecule layer on the MgF₂ surface at MoF₆ pressures from 10^?4 to 1 kPa, and the MoF₆ partial pressure in UF₆ could be decreased below 1 × 10^?4 kPa, which is the specific MoF₆ concentration for the reenrichment process.     

Deposition of UF5 on the surface of steel material having oxide coating in contact with gaseous UF6

When UF6 is handled in leak tight system, the amount of uranium compound formed on the surface of containers, valves, and others in UF6 handling facilities is not so significant for a short term that special attention has not been given to this problem. The present work was done to throw some more light on this problem based on the recent experiment. We discuss the possibility that the intermolecular transfer of a fluorine atom from UF6 to UF5 may participate in the formation of uranium compound. The discussion includes also the unique features contained in the experimental result, the reaction processes assumed in this problem, and the derivation of a rate equation for expressing the deposition of uranium compound. Furthermore, we propose a new method for determining nonlinear parameters included in a governing differential equation having two variables for expressing the deposition of uranium compound from experimental raw data.     

Development of unfolding method to obtain pin-wise source strength distribution from PWR spent fuel assembly measurement

An unfolding method has been developed to obtain a pin-wise source strength distribution of a 14 × 14 pressurized water reactor (PWR) spent fuel assembly. Sixteen measured gamma dose rates at 16 control rod guide tubes of an assembly are unfolded to 179 pin-wise source strengths of the assembly. The method calculates and optimizes five coefficients of the quadratic fitting function for X–Y source strength distribution, iteratively. The pin-wise source strengths are obtained at the sixth iteration, with a maximum difference between two sequential iterations of about 0.2%. The relative distribution of pin-wise source strength from the unfolding is checked using a comparison with the design code (Westinghouse APA code). The result shows that the relative distribution from the unfolding and design code is consistent within a 5% difference. The absolute value of the pin-wise source strength is also checked by reproducing the dose rates at the measurement points. The result shows that the pin-wise source strengths from the unfolding reproduce the dose rates within a 2% difference.     

Application of the hierarchical domain decomposition boundary element method to the simplified P3 equation

In this paper, the hierarchical domain decomposition boundary element method (HDD-BEM), which has been developed to solve the diffusion equation, is applied to the simplified P₃ (SP₃) equation. The HDD-BEM solution for the SP₃ equation is provided in the present paper. A computer program, ABEMIE, based on the HDD-BEM is developed, and a two-dimensional one-group anisotropic-scattering benchmark problem is solved with it to verify the present HDD-BEM for the SP₃ equation.     

Thermal conductivity degradation analyses of LWR MOX fuel by the quasi-two phase material model

The temperature measurements of mixed oxide (MOX) and UO₂ fuels during irradiation suggested that the thermal conductivity degradation rate of the MOX fuel with burnup should be slower than that of the UO₂ fuel. In order to explain the difference of the degradation rates, the quasi-two phase material model is proposed to assess the thermal conductivity degradation of the MIMAS MOX fuel, which takes into account the Pu agglomerate distributions in the MOX fuel matrix as fabricated. As a result, the quasi-two phase model calculation shows the gradual increase of the difference with burnup and may expect more than 10% higher thermal conductivity values around 75 GWd/t. While these results are not fully suitable for thermal conductivity degradation models implemented by some industrial fuel manufacturers, they are consistent with the results from the irradiation tests and indicate that the inhomogeneity of Pu content in the MOX fuel can be one of the major reasons for the moderation of the thermal conductivity degradation of the MOX fuel.     

Application of Nuclear Recoil to Radioisotope Enrichment of Copper-64

A new target material has been proposed for nuclear recoil enrichment of ⁶⁴Cu. This radionuclide was hitherto produced by tedious chemical processing using concentrated sulfuric acid solution of neutron irradiated copper phthalocyanine. Simplified chemical processes using a water-soluble copper phthalocyanine derivative, octacarboxyphthalocyanato copper (II), have been adopted in this work. Two methods of separation involving solvent ex-traction and precipitation have been tested. Higher specific activity of ⁶⁴Cuhas been obtained for the extraction method than for the precipitation method. Feasibility of the extraction method in ⁶⁴Cu enrichment is discussed.     

Capabilities of TRANSURANUS code in simulating power ramp tests from the IFPE database

Investigations of fuel behavior are carried out in close connection with experimental research, operation feedback and computational analyses. OECD NEA sets up the “International Fuel Performance Experiments (IFPE) database”, a public domain database on nuclear fuel performance experiments with the purpose of model development and code validation. The objective of the activity (performed in the framework of the IAEA CRP FUMEX-III project) is to investigate the pellet-clad interaction mechanism and the capability of TRANSURANUS code in simulating the phenomena, processes occurring in the fuel rod during the power ramps, with focus on the parameters influencing the cladding failures. The experimental database adopted is the Studsvik PWR Super-Ramp subprogram, part of the IFPE database, which consists of 28 pressurized water reactor fuel rods power ramped at burnup from 28 to 45 MWd/kgU. Relevant results by TRANSURANUS are presented in connection with the experimental evidences. Focus is given on the PCI/SCC failure, demonstrating that the failure threshold, available in TRANSURANUS, results conservative both in case of KWU and W rods.