Tomofumi Sakuragi

Development of New Waste Forms to Immobilize Iodine-129 Released from a Spent Fuel Reprocessing Plant

I-129 is a very long-lived radionuclide that is released to an off-gas stream when spent fuels are dissolved at a reprocessing plant. An iodine filter can capture I-129 in the form of AgI. However, because AgI is unstable under the reducing conditions of a geological repository and I-129 has a very long half-life, I-129 can migrate to the biosphere. These characteristics make I-129 a key radionuclide for the safety assessment of a geological disposal of radioactive wastes generated from a reprocessing plant (TRU wastes). To improve disposal safety, several new waste forms have been developed to confine I-129 for a very long period in order to reduce the leaching of I-129 from radioactive wastes. These new waste forms have technical objectives of solidifying more than 95% of I-129 into the waste form and achieving a leaching rate of less than 10-5/y. Several iodine immobilization techniques have been examined. This paper presents experimental results concerning the treatment process, leaching behavior, modeling, and related elements of these immobilization techniques.

Further Development of Iodine Immobilization Technique by Low Temperature Vitrification With BiPbO2I

The authors describe progress in the development of low temperature vitrification with BiPbO2 I (BPI) as a promising immobilization technique by which Iodine-129 is recovered by BiPbO2 NO3 to form BPI, and then solidified into a lead-boron-zinc glass matrix (PbO-B2 O3 -ZnO) using a low temperature vitrification process. The microscopic structure of BPI glass was analyzed by various analytical techniques, such as XRD (X-ray diffraction), NMR (nuclear magnetic resonance analysis), and XPS (X-ray photoelectron spectroscopy), using several types of glass samples. The results obtained provide structural information on key elements in BPI glass and can be applied for modeling the structure of the BPI glass, simulated by molecular dynamics. The previous work suggested that the leaching behavior of iodine from BPI glass depended upon the chemical conditions of the solution. Further leaching tests using solutions under varying conditions were carried out in order to predict mechanisms of iodine leaching. Normalized elemental mass loss values of iodine in simulated seawater and bentonite pore water are almost the same as those of boron, showing that iodine dissolves congruently with BPI glass, whereas iodine dissolves incongruently in Ca(OH)2 solutions of pH 9 and 11. To demonstrate the feasibility of the BPI vitrification process, recovery tests of iodine from spent iodine filters were conducted and a prototype melting furnace was developed for scale-up tests of glass sample. It was found that more than 95% of iodine can be recovered from the spent iodine filter and that the prototype furnace can produce approximately 0.5 liters of homogeneous glass.Copyright

Estimation of Carbon 14 Inventory in Hull and End-Piece Wastes From Japanese Commercial Reprocessing Operation

Hull and end-piece wastes generated from reprocessing plant operations are expected to be disposed of in a deep underground repository as Group 2 TRU wastes under the Japanese classification system. The activated metals that compose the spent fuel assemblies such as Zircaloy claddings and stainless steel nozzles are mixed and compressed after fuel dissolution, and then stuffed into stainless steel canisters.Carbon 14 is a typical activated product in the hulls and end-pieces and is mainly generated by the 14N(n,p)14C reaction. In the previous safety assessment of the TRU waste in Japan, the radionuclides inventory was calculated by ORIGEN-2 code. Some conservative assumptions and preliminary estimates were used in this calculation. For example, total radionuclides generated from a single type of fuel assembly (45 GWd/tU for a PWR unit), and the thickness of the Zircaloy oxide film on the hulls (80 μm) were both overestimated. The second assumption in particular has a large effect on exposure dose evaluation. Therefore, it is essential to have a realistic source term evaluation regarding such items as the C-14 inventory and its distribution to waste parts.In the present study, a C-14 inventory of the hull and end-piece wastes from the operation of a commercial reprocessing plant in Japan corresponding to 32,000 tU (16,000 tU in each BWR and PWR) was calculated. Analysis using individual irradiation conditions and fuel characteristics was conducted on 6 types of fuel assemblies for BWRs and 12 types for PWRs (4 pile types × 3 burnup limits). The oxide film thickness data for each fuel type cladding were obtained from the published literature. Activation calculations were performed by using ORIGEN-2 code. For the amount of spent assembly and other waste characteristics, representative values were assumed based on the published literature. As a preliminary experiment, C-14 in irradiated BWR claddings was measured and found to be consistent with the calculated activation.The total C-14 inventory was estimated as 4.46×1014 Bq, consisting of 2.58×1014 Bq for BWRs and 1.87×1014 Bq for PWRs, and is consistent with the safety assessment of 4.4×1014 Bq. However, the distribution of the C-14 inventory to hull oxide, which was estimated under the assumption of instantaneous radionuclide release in the safety assessment, decreased from 5.72×1013 Bq (13% of the total) in the previous assessment to 1.30×1013 Bq (2.9% of the total; consisting of 1.48×1012 for BWRs and 1.15×1013 for PWRs). In other words, the exposure dose peak is reduced to approximate 25% of its previous value due to the use of detailed oxide film data that the BWR cladding has a thin oxide film. Other instantaneous release components for C-14 such as the fuel residual were negligible.Copyright

Radiolytic decomposition of organic C-14 released from TRU waste

It has been found that metallic TRU waste releases considerable portions of C-14 in the form of organic molecules such as lower molecular weight organic acids, alcohols and aldehydes. Due to the low sorption ability of organic C-14, it is important to clarify the long-term behavior of organic forms under waste disposal conditions. From investigations on radiolytic decomposition of organic carbon molecules into inorganic carbonic acid, it is expected that radiation from TRU waste will decompose organic C-14 into inorganic carbonic acid that has higher adsorption ability into the engineering barriers. Hence we have studied the decomposition behavior of organic C-14 by gamma irradiation experiments under simulated disposal conditions. The results showed that organic C-14 reacted with OH radicals formed by radiolysis of water, to produce inorganic carbonic acid. We introduced the concept of “decomposition efficiency” which expresses the percentage of OH radicals consumed for the decomposition reaction of organic molecules in order to analyze the experimental results. We estimated the effect of radiolytic decomposition on the concentration of organic C-14 in the simulated conditions of the TRU disposal system using the decomposition efficiency, and found that the concentration of organic C-14 in the waste package will be lowered when the decomposition of organic C-14 by radiolysis was taken into account, in comparison with the concentration of organic C-14 without radiolysis. Our prediction suggested that some amount of organic C-14 can be expected to be transformed into the inorganic form in the waste package in an actual system.Copyright