A simple and rapid separation method was developed for actinides in low-level radioactive waste. Extraction chromatography columns were used for the separation of U, Np, Pu, Am, and Cm in the simulated solidified product and simulated waste solutions. In the investigation of the separation procedure, we tried to construct a scheme using relatively noncorrosive reagents, aiming to apply it to the routine analysis of radioactive waste. The recoveries and decontamination factors of actinides in the solution of simulated waste were high enough to determine the actinides in the radioactive waste by alpha-spectrometry, mass spectroscopy. The time required for the separation was 2-3 hours. The chromatography method was applied to the analysis of actinide in an actual waste solution; high recoveries and decontamination factors were obtained, which indicated that the extraction chromatographic separation method could be adopted for the simple and rapid separation of actinide from waste.
This paper is concerned with the gas leakage fault detection of electromagnetic valve using a neural network filter. In modern plants, the ability to detect and identify gas leakage faults is becoming increasingly important. The main difficulty in detecting gas leakage faults by sound signals lies in the fact that the practical plants are usually very noisy. To solve this difficulty, a neural network filter is used to eliminate background noise and raise the signal noise ratio of the sound signal. The background noise is assumed as a dynamic system, and an accurate mathematical model of the dynamic system can be established using a neural network filter. The predicted error between predicted values and practical ones constitutes the output of the filter. If the predicted error is zero, then there is no leakage. If the predicted error is greater than a certain value, then there is a leakage fault. Through application to practical pneumatic systems, it is verified that the neural network filter was effective in gas leakage detection.
Scatter of data found in a sorption database may bring about possible risk dilution in the result drawn by the probabilistic safety analysis unless inappropriate data were carefully examined and excluded if necessary. To reduce such scatter, the advantage of data analysis by adopting theoretical sorption models was demonstrated. The sorption coefficients (Kd) of various combinations of bentonite, minerals and rocks with elements in the database were examined at first to extract the data sets that have enough information for the analysis using the mechanistic sorption models such as the ion exchange and the surface complexation. The Kd of Cs on bentonite was selected as an example and the factors dominating the data scatter were found to be the concentrations of Cs and coexisting ions. The procedure to estimate the parameters that were not given in the database, such as the selectivities of ion exchange reactions among Cs and coexisting ions, and the cation exchange capacity were also suggested to assist in the model analysis. Although the Kd of Cs varied in several orders of magnitude, the scatter could be reduced within the same order of magnitude after appropriate interpretation of the reason for the scatter and data selection.
Data scatter found in a database of Cs sorption coefficients (Kd) was investigated to demonstrate the usage of the mechanistic sorption model. A large difference found between the data given by the batch sorption and diffusion experiments was of great interest and the analyses of the pH dependence of the Kd by adopting either PHREEQC or the revised version of MINEQL had been carried out to determine the reason for the disagreement among the sorption data. Both of them were able to treat the ion exchange reaction and the surface complexation simultaneously. In particular, the latter was a largely modified form of the original version suitable for the compacted form. The same parameter set was used in the analyses even under extremely different experimental conditions. The results indicated that the Kd for both cases could be reproduced by the model calculation and the perturbation of environmental condition such as the Na concentration, and the pH has a strong influence on the data stability. The results also suggested that the scatter of the Kd of Cs could be understandable and minimized through the analysis by adopting the mechanistic sorption models.
The radon diffusion coefficient for soil, D, is a very important parameter used to estimate radon dose for uranium-bearing waste. Many Ds were measured in the uranium mill tailing remediation action project in the US, and a formula for the estimation of diffusion coefficient, Rogers's formula, was proposed. However, it is uncertain whether Rogers's formula is applicable to Japanese soils because most of them have come from volcanic ash and contain much water. This paper describes the development of a measurement apparatus for D using a lump response transient method and a step response transient method, and presents measured D values for Japanese surface soils. Measured alpha ray count curves are good in agreement with those of theory. This shows that radon transportation in soil can be described using Fick's law. Furthermore, the measured effective D values are good in agreement with those of Rogers's formula. This means that Rogers's formula can be applied to Japanese soils.
The Japan Atomic Energy Agency has been developing a nuclear hydrogen production system by using heat from the Very High Temperature Reactor (VHTR). This system will handle a large amount of combustible gas and toxic gas. The risk from fire, explosion and acute toxic exposure caused by an accident involving chemical material release in a hydrogen production system is assessed. It is important to ensure the safety of the nuclear plant, and the risks for public health should be sufficiently small. This report provides the basic policy for the safety evaluation in cases of accident involving fire, explosion and toxic material release in a hydrogen production system. Preliminary safety analysis of a commercial-sized VHTR hydrogen production system, GTHTR300C, is performed. This analysis provides us with useful information on the separation distance between a nuclear plant and a hydrogen production system and a prospect that an accident in a hydrogen production system does not significantly increase the risks of the public.
Nuclear hydrogen production is necessary in an anticipated hydrogen society that demands a massive quantity of hydrogen without economic disadvantage. Japan Atomic Energy Agency (JAEA) has launched the conceptual design study of a hydrogen production system with a near-term plan to connect it to Japan's first high-temperature gas-cooled reactor HTTR. The candidate hydrogen production system is based on the thermochemical water-splitting iodine sulphur (IS) process. The heat of 10 MWth at approximately 900°C, which can be provided by the secondary helium from the intermediate heat exchanger of the HTTR, is the energy input to the hydrogen production system. In this paper, we describe the recent progresses made in the conceptual design of advanced process heat exchangers of the HTTR-IS hydrogen production system. A new concept of sulphuric acid decomposer is proposed. This involves the integration of three separate functions of sulphuric acid decomposer, sulphur trioxide decomposer, and process heat exchanger. A new mixer-settler type of Bunsen reactor is also designed. This integrates three separate functions of Bunsen reactor, phase separator, and pump. The new concepts are expected to result in improved economics through construction and operation cost reductions because the number of process equipment and complicated connections between the equipment has been substantially reduced.
The second version of WSPEEDI (WSPEEDI-II) is developed. It has functions to predict the radiological impact of nuclear accident abroad on Japan by quick calculations of air concentration, surface deposition, and radiological doses. WSPEEDI-II has the following new functions for better predictions and practical use: (1) high-performance prediction of atmospheric dispersion of radionuclides from local to regional ranges with appropriate resolutions by introducing a nonhydrostatic atmospheric dynamic model, (2) source term estimation by coupling calculation results and monitoring data for the case that no source information is available from abroad, (3) on-line prediction data exchanges with major emergency response systems in the United States and Europe having similar functions to WSPEEDI-II, (4) web-based graphical user interface system for easy operations of WSPEEDI-II, and (5) preset East-Asian database for quick start against a nuclear accident in Eastern Asia. In this paper, we describe these new functions of WSPEEDI-II.
Fluorescent liquid penetrant test equipment, for testing welded parts of the evaporator for phosphate liquid waste generated by solvent waste treatment at Tokai Reprocessing Plant, was developed. It is composed of a CCD camera and UV light for observation, nozzles for fluorescent liquid, water for washing, and air for drying, and a positioning mechanism that can adjust the position of this equipment relative to the inspection object by three-axis operation, varying the insert length, bending angle and turn angle. Also, the equipment can be inserted by way of the nozzle for inspection, whose inside diameter is 60 mm, into the evaporator. It was confirmed that the standard faults, as defined in JIS, could be observed by the equipment. Then, a fluorescent liquid penetrant test of an evaporator that has treated phosphate liquid waste for 18 years with a heating temperature of about 105°C was performed. The result of the test indicates the absence of faults in the evaporator.
In this paper, we focus on the kind of information the nuclear industry should provide for the general public to understand and trust in the plan of plutonium utilization for mixed-oxide fuel in light water reactors, i.e., Pu thermal utilization. We conducted an interview survey for 30 people who live in the Tokyo metropolitan area to analyze what they know and how they feel about Pu thermal utilization, and to compare three information materials based on their subjective evaluation of the degree of understanding, trust and so on. The content analysis of interviewees' comments regarding Pu thermal utilization shows that they have vague but correct knowledge, that is, “Pu thermal utilization is to recycle nuclear fuel or nuclear waste.” However, people do not have background information concerning the necessity of Pu utilization, such as the resource limitation of uranium. According to the comparative analysis of the three materials, a material that presents the necessity and usefulness of Pu thermal utilization using figures and graphs was evaluated most understandable, informative, trustful, and persuasive. The material including information of risk was evaluated more informative, but the evaluation of trust indicated a divided opinion. People who feel anxiety about nuclear power generation evaluated the material including risk messages more trustful than other materials. Others evaluated it less trustful because the risk management of Pu thermal utilization and the process to solve remaining problems, such as HLW disposal, are uncertain.
In the vitrification of high-level radioactive liquid wastes, platinum group particles originally included in the liquid wastes are mixed into molted glass and convected with the glass flow in vitrification melters. Though behaviors of the platinum group particles in the melters should be investigated to establish efficient melter operation methods, difficulties in observations or measurements of the particles in the actual melters due to high radioactivity have prevented enough investigation. In this paper, we present a numerical analysis method to evaluate the behaviors of the particles in the melters. Since the thermal convection field of the molten glass, electrical potential and particle distribution are formed by the interaction of the particles in the melters, we coupled these physical behaviors in the numerical method to evaluate the melter status with high accuracy. A numerical simulation was conducted and transient behaviors of the particles in the melters were determined. The results of simulation for ten batches of operation (each batch consists of melting, bottom heating and discharging operations) showed that an effective particle discharge can be achieved, as observed in mock-up experiments. These results imply that our numerical method can evaluate transient behaviors of the platinum group particles in vitrification melters.
Evaluation of the external radiation dose is essential for the judgment of evacuation directive in a nuclear accident. We have newly developed an external environmental dose evaluation system to estimate the radiation dose from a source term in a nuclear facility as a function of the SPEEDI network system. With this system, which is composed of local clients and central server, a realistic external dose profile can be quickly evaluated by correcting the dose calculated already with measured values during the monitoring. By the end of the 2006 fiscal year, we developed and validated the system that could be applied to almost all of the LWRs and principal nuclear fuel cycle facilities in Japan.