A social conflict over nuclear technology arises from the different interactions between society and nuclear technology. The purpose of this review is to grasp the essential points of this social conflict from a social viewpoint. These essential points can be discerned by interpreting results of polls about nuclear technology and the future of society in general. As a result, attitudes towards nuclear technology can be explained in terms of differences of general views on society such as social order or social progress. The attitudes of people toward nuclear technology were divided into strong agreement, weak agreement, weak objection and strong objection in order to obtain useful information for clarification of social conflict on this issue. Results of polls of people who have weak agreement for nuclear technology reveal their ambivalence about nuclear technology. This raises concern that further implementation of nuclear technology might cause these people to shift their views to objection.
Interest in the future hydrogen economy has prompted the research and development of the Very High-Temperature Gas-Cooled Reactor (VHTR). To achieve the targeted outlet gas temperature exceeding 950°C, material problems have yet to be solved. The development of advanced coated particle fuel is also due in view of the vulnerability of the SiC layer of conventional TRISO-coated particle fuel at temperatures exceeding 1,600°C. The coated particle fuel employing ZrC instead of SiC has been developed in JAEA. Although the past irradiation tests on the ZrC-coated particle fuel were exclusively on samples from the laboratory scale production, the promising results have been obtained. The properties, fabrication and inspection techniques as well as the results of irradiation and post-irradiation tests are reviewed. The post-irradiation heating tests at accident temperatures above 1,600°C revealed the durability of the ZrC-layer, which maintained the tightness to noble-gas and volatile metal fission products. From 2004, JAEA started (1) ZrC-coating process development by large-scale coater, (2) inspection method development of ZrC coating and (3) irradiation test and post irradiation experiment of ZrC coated particles under contract research which is entrusted to the JAEA and MEXT.
The design framework and operational guidelines for conducting repetitive dialogue between public and nuclear engineers are described in this paper. An action research project named repetitive dialogue forum has been conducted in two municipalities where nuclear facilities were sited. The qualitative evaluation by public participants indicated that the public trust in the nuclear experts, known as the crucial factor for meaningful communication, was successfully established through the dialogue forum. In addition, the expert showed a marked psychological change from distrust to trust in public. Through a detailed analysis of the comments of the participants raised during the forums, the nuclear risk recognition scheme of the public was clarified. The constituents of the risk recognition scheme about nuclear facilities were identified as follows. The first is related to the technical risk recognition factor including purely technical risk, organizational elements and regulatory elements. The second is the social risk recognition factor including economical and mental elements. The last is the communication factor including the influence of mass media, difficulty in frank communication in local community etc. It became clear that the information provision activities conducted by the government and the nuclear industry were lack of in-depth understanding of actual information needs in the public. Provision of information contents consistent with our observations is recommended for reestablishment of public trust in expert and for more informative dialogical interactions.
Japan Atomic Energy Agency (JAEA) has been carrying out design studies of the Gas Turbine High Temperature Reactor 300 for Cogeneration (GTHTR300C). One of the key components in the GTHTR300C is an intermediate heat exchanger (IHX). The IHX for the GTHTR300C is rated 170MWt and designed on the basis of the shell & tube type IHX adopted in the High Temperature engineering Test Reactor (HTTR). As a conceptual design, we selected the heat exchanger tube size which is based on heat transfer calculation, designed conceptual structure, and evaluated the primary stress in design condition. By the results of this study, the technical feasibility of the large size shell & tube type IHX for the GTHTR300C was verified.
Sodium combustion tests were performed to investigate both the pool burning behavior and the maximum temperature of a steel-made floor liner in the case of small-scale sodium leakage from a secondary loop of a fast breeder reactor. Sodium leakage heights, humidity in an atmosphere and sodium leakage rate were changed as parameters in the tests. This paper mainly described thermal influence to the steel liner. Main conclusions are as follows. The maximum liner temperature becomes higher with the increase in the leakage height and humidity. When the leakage rate decreases, the maximum temperature tends to become lower, and this behavior has the relation with final pool size that affects to heat release toward radial direction. Sodium leakage rate has no dependency on the sodium burning rate and pool expanding speed while it affects to the final pool size. The surface shape of sodium combustion pool has influence on both the pool expanding speed and the final pool size.
As a part of the materials aging degradation and structural integrity research for LWR components, the probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed in JAEA. This code evaluates the conditional probabilities of crack initiation and fracture of a reactor pressure vessel (RPV) under transient conditions such as pressurized thermal shock (PTS). A standardized analysis method is proposed on the basis of the development of PASCAL ver. 2 and results of sensitivity analyses. Graphical user interface (GUI) including the standardized analysis method as default settings and values has been also developed for PASCAL ver. 2. A case study showed that non-destructive examination with good performance had a more significant effect on the probability of failure than non-destructive examinations repeated with low performance.
Fifty-five nuclear power plants of light water reactors (LWRs) in Japan including seven plants which have already reached a 30-year operation have been operated as of September 2006. The probabilistic fracture mechanics (PFM) method has been recently highlighted to rationally incorporate the uncertainties arising from the material properties, defect distribution, inspection quality and so on, unlike the conventional deterministic method. As a part of the materials aging degradation and structural integrity research for LWR components, the PFM analysis code PASCAL has been developed in JAEA. This code evaluates the conditional probabilities of crack initiation and fracture of a reactor pressure vessel (RPV) under transient conditions such as pressurized thermal shock (PTS). Sensitivity analyses for pre-service and in-service inspections have been performed according to JSME S NA1-2004 and the correlation between the results by deterministic analysis according to the JEAC 4206-2004 and probabilistic analysis based on PFM for the integrity of RPV under PTS has also been described.
We studied the effect of the flow rate of reactor water on crack growth rates (CGRs) in a boiling water reactor (BWR) coolant system using 1/4TCT specimens made of sensitized 304 stainless steel (SUS304). Specimen electrochemical corrosion potentials (ECPs) in each oxidant concentration increased with increasing of water flow. This was caused by the increase of the mass transfer amount from bulk water to the specimen surface through a diffusion layer determined by the water flow. A relative CGR was defined as the ratio of the CGR under a given set of conditions to that under simulated normal water chemistry (NWC) at a low flow rate. The relative CGRs at high flow rates were higher than those at low flow rates under NWC. However, when the concentration of oxidants was below 10 ppb and the specimen ECP was below −0.15 V vs. Standard Hydrogen Electrode, the relative CGRs at high flow rates became as low as those at low flow rates. The mass flux of the oxidant was one dominant parameter, which controlled the CGRs at high flow rates. Although the flow rates in a BWR vessel and piping are very high, the CGRs should decrease when the oxidant concentration is adequately reduced by applying hydrogen water chemistry.
The process concept of separation system of high-level radioactive liquid waste (HLLW) using an octylphenyl-N,N-diisobutylcarbomoyl phosphine oxide impregnated silica adsorbent (CMPO-adsorbent) was designed for separation of Cs and Sr. The process consists of two separation steps. Each step is composed of three CMPO-adsorbent columns, and the adsorbents are used in rotation for separation of HLLW. That is, the system has six columns. Uranium and Zirconium inhibit the adsorption of minor actinoids (MA) and lanthanoids. Therefore, the adsorption and removal of U and Zr from the HLLW occur in the 1st step. Adsorbed U and Zr are eluted using citric acid. The HLLW from the 1st step is sent to the 2nd step to adsorb MA, lanthanoids and other transition metals. MA and lanthanoids are eluted using 0.01 mol/dm3 HNO3, and Mo is recovered by using oxalic acid. The process required a large amount of citric acid and oxalic acid, but these eluents can be reused, therefore the amounts of spent acids are reduced. The column adopted a cartridge form in consideration of the exchange of adsorbent. The bed volume of the adsorbent is 0.67 m3/column. The size of separation plant has 60 m length, 80 m breadth and 36 m height.
Theoretical study has been performed to clarify the ability of colloid release form the montmorillonite gel by the flowing groundwater. Evaluation of montmorillonite colloidal particles release from the bentonite buffer material is important for the performance assessment of radioactive waste disposal because the colloids may influence the radionuclide transport. In this study, the minimum groundwater flow rate required to tear off montmorillonite particles from surface of bentonite buffer was estimated from the shear stress on the gel front, which was calculated by the DLVO theory. The estimated shear force was converted to corresponding groundwater velocity by using Stoke's equation. The results indicated that groundwater velocity in a range of about 10−5 to 10−4 m/s would be necessary to release montmorillonite particles. This range is higher than the groundwater flow velocity found generally in deep geological media in Japan. This study suggests that the effect of montmorillonite particles release from the bentonite buffer on radionuclide transport is likely to be negligible in the performance assessment of high-level radioactive waste geological disposal.
Super-DIREX (Supercritical Fluid Direct Extraction) method is a technology that enables to extract U and Pu directly from decladded spent fuel, using supercritical carbon dioxide. Main process of reprocessing plant for LWR and FBR spent fuel by Super-DIREX method was designed conceptually. The number of main equipments at processes from head end to denitration decreased to about 0.66 times of that of the PUREX plant in case of a capacity of 800 t/y(LWR) and decreased to about 0.43 and 0.66 times in case of 200 t/y(LWR) and 200 t/y(FBR), respectively. The volume of acid liquid waste discharged from main process in the Super-DIREX plant decreased to about 0.2 times of that of PUREX plant in case of a capacity of 800 t/y(LWR). The equipment cost decreased to about 0.26 times of that of PUREX plant in case of 200 t/y(FBR).
The interim storage facility for spent fuel metallic cask is designed as a concrete building structure with air inlet and outlet for circulating the natural cooling. The feature of the interim storage facility is big capacity of spent fuel at several thousands MTU and restricted site usage. It is important to evaluate realistic dose rate in shielding design of the interim storage facility, therefore the three-dimensional continuous-energy Monte Carlo radiation transport code MCNP that exactly treating the complicated geometry was applied. The validation of dose evaluation for interim storage facility by MCNP code were performed by three kinds of neutron shielding benchmark experiments; cask shadow shielding experiment, duct streaming experiment and concrete deep penetration experiment. Dose rate distributions at each benchmark were measured and compared with the calculated results. The comparison showed a good consistency between calculation and experiment results.