Based on an extended literature survey covering recent studies on high-level radioactive glass dissolution under nearly saturated conditions, we have reached the conclusion that the slow dissolution is controlled by the diffusion of oxonium or boron ions in an altered layer formed on the glass surface. Experimental approaches, such as an elaborate and systematic diffusion experiment using isotopes, were proposed to validate the mechanism. It was also pointed out that the dissolution model applicable to glass waste form performance evaluation takes into account the surface area evolution, the stability of the altered layer, and the interactions with near-field materials.
For the sake of more efficient operation of nuclear power plants and to reduce the number of spent nuclear fuel assemblies, increasing uranium enrichment is one of the rational options. However, current uranium enrichment for the whole fuel cycle infrastructure is limited to no greater than 5 wt% from the view point of criticality safety. In this review, three main topics are discussed: 1) the necessity of increasing the uranium enrichment above 5 wt%; 2) current status and challenges to go over 5 wt% enrichment; and 3) proposal of the “Erbia Credit Super High Burnup Fuel” as a measure to break the “5 wt% enrichment barrier.” The third topic is further elaborated on by discussing the introduction scenarios before concluding with a mention of the necessity of a best-mix analysis for this concept in the fuel cycle supply chain.
Replacement of nuclear power plants has the possibility of affecting the management of electric power suppliers. Therefore, in the nuclear policy, a depreciation method as an equalization method, which means that part of the investment cost is accumulated as an allowance, and after the start of operation, the depreciation cost in the replacement project is equalized, has been introduced in Japan. In this paper, we evaluate the replacement of nuclear power plants by taking into account the uncertainty of operating costs and the depreciation cost in order to examine the effect of the depreciation method on the decision criteria of the replacement. We found that the equalization method is effective for inducing the acceleration of the replacement. Furthermore, we show the relationship between the uncertainty and the depreciation method. It turns out that as uncertainty increases, the difference in investment threshold between the equalization method and the existing depreciation method decreases, and that in option value increases.
The simplified decommissioning cost estimation code for nuclear facilities (DECOST code) was developed in consideration of features and structures of nuclear facilities and similarity of dismantling methods. The DECOST code could calculate 8 evaluation items of decommissioning cost. Actual dismantling in the Japan Atomic Energy Agency (JAEA) was evaluated; unit conversion factors used to calculate the manpower of dismantling activities were evaluated. Consequently, unit conversion factors of general components could be classified into three kinds. Weights of components and structures of the facility were necessary for calculation of manpower. Methods for evaluating weights of components and structures of the facility were studied. Consequently, the weight of components in the facility was proportional to the weight of structures of the facility. The weight of structures of the facility was proportional to the total area of floors in the facility. Decommissioning costs of 7 nuclear facilities in the JAEA were calculated by using the DECOST code. To verify the calculated results, the calculated manpower was compared with the manpower gained from actual dismantling. Consequently, the calculated manpower and actual manpower were almost equal. The outline of the DECOST code, evaluation results of unit conversion factors, the evaluation method of the weights of components and structures of the facility are described in this report.
An application of a cathodic protection method with an impressed current system to control the corrosion of austenitic stainless steel in a boiling nitric acid solution was studied to improve corrosion resistance and to extend the operation life of components in a fuel reprocessing plant. Plate-type specimens made of ultralow carbon type 304 stainless steel (SUS304ULC) were immersed in 3 mol·dm−3 boiling nitric acid solutions including 10 and 1.7 g·dm−3 vanadium ions. Electrochemical potentiostatic tests and cathodic protection tests were performed using electrochemical test cells. The selected protective potential was below the transition potential between the passive and trans-passive states based on anodic polarization measurement. Corrosion rates in the solution with and without the protection were measured by potentiostatic tests. Additionally, the outer surface of the tube-type specimen of SUS304ULC was studied under the same condition. From the obtained results, corrosion rates of plate-type specimens with cathodic protection were observed to decrease by 1/40 and 1/10 those of the specimens without cathodic protection in the solutions including 10 and 1.7 g·dm−3 vanadium ions, respectively. In the case of tube-type specimens, outer surface thickness loss was decreased from 24 to 3 μm by the protection, and platinum was chosen as the anode because it showed no corrosion loss like gold and no cracking like zirconium. Authors concluded that the cathodic protection method can be expected as one of the methods of maintaining components in a fuel reprocessing plant.
The Japan Atomic Energy Agency has been conducting R&D on the thermochemical iodine-sulfur process for large-scale hydrogen production using nuclear heat supplied by high-temperature gas-cooled reactors. Since the IS process uses strong acids such as sulfuric acid, it is important to examine the applicability of equipment such as pipes, pumps and instrumentations exposed to corrosive environments, and, therefore, flow tests of concentrated sulfuric acid have been carried out using a test apparatus made of candidate materials of construction at temperatures of up to 300°C for ca.150 h. Glass lining pipes and PTFE gaskets used for high-temperature service were sound during the test term, whereas high-silicon-containing austenitic stainless-steel pipes suffered general corrosion.
The least-squares inverse kinetics method (LSIKM) has been frequently applied to constant-reactivity insertion experiments, such as a control-rod drop experiment conducted to determine not only the reactivity but also the effective source strength of a subcritical system driven by a neutron source. When a large negative reactivity is inserted into such a loosely coupled core system consisting of two-core and one-large-core reactors, spatial higher-harmonic modes are probably significantly excited. Consequently, the excitation leads to the failure of the one-point kinetics model, and the LSIKM based on the model infers spatially dependent reactivity and source strength. In this paper, we present a reduction technique for the apparent spatial dependence and a demonstration of the applicability of such a technique to rod drop experiments carried out in both the subcritical and critical states of the two-core reactor.
The fabrication of fast neutron reactor cycles is intended for next-generation nuclear energy systems. This is in line with the fact that the amount of plutonium, which should be reprocessed, increases significantly. Techniques for increasing the nuclear proliferation resistance, especially extrinsic measures including safeguards, are essential for such systems to be accepted by the international community. A highly resistant system with a high detection capability, as well as satisfying current safeguards requirements, was studied for an advanced aqueous reprocessing, and its technical practicability and operational compatibility were discussed. The effect of the proposed safeguards system in this paper was evaluated using the Markov model approach developed by the GIF Proliferation Resistance and Physical Protection Working Group (PR & PP WG). The proposed safeguards system includes a high-detection-capability system and the accountancy/verification measures based on the monthly interim inventory taking or verification that should be performed at a similar level of quality to the normal physical inventory verification with very little impact to the practical plant operation. This can only be realized with “safeguards by design.”
Some pressurized water reactor (PWR) plants have switched secondary system feed water treatment to ethanol-amine (ETA) injection from all-volatile treatment (AVT) to reduce iron transfer in the steam generator (SG). However, the effect of ETA injection on FAC rate has not been studied systematically. To assess the effect of ETA injection on FAC rate, the water chemistries in secondary systems were calculated by considering the thermal decomposition of hydrazine in SG and the vapor/liquid partition of ammonia, ETA, and hydrazine in SG and in a moisture-separator-and-reheater (MSR). Then, we measured the FAC rate experimentally by rotation tests to examine the effect of ETA injection. The high pH condition of ETA injection reduced the FAC rate more than the low pH condition of AVT. No chemical effect on the FAC rate was observed between ETA injection and AVT at 180°C. We also evaluated the FAC rate using magnetite solubility with and without ETA injection. The evaluation showed that ETA injection reduces the FAC rate of the secondary system.
The Japan Atomic Energy Agency is performing a research project in the Mizunami Underground Research Laboratory (MIU) to build a firm scientific and technological basis for the studies of the deep underground environment in crystalline rock. In the project, it is necessary to reduce the fluorine and boron concentrations in groundwater pumped from the MIU shafts to levels below the environmental standards. This is done at the MIU water treatment facility using coagulation and ion exchange treatment for fluorine and boron, respectively. In addition, in 2006, research started on the efficient treatment of groundwater for removal of fluorine and boron using a radiation-induced graft polymerization adsorbent. The adsorbent removed boron at a flow rate (space velocity (SV)=120 h−1) higher than that of a general ion exchange resin (SV=10 h−1) and the adsorbent could be used repeatedly. It was also apparent that the pH of groundwater had an influence on adsorption performance. With respect to fluorine removal, more than 90% of fluorine was removed. However, the adsorbent for fluorine showed a lower adsorption capacity than that for boron. The reason for this difference is considered to be related to the initial concentration difference between fluorine and boron in the groundwater. Therefore, it is necessary to define the initial concentrations of dissolved materials, which can be used as better indicators of the performance of the adsorbent.
A special committee of “Research on analysis methods for accident consequence in nuclear fuel facilities” was organized by the Atomic Energy Society of Japan under the entrustment of Japan Atomic Energy Agency (JAEA). The research results were summarized in a series of six reports, which consist of the general report and five technical ones. This is the fifth technical one dealing with airborne release fraction (ARF) in the case of accidental leakage of molten glass from a vitrification melter. In the effective dose assessment of the design basis accident of the molten glass leakage at the Rokkasho Reprocessing Plant, ARFs of 1 for Ru and Cs and of 0.1 for nonvolatile elements, such as Sr and minor actinides, were used. In the present report, more realistic values of 1 × 10−3 for Cs, 5 × 10−5 for Ru, and 10−8 for nonvolatile elements, such as Sr and minor actinides, were proposed for risk assessment; these values were obtained by calculation using the measured volatilities of these elements at high temperatures of 800-1,100°C reported in the literature. Although these values depend on the scale and conditions of the leakage, they are given as maximums irrespective of the conditions.