The purpose of this paper is to extract problems of the regulations on nuclear power plants in Japan in order to indicate the direction of the solution as much as possible. This paper includes problems about the duplicated application of Nuclear Reactor Regulation Law and Electricity Utility Law, the examination criteria of nuclear reactor establishment permission, the relation between the establishment permission and the regulations, the double check evaluation system by the Minister of Economy, Trade and Industry, and Nuclear Safety Commission, fuel inspection and in-service inspection, management of the information about insignificant events, and so on.
Design studies of the hydrogen cogeneration high temperature gas cooled reactor (GTHTR300C), which can produce both on electric power and hydrogen, have proceeded in Japan Atomic Energy Agency. In future, it will be obliged to operate using not enriched uranium but plutonium to coexist with fast reactors after the full deployment of a fast reactor cycle. Therefore, a nuclear and thermal design has been performed to confirm the feasibility of the reactor core using Mixed-Oxide (MOX) fuel. The reactor core with operation period of 450days and average burn-up of 123 GWd/ton for discharged fuel was designed. The reactor core met safety requirements of the maximum fuel temperature of less than 1,400°C during normal operation, the maximum power density of less than 13 W/cm3, shutdown margin of more than 1.0%Δk/kk′ and negative reactivity coefficient. The results proved that it is possible to operate the GTHTR300C using MOX fuels without consuming natural uranium resources.
A HTTR (High Temperature Engineering Test Reactor), which has a thermal output of 30 MW, a coolant inlet temperature of 395°C and a coolant outlet temperature of 850°C/950°C, is a first high-temperature gas-cooled reactor (HTGR) in Japan. The HTGR has a high inherent safety potential to accidents. The safety demonstration tests such as the reactivity insertion and the coolant flow reduction tests using the HTTR are underway in order to demonstrate such excellent inherent safety features of HTGRs. These tests demonstrate that the rapid increase or decrease in the reactor power is restrained by only the negative reactivity feedback effect without an operation of the reactor power control system, and the temperature transient of the reactor is slow. A one-point core dynamics approximation with one fuel channel model could not simulate accurately the reactor power behavior. On the other hand, an original new method using region temperature coefficients and a connection between some fuel channel models and whole core component model for calculating radial heat transfer in the core is adopted. It is crucial to evaluate this method precisely to simulate a performance of HTGR with showing excellent inherent safety features under the reactivity incident by gas circulator tripping error.
The design studies on High Temperature Gas Cooled Reactor with Gas Turbine (HTGR-GT) have been performed, which were mainly promoted by Japan Atomic Energy Agency (JAEA) and supported by fabricators in Japan. HTGR-GT plant feature is almost determined by selection of power conversion system concepts. Therefore, plant design philosophy is observed characteristically in selection of them. This paper describes the evaluation and analysis of the essential concepts of the HTGR-GT power conversion system through the investigations based on our experiences and engineering knowledge as a fabricator. As a result, the following concepts were evaluated that have advantages against other competitive one, such as the horizontal turbo machine rotor, the turbo machine in an individual vessel, the turbo machine with single shaft, the generator inside the power conversion vessel, and the power conversion system cycle with an intercooler. The results of the study can contribute as reference data when the concepts will be selected. Furthermore, we addressed reasonableness about the concept selection of the Gas Turbine High Temperature Reactor GTHTR300 power conversion system, which has been promoted by JAEA. As a conclusion, we recognized the GTHTR300 would be one of the most promising concepts for commercialization in near future.
We conducted the experimental studies on the water hammer caused by striking of a water mass pushed up by a rapidly growing steam bubble, using a cylindrical model containment vessel of 0.4286 m in diameter. In the experiments, a rapid gas growth was simulated by injecting high-pressure steam into a water pool. It was clarified that coherency of the water mass movement and its water hammer caused by the condensable gas production considerably decreased in comparison with the case of the non-condensable gas production because the rising velocity of the water mass was suppressed due to the steam bubble condensation. On the basis of the data, experimental correlations for estimating the water hammer on the structures in the containment vessel were proposed.
For numerical prediction of gas entrainment from a free surface in an upper plenum of a sodium-cooled fast breeder reactor (FBR), accurate computation of a swirling flow or vortices is one of the crucial ingredients. In this study, a multiscale large-eddy simulation (LES) code, MISTRAL, is applied to a swirl flow in a cylindrical vessel to examine the effects of the multiscale turbulence models on the accuracy of vortex prediction. The numerical results are compared with experimental results and another numerical results computed with a conventional LES code (SMART-fem). The comparison indicates that the multiscale LES code is able to resolve the fine vortex core structure more accurately than the conventional LES code. Especially, elimination of eddy viscosity in the large-scale equations (or introduction of eddy viscosity only in the small-scale equations) can improve the accuracy in computing a circumferential velocity around the vortex core. It has also turned out that a use of coarse mesh suppresses the peak value of circumferential velocity, which suggests us to employ an extrapolation method such as a vortex model in practice for quantitative estimate of unresolved vortex features.
A reduction of the maintenance cost and an improvement of utilization in facilities have been performed by introducing Condition-Based Maintenance (CBM) to nuclear power plants since the 1980s in USA. In Japan, also, maintenance methods combined with TBM (Time-Based Maintenance) and CBM have a tendency to be applied. In the present work, an estimation and optimization model for extending TBM time interval is developed considering spurious diagnosis of measuring devices in the maintenance method combined with TBM and CBM. Using the present model, the life time of applied measuring devices can be estimated in order to prolong the time interval of TBM execution and satisfy the required one. Moreover, TBM time interval can be given for minimizing the maintenance cost under the condition of the target safety. From a sensitivity analysis of the parameters including the present model, it is clarified that leaving the facilities in an abnormal state has the largest effect on the cost of damage for the maintenance.
In consideration of applying methods of criticality safety control based on neutron moderation management to a new uranium oxide storage building which Japan Nuclear Fuel Ltd. (JNFL) plans to construct at the Rokkasho Reprocessing Plant in the future, it is necessary to verify that water content of UO3 powder controlled by the process will not increase, (1) on condition that, after it is filled to storage canisters, which are sealed and stored for a long term, and (2) in the process that it is filled to these canisters. Then, concerning the former (1), we investigated operations in the uranium denitration facility at Japan Atomic Energy Agency similar to the one at JNFL and the actual values of water content of UO3 powder about its storage. On the other hand, concerning the latter (2), it is possible to verify that basically by sampling and analyses of water content of UO3 powder in the filling process. But, for the purpose of increasing reliability for criticality safety control more, we newly developed the system which continuously measures water content of UO3 powder with infrared moisture meters and we confirmed the effectiveness in a mock-up test. As a result, it was found that water content of UO3 powder in the filling process was stable below 0.5% and that it did not tend to increase even after storage for about 15 years. Furthermore, we developed two systems which continuously measure water content of UO3 powder with infrared moisture meters, and then we got prospects to apply to the real plant. Based on these results, we suggested a basic policy of criticality safety control based on the neutron moderation management for the new uranium oxide storage building as an example.
The distribution coefficient for U(VI) between an aqueous phase and a supercritical one was determined with an absorption spectrophotometer, which is corresponding to two phase measurements at a time in a high pressure vessel through sapphire windows. The distribution feature of an aqueous/supercritical phase shows the same pattern with that of an aqueous/organic phase, so, the mechanism of U(VI) extraction in TBP of both systems will be identical. The distribution of U(VI) is influenced a great deal by the concentration of nitric acid. The distribution of nitric acid was determined to fix the nitric acid concentration and the distribution of U(VI) because the nitric acid coordinated to TBP in a supercritical fluid of carbon dioxide is also stripped into an aqueous phase. The distribution coefficient for nitric acid between an aqueous phase and TBP in a supercritical fluid of carbon dioxide consists with the one between an aqueous phase and TBP. The stripping of U(VI) from a supercritical fluid of carbon dioxide to aqueous nitric acid with a countercurrent column was demonstrated and about 80% of U(VI) was collected in the aqueous phase through a column of 380 mm high.
Spent zircaloy cladding and channel boxes from Boiling Water Reactors (BWR), which constitute spent zircaloy waste, are not recycled in spent fuel reprocessing plants. The zircaloy waste should be pressed and packed into cans for radioactive waste disposal. The disposal costs of zircaloy waste were estimated to be about 440 billion yen because they require the same deep disposal as the high-β•γ waste generated from the spent fuel reprocessing plant at Rokkasho in Aomori Prefecture. The recovery and recycle of expensive zirconium from the spent zircaloy waste is possibly one of the most effective solutions to the problem regarding environmental issues in the nuclear field. Electrorefining in molten salts for the zirconium recovery process is a promising technique, which has high decontamination factors (DFs) and reduces the secondary waste. This study aims to confirm the feasibility of the zirconium recovery process from the spent channel boxes after eliminating radioactive nuclides and to illustrate the economic feasibility of zirconium recycle. Zirconium was deposited on the cathode and recovered by electrorefining in molten salts to separate radioactive cobalt and antimony. The zirconium recycle cost was estimated to be about 1/10 of direct disposal cost.
Continuous stripping of rare earth elements (REs) with the diethylenetriaminepentaacetic acid (DTPA) solution from the diisodecylphosphoric acid (DIDPA) solvent was examined with lab-scale mixer-settlers to develop a process for the separation of trivalent actinides (An(III)) from the mixture of An(III) and REs. DIDPA is an extractant applied to separate transuranic elements from high-level liquid waste in the 4-Group Partitioning Process developed in Japan Atomic Energy Research Institute (presently: Japan Atomic Energy Agency). The experiment on continuous stripping with the DTPA solution of a little higher pH than an appropriate pH for selective stripping of An(III) showed that Nd could be selectively stripped from the DIDPA solvent containing La and Nd. The stripping yield of Nd was found to be higher than 99.8%, whereas 97.6% of La was left in the solvent. The pH of the DTPA solution decreased in the stages near the solvent feed, which was explained by the extraction of ammonium ion. These findings make it possible to predict quantitatively the behavior of An(III) and REs in the various conditions of the DIDPA-DTPA continuous stripping system.
The first wall of the International Thermonuclear Experimental Reactor (ITER) will be fabricated by means of the Hot Isostatic Pressing (HIP) method for the bonding of cooling tubes and a copper alloy heat sink. A ultrasonic testing (UT) is adopted as a non-destructive inspection method for the bonding interface as one of the acceptance tests of the first wall components. Therefore, clarification of a defect size criteria is one of the critical issues for the soundness of the first wall. In this paper, a thermo-mechanical behavior of an initial defect at the bonded interface of the ITER first wall was numerically analyzed. An elastoplastic fracture mechanical parameter, J-integral, was calculated to evaluate the propagation behavior of the interfacial defects under thermal loading. As a result, it was found that the initial defect size of 10 mm×20 mm in semi-elliptic shape was unlikely to propagate under the thermal loading of ITER. This defect size is more than ten times larger than a detection limit of present UT techniques, and it can be resulted that the UT method presently available is sufficient to detect such harmful initial defects of the ITER first wall.
A source efficiency was measured experimentally for various materials, such as metals, nonmetals, flooring materials, sheets and other materials, contaminated by alpha- and beta-emitter radioactive nuclides. Five nuclides, 147Pm, 60Co, 137Cs, 204Tl and 90Sr-90Y, were used as the beta emitters, and 241Am was used as the alpha emitter. The source efficiencies of nonpermeable materials were higher than those given by the Japanese Industrial Standards (JIS). In contrast, the source efficiencies of some permeable materials were lower than those given by the JIS because the source efficiency varies depending on whether the materials or radioactive sources are wet or dry. This study provides basic data on the source efficiency for estimating the surface contamination level of materials.
Accidents such as oil-tank fires and breakage of members of civil structures caused by the earthquakes have occurred during the past several years. Reliability is expected to maintain the safety of infrastructures like nuclear plants. We have focused on the construction of an analysis system called the “three-dimensional virtual plant vibration simulator,” which is a numerical simulation system for a nuclear plant which considers the interconnection of machines, pipes, buildings, and their foundations under real operating conditions. In this paper, as the first step of this simulation system, the “assembled structural analysis” is proposed in which each structural component is treated independently and analyzed as an assembly structure. The system configurations in a parallel distribution environment are described. This study shows an example of the application of this method to a real nuclear-plant cooling system that has tens of millions of degrees of freedom.