We found loopholes of laws and regulations for supervising radioactive materials. It is not obliged to measure the soil radioactivity of the sites that were formerly used as scientific or engineering institutes, or hospitals with a radioactive material control area. If the former institutes or hospitals made studies with radioactive materials before the enforcement of the law concerning prevention from radiation hazards due to isotopes and its detailed regulations, it is concluded that there was the period when the radioactive materials were not under management. If it is found that the radioactive materials were applied at the former site before the enforcement of the related laws and regulations, the radioactivity in the soil of the redeveloped area should be examined, which should be obliged by some laws or regulations.
This paper deals with the heat removal performance of the BWR drywell local cooler (DWC) applied as a Japanese phase-II accident management. Separate effect tests were conducted using a single DWC unit of a typical BWR plant under severe accident (SA) condition. It was demonstrated that noncondensable gas mixture with nitrogen and helium was constantly vented from the DWC casing by natural circulation due to the density difference between DWC tube space and outside. Favorable steam condensation rate was maintained even under the highest assumed noncondensable gas condition and obtained condensation data sets were reduced to empirical correlations for the DWC heat removal model. The dependence of the DWC heat removal performance on the noncondensable species was discussed by analyzing the computational fluid dynamic code of STAR-CD. In conclusion, the DWC was found to be promising even under wide range of SA conditions.
This paper deals with the system interaction performance of the BWR drywell local cooler (DWC) in combination with containment spray as a Japanese Phase-II accident management (AM). By using almost full height simulation test facility (GIRAFFE-DWC) with volumetric scaling ratio of 1/600 for a typical BWR containment, the system integral tests simulating BWR low pressure vessel failure sequence were accomplished during about 14 hours. In case of DWC application, the containment pressure increase was found milder due to DWC heat removal performance. Initial spray timing was delayed about 3 hours and each spray period was reduced almost by half, which contributed to the containment gas compression. These containment pressure transients were confirmed by analyzing the severe accident analysis code of MELCOR. It was concluded that the application of a BWR DWC to Phase-II AM measure is quite promising from the point of delaying or preventing the containment venting.
A new approach for the numerical simulation of the Molten Salt Reactor (MSR) is described in this paper. The MSR is a thermal neutron reactor with graphite moderation and employs fissile salt as fuel material. Within the MSR, the fission reaction, the fuel salt flow and the heat transfer are inseparably linked each other due to fission fuel circulation in the reactor. A numerical model developed here consists of two-group diffusion equations for fast and thermal neutron fluxes, balance equations with convective terms for six-group delayed neutron precursors, an energy equation for the heat convection and conduction of fuel salt and an energy equation for the heat conduction of graphite moderator. These equations were iteratively solved by the SOR algorithm. The verification of the model of the fission reaction was performed using the Standard Reactor Analysis Code (SRAC). The results showed that the model has comparable accuracy with the SRAC. Moreover, the interactions of the fission reaction, the fuel salt flow and the heat transfer were analyzed. As a result, the fuel salt flow affected the distributions of the precursors significantly, but those of neutron fluxes slightly. And both the precursors and the neutron fluxes decreased in the high temperature region of the reactor. In contrast, they increased in the low temperature region.
During a severe accident of a light water reactor, the failure of piping of the reactor cooling system could occur due to a thermal load, resulted from the heat transfer from a high temperature gas generated in the reactor core and decay heat released from fission product deposits. It is considered that, under such a condition, the short-term creep at a high temperature causes the piping failure. The objective of the present study is to predict the piping failure quantitatively. For this purpose, the development of an analytical method for the accurate prediction of the creep deformation is required, in which a creep constitutive equation taking the creep damage into account should be used, in order to evaluate the structural integrity of the piping during the severe accident. In this paper, creep constitutive equations considering the tertiary creep was fabricated for cold-drawn type 316 stainless steel (SUS316) based on the isotropic damage rule proposed by Kachanov-Rabotnov. In addition, creep analyses were performed for a pipe made of cold-drawn SUS316 under a condition that elevated temperature distribution was established in the pipe wall. The numerical results show that the damage of the pipe is quantitatively described by the damage variable introduced in the finite element analyses, and the failure characteristics are in reasonable agreement with those observed in a piping failure test. The failure time does not agree well with that the time of the piping failure test. It is, however, indicated that we can estimate the state of the failure of the coolant piping under severe accident by the accurate estimation of the temperature.
Japan Nuclear Cycle Development Institute (JNC) is studying the electro-winning process using NaCl-2CsCl molten salt as a dry reprocessing technology development. It would be necessary to repeatedly purify the used salt after the MOX co-deposition electrolysis and the MA recovery in this process. JNC had selected the phosphate precipitation method as a purification technique of the used salt, and we carried out the phosphate precipitation experiment. In this experiment, Na3PO4 was added into the NaCl-2CsCl molten salt including fission product (FP) ions (Ce3+ Nd3+, Sr2+) of non-radioactive, and we investigated the removing performance of FP ions from the NaCl-2CsCl molten salt by the phosphate precipitation. In conclusion, we confirmed most of FP ions in the molten salt were removed by adding Na3PO4 of 1.5 times the mole amount of the FP content.
The observed change of public attitude around the time of inhabitants' poll in Japan was compared with model calculation to investigate its non-linear behavior. Two inhabitants' polls regarding nuclear issues, the approval and disapproval of the construction of Maki nuclear station, and of the MOX fuel use at Kashiwazaki-Kariwa nuclear station, were considered together with the poll on the reconstruction of the tenth weir in Yoshino river carried out in Tokushima. By using a mathematical model such that the individual attitude is mainly subject to two factors of the information environment and the mutual communication between the public, it was found that the change and the unification of public attitude around the time of inhabitants' poll can be interpreted as a manifestation of self-organization resulted from the cooperative phenomenon of those two factors. Moreover, it was also found that the abrupt change of public attitude just before the poll can be interpreted as a result of positive feedback of the information environment formed by the various types of propaganda activities to the attitude change, though the extent of such non-linear effects differs from case to case.
Research and development program for helium gas compressor aerodynamics was planned for the power conversion system of the Gas Turbine High Temperature Reactor (GTHTR300). The axial compressor with polytropic efficiency of 90% and surge margin more than 30% was designed with 3-dimensional aerodynamic design. Performance and surge margin of the helium gas compressor tends to be lower due to the higher boss ratio which makes the tip clearance wide relative to the blade height, as well as due to a larger number of stages. The compressor was designed on the basis of methods and data for the aerodynamic design of industrial open-cycle gas-turbine. To validate the design of the helium gas compressor of the GTHTR300, aerodynamic performance tests were planned, and a 1/3-scale, 4-stage compressor model was designed. In the tests, the performance data of the helium gas compressor model will be acquired by using helium gas as a working fluid. The maximum design pressure at the model inlet is 0.88MPa, which allows the Reynolds number to be sufficiently high. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan.
The fuel assembly design with tightened fuel rod pitch gives a possibility of higher conversion ratio for Boiling Water Reactor (s). Pressure drop is one of the key design parameters to evaluate the possibility of BWR core with higher conversion ratio. The pressure drop test for tight lattice rod bundles under steady conditions has been done under BWR fluid conditions. The following results were obtained: (1) The present prediction method has the reasonable performance under 25% of quality at the rated mass flux, but the predicted value is larger than the measured ones for larger mass flux. (2) This method can be applied to evaluate the pressure drop for tight lattice bundles from the conservative point of view, but it is necessary to improve the predicted accuracy for larger mass flux.
The design targets of main control boards of nuclear power plants are as follows. (1) To make a good working environment where operators can operate plants easily. (2) To reduce the work load and operators human error. To this end, Mitsubishi group and Japanese PWR utilities have been improving main control boards design for next PWR plants with full digital instrumentation and control system. The main control boards consist of a soft operation console and a large display panel. According to our evaluation, the work load and human error rate of the new main control boards are reduced much compared with the latest plants. In this study, in order to establish the standard specification of the advanced main control boards, validation test of mock-up for the actual plant, from 1998 to 2001, the dynamic operator verification linked with a simulator had been executed, as well as operational know-how is being collected, and detailed standardization is executed.
Basic policy of maintenance was determined for major equipment in the power conversion system of the Gas Turbine High Temperature Reactor 300 (GTHTR300). It was developed based on the current maintenance practice in Light Water Reactors (LWRs), High Temperature Engineering Test Reactor (HTTR) and conventional combined cycle power plants while taking into account of unique design features of GTHTR300. First, potential degradation phenomena in operations were identified and corresponding maintenance approaches were proposed for the equipment. Such degradations encountered typically in LWRs as corrosion, erosion and stress corrosion cracking are unlikely to occur since the working fluid of GTHTR300 is inert helium. Main causes of the degradations are high operating temperature and pressure. The gas turbine, compressor, generator, control valves undergo opening and dismantling maintenance in a suitable time interval. The power conversion vessel, heat exchanger vessel, primary system piping and heat exchanging tubes of precooler are subjected to in-service inspections similar to those done in LWRs. As turbine blades represent the severest material degradation because of their high-temperature and high-stress operating conditions, a lifetime management scheme was suggested for them. The longest interval of open-casing maintenance of the gas turbine is estimated to be six to seven years from technical point of view. Present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan.
The field migration test using TRU nuclide was carried out as a cooperative research project between JAERI (Japan Atomic Energy Research Institute) and CIRP (China Institute for Radiation Protection). This report introduced the outline of the field migration test and described the outline of the series of "Field Test on Migration of TRU-nuclide" and main results as a summary report.
Field tests on migration of radionuclides for engineered barrier materials such as bentonite and cementitious materials were performed. The tests were run under both wet conditions with artificial rainfall and dry conditions with natural rainfall. Laboratory experiments such as batch adsorption tests were also conducted to analyze the result of field test. The results of field tests agreed with the predicted moisture conditions and the migration behaviors observed at the laboratory experiment that is reported so far. For bentonite material, the movements of the tracer were calculated using known information such as the results of batch sorption tests and migration mechanism. Comparing the result of field test and calculations, it is suggested that tracer migration behavior in bentonite material in field can be evaluated quantitatively by the known migration mechanism and the results of laboratory experiments such as batch sorption test.
Evaluation of radionuclide migration in geologic media is of great importance in safety assessment for shallow land disposal specially for TRU nuclides because of inadequate data based on field test. As a cooperative research between JAERI and CIRP, a field test of radionuclide migration was conducted under natural condition using 90Sr, 237Np and 238Pu to obtain migration data of the radionuclides in actual aerated layer under natural rainfall condition. Values of input parameters to an existing evaluation equation of nuclide migration were determined on the basis of the results of laboratory experiments of batch and column methods and field investigation. Migration distribution of the radionuclides calculated with the determined values showed reasonable agreement with the measured distribution of the field test. This confirmed an applicability of the evaluation equation for radionuclide migration to actual aerated layer under natural condition.
Migration data of 90Sr, 237Np and 238Pu in natural aquifer were collected by a field test, which was performed in an aquifer at 30m under ground surface of the field test site. Migration parameters for analysis of the results obtained from the field test were measured by laboratory column tests and batch tests. The large retardation effects of the soil for the radionuclides were confirmed in the natural aquifer. Dispersion coefficient corresponding to water velocity was determined from the relationship between water velocity and dispersivity, which obtained from the column tests. Distribution coefficient was determined by considering reliability of data, test conditions, and environmental conditions. One dimensional migration behavior of the radionuclides in aquifer, calculated by using the migration parameters obtained from the batch and column tests, agreed with the results obtained from the field test. It was confirmed that the migration behavior of α-nuclides could be evaluated by applying the conventional equation for evaluating the radionuclide migration and the migration parameters obtained from laboratory tests.
The parameters for radionuclide migration were obtained from field environmental investigations, artificial barrier tests, column tracer tests and aquifer tracer test carried out in CIRP site. The validation study of Safety Assessment Code System of Shallow Land Disposal (GSA-GCL) was performed through the comparison between experimental data and calculated results. Analytical results of ground water flow and radionuclide migration could explain experimental results reasonably by GSA-GCL.
Data obtained in reactor dosimetry, particularly from reactors exclusively serving for irradiation, can today adequately treated making use of the MCNP or other code of the Monte Carlo method, although the need to prepare the benchmark problem to validate the data for Monte Carlo treatment will limit the range of experiments. The present paper discusses the usefulness and effectiveness of the Monte Carlo method, citing typical problem treated by this means in the materials testing reactor JMTR, the experimental fast reactor "JOYO" and PWRs/BWRs in general.