In order to analyze large amounts of trouble information of overseas nuclear power plants, it is necessary to select information that is significant in terms of both safety and reliability. In this research, a method of efficiently and simply classifying degrees of importance of components in terms of safety and reliability while paying attention to root-cause components appearing in the information was developed. Regarding safety, the reactor core damage frequency (CDF), which is used in the probabilistic analysis of a reactor, was used. Regarding reliability, the automatic plant trip probability (APTP), which is used in the probabilistic analysis of automatic reactor trips, was used. These two aspects were reflected in the development of criteria for classifying degrees of importance of components. By applying these criteria, a method of quantitatively and simply judging the significance of trouble information of overseas nuclear power plants was developed.
The hydrogen moderator of 1MW pulse spallation neutron source (JSNS) in J-PARC must be designed to get effective coolability with liquid hydrogen under such sever conditions as compact vessel size and high heat load deposited with high neutron flux irradiation. To solve this problem and to clarify the heat transfer characteristics of the impinging jet in a limited narrow space, heat transfer experiments have been carried out using an actual-scale simulation test vessel under water flowing condition. And, an evaluation method of the heat transfer coefficient from a heat conduction equation was established. From the experimental results and the evaluated heat transfer coefficients, the heat transfer characteristics of the bottom surface of moderator vessel were clarified. The operating flowing condition of the hydrogen moderator was determined from the predicted heat transfer curve for hydrogen based on those results.
Application of latent heat storage technology using phase change materials (PCM) to reduce temperature changes in heat transfer fluid was studied. Temperature characteristics of heat transfer fluid flowing in a copper pipe surrounded by pure PCM or composite PCM where high porosity porous metals were saturated by the PCM were investigated by experiment and analyses. Experiment was performed by using air as a heat transfer fluid, octadecane as a PCM and copper and nickel foams as porous metals. Experimental results were compared with numerical and approximate analyses. The results show that composite PCM with increased effective thermal conductivity augmented temperature change reduction of the heat transfer fluid. The highest reduction rate was performed by the copper foam, the second and the third were the nickel foams with fine and large pore sizes respectively. Reduction rate is increased with the increase in Reynolds number. Numerical results agreed comparatively well with the experimental results, especially agreement was well for case of the composite PCM. This implies natural convection in liquid PCM is suppressed by inserting the metal foams. Use of latent heat storage technology at higher Reynolds number region will be effective for the reduction of temperature changes in heat transfer fluid.
Seismic probabilistic safety assessment (PSA) is an available method to evaluate residual risks of nuclear plants that are designed on definitive seismic conditions. From our preliminary seismic PSA analysis, horizontal shaft pumps are important components that have significant influences on the core damage frequency (CDF). An actual horizontal shaft pump and some kinds of elements were tested to evaluate realistic fragility capacities. Our test results showed that the realistic fragility capacity of horizontal shaft pump would be at least four times as high as a current value, 1.6×9.8m/ss2, used for our seismic PSA. We are going to incorporate the fragility capacity data that were obtained from those tests into our seismic PSA analysis, and we expect that the reliability of seismic PSA should increase.
Analytical equipment, which consists of a microchemical chip and a desktop-sized thermal lens microscope (DTTLM), is being developed to analyze solutions in PUREX reprocessing of spent nuclear fuels. Radiation degradation by gamma-rays of the microchemical chip and capillary tubes used in this equipment were studied. The decreased thermal lens signal of a colored microchemical chip made of Pyrex (Corning#7740) glass by the irradiation can be corrected by using empirical correlations of the light transmittance. The usable dose of the EXLON PFA capillary tube was less than 30kGy. The microchemical chip made of Pyrex (Corning#7740) glass and the EXLON PFA capillary tubes can be applied to the analyses of high radioactivity samples since the sample quantity required for analysis is very small. Radiation degradation of the microchemical chip made of synthetic quartz (SUPRASIL-P) and the VICTREX PEEK capillary tubes was not observed for the dose studied here.
A new high-speed multiple pulse time data registration, processing and real-time display system for time interval analysis (TIA) was developed for counting either β-α or α-α correlated decay-events, The TIA method has been so far limited to selective extraction of successive α-α decay events within the milli-second time scale owing to the use of original electronic hardware. In the present pulse-processing system, three different high-speed α/β(γ) pulses could be fed quickly to original 32 bit PCI board (ZN-HTS2) within 1 μs. This original PCI board is consisting of a timing-control IC (HTS-A) and 28 bit counting IC (HTS-B). All channel and pulse time data were stored to FIFO RAM, followed to transfer into temporary CPU RAM (32MB) by DMA. Both data registration (into main RAM (200MB)) and calculation of pulse time intervals together with real-time TIA-distribution display simultaneously processed using two sophisticate softwares. The present system has proven to succeed for the real-time display of TIA distribution spectrum even when 1.6×105 cps pulses from pulse generator were given to the system. By using this new system combined with liquid scintillation counting (LSC) apparatus, both a natural micro-second order β-α correlated decay-events and a millii-second order α-α correlated decay-event could be selectively extracted from the mixture of natural radionuclides.
This paper describes a method of estimating source term, i.e., location, period and amount of atmospheric release of radioactive material in real-time during nuclear emergency. This method consists of: (1) trial simulations of atmospheric dispersions on the possible combinations of these parameters and (2) statistical comparison of model predictions with offsite measurements of air concentrations of radionuclides and/or air dose rates from monitoring stations, to find a set of release condition providing model prediction that fits best to the measurement. A parallel execution method for efficiently processing many possible initial conditions is also developed. The performance of this method is favorably evaluated by a verification study using the dataset from European Tracer Experiment.
It was shown from the annular core critical experiment of the HTTR that its excess reactivity values tend to be overestimated about 3%Δk/k by the HTTR core design method. Therefore, assessment of the calculation model for the annular core of the HTTR was systematically undertaken. The revised model was investigated which was based on the SRAC code system. It was concluded that the multiplication factor of the annular core is considerably affected by the three factors listed below: (1) The Benoist's anisotropic diffusion coefficients (2) The mesh interval in the whole core diffusion calculation (3) The mesh structure of graphite region in fuel lattice cells. The significantly large discrepancy previously reported was reduced down to 1.6%Δk/k by the revised annular core model.