The SIRIUS-F facility was designed and constructed for highly accurate simulation of channel, core-wide and regional instabilities of an ABWR. A real-time simulation is performed for the modal-point kinetics of reactor neutronics and fuel-rod conduction on the basis of a measured void fraction in a reactor core section of the facility. A noise analysis method was performed to calculate decay ratios from dominant poles of transfer function on the basis of the AR method by applying time series of a core inlet flow rate. By utilizing this method, one can estimate stability at any specific operating point online without assuming excess conservative conditions. Channel and regional stability experiments were conducted for a wide range of operating conditions including maximum power points along the minimum pump speed line and the natural circulation line of the ABWR. The decay ratios and the resonance frequencies are in good agreement with those from the design analysis code, ODYSY. The SIRIUS-F experimental results demonstrated stability characteristics such as a stabilizing effect of the power, and reviled a sufficiently large stability margin even under hypothetical conditions of power enlargement.
The reduced-moderation water reactor (RMWR) core adopts a hexagonal tight-lattice arrangement with about 1mm gap between adjacent fuel rods. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the water-vapor two-phase flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of the RMWR core using an advanced numerical simulation technology. As a part of this technology development, we are developing an advanced interface tracking method to improve the conservation of volume of fluid. The present paper describes a part of the development of the two-phase flow simulation code TPFIT. The vectorization and parallelization of TPFIT code was conducted to fit the large-scale simulations. And the modified TPFIT code was applied to large-scale water-vapor two-phase flow in tight lattice rod bundle. In the results, calculated void fraction distribution in horizontal plane agreed quantitatively with the measured one obtained from the advanced neutron radiography technique including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.
Control-rod withdrawal tests simulating reactivity insertion are carried out in the HTTR to verify the inherent safety features of HTGRs. This paper describes pre-test analysis method using a neural network (NN) to predict the changes of reactor power and reactivity. The inputs of the network are the changes of the central control rods position and other significant core parameters. The actual tests data, which were previously carried out in the HTTR, were used for leaming the model of the plant dynamics. After the learning, the network can predict the changes of reactor power and reactivity in the following tests. Furthermore, the following tests are carried out at higher initial power stage; therefore, it is needed to enhance the capability of extrapolation of the network model. In this study, we introduced new guidelines for enhancing the extrapolation capability of NN. The network model applied in this study was designed according to the guidelines. From the results with new test data which are carried out at higher power stage, it is shown that the network is able to predict the changes of reactor power and reactivity precisely and the guidelines are effective in enhancing the extrapolation capability of NN.
The fractional distillation characteristics of the materials used for the reactor pressure vessel made of ASTM A302B and the structures in reactor made of SUS304 which are the radioactive metallic waste of Japan Power Demonstration Reactor (JPDR) were analyzed numerically. In the simulation, the vaporization rates of the components of the waste were calculated by using the Langmuir's equation and Henry's law. As the result of simulation, it was calculated that 152Eu, 154Eu, 14C and 94Nb can be reduced to less than clearance level by the fractional distillation. On the ASTM A302B case, it was pointed out that the other radioactive nuclei which are 54Mn, 60Co, 59Ni and 63Ni satisfy clearance level after 77 years cooling down. On the SUS304 case, it was pointed out that 59Ni and 63Ni must be separated to satisfy clearance level using isotope separation.
A refinement of a simplified QAD-SKYSHINE calculation for gamma-rays skyshine emitted from a turbine building of nuclear power plant is performed by the hybrid method using the MCNP Monte Carlo code. In this method, a detailed Monte Carlo transport calculation is firstly used to determine the energy and emission angle of an crossing point of the gamma-rays through the turbine building wall into atmosphere. Then, the computationally efficient integral Line-Beam method is used to transport gamma-ray from this pseudosource to the distant location at which the skyshine dose is to be calculated. Where, the Line-Beam response functions are generated by the MCNP and fitted to the four-parameter formula. The present developed method is able to estimate precisely skyshine dose within short time.
High Temperature Engineering Test Reactor (HTTR) of high temperature gas-cooled reactor at Japan Atomic Energy Research Institute achieved the reactor outlet coolant temperature of 950°C for the first time in the world at Apr. 19, 2004. To remove generated heat at reactor core and to hold reactor inlet coolant temperature as specified temperature, heat exchangers in HTTR main cooling system should have designed heat exchange performance. In this report, heat exchanger performance is evaluated based on measurement data in high temperature test operation. And it is confirmed the adequacy of heat exchanger designing method by comparison of evaluated value with designed value.
As a part of the feasibility study on commercialized fast reactor cycle systems, Japan Nuclear Cycle Development Institute (JNC) has been developing the fuel decladding technology for the dry reprocessing process (oxide electrowinning process) and aqueous reprocessing process. Particularly, in the oxide electrowinning process, the spent fuel should be reduced to powder for quick dissolution in the molten salt at electrolyzer. Therefore, JNC proposes new decladding system with innovative mechanical decradding devices. The decladding system consists of fuel crushing stage, hull separation stage and hull rinsing stage. In the fuel crushing stage, disassembled spent fuel pins are crushed and powdered by mechanical decladding device, then the following stage. the hull and the fuel powder are separated by magnetic separator. Only the fuel powder is fed to the electrolyzer. On the other side, the separated hull is melted by induction heating method, and the small amount of oxide included in the hull fragments is recovered at the hull rinsing stage. The recovered oxide fuel is fed back to the electrolyzer. In this paper, the basic performance of the element equipment that composes this new decladding system will be descried.
Noriyosu HAYASHIZAKI, Minoru TAKAHASHI, Takafumi AOYAMA and Shoji ONOSE Nuclear engineering experiments using outside facilities of the campus have been offered for graduate students in the nuclear engineering course in Tokyo Institute of Technology (Tokyo Tech.). The experiments are managed with the collaboration of Japan Nuclear Cycle Development Institute (JNC), Japan Atomic Energy Research Institute (JAERI) and Research Reactor Institute, Kyoto University (KUR). This report presents the new curriculum of the nuclear engineering experiments at JNC since 2002. The change is due to the shutdown of Deuterium Criticality Assembly Facility (DCA) that was used as an experimental facility until 2001. Reactor physics experiment using the training simulator of the experimental fast reactor JOYO is continued from the previous curriculum with the addition of the criticality approach experiment and control rods calibration. A new experimental subject is an irradiated material experiment at the Material Monitoring Facility (MMF). As a result, both are acceptable as the student experiments on the fast reactor.
Typical weight estimation methods with Monte Carlo method such as MCNP default, empirical formula, monoenergy neutron attenuation curve, MCNP Weight Window Generator and adjoint flux are described. The mono-energy neutron attenuation curve method is proposed by authors. Weights estimated by methods by empirical formula and monoenergy neutron attenuation curve are compared with those calculated by MCNP WWG method.