The purpose of this study is to develop a steam separator for BWR with low pressure loss by CFD and a scale model test. The swirler is the key component for the separation performance and pressure loss. Therefore, we have focused on the improvement of the swirler and tried to improve the swirler geometry to reduce the swirler pressure loss while maintaining the separation performance. This study consists of two stages. In the first stage, we measure pressure loss and water discharge ratio using the 1/2.22-scale test under BWR operating pressure, and then validate the CFD method by comparison with the scale test results. In the second stage, we improve the swirler geometry by the survey of some geometry factors using CFD. We predict that the inlet and outlet pressure losses of the swirler are dominant in the separator. By improving the geometry at the inlet and outlet of the swirler, we show from the results of CFD and a scale test that the improved swirler can reduce pressure loss by about 50% compared with the conventional swirler.
The purpose of this study is to develop a new BWR steam separator with a lower pressure loss of more than 20% than the current steam separator while maintaining the steam separation performance. Reducing the pressure loss of steam separators is advantageous in that it reduces the required recirculation pump power and hence increases the net electrical power output. In previous studies, a low-pressure-loss steam separator was developed by three-dimensional two-phase flow simulations verified by 1/2.22-scale model experiment. In this study, performance tests for a full-scale steam separator with an improved swirler were conducted under actual reactor operating conditions of 7.17 MPa and 287°C. The obtained results demonstrate that pressure loss decreases by about 25% and the separation performance is similar to that of the current steam separator under rated reactor conditions.
A number of probabilistic safety assessment (PSA) studies have been conducted to optimize maintenance activities in nuclear power plants from the viewpoint of safety focusing on the risk of core meltdown. However, even a small-scale incident of component, which never causes the core meltdown, resulted in reactor shutdown and economic losses. Accordingly, in addition to the safety analysis focusing on the risk of core meltdown, it is very useful to develop a simulator that can establish maintenance strategies in terms of availability and economic efficiency of nuclear power plants. We have developed an integrated simulator for the maintenance optimization of light water reactors (LWRs). The concept of the simulator is to provide a method for optimizing maintenance activities for representative components and piping systems comprehensively and quantitatively in terms of safety, availability, and economic efficiency (both from cost and profit) under various maintenance strategies including altering inspection frequency and inspection accuracy, conducting sampling inspection, repairs and/or replacements, introducing various maintenance rules, and long-term fuel cycles. In addition, a function of visualization of the simulated results by a divided multidimensional visualization method has also been developed in order to support a decision-making process for optimizing the maintenance activities.
A reduced-moderation water reactor (RMWR) is an innovative light water reactor comprising tight-lattice fuel assemblies with a gap clearance of around 1.0 mm for a reduction in water volume ratio to achieve a high conversion ratio. By taking into account recent experimental findings from 37-rod tight-lattice bundle thermal-hydraulic tests indicating that the critical power tends to be higher for a peripheral peak local power distribution than for a flat power distribution, the viability of fuel assembly designs with fewer types of plutonium enrichment of MOX fuel, which may result in a large local peaking factor in peripheral rods, was assessed in this report. Critical powers of 217-rod bundles with peripheral peaks for upper and lower MOX regions of the double-flat core of the RMWR were calculated by a subchannel analysis code NASCA. Peripheral peaking with the corresponding local peaking factor for a uniform plutonium enrichment design yields almost the same critical power as that for a flat power distribution. Thus, a reduction in fuel fabrication burden may be possible by decreasing the number of types of plutonium fuel enrichment while maintaining the same thermal-hydraulic margin as the fuel assembly design with five enrichment types of MOX fuel.
Toward resolving nuclear conflict, deep dialogue sessions among stakeholders having different opinions concerning nuclear technology are strongly required. In order to realize a fair and constructive dialogue session, it is necessary to pay attention to the design of the dialogue session, e.g., participant selection, agenda setting, facilitation, and rules of dialogue. The basic requirements for a dialogue session have been proposed in this study based on theoretical and empirical analyses of previous dialogue sessions on nuclear issues. The theoretical analysis has been performed based on various aspects concerning defects in nuclear communication mentioned in science and technology studies. The empirical analysis has been performed by analyzing participants' responses and by systematizing practical findings of previous dialogues. The proposed requirements for a dialogue session were utilized for the design and operation of a preliminary attempt of a dialogue session named “Open Forum for Nuclear Communication.” Through the analysis of data, such as minutes and recorded conversations collected after the session, it has been confirmed that the prespecified requirements have been satisfied. In addition, the participants' responses have shown a high acceptance for an open forum. According to these results, the basic validity and effectiveness of the proposed requirements in the design of a dialogue session have been successfully demonstrated.
Electron backscatter diffraction (EBSD) measurements and an indentation test were applied to measure the plastic strain induced by tensile and fatigue tests using Type 316 stainless steel. It was shown that the local misorientation obtained by EBSD measurements correlated well with the degree of plastic strain. Furthermore, by investigating the statistical distribution of the local misorientation averaged for each grain, it was possible to distinguish the source of strain (cyclic or uniform strain). By the indentation test, it was shown that the hardness could be a parameter for plastic strain estimation, the accuracy of which was about ±1.5%. A detailed investigation using the nanoindentation method revealed that the variation in hardness was attributed to microstructural inhomogeneity, which was evinced by the local misorientation. Also, the estimation of the standard deviation of hardness allowed the source of strain to be identified. It was concluded that the indentation test can be used for estimating the degree of plastic strain induced by seismic loading in nuclear power plant components.
It is important to control the chemistry of the helium coolant used in high-temperature gas-cooled reactors (HTGRs). The effect of a decarburizing environment on the creep rupture properties tends to decrease the creep rupture life of the heat-resistant alloy used in heat exchangers. In this paper, we describe an active control method for the concentration of impurities using the existing helium purification system, which consists of a helium heater, a copper oxide trap (CuOT), a molecular sieve trap, a cold charcoal trap, and a bypass line. Analysis showed that the efficiency control of CuOT is effective in improving the decarburizing atmosphere. The efficiency control of CuOT increases the concentrations of carbon monoxide and hydrogen. It was found that both the enrichment of carbon monoxide suggested in previous studies and the enrichment of hydrogen are also effective in forming the carburizing atmosphere.
To confirm the sealing performance of a metal cask subjected to impact force due to commercial aircraft crash against a spent fuel storage facility, a vertical impact test was carried out. In this test, a simplified deformable missile was used by considering the rigidity of the actual aircraft engine and accelerated to the specified impact velocity (60 m/s) to hit the full-scale lid structure with the primary and secondary lids. Then, the leak rate, the inner pressure between the lids, and the displacement of the lids were measured. The leak rate of the secondary lid exceeded 1.0×10−3 Pa·m3/s upon impact. However, because no residual lid opening displacement occurred after loading, the leak rate recovered to less than 1.0×10−6 Pa·m3/s after 3 h from the impact test. In addition, to clarify the impact behaviour of the lid structure, the impact analysis using the LS-DYNA code was executed. It was found that the lid bolts maintained the good tightening force after impact loading, and the sealing performance of the full-scale metal cask would not be affected immediately by the vertical impact of the aircraft engine with a speed of 60 m/s.
To release materials with considerably low radioactive concentrations arising from decommissioning activities of nuclear installations, it is required to confirm that the sum of “the ratio of radioactive concentration (D)”/“clearance level (C)” is lower than or equal to the reference value by the reliable measurement and evaluation of radioactive concentration. When the radioactive concentration is evaluated by using a statistical method where samples are taken from the materials, the following two points should be taken into account: (1) a conservative evaluation that prevents underestimation caused by statistical uncertainties and (2) an error that may lead to a wrong decision where the materials are not released due to a highly conservative evaluation. In this paper, we propose a method to determine the number of samples required for clearance verification based on a statistical theory in a consistent manner, where uncertainties in the sum of D/C are taken into account.
In order to develop an innovative fuel fabrication method for americium-containing oxide fuels, a feasibility study of metallic U- and Mo-doped oxide pellet fuels with extruding granulated oxides was conducted using UO2. In this study, the hot-press sintering method was adopted as a fuel pelletizing process. To investigate the thermochemical characteristics of the pellets, a hot-press sintering test of the U- and Mo-doped UO2 powder samples was carried out. The thermal conductivity of the sintered material was evaluated. Results showed that the U and Mo doping processes reduce oxygen potential and improve thermal conductivity, respectively. It is, therefore, concluded that the U- and Mo-doped oxide fuel pellets can be successfully fabricated using the hot-press method. Also in this study, to investigate the sinterability of the extruding granulated oxides, a hot-press sintering test of the U- and Mo-doped UO2 granules formed by the extruding granulation method was carried out. Results revealed that the granulated substances formed by the extruding granulation method have preferable sintering characteristics.
The helium/helium heat exchanger (i.e., intermediate heat exchanger: IHX) of a high-temperature gas-cooled reactor (HTGR) system with nuclear heat applications is installed between a primary system and a secondary system. IHX is operated at the highest temperature of 950°C and has a high capacity of up to 600 MWt. A plate-fin-type heat exchanger is the most suitable for IHX to improve construction cost. The purpose of this study is to develop an ultrafine plate-fin-type heat exchanger with a finer pitch fin than a conventional technology. In the first step, fabrication conditions of the ultrafine plate fin were optimized by press tests. In the second step, a brazing material was selected from several candidates through brazing tests of rods, and brazing conditions were optimized for plate-fin structures. In the third step, tensile strength, creep rupture, fatigue, and creep-fatigue tests were performed as typical strength tests for plate-fin structures. The obtained data were compared with those of the base metal and plate-fin element fabricated from SUS316. Finally, the accuracy of the creep-fatigue life prediction using both the linear cumulative damage rule and the equivalent homogeneous solid method was confirmed through the evaluation of creep-fatigue test results of plate-fin structures.