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[en] RTP cooling system was designed to provide sufficient cooling to the reactor core. Heat generated in the reactor core were carried by the primary cooling system to the secondary cooling system through the heat exchangers as a transfer medium. The performance of primary and secondary cooling system in different operation condition; start-up, steady state and shutdown were characterized and evaluated. The characterization of operational parameter has been done at nominal power level in normal operation. The aims of this study are to evaluate the response of the primary and secondary cooling system in start-up, steady state and shutdown condition and to determine the nominal range of parameters at each operating condition. (author)
[en] During severe accidents in nuclear reactors, the core can melt and get relocated in the reactor cavity/core catcher. If there is water present in the cavity, the melt forms a heat generating debris bed. Ensuring further long term coolability of such heat generating debris beds is very much important in the context of safety with respect to coolability as well as failure of cooling, yielding re¬melting and subsequent attack of structures. To achieve long term coolability of the configuration, all evaporated water has to be replaced by water inflow due to natural forces. At the same time the produced steam must escape the porous structure driven by buoyancy forces. If the heat generation is too high and the coolant is unable to take out the heat, the bed reaches dryout condition where the temperature of the bed increases sharply. The heat flux corresponding to this is called as dryout heat flux (DHF). It is important to determine the diyout heat flux of the bed in order to assess the coolability of the beds.The heat transfer behaviour in debris bed is very complex and is influenced by many factors like the mean size of the particles, porosity, operating conditions such as water entry from the top or the bottom of the debris bed, water temperature, Magnitude of non-condensable gases generated during MCCI, Spatial distribution of the bed porosity etc. In the present work, experimental results an determination of dryout heat flux in a realistic debris bed with irregularly shaped particles mixed with stainless steel balls have been reported. Influence of pressure and bottom injection on dryout heat flux has also been presented. It is observed that dryout heat flux increases with pressure. It also increases substantially with a small inflow. (author)
[en] When a research reactor which has a characteristic of the core down flow is designed, some important components like pump are located at a lower height than the core is. It is because of siphon phenomenon. It happens through a pipe when the main pipe of the primary cooling system is ruptured. As coolant leaks from the reactor pool, the water level of the pool gets lower as much as the coolant leaks. Thus, a core is exposed to air and this can lead to dangerous situations. To prevent it, siphon breaker is developed. However, as it is difficult to predict the results, a siphon breaker simulation program(SBSP) was designed by Lee et al. In this study, by using the SBSP, a small scales siphon breaker was designed to verify the SBSP. Range of experiments included previous experimental range by Kang et al., and expended to improve the SBSP. The results of experiments follow the SBSP’s one except for the extrapolation range. As a result, the SBSP is a good estimate for designing general siphon breakers, but it requires the model improvement for satisfying the wider range. (author)
[en] Separation level of non-volatile radioactive substances was analyzed by theoretical method, and specially designed purification tower was tested. The results showed that the experimental result was unanimous of which was gained by theoretical method. So the theoretical method and calculation model were rational. The theoretical model was applied on the evaporation equipment in the liquid waste treatment TEU system of a nuclear power plant, and the results showed that 3 sieve plates with wire mesh demister was rational and it could satisfy the requirement of nuclear power plant with a high safety margin. (authors)
[en] Aerosol is the main carrier of radioactive fission products releasing from severe accident. Test method of aerosol migration mechanism affected by passive containment cooling system was established and carried out with a reference of the GRACE test in the European Joint Research Centre. The test results agree with both the GRACE results and theoretical calculations, thus confirming the reliability of the method. (authors)
[en] The test facility HYMIT (HYdrogen MItigation Test) focuses on hydrogen mitigation efficiency of ignitor and PAR under condition of spray, steam and iodine aerosols in the containment of light water reactors during severe accidents. It can be used for containment safety research under severe accident conditions too.
[en] Coolant channels of nuclear reactor systems may experience flow decreasing transient under accident conditions, especially during loss of coolant accident (LOCA). Experimental study on transient CHF in horizontal channels under low pressure and low flow conditions is carried out at the test facility built at IIT Bombay. The objective of the study is to generate CHF data, develop correlation to predict the CHF for flow decreasing transient and to compare the data with the steady state CHF values. CHF measurement under flow decay and steady state conditions were carried out in horizontal tube with water as the coolant. Significant reduction in the transient flow CHF values is observed as compared to the steady flow CHF data. The setup is modeled and the transient is analyzed using system thermal hydraulic code RELAP5 to understand the thermal hydraulic phenomena associated with the CHF under flow decreasing conditions. The analysis predicts the experimental data within 20% deviation. The effect of various parameters on CHF analysed are half flow decay time t1/2, test section diameter d, length L, initial mass flux GINI, and upstream flow restriction pressure drop (stiffness). The results of the analysis will be presented in the full manuscript. (author)
[en] Full text: Accurate estimation of the through-wall mechanical property profile (attenuation) in reactor pressure vessel (RPV) steels is the key technology to evaluate the structural integrity of the RPVs of light water reactors. Prediction of the amount of embrittlement of RPV steels due to neutron irradiation during plant operation is the most important part of the estimation, and huge amount of efforts have been devoted worldwide to develop accurate and reliable embrittlement trend curves (ETCs). On the other hand, consideration of the profile of initial mechanical property through the wall thickness of RPV steels is also very beneficial in evaluating the actual structural integrity of RPVs. In this talk, we will review the development of ETCs for the RPV steels in Japan, and demonstrate recent results on the research activity on through-wall attenuation. (author)
[en] In 2015, the Risk-Informed Safety Margin Characterization (RISMC) Pathway, as part of the DOE Light Water Reactor Sustainability (LWRS) Program, initiated a set of demonstration activities to focus the LWRS risk analysis research and development (R&D) activities onto industry issues. RISMC is working to develop and provide methodologies and tools to plant operators/owners to support plant decisions for risk-informed margins management. These decisions relate to improved economics, prioritized reliability, and sustained safety of current nuclear power plants. Goals of the RISMC pathway are to: • develop and demonstrate a risk-assessment method coupled to safety margin quantification. Nuclear plant owners and decision makers could then use the developed methods to help with their margin recovery strategy. • create an advanced RISMC Toolkit. This toolkit can help nuclear plant owners to represent safety margins more accurately by reducing conservatisms deemed to be excessive.
[en] Periodic sampling of the discharged seawater effluent from Madras Atomic Power Station (Kalpakkam, Tamil Nadu, India) was carried out during 2013–2017 to assess the residual chlorine and trihalomethanes content in the outfall discharge water. The variations in dissolved oxygen, temperature, and pH were correlated with the residual chlorine and trihalomethanes content in the discharged effluent. The difference in temperature (ΔT) between influent and effluent seawater samples ranged from 1.95 to 11.0 °C (6.47 ± 1.87). More than 95% of the ΔT values were within the guideline value of 7 °C. The discharge water was associated with a marginal reduction in dissolved oxygen and a marginal increase in conductivity values. The total residual chlorine content in the discharged seawater at outfall ranged from 0.06 to 0.42 (0.16 ± 0.08) mg/L, which was within the stipulated values of 0.5 mg/L. Trihalomethanes values ranged from 0.04 to 65.03 (13.06 ± 14.38) μg/L. In addition to bromoform as the major constituent, occurrence of significant amount chloroform of was occasionally observed in the discharge water.