[en] In the severe accidents of nuclear power plants, hydrogen can be generated and in the reactor cavity by molten core - concrete interactions (MCCI). The cavity (CAV) package model of MELCOR code has several options on compositions and thermal properties of concrete in order to estimate the impacts of MCCI. Concretes types and its thermal properties are summarized well in the VTT research report. According to the different characteristics they give different amounts of hydrogen generation when they interact with molten core materials ejected from the reactor vessel. The amount of hydrogen generation and thermal responses in cavity are compared for the seven different kinds of concrete compositions for the 20 inches LOCA case of APR1400 as an example case

[en] After hydrogen explosion accident at Fukushima Daiichi nuclear power plant site occurred in March 11, 2011 in Japan, the hydrogen explosion accident in nuclear power plants has focusing during last one year in worldwide. Design features of APR1400 against severe accident and modeling procedure of MELCOR version 1.8.6 input deck is explained first. Core melt progression scenarios, especially hydrogen generation mechanisms by metal water reaction, in core are analyzed and compared in the point of the amount of hydrogen generation and reactor vessel failure timings for typical severe accident sequences such as LOCAs and SBO scenarios. Zirconium and stainless steel oxidation process is identified in APR1400 reactor

[en] In the Fukushima Accident occurred in March 11, 2011, the containment performance is very important in the long-term station blackout (SBO) situations. For the APR1400, containment performance analyses are performed for the hypothetical long-term SBO situations

[en] In the severe accidents of nuclear power plants, hydrogen can be generated in the reactor pressure vessel (RPV) by the chemical reaction of the metals with water called metal-water reaction (MWR). The amount of metals such as zirconium (Zr) and stainless steel (SS) in RPV and of water available which depends on the accident sequence will affect the hydrogen generation along the accident progression. The initial mass of structural materials including Zr and SS in RPV are known to be one of the most important sources of uncertainties in the input parameters. In this paper, the MELCOR code version 1.8.6 is used to estimate how much hydrogen will be generated for the various accident scenarios. A sensitivity analysis has been performed to study how the initial mass of SS in lower plenum affects the hydrogen generation in RPV of APR1400 for typical severe accident sequences such as loss of coolant accident (LOCA) and station blackout (SBO)

[en] This paper performed research on the use of Nuclear Economics Support Tool (NEST) to gain a basic understanding on the algorithm, necessary input parameters, and output parameters of this tool. Then, the application of NEST version 4 for a case study to calculate the economic parameters for a 1000MW(e) PWR nuclear reactor operating with a once-through fuel cycle 1 was also carried out. The purpose is to initially build the capacity in the economic assessment of the nuclear energy system. The calculated parameters including levelized unit energy cost (LUEC), internal rate of return (IRR), Return on Investment (ROI), Net present value (NPV), and total investment were compared with those in IAEA No.NG-T-4.4. The comparative results showed that the calculated results are in good agreement with IAEA results. The sensitivity evaluations were also performed to identify the most sensitive parameters that influent the LUEC. The sensitivity analysis results showed that the LUEC is highly dependent on the discount rate, the construction cost, the reactor lifetime, and the capital investment schedule. However, LUEC is almost insensitive to the cost of natural uranium. The analysis also shows that the discount rate and construction cost should be kept to a minimum and the reactor lifetime should be increased to a maximum to minimize the value of LUEC, then enhance the competitiveness of NES. In addition, the distribution of capital investment during the construction period should be carefully considered because it influences the electricity cost considerably. (author)

[en] One of the key objectives of severe accident management is to maintain the integrity of the reactor pressure vessel (RPV) during a severe accident. The timing of an RPV lower head failure has a controlling effect on the subsequent accident progression and is an important aspect for operator recovery actions. In this paper, a sensitivity study on a lower head failure was performed to obtain insight into the operator recovery actions. Analyses on a severe accident progression were performed for a station blackout (SBO) and small break LOCA (SBLOCA), initiating severe accidents, using MELCOR 1.8.6. Operator recovery actions for decay heat removal and inventory make up using a depressurization system and safety injection pump were simulated in parallel with a simulation of the accident progression. Sensitivity studies on the timing of operator actions and variations of heat transfer coefficients from the debris to the penetrations and to the lower head were performed to investigate the changes in the timing of the vessel failure. The results show that lower head failure time is sensitive to the heat transfer from debris to the penetration and to the lower head. The change in the timing of the lower head failure was more prominent in the case of an SBLOCA initiated accident in comparison with SBO. It was also found that the formation of a molten pool plays an important role in the failure of a lower head when heat transfer from the debris to the penetration and lower head is very low. (authors)

[en] Since the crisis at the Fukushima plants, the severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. A station blackout (SBO) scenario for an APR1400 nuclear power plant is simulated using the MELCOR computer code. A reactor coolant system (RCS) depressurization by a safety depressurization system (SDS) for water to be injected into the reactor pressure vessel (RPV) is performed as a mitigation action. The purpose of this study is to investigate the effect of the SDS actuation timing on the accident progression and determine the optimum depressurization strategy to prevent core damage and a reactor vessel failure. SBO without SDS actuation is analyzed as a base case to understand the main phenomena during SBO accident. In base case, the RPV lower head will fail after 4.2 hours since SBO occurs. SBO with SDS actuation is performed by changing the SDS actuation timing to inject water into RPV. The results show that the RPV lower head failure can be prevented if SDS is opened no later than 3.5 hour. Sensitivity study on some main parameters in MELCOR is also performed to see the effect of these parameters on the failure time of the lower head in an SBO accident. (author)

[en] A small break LOCA (SB LOCA) with a break size equivalent to a 3 inch cold leg break of the APR1400 nuclear power plant is analyzed using MELCOR computer code. The objective of this study is to estimate the effectiveness of alternate injection on the recovery action during a severe accident initiated from a loss of coolant (LOCA) accident. By changing the injection timing of safety injection pump (SIP) in to Reactor Coolant System (RCS), sensitivity studies are performed on the accident progression such as the time of core damage, amount of hydrogen production, time when core relocation starts, the time of reactor vessel failure. From the analyses results, the optimum time to actuate the SIP are estimated. (authors)

[en] A LBLOCA scenario for an advanced pressurized water reactor, call APR1400, developed in Korea is analyzed in order to obtain an overall insight into a severe accident progression from an initiating event to the reactor vessel failure in detail by using the MELCOR computer code Versions 1.8.5 and 1.8.6. The present results (the amount of molten corium and vessel failure timing) would be used as input for the establishment of severe accident management strategies or for the design of a core catcher for the APR1400. The MELCOR results showed that the lower head instrumentation tube penetration failure model and internal structure in the reactor vessel had influence on the amount of corium ejected and the timing of reactor vessel failure