A severe accident at TEPCO’s Fukushima Dai-ichi Nuclear Power Plant (FDNPP) occurred due to a strong earthquake and massive tsunami that struck Japan’s eastern region on 11 March 2011. During this accident, the reactor core of Units 1, 2, and 3 of the 1F melted and fuel debris formed in the reactor pressure vessel (RPV) and primary containment vessel (PCV). The Japanese Government has drawn a mid- and long-term roadmap towards 1F decommissioning. However, three reactors’ decommissioning work has faced difficulties due to the lack of realistic information about the damaged cores such as the distribution of fuel debris. Since the radiation levels inside the reactor buildings have been too high for human access, the robots and cameras have been using to investigate the internal status of PCV. To accelerate the decommissioning process of 1F, determining of fuel debris information and contaminated water leakage is crucial to decide further decommissioning steps and strategies. Therefore, in this study, we are developing an ultrasonic measurement system that can obtain object’s shape and the velocity around it. Because ultrasonic transducers are robust against high radiation dose and applicable in opaque liquids. The ultrasonic system used in the present study consists of two sectorial array transducers, which can measure object shape and 2D vector flow field. To measure the object’s velocity and shape that represents fuel debris, Ultrasonic Velocity Profiler (UVP) and Total Focusing Methods (TFM) are employed, respectively. Based on measurement results of the 2D vector field and shape of the object, we can predict contaminated water leakage and distribution of object at the bottom of the PCV in 1F.

In the foreseen decommissioning and debris removal plans of the damaged Fukushima Daiichi reactors, the fuel debris will be broken into small pieces using laser or mechanical cutting techniques prior to removing them from the reactor buildings. Regardless of the technique to be employed, submicron radioactive aerosol particles will be generated and dispersed in the gas space of primary containment vessel during cutting operations. The water spray system has been proven to be an applicable method in removing aerosol particles. However, it cannot remove Greenfield-gap aerosol particles (with diameters between 0.1–1 μm) so effectively. To solve this problem, a new agglomeration method by addition of water mist before spray injection was developed. With preexisting water mist, aerosol particles were expected to aggregate with water mist and form larger-sized agglomerated aerosol-mist particles, which increased the effect of inertial impaction mechanism leading to higher scavenging efficiency. The new method has been verified to be capable of improving the spray scavenging efficiency for the Greenfield gap particles by conducting aerosol scavenging experiments without and with mist in the newly built UTARTS facility in the University of Tokyo. The experiment results showed that the aerosol removal rate increased along with the increasing of mist concentration level. To verify the new agglomeration method in different experiment facilities and to investigate the effects of vessel’s size scale on aerosol collection efficiency, similar experiments were repeated in the TOSQAN facility of IRSN, France. Though the cylindrical vessel in two facilities have same internal diameter, the vessel’s height of TOSQAN facility is 4.8 m, which is larger than the one in the UTARTS facility (2.5 m). The experiment results in TOSQAN facility also showed that water mist has potential to improve aerosol spray scavenging efficiency. The corresponding numerical simulations about aerosol removal by spray droplets without mist in both UTARTS and TOSQAN facilities were conducted to better understand the aerosol removal process, including time evolution of aerosol mass fraction and flow field of the gas phase.

In January 2012, Toshiba Energy Systems & Solutions Corporation inserted a camera into the damaged Primary Containment Vessel (PCV) of Fukushima Daiichi Nuclear Power Station unit 2 for the first time since the Great East Japan Earthquake. Since then, we, as a member of International Research Institute for Nuclear Decommissioning (IRID), has developed several devices and has investigated inside of PCV. The investigation devices for unit 2 are roughly categorized into three types. For the investigations in early stages, just a camera and sensors with simple support tools were inserted through a small hole. Then small crawler type robots were developed to expand the investigation area to inside of Pedestal which exists the central part of PCV. Since the devices found that there were not enough areas to run around by the robots in the Pedestal, a long telescopic guiding pipe device was chosen to deliver cameras, sensors, and tools into the Pedestal. This type of device took movie of wide area of inside of the Pedestal. An image processing software was also developed to get a high quality and extensive picture from the movie taken by a low resolution camera and insufficient lighting. In February 2019, the other telescopic guiding pipe device finally delivered a camera and fingers like tong to grab deposits fallen on the bottom of pedestal for the first time after the earthquake, and found that some of deposits can be lift up. Some of projects were implemented as subsidized projects of “Decommissioning and Contaminated Water Management” founded by the Ministry of Economy, Trade and Industry.

There are a number of applications in nuclear energy and hazardous waste disposal that require monitoring of fluids under extreme environments, including high levels of temperature, pressure, toxicity and radioactivity. Many of these applications will benefit from a monitoring technique that is external and non-invasive. Currently the sensors used are invasive, must reside inside the pressurized vessels and must penetrate the vessel walls, which can create a weakness in the vessel. Additionally, instruments that are used inside such containers must be exceptionally hardened to the environment. Information Systems Laboratories (ISL) has developed an external mass (gravimetric) measuring technique for monitoring nuclear coolant in Small Modular Reactors (SMRs), which will also work for measuring fluid levels in waste tanks, that avoids the problems inherent in invasive sensors. It utilizes a COTS gravitational sensor of unprecedented accuracy, leveraged via proper sensor placement geometry, to detect fluid changes of small amplitude from an outside position, obviating the need to penetrate the vessel. The technique is called Gravisense™. ISL has proven via simulation and experiment that this concept can be usefully applied to monitoring fluid levels in both nuclear reactors and large waste tanks. Numerical simulation algorithms were developed to calculate the gravity effect of small changes in water level, which were verified by experiments at the NIST Physical Simulator facility at the Oregon State University. The measured ultra-low noise levels of the superconducting gravimeter type which utilizes a Niobium sphere suspended in a magnetic field to attain its phenomenal accuracy, demonstrated that fluid levels in SMRs can be measured at least to within 3 cm. Furthermore, the method can distinguish between a contained leak (from reactor to containment vessel) from an external leak (from reactor to outside of containment). Additionally, simulations of waste canisters that hold spent fuel rods show that the fluid level measuring accuracy can potentially do better than 1 cm accuracy by measuring from below the vessel, and judicious placement of sensors on top of large waste tanks can potentially achieve a very impressive 2 mm measurement accuracy. These encouraging results prove that the Gravisense™ technique for fluid determination can be very useful in nuclear energy generation, testing, and research, as well as in waste monitoring situations that are difficult to monitor via traditional sensing technology. We believe that the next step should be to test the technique on canisters of the type that are currently storing waste in various DOE locations.

From the Fukushima accident, Thailand has studied fission product behavior in containment vessel using ART Mod 2 code. Cesium iodide (CsI), cesium hydroxide (CsOH) and iodine (I 2 ) behaviors are studied using modified ART Mod 2 code. However, there are other compounds which are not included in the codes especially cesium and iodine compounds such as from Phébus FPT3 experiment including cesium molybdate (Cs 2 MoO 4 ), cesium telluride (Cs 2 Te), methyl iodide (CH 3 I) and iodine pentoxide (I 2 O 5 ). The paper objective is to add the four compounds in the codes in order to enlarge the coverage of the code in evaluation fission product behavior in the containment vessel. Physical parameters and models of the four compounds are updated in the codes. It is found that deposition phenomena of Cs 2 MoO 4 , Cs 2 Te CH 3 I and I 2 O 5 are close to the experiment in case of no chemical reaction.

Sodium-concrete reaction (SCR) is one of the important phenomena during severe accidents in sodium-cooled fast reactors (SFRs) owing to the generation of large sources of hydrogen and aerosols in the containment vessel [1–3]. In this study, SCR experiments were performed to investigate the chemical reaction beneath the internal heater (800 °C), which was used to simulate the obstacle and heating effect for SCR. Furthermore, the effects of the internal heater on the self-termination mechanism were discussed. The internal heater on the concrete seemed to hinder the transport of Na into the concrete. Therefore, Na could react with the concrete at the periphery of the internal heater. The concrete ablation depth at the periphery was larger than under the internal heater. However, the Na concentration around the reaction front was about 30 wt.% despite the position of the internal heater. The Na concentration was similar to that of Na 2 SiO 3 , which was almost same as that in our past study [4–5]. It was found that the Na concentration condition was one of the dominant parameters for the self-termination of SCR, even in the presence of the internal heater.

A large amount of hydrogen is generated by the metal water-reaction in the Primary containment vessel (PCV) of light water reactors in the severe accident (SA). In the present accident management for boiling water reactor (BWR), vent of mixing gas with filtered vent is regarded as the most likely method that prevents the PCV overpressure. However, it is difficult to vent in early stage of SA because of high radioactive dose. Then we have been developing the hydrogen treatment system to prevent excessive pressure without PCV vent. In focusing on the oxidation-reduction reaction of metal oxides (MOs) with high reaction rate, we have been studying hydrogen treatment system using MOs as effective device under oxygen deficit conditions like PCV of BWR. In the previous studies, we evaluated the hydrogen treatment rate using a couple of MOs, and confirmed that CuO, Co 3 O 4 , and MnO 2 were effective for the hydrogen oxidation under the oxygen-free condition. We also found that granules of these three MOs could achieve the goal of hydrogen treatment rate with reactor of hydrogen treatment system. We predicted that the performance of MOs decreased with exposure to steam and fission products (FPs) in the PCV during the hydrogen treatment, and investigated their influence. The objective of the present research is to investigate how the steam and FPs, which is supposed to be a reaction-inhibiting-factor, influence hydrogen treatment rate. Then, we conducted hydrogen treatment experiments using a fixed bed reactor with MOs layer. As the results, we confirmed that the hydrogen treatment rate of MnO 2 decrease from 70 g/s/m 3 to 15 g/s/m 3 when partial pressure of vapor went above 0.1 MPa-abs, though, that of CuO didn’t depend on the partial pressure of vapor and sustain the same rate about 40 g/s/m 3 . We also confirmed that the hydrogen treatment rate was decreased with the consumption of granulated MOs faster than our expectation estimated with unreacted-core model*. We also estimated that CsI selected as typical FPs could not affect the hydrogen treatment rate of CuO. From these results, we have evaluated the reaction rate equation including the steam influence in CuO, which could estimate the hydrogen treatment rate of reactor unit. *Gas reacts only on the surface of solid and generates shell of products around reactants core. The core shrinks with reaction.

A large amount of hydrogen is generated by the metal water-reaction in the Primary Containment Vessel (PCV) of Light Water Reactors in the severe accident. Then we have been developing the Hydrogen Treatment System to prevent excessive pressure without PCV vent. By focusing on the oxidation-reduction reaction of metal oxides with high reaction rate, we have been studying hydrogen treatment system using metal oxides as effective device under oxygen deficit conditions of PCV of Boiling Water Reactor (BWR). We have been considering a hydrogen treatment unit with a lot of pipes in which metal oxides are filled. We have already investigated experimentally the basic trend of metal oxides temperature and gas concentration as well as hydrogen processing rate dependency on gas temperature and concentration in the absence of steam. However, the influence of steam on hydrogen processing characteristics has not been clarified yet. The objective of the present research is to investigate how the steam, which is supposed to be a reaction-inhibiting-factor, affects hydrogen processing characteristics of the hydrogen treatment unit. We conducted experiments using a test section of one pipe simulating a part of the hydrogen treatment pipes. The granulated CuO, which is a candidate material for the actual system was used. The hydrogen concentration of 10 wt% at the inlet of the pipes was decreased to 0 wt% at the outlet even in high steam concentration conditions (30–50 wt%) when the gas temperature was 250 °C, therefore, it was confirmed that hydrogen was treated with high processing rate under steam circumstances. It was also found that cumulative amount of treated hydrogen was strongly correlated with temperature and relative humidity. We have been developing the thermal-chemical model of hydrogen treatment unit. The prediction margin of error was decreased to 30 % from over 100 % by improving degradation model based on the experiment results, therefore, it reached the practical level.

Kashiwazaki-Kariwa nuclear power station of TEPCO is the largest nuclear power station in the world, and it has seven nuclear power plants. As the experience at Fukushima Daiichi nuclear power station accident in March 2011 involving concurrent core damage at multiple units, it is considered that the risk derived from hazards of Earthquakes and Tsunamis is relatively significant in Japan, and these events have a high likelihood of damaging multiple units simultaneously. Therefore, it is very important to grasp the multi-unit specific risk. Although there are some unique accident scenarios of Multi-Unit PRA, this paper focuses on the influence of radioactive materials released outside the containment vessel on the accident management of the adjacent unit. The events including core damage and loss of containment function should be considered as the causes of the release of radioactive substances, and operator’s operation or the like should be considered as objects to be adversely affected by them. It is necessary to incorporate that into PRA to confirm the effect on risk. It is very difficult in terms of the maturity of evaluation method and the calculation load to accurately incorporate consequences derived from time series of various events and complicated interaction into PRA model. Therefore, as the first step in evaluating the risk of influence of radioactive material release on the accident management, some streamlining efforts are implemented according to the purpose. For example, Kashiwazaki-Kariwa unit 6 and unit 7 were set as the target units for model simplification. We also assume the earthquake as the initiating event due to the strong common factor for multi units. Whether or not to be operable in the adjacent plant is set conservatively based on deterministic evaluation. PRA taking into consideration the radiation influence by multi-unit accident is compared with normal PRA. Some kind of Core Damage Frequency (CDF) such as CDF 1 (Core Damage Frequency at which the damage of one or the other of two unit occur), CDF 2 (CDF at which the damage of both of units occur) and CDF TOTAL (CDF at which the damage of one or more units occur: CDF 1 + CDF 2 ) are quantified, and the degree of this issue is provided. Although the change of CDF TOTAL was insignificant, the necessity of further study was shown from the viewpoint of the amount and timing of radioactive substance released due to an approximately 1.5-fold increase in CDF 2 .

In order to enhance Generation II reactors safety, Generation III+ reactors have adopted passive mechanisms for their safety systems. In particular, the AP1000 ® reactor uses these mechanisms to evacuate heat from the containment by means of the Passive Containment Cooling System (PCS). The PCS uses the environment atmosphere as the ultimate heat sink without the need of AC power to work properly during normal or accidental conditions. To evaluate its performance, the AP1000 PCS has been usually modeled with a Lumped Parameters (LP) approach, coupled with another LP model of the steel containment vessel to simulate the accidental sequences within the containment building. However, a 3D simulation, feasible and motivated by the current computational capabilities, may be able to produce more detailed and accurate results. In this paper, the development and verification of an integral AP1000 ® 3D GOTHIC containment model, taking into account the shield building, is briefly presented. The model includes all compartments inside the metallic containment liner and the external shield building. Passive safety systems, such as the In-containment Refueling Water Storage Tank (IRWST) with the Passive Residual Heat Removal (PRHR) heat exchanger and the Automatic Depressurization System (ADS), as well as the PCS, are included in the model. The model is tested against a cold leg Double Ended Guillotine Break Large Break Loss of Coolant Accident (DEGB LBLOCA) sequence, taking as a conservative assumption that the PCS water tank is not available during the sequence. The results show a pressure and temperature increase in the containment in consonance with the current literature, but providing a greater detail of the local pressure and temperature in all compartments.

Accident analyses of unit 3 in the Fukushima-Daiichi nuclear power station were performed with MAAP (Modular Accident Analysis Program) 5.03. In the analysis, the operations of RCIC, HPCI, SRV were simulated and assuming the operation of alternative water injection to reproduce the measured pressure and temperature. As a result of parametric evaluation, analysis results consistent with measured values are obtained. In this paper, the results of the sensitivity analysis are reported. Simultaneously, attempts were made to analyze the transient and deposition amount of fission product (FP) in the reactor building (R/B) and a model dividing each room of R/B in detail was created to confirm the FP behavior. Based on these analyses, the deposition amount of the primary containment vessel (PCV) (drywell (D/W) and wetwell (W/W)) and R/B could be estimated using detailed model of the R/B dividing into nodes in the MAAP simulation.

The natural circulation small modular reactor (NCSMR) is a 330 MW reactor which has no reactor coolant pumps (RCP) and no active safety injection systems at all. The reactor is mainly comprised of the reactor pressure vessel (RPV) with integral pressurize r and steam generator. RPV is enclosed by a vacuumed pressure containment vessel (PCV) and the PCV is submerged in the underground containment pool. A MELCOR model and corresponding input deck are developed for the RPV, PCV, and containment pool. The containment pool takes the role of ultimate heat sink (UHS) in accident situations. The containment pool may crack and leak in some critical accidents as the earthquake, leading to the severe accident of the reactor. A TMI-2 like SBLOCA in the RPV (stuck open RVVs) along with the containment pool crack (loss of ultimate heat sink) is simulated in the work. So me key parameters as the RRVs stuck open fraction, the PCV-SRVs open or not, the containment pool crack position would have large influence on the severe accident sequence. The sensitivity of these parameters to the accident sequence is analyzed in the work. According to the simulation results, the RPV pressure decreased with the RRVs stuck open. The depressurization of RPV accelerated with the RPV-SRV open fraction increase. The PCV pressure increased after that. Two cases as the PCV-SRV open after PCV pressure increase to 5 MPa, and PCV break while the RV d id not open, are analysis. The coolant discharge mass flo wrate in RPV and PCV were different in two cases, leading to the different degradation situation of the core. Since the containment pool is so important for the accident mitigation, sensitivity analysis is done for the containment pool crack position in the pool. The work will be meaningful in gaining an insight into the detailed process involved. One of the final goals of this work would be to identify appropriate accident management strategies and countermeasures for the potential extreme natural hazard induced severe accidents during the design process of NCSMR.

Focusing on the general and localized elastoplastic buckling of the cylindrical section of steel containment under axial pressure, nonlinear finite element method (FEM) and small-scaled experiments are applied to analysis. First, FEM analysis is conducted considering nonlinear items caused by geometric shape imperfection and elastoplastic constitutive model by the arc-length method RIKS procedure. Parameter sensitivity of the buckling is revealed. Then, small-scaled experiments are carried out. Buckles status is observed, and key geometrical parameters’ influence are found. The results show that cylindrical buckling under axial pressure is sensitive to geometrical parameters and imperfection. It is necessary to employ more realistic parameters to the FEM analysis via accurate geometrical measurement. This research has reference value for the design and fabrication of AP series steel containment vessel.

Hydrogen may be released by Fuel and Cladding interaction with the steam at very high temperatures into the containment during a severe accident at the nuclear power plant (NPP). Locally, high hydrogen concentration may be achieved that might probably result in detonation or fast deflagration and challenges the reliability of the containment. The mixing and distribution of hydrogen is a serious safety issue for scientists to preserve the structural reliability of the containment. Although the Three Miles Island (TMI) accident in 1979 was the initiator to study the production and buildup of hydrogen in the containment. Subsequently, the hydrogen explosion in the Fukushima Dai-ichi NPP accident (2011), modeling of the gas behavior turn out to be a significant topic in the nuclear safety analysis. Computational fluid dynamics (CFD) codes can be used to investigate the hydrogen distribution in the containment during accidental scenarios and predicts the local hydrogen concentration in different regions of the containment vessel. In such a way, the associated risk of hydrogen safety can be determined, and safety assessment and procedures can be measured. The current paper presents the results of systematic work done by using the HYDRAGON code, developed by Department of Engineering Physics, Tsinghua University, to study (I) Mesh sensitivity (II) Buoyancy-driven flows analysis, for various turbulence models i.e., a standard k-ε (SKE) model, a renormalization k-ε (RNG) model and a realizable k-ε (RLZ) model and, to demonstrate the HYDRAGON code thermal-hydraulic simulation capability during a severe accident at the NPP. The HYDRAGON code simulation results were compared to the published data performed by Jordan et al., and it has been observed that the simulation results obtained by refined mesh have given generally satisfactory results. However, the global effect of the buoyancy term in the turbulence equations on the HYDRAGON code simulated results is very small.

Even after 6 years since the accident, the exact accident progression for each unit and location of core debris has not been clarified. Currently efforts are directed towards robotic inspection with remote cameras, as well as dose and temperature measurements of the environment inside of the Primary Containment Vessels (PCV). In spite of their effort, the observed environmental data do not support the existence of a large radiation heat sources attributable to the molten core at the bottom of the Primary Containments. Under this situation the author has conducted a forensic engineering study (i.e., different fields of science work together to collect and integrate independent evidences) to clarify the most likely accident scenarios of the Fukushima Daiichi accident. Through this study the author found that the environmental contamination and public exposure could have been substantially mitigated should the following vulnerabilities have been removed before the accident: (1) The threat of hydrogen generation through “radiation-induced electrolysis”. (2) Potential threat of “internal hydrogen explosion” in the suppression pools. The atmosphere on top of the suppression pool water (cover gas) should have been nitrogen. (3) Potential threat of “internal hydrogen explosion” in pipes which had occurred in the case of the Hamaoka Unit 1 accident. (4) The leak rate of PCV should have been testable at its design pressure. Intrinsic safety factor of the containment flanges against effluent leakage should have been rated as a 3 for functional integrity of the PCV. (5) Spread of hydrogen gas from vent lines through duct networks connected to SGTS. The hardened vent line should be independent and provided with filters for Dry Well venting. No rupture disks should have been installed in the vent lines.

While hydrogen release into large compartment from confined compartment, hydrogen diffusion flame is easy to occur. There is intense heat radiation effect on electrical penetration from diffusion flame. Aim to evaluate the influence of diffusion flame on electrical penetration, systematic method is constructed, including computing view factor of electrical penetration, assessment of hot spot of containment vessel and research of heat transfer for electrical penetration. Research results give theory basis for determining location of venting which can generate hydrogen diffusion flame. The method can be extended to use in the influence evaluation of personnel hatch and equipment hatch in the containment vessel.

This paper explains the strategy of our company (Tokyo Electric Power, TEPCO) regarding means of long-term heat removal from the primary containment vessel (PCV) of Units 6 and 7 (ABWR) of the Kashiwazaki-Kariwa Nuclear Power Station in a severe accident. If the PCV continues in a high-temperature state for a long time, the strength of the PCV concrete will decline, and the risk of being affected by an earthquake will increase. Therefore, it is crucial for safety to cool the PCV and reduce its temperature to the maximum working temperature or lower. TEPCO provides a means of cooling the reactor pressure vessel (RPV) and PCV called the alternative coolant circulation system (ACCS). This system uses the heat exchanger of the residual heat removal (RHR) system, the make up water condensate (MUWC) pump, and alternative heat exchanger vehicles. By using these measures, it is possible reduce temperature in the PCV over the long term to the maximum working temperature (design value) or less, even in severe accident scenarios such as a large LOCA + ECCS function failures + SBO (station blackout). This function has quite high reliability, but in a scenario where these measures cannot be used, expectations are placed on the filtered vent (FV). However, due to FV characteristics, it is impossible to reduce to below the saturation temperature of 100°C at atmospheric pressure using FV alone, and it will be necessary in the medium/long-term to cool the PCV while also restoring the cooling equipment. Therefore, the following restoration operation of PCV cooling and its dose evaluation were studied. (1) RPV heat removal by restoring the RHR system (2) RPV and PCV heat removal using a portable pump employing a portable heat exchanger (3) RPV and PCV heat removal using the suppression pool water clean up system (SPCU) employing portable heat exchangers (4) RPV heat removal using the clean up water system (CUW) By clarifying beforehand issues such as feasibility of these systems, the on-site environment for restoration measures, and the necessary gear/systems, the authors were able to secure means of long-term cooling of the PCV, and further enhance PCV reliability.

Typically, the concentration of natural radioactive aerosols in the containment vessel of nuclear power plant is low and in equilibrium. But when a serious nuclear accident occurred, the massive radioactive aerosols would be rapidly released. In order to ensure the integrality of the containment, the pressure inside the containment must be reduced by reducing the concentration of the aerosol. It can cause a serious damage to the atmospheric environment if such radioactive aerosol directly release. In this paper a new mesoscopic impactor filter has been developed which not only can filtrate and collect the aerosol particles but also can decrease the flowing resistance of gas. This paper intends to make numerical simulation to study the regularity of the deposition of aerosols under the laminar condition at different working condition in mesoscopic impactor filters. The 3D model of the filter was established with the commercial software of ICEM CFD and the meshes were divided accurately. The gas phase uses the laminar model and the particles use DPM (Discrete Phase Model). The detailed modeling method is given and the simulation results are analyzed.

Hydrogen combustion including deflagration and detonation could become a significant threat to the integrity of containment vessel or reactor building in a severe accident of nuclear power stations. In the present study, numerical analyses were carried out for the ENACCEF No.153 test to develop computational techniques to evaluate the flame acceleration phenomenon during the hydrogen deflagration. This experiment investigated flame propagation in the hydrogen-air premixed gas in a vertical channel with flow obstacles. The reactingFoam solver of the open source CFD code, OpenFOAM, was used for the present analysis. Nineteen elementary chemical reactions were considered for the overall process of the hydrogen combustion. For a turbulent flow, renormalization group (RNG) k-ε two-equation model was used in combination with wall functions. Three manners of nodalization were applied and its influences on the flame propagation acceleration were discussed.

The Japan Atomic Energy Agency (JAEA) is pursuing the development and application of the methodologies on fission product (FP) chemistry for source term analysis by using the integrated severe accident analysis code THALES2. In the present study, models for the eutectic interaction of boron carbide (B 4 C) with steel and the B 4 C oxidation were incorporated into THALES2 code and applied to the source term analyses for a boiling water reactor (BWR) with Mark-I containment vessel (CV). Two severe accident sequences with drywell (D/W) failure by overpressure initiated by loss of core coolant injection (TQUV sequence) and long-term station blackout (TB sequence) were selected as representative sequences. The analyses indicated that a much larger amount of species from the B 4 C oxidation was produced in TB sequence than TQUV sequence. More than a half of carbon dioxide (CO 2 ) produced by the B 4 C oxidation was predicted to dissolve into the water pool of the suppression chamber (S/C), which could largely influence pH of the water pool and consequent formation and release of volatile iodine species.