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Background: Microdialysis sampling after drug microdosing may provide tissue pharmacokinetic data early in clinical drug development. However, low administered doses and small sample volumes pose an analytical challenge, particularly for highly protein-bound drugs. Materials & methods: Carbon-14 [14C]diclofenac was used as a model drug to assess the technical and analytical feasibility of in vivo microdialysis after microdose administration in an in vitro setup. Results: [14C]diclofenac dialysate concentrations were accurately quantified with accelerator MS. [14C]diclofenac dialysate recoveries were similar in the presence and absence of therapeutic diclofenac concentrations but were considerably decreased when albumin was added to the immersion solution, suggesting high protein binding. Conclusion: These results demonstrate the feasibility of combining microdosing and microdialysis to assess tissue pharmacokinetics.

This article introduces an artificial neural network (ANN) model to determine cycle-times for forming curved hull plates when the target shape is known. The proposed model aids shipbuilding companies in predicting the cycle-times required for ship fabrication. The input parameters are geometric information extracted from the target shape (curvedness, Gaussian curvature, width, and height of the hull plate), and the output parameter is the heating duration per unit area. The structure of the proposed model, which predicts cycle-times for line heating after the cold forming case, consists of two hidden layers. The proposed model is convenient to use and flexible because it only requires retraining when the dataset is changed. The performance of the proposed model was analyzed by five-fold cross-validation and compared with that of a mathematical model obtained from the linear regression analysis method and predefined formulas. The results show that the ANN model is reliable and accurate for the cycle-time prediction of curved hull plates in shipbuilding applications. Introduction Shipbuilding companies generally estimate the production cost of a ship based on their previous ships for various purposes before the production planning department begins to optimize the fabrication process. They use the estimated value to refine the overall fabrication process or improve it by reducing unnecessary tasks and maximize the overall production efficiency.

Shingled string photovoltaic (PV) cells have a larger active area, and they significantly reduced damage from the interconnection process using electrically conductive adhesives. This new method using a shingled design provides increased PV module power compared with the conventional PV module for the same installed area. In this work, we study the optimization of shingled string PV cells for high-power shingled design modules (SDMs). To successfully fabricate shingled string PV cells and high-power SDMs, we adjusted the overlap width between solar cells and achieve conversion powers of 406.5 W for a 1100 mm × 1980 mm area. Finally, we perform a reliability tests to verify the durability of our SDMs.

This paper presents a study on the effects of heat treatment conditions on electrically conductive adhesives. Among the advantages of the shingled solar cells include larger active area and smaller current density since one of the main factors of the power loss is due to a decrease in current density. Therefore, when there is a small current, there is a benefit in regards to the power loss. The advantage of this new technique of developing photovoltaic modules is the increase of module power using the same installed area. Electrically conductive adhesives play an important role in the manufacture of shingled solar cells and understanding the effects of its curing condition is necessary to maximize its output power. Through changing the curing time and temperature, the optimized curing conditions for electrically conductive adhesives and fabricated shingled strings for development of a module could be established. Finally, we demonstrated a 500 mm × 500 mm photovoltaic module with a conventional and the other using the shingled method for purposes of comparison and a shingled module showed about 29% increase in maximum output power compared to a conventional module with the same installed area.

The Fukushima accident is characterized by the fact that three reactors at the same site experienced reactor vessel failure and the accident resulted in significant radiological release to the environment, which was about 1/10 of the Chernobyl releases. The safe removal of fuel debris in the reactor vessel and Primary Containment Vessel (PCV) and treatment of huge amount of contaminated water are the major issues for the decommissioning in coming decades. Discussions on the new researches efforts being carried out in the area of investigation of the end state of fuel debris and Boling Water reactor (BWR) specific core melt progression, development of technologies for the mitigation of radiological releases to comply with the strengthened safety requirement set after the Fukushima accident are discussed.

An analysis on the thermal hydraulic behavior, core damage progression and radiological releases of a steam generator tube rupture event initiated by a station black out event for APR1400 is provided in this paper. The analysis was performed by using MELCOR, where counter-current natural circulation, mixing in the steam generator inlet plenum, creep ruptures of reactor coolant system components and U-tube, and mitigation actions by the operator are modeled. A station black out accident with an opening of atmospheric steam dump valves (ADVs) at the time of severe accident management strategy entry condition with a failure of a supply of auxiliary feed water to the steam generators selected as a base case resulted in a creep rupture of U-tubes with a direct release of radioactive materials into the atmosphere. It was shown that a mitigation action by the operator for an aggressive cooldown of the reactor coolant system using ADVs and supplying the water to the steam generator even after the creep rupture of U-tubes was effective. It arrested core melt progression and significantly reduced the amount of radiological releases to the atmosphere.

SMART is a small-sized integral type PWR containing major components within a single reactor pressure vessel. Advanced design features implemented into SMART have been proven or qualified through experience, testing, or analysis according to the applicable approved standards. After Fukushima accident, a rising attention is posed on the strategy to cope with a Station Blackout (SBO) accident, which is one of the representative severe accidents related to the nuclear power plants. The SBO is initiated by a loss of all offsite power with a concurrent failure of both emergency diesel generators. With no alternate current power source, most of the active safety systems that perform safety functions are not available. The purpose of SBO analysis in this paper is to show that the integrity of the containment can be maintained during a SBO accident in the SMART (System-integrated Modular Advanced ReacTor). Therefore, the accident sequence during a SBO accident was simulated using the CINEMA-SMART (Code for INtegrated severe accidEnt Management and Analysis-SMART) code to evaluate the transient scenario inside the reactor vessel after an initiating event, core heating and melting by core uncovery, relocation of debris, reactor vessel failure, discharge of molten core, and pressurization of the containment. It is shown that the integrity of the containment can be maintained during a SBO accident in the SMART reactor. It has to be mentioned that the assumptions used in this analysis are extremely conservative that the passive safety systems of PSIS and PRHRS were not credited. In addition, as ANS73 decay heat with 1.2 multiplier was used in this analysis, actual progression of the accident would be much slow and amount of hydrogen generation will be much less.

A new empirical model is proposed for estimating the amount of volatile iodine in an aqueous phase. The volatile iodine concentration is estimated for highly irradiated CsI solutions in which the p...

Melt behavior of a BWR in-core monitor (ICM) guide tube, which penetrates a RPV bottom wall, under a severe accident condition was tested using a corium with a mixture of UO 2 , ZrO 2 and Zr. There are several BWR-ICMs; two of them are a short range monitor (SRM) and an intermediate range monitor (IRM), which are made of stainless steel, have the same tube dimensions and the bottom ends of the tubes open to the drywell (or, the pedestal region). In the test, the full scale SRM/IRM guide tube with the same dimension as one of Fukushima Daiichi Unit-1 but short length, was installed in the lower crucible which simulated the lower plenum and which bottom simulated the RPV bottom wall with about 200 mm thickness. The melt corium was generated in the upper crucible and it was discharged into the lower crucible. The test result showed that: (1) The corium discharged into the lower plenum attacked the guide tube first resulting in the tube melt from the top and then accumulated in the lower plenum. (2) Small amount of corium had fallen out from the bottom opening of the tube into the pedestal as particles, and such outflow stopped soon because a stainless steel melt had solidified in the tube and blocked the flow path. (3) The depth of the melt ingression into the tube was 60–65 mm from the inner surface of the RPV bottom wall, where the tube was completely blocked by the solidification of steel melt and no oxide melts had ingressed there.

Present paper investigates the potential configuration of post-accident fuel debris for a severe accident in a nuclear power plant. The physical and chemical behaviors of molten core material are investigated by experimental efforts. A series of melting and solidification experiments utilizing a cold crucible technique was performed using corium of 4.5–15 kg at various compositions. Considering that U/Zr atom ratio and Zr oxidation index could vary substantially during the accident progression, corium melts at compositions with U/Zr in the range of 0.94–1.14 and Zr oxidation index in the range of 36.3–100% were investigated. It was found that the solidified corium separated into a metal rich and an oxide rich layer in cases of partially oxidized corium. Depending on the compositions, metal rich layer was placed either on top or at the bottom of the oxide rich layer. Inductively Coupled Plasma-Atomic Emission Spectroscopy was used to estimate the chemical compositions of the layers and X-Ray Diffraction was used to find the crystal structure of the layers. In the second series of experiments, a molten corium of 13 kg in the form of liquid jet is poured into a pool of water. Melt water interaction resulted in fragmented and quenched particles with a range of size distributions. Scanning Electron Microscope and Electron Probe Micro-Analysis were used to analyze the physical and chemical behaviors of the particles. Partially oxidized corium melt resulted in a quenched particles with irregular shape. The composition of the particles varied among different particles and also within a particle. Fully oxidized melt resulted in quenched particles with homogenous composition among different particles and homogenous distribution of composition within a particle. The shape of these particles was spheroid in most of cases.

An analysis on the behavior of fission products during a station blackout (SBO) accident for the APR1400 nuclear power plant is performed using MELCOR version 2.1. The analysis is focused on investigating the characteristics of radiological releases from the damaged fuel in the reactor core and molten corium in the reactor cavity, transport of the radioactive materials in the reactor coolant system and containment, and release of the radioactive materials to the environment. It is shown that the release characteristics of radionuclides from the fuel, distributions of radionuclides in the reactor coolant system and containment strongly depend on the species of radionuclides. In cases of volatile radionuclides such as iodine and cesium, more than 95% of initial inventory is released before the reactor vessel failure. The amount of radionuclides released to the environment after a failure of the containment was significantly reduced due to the deposition of radionuclides in the form of aerosols on the surfaces of heat structures and a pool of water in the containment. A sensitivity study on the presence of water in in-containment refueling water storage tank (IRWST) shows that the water in the IRWST significantly reduced the amount of radionuclides released to the environment.

Filtered containment venting system (FCVS) has been developed since 2014 in Korea, and an integration performance verification test has been performed at the Korea Atomic Energy Research Institute (KAERI). The experimental facility was named ARIEL: Aerosol Removal & Iodine Elimination (ARIEL). The ARIEL facility is composed of four major parts; venturi scrubber, cyclone, metal fiber filter (MFF), and molecular sieve (MS). In addition, a generation and sampling system of aerosol and iodine were equipped to the system. In this study, the current status of the FCVS experimental facility in KAERI is introduced along with a description of the aerosol generation and sampling system. Aerosol removal test results, performed 7 bar abs. with non-condensable gas, is also presented. The analysis of the results was conducted with filter and aerosol measurement equipment, and is included in the study.

Evaluation of radionuclide releases to the environment from operating nuclear power plants has increased in importance in the Republic of Korea following the 2015 amendment to its Nuclear Safety Act that includes not only design basis accidents but also severe accident management. Containment bypass accidents are among the crucial scenarios that should be considered under the new circumstances due to their high consequences, even though the probability of such accidents is low. In this study, analysis of a thermally induced steam generator tube rupture (TI-SGTR) accident caused by a severe accident has been conducted with MELCOR. To increase the reliability of the analysis results, qualification processes were performed in the process of producing the input. In addition, computational fluid dynamics analysis was performed to simulate the natural circulation phenomena in the reactor coolant system (RCS) during the accident and major flow parameters were obtained for the target plant. Thermal-hydraulic and fission product behavior was found during the accident for the base case scenario, and uncertainty band for CsOH release fraction to the environment were presented through the uncertainty analysis on the natural circulation flow parameters. Sensitivity calculations were also performed on uncertain parameters such as the number of ruptured tubes and the depressurization time of the secondary system, and the CsOH release fraction to the environment without passing through the containment were evaluated.

In order to develop a domestic research roadmap for severe accidents, a special committee was established by the Korean Nuclear Society. One of the subcommittees discussed the characteristics and the relevant technical issues in the stages of fission product release and physical forms of radionuclide release and transport. The group members developed a tree to identify fission product release phenomena by tracing failures of individual defense-in-depth barriers and added possible countermeasures against failure. For each elemental issue, they searched for technical problems by examining the phenomena, accident management actions, and regulatory aspects relevant to the mitigation features for containment, including mitigation strategies against containment bypass accidents. Regulatory concerns, including the source term and the acceptance criteria for radionuclide release, were also considered. They identified further research needs regarding important technical issues based on the degree of the current knowledge level in Korea and in foreign countries, looking at the significance and urgency of issues and the expected research period required to reach an advanced level of knowledge. As a result, the group identified the 12 most important and urgent issues, most of which were expected to require mid-term and long-term research periods.

VESTA (Verification of Ex-vessel corium STAbilization) and VESTA-S (-small) test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging ZrO 2 melt jet on a sacrificial material were performed to investigate the ablation characteristics. ZrO 2 melt in an amount of 65–70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40), and the other is a stainless steel (SUS304) melt. Metallic melt in an amount of 1.5–2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. ZrO 2 melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is UO 2 60%, Zr 10%, ZrO 2 15%, SUS304 14%, and B 4 C 1%, was melted in a cold crucible using an induction heating technique.

As a part of the integrated severe accident code development project in Korea, KAERI (Korea Atomic Energy Research Institute) has been developing a stand-alone severe accident analysis code, COMPASS (COre Meltdown Progression Accident Simulation Software), which simulates the in-vessel severe accident phenomena including the core heat up, material melting and relocation, corium behavior in the lower plenum, and vessel failure. For the purpose of COMPASS code validation, a numerical simulation was conducted for the PHEBUS FPT3 experiment, which is an integrated severe accident experiment, initiated in 1988 by IRSN. In the COMPASS code, the core and surrounding structure in a test section consist of a two-dimensional node system, and the mass and energy equations are established for the main component in the core and surrounding structures. The temperature evolution of the core and surrounding structure at various locations of the test section are pursued and compared with the experimental data as well as the numerical results of the MELCOR code. The hydrogen generation rate and final mass distribution are compared among the experimental data and the numerical results of COMPASS and MELCOR. Generally, the COMPASS and MELCOR codes predicted the temperature evolution and hydrogen generation rate during the experiment well. In addition, COMPASS predicted a slightly smaller mass relocation compared with the experimental data, whereas MELCOR has a larger mass relocation.

An analysis of the responses of the containment during a station blackout accident is performed for the APR1400 nuclear power plant using MELCOR 2.1. The analysis results show that the containment failure occurs at about 84.14 h. Prior to the failure of the reactor vessel, the containment pressure increases slowly. Then, a rapid increase of the containment pressure occurs when a large amount of hot molten corium is discharged from the reactor pressure vessel to the cavity. The molten corium concrete interaction (MCCI) is arrested when water is flooded over a molten corium in the cavity. The boiling of water in the cavity causes a fast increase in the containment pressure. During the early phase of the accident, a large amount of steam is condensed inside the containment due to the presence of the heat structures. This results in a mitigation of a containment pressure increase. During the late phase, the containment pressure increases gradually due to the addition of steam and gases from an MCCI an...

Ex-vessel core catcher cooling system driven by natural circulation is designed using a full scaled air-water system. A transparent half symmetric section of a core catcher coolant channel of a pressurized water reactor was designed with instrumentations for local void fraction measurement and flow visualization. Two designs of air-water top separator water tanks are studied including one with modified ‘super-step’ design which prevents gas entrainment into down-comer. In the experiment air flow rates are set corresponding to steam generation rate for given corium decay power. Measurements of natural circulation flow rate, spatial local void fraction distribution and re-wetting time near the top wall are carried out for various air flow rates which simulate boiling-induced vapor generation. Since heat transfer and critical heat flux are strongly dependent on the water mass flow rate and development of two-phase flow on the heated wall, knowledge of two-phase flow characteristics in the coolant channel is essential. Results on flow visualization showing two phase flow structure specifically near the high void accumulation regions, local void profiles, rewetting time, and natural circulation flow rate are presented for various air flow rates that simulate corium power levels. The data are useful in assessing the cooling capability of and safety of the core catcher system.

In this paper, we propose a method for registering unorganized point clouds without using targets or markers. Motivated by the 4-points congruent sets (4PCS) algorithm, which is a nontarget-based registration method commonly used in the related fields, we develop a feature-based 4PCS algorithm (F-4PCS). The method combines the basic approach of the 4PCS algorithm with geometric feature information to produce consistent global registration results efficiently. We use the features from the point feature histogram descriptor and the ones that capture the surface curvature. The experimental results show that the proposed method successfully registers point clouds of both the outdoor and indoor scenes and demonstrates better performance than the existing 4PCS-based registration methods.

The operating system fingerprint is a factor that can help determine a target operating system and version through network scanning. There are two methods of discrimination: Internet Control Message (ICMP) and Transmission Control Protocol (TCP). In this study, we analyzed whether it is possible to categorize the operating system version (e.g., Windows 7, 8.1, 10) in a manner that the operating system can be determined using ICMP. Using ICMP, we could successfully classify the operating systems into Windows and Linux.

Some advanced reactors adapt the in-vessel corium retention concept by cooing the outside wall of the reactor vessel in severe accidents. If a reactor vessel failure happens in this case, the molte...

International audience; The Great East Japan earthquake occurred on March 11, 2011, at 1446, and the subsequent tsunami led Tokyo Electric Power Company's (TEPCO's) Fukushima Daiichi Nuclear Power Station (NPS) beyond a design-basis accident. After the accident, the Japanese government and TEPCO compiled a roadmap toward an early resolution to the accident including, among the main activities, the employment and improvement of existing severe accident (SA) computer codes. In the member countries of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA), SA codes were developed after the accident at Three Mile Island Unit 2 and widely employed to assess NPS status in the postulated SA conditions. Therefore, working plans have been set up with the country members of the OECD/NEA to apply existing SA codes to analyze the accidents at the Fukushima Daiichi NPS Units 1, 2, and 3 and support the decommissioning, constituting an international program named Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF). The objectives of the BSAF project are to analyze the accident progression of Fukushima Daiichi NPS, to raise the understanding of SA phenomena, to contribute to the improvement of the methods and models of the SA codes, and to define the status of the distribution of debris in the reactor pressure vessels and primary containment vessels for decommissioning. The present technical paper summarizes the achievements obtained through a comparison of the results, emphasizing the portions of the accident where all the participants reached a common consensus and identifying still open questions where future work should be directed. Consensus exists on the current condition of Unit 1, where a large fraction of the fuel is assumed to have relocated ex-vessel. On the other hand, larger uncertainties exist for Units 2 and 3, where in-vessel and ex-vessel scenarios produce a reasonable prediction of the accident progression.

An experimental research platform using corium melts is established for the understanding of safety related important phenomena during a severe accident progression. The research platform includes TROI facility for corium water interaction experiments and VESTA facility for corium-structural material interaction experiments. A cold crucible technology is adapted and improved for a generation of 5–100 kg of corium melts at various compositions. TROI facility is used for experiments to investigate premixing and explosion behaviors during a fuel coolant interaction process. More than 70 experiments using corium at various compositions were performed to simulate steam explosion phenomena in a reactor situation. The results indicate that the conversion efficiency of steam explosion for corium is less than 1%. VESTA facility is used to investigate molten corium-structural material interaction phenomena. VESTA facility consists of two cold crucibles. One crucible is used for the melting of charged material and pouring of corium melt. The other crucible is used for the corium-structural material interaction while providing an induction heating to simulate the decay heat. The results of an experiment on the interaction between corium melt and a specimen made of Inconel performed in the VESTA facility is reported.

A configuration of molten core in the Fukushima Daiichi NPP (nuclear power plant) was investigated by a melting and solidification experiment. About 5 kg of a mixture, whose composition in terms of weight is UO 2 (60%), Zr + ZrO 2 (25%), stainless steel (14%), B 4 C (1%), was melted in a cold crucible using an induction heating technique. It was shown that the solidified melt consists of upper crust and lower solidified ingot. The solidified ingot was separated into two layers. A physical and chemical analysis was performed for the samples taken from the solidified melt to investigate the morphology and chemical characteristics. It was found that the solidified ingot consists of a metal-rich layer on the top and an oxide-rich layer at the bottom. In addition, the oxide layer at the bottom has composition close to the initial charge composition and surrounded by a thin crust layer. It turned out that B 4 C was more concentrated in the upper metal-rich layer. These findings provide important insights for understanding the core melt progression and taking proper post-accident recovery actions for the Fukushima Daiichi NPP.

Steam explosion experiments are performed at various modes of melt water interaction configuration using prototypic corium melt. The tests are performed to simulate both melt water interaction in a...

A scaling method is proposed for the design of a reduced-scale experimental facility for testing the performance of a newly proposed filtered containment venting system (FCVS). A full-height facili...

A severe accident in Fukushima occurred on March 11, 2011 and units 1, 2 and 3 were damaged severely. A tsunami following an earthquake made the supply of electricity power stop, and the safety systems, which use AC or DC power in plants could not operate properly. It is supposed that the degree of core degradation of unit 2 is less serious than in the other plants, and it was estimated that the operation of reactor core isolation cooling (RCIC) system at the initial stage of the accident minimized the core damage through decay heat removal. Although the operating conditions of the RCIC system are not known clearly, it can be important to analyze the accident scenario of unit 2. In this study, best case of the Fukushima unit 2 accident was presented considering the operating conditions of the RCIC system. The effects of operating condition on core degradation and fission product release rate to environment were also examined. In addition, importance of torus room flooding level in the accident analysis was discussed. MELCOR 1.8.6 was used in this research, and the geometries of plant and operating conditions of safety system were obtained from TEPCO through OECD/NEA BSAF Project.

This paper investigates risk significance of Steam Generator Tube Rupture (SGTR) events accompanying a core damage and a direct release of radionuclides into the atmosphere bypassing the containment. Analyses on a spontaneous steam generator tube rupture (S-SGTR) event leading to a core damage and a consequential steam generator tube rupture (C-SGTR) event caused by a station black out (SBO) event are performed for APR1400 using MELCOR 2.2. Both events led to a direct radiological release to the atmosphere. The amount of radiological release was more than that of the Chernobyl and the Fukushima accident. Operator actions to mitigate the radiological releases are investigated. For the S-SGTR scenario, recovery of either auxiliary feed water (AFW) or safety injection alone was not able to prevent significant radiological releases. Combined action of opening of atmospheric dump valve (ADV) and supply of AFW to the steam generators (SGs) resulted in an aggressive depressurization of the primary system and an actuation of safety injection tank. Recovery of liquid inventory and minimizing the U-tube leakage prevented core damage. For C-SGTR scenario, delayed operator action until the time of U-tube creep rupture was not able to prevent significant radiological releases. C-SGTR had to be prevented by a supply of AFW to SGs or rapidly depressurizing the reactor coolant system before an occurrence of U-tube creep rupture.

Much is still not known about the end-state of core materials in each of the units at Fukushima Daiichi Nuclear Power Station (Daiichi) that were operating on March 11, 2011. The Nuclear Energy Agency of the Organization for Economic Development has launched the Preparatory Study on Analysis of Fuel Debris (PreADES) project as a first step to reduce some of these uncertainties. As part of the PreADES Task 1, relevant information was reviewed to confirm the accuracy of graphical depictions of the debris endstates at the damaged Daiichi units, which provides a basis for suggesting future debris examinations. Two activities have been completed within the PreADES Task 1. First, relevant knowledge from severe accidents at the Three Mile Island Unit 2 and the Chernobyl Nuclear Power Plant Unit 4 was reviewed, along with results from prototypic tests and hot cell examinations, to glean insights that may inform future decommissioning activities at Daiichi. Second, the current debris endstate diagrams for the damaged reactors at Daiichi were reviewed to confirm that they incorporate relevant knowledge from plant observations and from severe accident code analyses of the BSAF (Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station) 1 and 2 projects. This paper highlights Task 1 insights, which have the potential to not only inform future Decontamination and Decommissioning activities at Daiichi but also provide important perspectives for severe accident analyses and management, particularly regarding the long-term management of a damaged nuclear site following a severe accident.

Investigations on the morphology and phase equilibrium characteristics of corium were performed by a series of experiments in parallel with thermodynamic phase equilibrium analyses. Melting and solidification experiments were performed using corium consists of U, Zr, ZrO2, SS, and B4C. TROI-49 and TROI-50 experiments with corium compositions representing Pressurized Water Reactor (PWR), whose compositions are similar to those of MA-3 and MA-4 of OECD MASCA (Material Scaling) Program while amount of mass is 5 times more, resulted in two-layered structure with oxide rich layer on top of metal rich layer. Experiments of FK-1 and FK-2 with a mixture of UO2, ZrO2, Zr, SS and B4C at a representative Fukushima Daiichi Nuclear Power Plant (FDNPP) corium composition resulted in a formation two-layered structure of corium with upper layer rich in metal and lower layer rich in oxide. Zirconium diboride phase trapped boron and mainly distributed in the metallic layer. Predictions by thermodynamic calculations for the equilibrium phase distribution and chemical compositions of each phase using the NUCLEA thermodynamic database with a focus on U-Zr-O-Fe system were in good agreement with the experimental results. Agreement in terms of U/(U+Zr) was quite close while there were dispersions in oxygen and steel components. The temperatures for the formation of two immiscible liquids calculated by the NUCLEA for FK-1 and FK-2 was 300 K higher than the experimental observation. This point has to be looked at for improving the NUCLEA models for corium containing B4C, by introducing a non-zero boron solubility in the ceramic phase. The agreement between the NUCLEA predictions and the results of experiments clearly indicate that thermodynamic equilibrium phases play an important role in governing the core damage progression, which is important not only for the severe accident management but also for decommissioning and defueling process for the Fukushima Daiich damaged reactors.

The Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF), which started in 2012 and continued until 2018, was one of the earliest responses to the accident at Fukushima Daiichi. The project, divided into two phases, addressed the investigation of the accident at Units 1, 2, and 3 by severe accident (SA) codes until 500 h, focusing on thermal hydraulics, core relocation, molten corium concrete interaction (MCCI), and fission product release and transport. The objectives of the BSAF were to make up plausible scenarios based primarily on SA forensic analysis, support the decommissioning, and inform SA code modeling. The analysis and comparison among the institutes have brought up vital insights regarding the accident progression, identifying periods of core meltdown and relocation and reactor pressure vessel (RPV) and primary containment vessel (PCV) leakage/failure through the comparison of pressure, water level, and containment atmosphere monitoring system (CAMS) signatures. The combination of code results and inspections (muon radiography, PCV inspection) has provided a picture of the current status of the debris distribution and plant status. All units present a large relocation of core materials and all of them present ex-vessel debris with Unit 1 and Unit 3 showing evidence of undergoing MCCI. Uncertainties have been identified, in particular on the time and magnitude of events such as corium relocation in the RPV and into the cavity floor and RPV and PCV rupture events. Main uncertainties resulting from the project are the large and continuous MCCI progression predicted by basically all the SA codes and the leak pathways from the RPV to the PCV and the PCV to the reactor building and environment. The BSAF project represents a pioneering exercise that has set the basis and provided lessons learned not only for code improvement but also for the development of new related projects to investigate in detail further aspects of the Fukushima Daiichi accident.

Jet impingement experiments were performed to investigate the ablation characteristics of special concrete, which has been developed as one of the candidate protecting materials for the EU-APR1400 ex-vessel core catcher. In order to simulate the jet impingement phenomenon owing to the reactor vessel failure during a severe core meltdown accident, the experimental facility was established and the experimental conditions were determined based on parametric studies. The special concrete specimen was manufactured in accordance with the standard procedures, and its microstructures and physicochemical properties were analyzed to verify the requirements for the qualification. An induction melting technique in a cold crucible was employed to generate the zirconium-dioxide melt as a simulant of the corium melt. The special concrete was ablated uniformly over the impact area by jet impingement, and the average ablation depth was measured to be 4.33 mm. The average ablation rate in depth was evaluated as 1.59 mm/s using the temperature measurements of the specimen. As compared with the predictions by the models based on the convective and radiative heat transfer analysis, both the measured ablation rate and depth were estimated appropriately within the bounds of their limits. However, the convective heat transfer model turned out to predict the ablation characteristics of the special concrete more reasonably during the jet impingement even though some water content within the special concrete could lead to a sudden generation of the steam layer through which the material ablation is attenuated substantially by the radiative heat transfer.

Dose-response trend between thyroid cancer and duration of employment in Financial and insurance activities (Reference: office workers of Financial and insurance activities regarding each respective duration of work). (DOCX 15Â kb)