Search

AbstractThe latest investigations of Fukushima Daiichi Unit 1 have demonstrated that corium attack to the pedestal walls and pedestal floor has occurred in Fukushima Daiichi Unit 1 to a certain extent. The results of past analytical benchmarks, such as the Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF project), have agreed with this finding. However, the latest investigation does not show evidence of unlimited molten core–concrete interaction (MCCI), which is one of the main discrepancies from the BSAF project.More recently a MCCI benchmark has been launched in the context of the OECD/NEA project ARC-F (Analysis of Information from Reactor Building and Containment Vessels of Fukushima Daiichi Nuclear Power Station). In the benchmark, common geometry, boundary, and initial conditions have been selected among all the participants. The results show an improved agreement among different codes for what concerns overall erosion, corium temperature, and hydrogen generation, confirming that to some extent, the earlier scatter found in these variables came from differences in the MCCI scenario modeled by each partner.However, common unlimited erosion, not observed by onsite visual inspections, is still predicted. Understanding the origin of this deviation might provide insights into boundary conditions, model drawbacks, or ill-posed assumptions that might need to be revisited (e.g. interfacial temperature, effective heat transfer coefficients, concrete heat transfer). In this paper, a summary of the overall results and a discussion of modeling and boundary conditions is presented to disclose the results of the activity and the future steps to be taken in the OECD/NEA project FACE (Fukushima Daiichi Nuclear Power Station Accident Information Collection and Evaluation). [ABSTRACT FROM AUTHOR]

Accomplishing the retrieval of fuel debris from Fukushima Daiichi Nuclear Power Plant (1F) Unit 3 (1F3) requires an understanding of its distribution. In this study, we performed real-scale corium spreading and sedimentation behavior analyses using Lagrangian moving particle hydrodynamics and large eddy simulation methods. These methods allowed us to calculate the spreading of corium with various shear viscosities under water conditions and to propose the best estimation for the fuel debris distribution in 1F3. To minimize uncertainties arising from unknown boundary conditions, we investigated relevant parameters through literature review. Our analyses showed that highly viscous corium tends to pile up within the pedestal region under strong convective vapor and boiling heat transfer, while low-viscosity corium spreads to the outside of the pedestal regions regardless of cooling efficiency. We identified three cooling modes based on initial shear viscosity and cooling efficiency and predicted the fuel debris distribution in 1F3 by comparing our results to those of the Tokyo Electric Power Company (TEPCO) and Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) project. The distribution estimation of highly viscous corium derived from oxidic corium is consistent with the three-dimensional reconstructed image by TEPCO and the calculated results by the OECD/NEA BSAF project. [ABSTRACT FROM AUTHOR]

The recent Nuclear Regulation Authority evaluation report suggests that at the Fukushima Daiichi Nuclear Power Station, the concrete shield plugs above the primary containment vessel (PCV) have exceptionally high radiation levels in Units 2 and 3, which may increase the risk of radiation exposure during decommissioning operations. During the cleaning and disassembly of such radiation hot spots, it is expected that a large amount of submicron-sized radioactive aerosol particles will be generated, which may influence the decommissioning operation. In the present study, laser cleaning experiments were conducted at the University of Tokyo Aerosol Removal Test Facility to simulate aerosol scavenging during the laser cleaning process. The facility can reproduce multiple phenomena expected in actual plant decommissioning, such as laser decontamination and simultaneous mist and spraying operations. Through the work, we have developed effective aerosol dispersion control methods and strategies based on the joint use of water mist and water spray to reduce radiation risk in either laser cutting or other means of decontamination methods. Preliminary laser cleaning experiments on various coated samples were conducted to assess the aerosol removal efficiency using water droplets and mist. It was verified that the proposed method effectively cleans the radiation hot spots during the decommissioning process. [ABSTRACT FROM AUTHOR]

For the Japan Sodium-cooled Fast Reactor (JSFR), should the hypothesized core disruptive accident (CDA) happened, the in-vessel retention (IVR) will be the main target to achieve. In the heat-removal phase of the CDA, the debris bed will be piled up on the debris catcher. The capability of stable cooling and avoiding recriticality on the debris bed will be the main issues for achieving IVR. Previous studies have shown that the homogeneous debris bed can attain stable cooling and eliminate the probability of recriticality. Besides, self-leveling, which is a mechanism redistributing and flattening the debris bed by the natural circulation or vaporization from surrounding coolant, can further suppress the debris bed's thickness to below the coolable thickness. However, in the real situation, the debris bed is composed of mixed-density debris particles. Hence, when these mixed-density debris particles start to redistribute due to self-leveling, the debris bed will form a heterogeneous density distribution. Under this scenario, the capability of coolability and the probability of recriticality could deviate from the previous study. Therefore, it is necessary to obtain a verified coupled model between the computational fluid dynamics (CFD) and the discrete element method (DEM) to track the mixed-density debris particles' movement under the phenomenon of self-leveling. In this paper, first, the experiments simulating self-leveling on the mixed-density particle bed are performed. Afterward, the random heavy particle movement's experimental data are extracted and transformed into the statistics form as the benchmark materials. Finally, the CFD-DEM model is validated via a series of sensitivity studies. The verified CFD-DEM can be expected to simulate the self-leveling behavior on the mixed-density debris bed and the real reactor case. [ABSTRACT FROM AUTHOR]

The Great East Japan Earthquake and tsunami on March 11, 2011, mark the start of the nuclear accident at the Fukushima Daiichi nuclear power plant. Progression of the accident has been analyzed with the SAMPSON code. SAMPSON was originally designed as a large-scale simulation system with the maximum use of mechanistic models and theoretically based equations. In the progression analysis done for Unit 2, SAMPSON could reproduce the pressure transient of the reactor pressure vessel (RPV) reasonably well by assuming partial load operation of the reactor core isolation cooling system (RCIC). The pressure transient of the primary containment vessel was reproduced reasonably well by assuming torus room flooding. After the RCIC trip and manual opening of the steam relief valve, SAMPSON predicted the damage to the upper part of the fuel assemblies near the core center and RPV failure due to creep rupture. More than 91 wt% of the core debris relocated to the lower plenum was as particles, and the major constituents were UO2, Zr, and ZrO2 by SAMPSON analysis. [ABSTRACT FROM AUTHOR]