Retraction: Li, J. Fault-Event Trees Based Probabilistic Safety Analysis of a Boiling Water Nuclear Reactor’s Core Meltdown and Minor Damage Frequencies. Safety 2020, 6, 28

The author would like to retract the article, published on 17 June 2020, cited above [...]

Liquid Film Behavior of Bottoming Liquid Jet in a Shallow Pool Measured by 3D-LIF

When the core meltdown accident occurs in the nuclear plant, molten corium falls into a coolant pool of the lower plenum. It is considered that the molten corium jet is broken up, cooled, and solidified with fuel-coolant interaction (FCI). However, the coolant pool could be a shallow condition by the leakage and evaporation of the coolant. In this situation, it is considered that the corium jet bottoms and spreads without the jet breakup. From the viewpoint of safety, understanding a jet behavior and estimating a cooling behavior are needed. The purpose of this study is to clarify the mechanism of the liquid jet behavior in a shallow pool as the fundamental process for estimating the cooling behavior in the real machine. In this paper, we discuss the spreading behavior of the liquid jet after bottoming. The jet injection experiment was conducted using test fluids. By using the 3D-LIF method, the 3D visualization of the liquid jet was Successfully implemented. From the visualization result, the following behaviors were seen. After bottoming, the jet spread radially with the liquid film. As the jet spreading behavior, the liquid film was rolled up to the inside, and the vortex was formed. After a certain time, the vortex was broken. Then the flow and the number density of the fragment were changed.

Evaluation of the Passive Reactor Shutdown System Performance in European Sodium Fast Reactor

The ESFR-SMART European project (Contract number: 754501) focuses on the development of innovative safety design options for European Sodium-cooled Fast Reactor (ESFR). The task of Work Package 1.3 is to assess the impact of the new safety measures on the reactor behaviour in the transients protected by either active or passive reactor shutdown systems. The aim of Task 4 in this Work Package is to evaluate the passive reactor shutdown system performance in the ESFR core. This paper deals with the results of this evaluation, which is based on the analysis of four transients passively protected by 12 Diversified Shutdown Device (DSD) rods. Simulations have been done with the SIM-SFR system code and demonstrated that DSD rods are capable to shutdown ESFR in a timely manner, in order to avoid the negative consequences of the analyzed transients. Although a total loss of heat sink transient is a practically eliminated event, it was included in the analysis to estimate the grace time before the core meltdown.

ESFR SMART PROJECT CONCEPTUAL DESIGN OF IN-VESSEL CORE CATCHER

Even before Fukushima accident occurred, the safety authorities have required that new power plant designs must take into account beyond design-basis accidents including possible core meltdown. Among the mitigation strategies, the corium retention must be ensured, so a core catcher is implemented in the design of the Generation IV Sodium-cooled Fast Reactor. An internal core catcher within the vessel (in-vessel retention) is the option chosen for the European Sodium-cooled Fast Reactor investigated in the H2020 ESFR-SMART project. The new core investigated in ESFR SMART with lower void effect has a better behavior in case of severe accident. The use of passive control rods is also an improvement for prevention of severe accident. Moreover, we have in the ESFR SMART core dedicated tubes for corium discharge that should allow discharging quickly the melted materials and should help to prevent large criticality. Calculations show that after several seconds, these discharge tubes begin to open, and the corium arrives by this preferential way on the core catcher, quicker and in limited quantities at the beginning of the accident. However, the core catcher is designed to be able to retain the whole core meltdown. Its design allows good possibilities of cooling by natural convection of sodium. Some thermal calculations were provided with a multi-layer concept but the global mechanical conception seems difficult. So a one layer core catcher in molybdenum, material compatible with sodium and used on the core catcher of the last SFR, started in 2016: BN 800, is investigated. Explanations are given on the choice of this material proposed for the catcher and used for thermal calculations. With the proposed design, the corium is spread on the core catcher and the residual power of the corium can be dispelled by natural convection by the sodium circulating around and above the core catcher without boiling of sodium if the melted core is less than about 25% of whole core. In case of bigger quantities of melted core, boiling of sodium could appear under the core catcher. Further less conservative calculations would be necessary to better know the limit.

Experimental and Numerical Investigation of Solidification of Gallium in an Initially Emptied Horizontal Pipe Flow

In a reactor core meltdown under postulated severe accidents, the molten material (corium) could be ejected or relocated through existing vessel penetrations (cooling pipe connections), thus potentially contaminating other locations in the power plant. There exists, however, a potential for plugging of melt flow due to its complete solidification, providing the availability of an adequate heat sink. Therefore, a numerical model was created to simulate the flow of molten metal through an initially empty horizontal pipe. The numerical model was verified using a previously developed analytical model and validated against experimental tests with gallium (low melting temperature) as a substitute for corium. The numerical model was able to predict the penetration length (length of distance travelled by the molten metal) after a complete blockage occurred with an average percent error range of 9%. Since the numerical model has been verified and validated, the model can be updated to predict the penetration length in the cooling pipe in case of a severe accident.

Criticality Characteristics and Sensitivity Analysis of Fukushima Debris Beds Containing MCCI Products

This study investigates the criticality characteristics of debris beds that may have been formed through the molten–core–concrete-interaction (MCCI) at the pedestal floor of the damaged reactors in Fukushima Daiichi Nuclear Power Station. These were modeled as UO2-concrete systems submerged in water. First, a conservative model was used to evaluate the impact that the presence of concrete has on the neutron multiplication factor (keff) of debris beds. The good moderation capacities of concrete were proved, and it was found that recriticality would be possible under the considered conservative assumptions. Second, a more realistic model was used to perform an uncertainty and sensitivity analysis of a wide range of debris parameters (debris porosity, core meltdown grade, debris size, debris composition, concrete erosion factor, etc.). In this case, the results indicate that the probability of a recriticality event is very remote. It was also found that the presence of boron (B4C) from the control rods within debris has by far the highest influence on keff.

Fault-Event Trees Based Probabilistic Safety Analysis of a Boiling Water Nuclear Reactor’s Core Meltdown and Minor Damage Frequencies

A systematic probabilistic safety assessment for a boiling water nuclear reactor core is performed using fault trees and event trees analysis models. Based on a survey of the BWR’s safety systems against potential hazards, eight independent failure modes (initiating events) triggered scenarios are modelled and evaluated in the assembled fault-event trees, obtaining the two key outcome probabilities of interest, i.e., complete core meltdown (CCMD) frequency and minor core damage (MCD) frequency. The analysis results indicate that the complete loss of heat sink accounts for the initiating accident most vulnerable to CCMD (with a frequency of 1.8 × 10 − 5 per year), while the large break in the reactor pressure vessel is the least susceptible one (with a frequency of 2.9 × 10 − 12 per year). The quantitative risk assessment and independent review conducted in this case study contributed a reference reliability model for defense-in-depth core optimizations with reduced costs, informing risk-based policy decision making, licensing, and public understanding in nuclear safety systems.

Experimental Investigation of Melt Coolability and Ablation Behavior of Oxidic Sacrificial Material at Prototypic Conditions in Scaled Down Core Catcher

To minimize the potential risk of design extension conditions (DEC) with core meltdown, some advanced reactors employ ex-vessel core catchers which stabilize and cool the corium for prolonged period by strategically flooding it. This paper describes the coolability of the melt pool and ablation process in a scaled down ex-vessel core catcher employing sacrificial material which reduces the specific volumetric heat, temperature, and density of the melt pool. To understand these phenomena, a simulated experiment was carried out. The experiment was performed by melting about 500 kg of corium simulant using thermite reaction at about 2500 °C. The bricks of oxidic sacrificial material were arranged in the core catcher vessel which was surrounded by a tank filled with water up to a certain level. After the time required for melt inversion, water was introduced to flood the test section from the top. The melt pool temperatures were monitored at various locations using “K” and “C” type thermocouples to obtain ablation depth at different elevations with time. The results show that the coolability of the molten pool in the presence of water for the present geometry is achievable with outside vessel temperatures not exceeding 100 °C. A ceramic stable crust was observed at the top surface of the melt pool, which prevented water ingression into the molten corium. The ablation rate was found to be maximum at the lower corners of the brick arrangement with the maximum value being 0.75 mm/s. An average rate of about 0.18 mm/s was obtained in the brick matrix.

INITIAL EVALUATION OF INDIVIDUAL DOSES IN THE EARLY PHASE OF A NUCLEAR REACTOR ACCIDENT BASED ON IN-VIVO MONITORING DATA AND SIMULATED RADIOLOGICAL CONSEQUENCES

In the early phase of a nuclear reactor accident, in-vivo monitoring of impacted population would be highly useful to detect potential contamination during the passage of the cloud and to estimate the dose from inhalation of measured radionuclides. However, it would be important to take into account other exposure components: (1) inhalation of unmeasured radionuclides and (2) external irradiation from the plume and from the radionuclides deposited on the soil. This article presents a methodology to calculate coefficients used to convert in-vivo measurement results directly into doses, not only from the measured radionuclides but from all sources of exposure according to model-based projected doses. This early interpretation of in-vivo measurements will provide an initial indication of individual exposure levels. As an illustration, the methodology is applied to two scenarios of accidents affecting a nuclear power plant: a loss-of-coolant accident leading to core meltdown and a steam generator tube rupture accident.