Abstract Surface-modified zirconium (Zr)-based alloys, mainly by fabricating protective coatings, are being developed and evaluated as accident-tolerant fuel (ATF) claddings, aiming to improve fuel reliability and safety during normal operations, anticipated operational occurrences, and accident scenarios in water-cooled reactors. In this overview, the performance of Zr alloy claddings under normal and accident conditions is first briefly summarized. In evaluating previous studies, various coating concepts are highlighted based on coating materials, focusing on their performance in autoclave hydrothermal corrosion tests and high-temperature steam oxidation tests. The challenges for the utilization of coatings, including materials selection, deposition technology, and stability under various situations, are discussed to provide some valuable guidance to future research activities.

https://www.degruyter.com/downloadpdf/j/corrrev.2017.35.issue-3/corrrev-2017-0010/corrrev-2017-0010.xml

Protective coatings on zirconium-based alloys as accident-tolerant fuel (ATF) claddings

Three new structural models of montmorillonite with differently distributed Al/Si and Mg/Al substitutions in the tetrahedral and octahedral clay layers are systematically developed and studied by means of MD simulations to quantify the possible effects of such substitutional disorder on the swelling behavior, the interlayer structure, and mobility of aqueous species. A very wide range of water content, from 0 to 700 mgwater/gclay is explored to derive the swelling properties of Cs–montmorillonite. The determined layer spacing does not differ much depending on the clay model. However, at low water contents up to 1-layer hydrate (∼100 mgwater/gclay) the variation of specific locations of the tetrahedral and octahedral substitutions in the two TOT clay layers slightly but noticeably affects the total hydration energy of the system. Using atom–atom radial distribution functions and the respective atomic coordination numbers we have identified for the three clay models not only the previously observed binding ...

Structural Arrangements of Isomorphic Substitutions in Smectites: Molecular Simulation of the Swelling Properties, Interlayer Structure, and Dynamics of Hydrated Cs–Montmorillonite Revisited with New Clay Models

Computational modelling of flash boiling flows: A literature survey

A nearly equal molar ratio AlCrMoNbZr high-entropy alloy (HEA) coating was deposited on N36 zirconium alloy substrates using magnetron co-sputtering technology to enhance the corrosion resistance of light water reactor (LWR) fuel cladding. The microstructure, mechanical properties, surface wettability and corrosion resistance of the AlCrMoNbZr coating were systematically investigated. The X-ray diffraction (XRD) and transmission electron microscopy (TEM) analyses showed that the AlCrMoNbZr HEA coating contained a composite of amorphous and bcc-structured nanocrystals. A nanoindentation test revealed that the coating had a high hardness of 11.8 GPa, and a scratch test indicated that the coating well adhered to the N36 substrate. The contact angle test showed that the static contact angle was 109°, which indicated the good hydrophobic property of the coating. The corrosion measurement showed that the AlCrMoNbZr coating remained effective after it was immersed in static pure water at 360 °C and 18.7 MPa for 30 days, and no N36 substrate oxides formed, which indicated a superior corrosion resistance. Furthermore, the corrosion behaviour of the coating was discussed. The weight gain measurement (8.8 mg/dm2 weight gain) proved that the coating experienced an oxidation process, and during the process, protective Nb2Zr6O17, ZrO2 and Cr2O3 formed on the surface of the coating. Hence, the AlCrMoNbZr HEA coating is a potential candidate material for accident-tolerant fuel (ATF) coatings.

Preparation, structure, and properties of an AlCrMoNbZr high-entropy alloy coating for accident-tolerant fuel cladding

Deposition, characterization and high-temperature steam oxidation behavior of single-phase Ti2AlC-coated Zircaloy-4

An unprecedented earthquake and tsunami struck the Fukushima Dai-ichi Nuclear Power Plants on 11 March 2011. Although extensive efforts have been continuing on investigations into the causes and consequences of the accident, and the Japanese Government has presented a comprehensive report on the accident in the IAEA Ministerial Conference held in June 2011, there is still much to be clarified on what happened during the accident and why. This article aims at identifying what should be clarified further about the progression of the accident at Units 1–3 through the review and analysis of information released from Tokyo Electric Power Company and government authorities. It also discusses the safety issues raised by the accident based on the insights gained, in order to contribute to establishing a new framework that pursues continuous improvement toward the highest standards of safety that can reasonably be achieved.

Insights from review and analysis of the Fukushima Dai-ichi accident

http://www.koreascience.or.kr:80/article/JAKO200931670167703.pdf

Overview of recent efforts through rosa/lstf experiments

Experiments simulating PWR intermediate-break loss-of-coolant accidents (IBLOCAs) with 17% break at hot leg or cold leg were conducted in OECD/NEA ROSA-2 Project using the Large Scale Test Facility (LSTF). In the hot leg IBLOCA test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing (LSC) induced by steam condensation on accumulator coolant injected into cold leg. Water remained on upper core plate in upper plenum due to counter-current flow limiting (CCFL) because of significant upward steam flow from the core. In the cold leg IBLOCA test, core dryout took place due to rapid liquid level drop in the core before LSC. Liquid was accumulated in upper plenum, steam generator (SG) U-tube upflow-side and SG inlet plenum before the LSC due to CCFL by high velocity vapor flow, causing enhanced decrease in the core liquid level. The RELAP5/MOD3.2.1.2 post-test analyses of the two LSTF experiments were performed employing critical flow model in the code with a discharge coefficient of 1.0. In the hot leg IBLOCA case, cladding surface temperature of simulated fuel rods was underpredicted due to overprediction of core liquid level after the core uncovery. In the cold leg IBLOCA case, the cladding surface temperature was underpredicted too due to later core uncovery than in the experiment. These may suggest that the code has remaining problems in proper prediction of primary coolant distribution.

https://www.jstage.jst.go.jp/article/jpes/6/2/6_87/_pdf

RELAP5 Analyses of OECD/NEA ROSA-2 Project Experiments on Intermediate-Break LOCAs at Hot Leg or Cold Leg

The Japan Atomic Energy Agency (JAEA) started OECD/NEA ROSA Project in 2005 to resolve issues in the thermal-hydraulic analyses relevant to LWR safety through the experiments of ROSA/LSTF in JAEA. More than 17 organizations from 14 NEA member countries have joined the Project. The ROSA Project intends to focus on the validation of simulation models and methods for complex phenomena that may occur during DBEs and beyond-DBE transients. Twelve experiments are to be conducted in the six types. By utilizing the obtained data, the predictability of codes is validated. Nine experiments have been performed so far in the ROSA Project to date. The results of two out of these experiments; PV top and bottom small-break (SB) LOCA simulations are studied here, through comparisons with the results from pre-test and post-test analyses by using the RELAP5/MOD3.2 code as a typical and well-utilized/improved best estimate (BE) code. The experimental conditions were defined based on the pre-test (blind) analysis. The comparison with the experiment results may clearly indicate a state of the art of the code to deal with relevant reactor accidents. The code predictive capability was verified further through the post-test analysis. The obtained issues in the utilization of the RELAP5 code are summarized as well as the outline of the ROSA Project.Copyright © 2008 by ASME

RELAP5/MOD3 Code Verification Through PWR Pressure Vessel Small Break LOCA Tests in OECD/NEA ROSA Project

/pdf/numerical-simulation-of-two-phase-critical-flow-with-the-98p2wzo1s4.pdf

Tadashi Watanabe

Papers

Insights from review and analysis of the Fukushima Dai-ichi accident

Overview of recent efforts through rosa/lstf experiments

RELAP5 Analyses of OECD/NEA ROSA-2 Project Experiments on Intermediate-Break LOCAs at Hot Leg or Cold Leg

RELAP5/MOD3 Code Verification Through PWR Pressure Vessel Small Break LOCA Tests in OECD/NEA ROSA Project

Numerical Simulation of Two-Phase Critical Flow with the Phase Change in the Nozzle Tube