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

A continuous jet changes to droplets where jet breakup occurs. In this study, two-dimensional numerical analysis of jet breakup is performed using the MPS method (Moving Particle Semi-implicit Method) which is a particle method for incompressible flows. The continuous fluid surrounding the jet is neglected. Dependencies of the jet breakup length on the Weber number and the Froude number agree with the experiment. The size distribution of droplets is in agreement with the Nukiyama-Tanasawa distribution that has been widely used as an experimental correlation. Effects of the Weber number and the Froude number on the size distribution are also obtained.

/pdf/numerical-analysis-of-jet-breakup-behavior-using-particle-1lkgna83x0.pdf

Numerical Analysis of Jet Breakup Behavior Using Particle Method

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

Heat transfer—A review of 2003 literature

The official report of the Fukushima Nuclear Accident Independent Investigation Commission : executive summary

Development of fuel-coolant interaction analysis code

In a severe light water reactor accident, aerosol deposition within a reactor coolant piping system could significantly influence the source term because of the reduction of released fission products and the piping structural integrity due to the decay heat emission from the deposited fission products. Aerosol deposition tests have been performed in the Wide Range Piping Integrity Demonstration (WIND) project at the Japan Atomic Energy Research Institute under a contract with the Atomic Energy Bureau in Science and Technology Agency of Japan. In addition to a scoping test, three tests have been conducted to clarify the deposition characteristics of cesium iodide (CsI) aerosols in horizontal straight piping.

Experimental study on aerosol deposition in horizontal straight piping

Source Term Analysis Considering B4C/Steel Interaction and Oxidation During Severe Accidents

It has been pointed out that the reactor coolant system piping could fail prior to the meltthrough of the reactor pressure vessel in a high pressure sequence of pressurized water reactor severe accidents. In order to apply to the evaluation of the piping failure which influences the subsequent accident progression, models for the strength of piping materials at high temperatures were examined. It was found that 0.2% proof stress and ultimate tensile strength above 1,073 K obtained from tensile tests was reproduced by a quadratic equation of the reciprocal absolute temperature. Short-term creep rupture time and minimum creep rate at high temperatures were well correlated by the modified Norton's Law as a function of stress and temperature, which implicitly expressed the effect of the precipitation and the resolution of precipitates on the creep strength. The modified Norton's Law gave better results than the conventional Larson-Miller method. Relating applied stress vs. minimum creep rate and tensile prope...

https://www.tandfonline.com/doi/pdf/10.1080/18811248.2000.9714925

Evaluation of High Temperature Tensile and Creep Properties of Light Water Reactor Coolant Piping Materials for Severe Accident Analyses

The present article is the second part of a two-part article, examining the feasibility of the method proposed in the preceding part. In this method, the ratio of the core uncovered from water x u, the ratio of the core flooded by water x f, the ratio of the core slumped into the pedestal area of the drywell x s, and the ratio of the injected water leaking before reaching the core x wl are the four important uncertain parameters. The base case study shows that water injection via the core spray line is more effective to cool the uncovered core and to reduce the amount of cesium hydroxide (CsOH) released from the drywell. The sensitivity study is conducted by introducing the dimensionless decay heat N qd, which combines the effects of x f, x s, and x wl on the steam generation rate associated with the forced convection cooling in the reactor pressure vessel (RPV). The results show that the temperatures of the uncovered core and the other structures increase with N qd. Consequently, the release rate of CsOH...

A simple method for estimating the structure temperatures and the cesium revaporization inside the reactor pressure vessel – II: Feasibility study for the Fukushima Daiichi Nuclear Power Plant

Yu Maruyama

Papers

Development of fuel-coolant interaction analysis code

Experimental study on aerosol deposition in horizontal straight piping

Source Term Analysis Considering B4C/Steel Interaction and Oxidation During Severe Accidents

Evaluation of High Temperature Tensile and Creep Properties of Light Water Reactor Coolant Piping Materials for Severe Accident Analyses

A simple method for estimating the structure temperatures and the cesium revaporization inside the reactor pressure vessel – II: Feasibility study for the Fukushima Daiichi Nuclear Power Plant