. On 11 March 2011, an earthquake occurred about 130 km off the Pacific coast of Japan's main island Honshu, followed by a large tsunami. The resulting loss of electric power at the Fukushima Dai-ichi nuclear power plant developed into a disaster causing massive release of radioactivity into the atmosphere. In this study, we determine the emissions into the atmosphere of two isotopes, the noble gas xenon-133 (133Xe) and the aerosol-bound caesium-137 (137Cs), which have very different release characteristics as well as behavior in the atmosphere. To determine radionuclide emissions as a function of height and time until 20 April, we made a first guess of release rates based on fuel inventories and documented accident events at the site. This first guess was subsequently improved by inverse modeling, which combined it with the results of an atmospheric transport model, FLEXPART, and measurement data from several dozen stations in Japan, North America and other regions. We used both atmospheric activity concentration measurements as well as, for 137Cs, measurements of bulk deposition. Regarding 133Xe, we find a total release of 15.3 (uncertainty range 12.2–18.3) EBq, which is more than twice as high as the total release from Chernobyl and likely the largest radioactive noble gas release in history. The entire noble gas inventory of reactor units 1–3 was set free into the atmosphere between 11 and 15 March 2011. In fact, our release estimate is higher than the entire estimated 133Xe inventory of the Fukushima Dai-ichi nuclear power plant, which we explain with the decay of iodine-133 (half-life of 20.8 h) into 133Xe. There is strong evidence that the 133Xe release started before the first active venting was made, possibly indicating structural damage to reactor components and/or leaks due to overpressure which would have allowed early release of noble gases. For 137Cs, the inversion results give a total emission of 36.6 (20.1–53.1) PBq, or about 43% of the estimated Chernobyl emission. Our results indicate that 137Cs emissions peaked on 14–15 March but were generally high from 12 until 19 March, when they suddenly dropped by orders of magnitude at the time when spraying of water on the spent-fuel pool of unit 4 started. This indicates that emissions may not have originated only from the damaged reactor cores, but also from the spent-fuel pool of unit 4. This would also confirm that the spraying was an effective countermeasure. We explore the main dispersion and deposition patterns of the radioactive cloud, both regionally for Japan as well as for the entire Northern Hemisphere. While at first sight it seemed fortunate that westerly winds prevailed most of the time during the accident, a different picture emerges from our detailed analysis. Exactly during and following the period of the strongest 137Cs emissions on 14 and 15 March as well as after another period with strong emissions on 19 March, the radioactive plume was advected over Eastern Honshu Island, where precipitation deposited a large fraction of 137Cs on land surfaces. Radioactive clouds reached North America on 15 March and Europe on 22 March. By middle of April, 133Xe was fairly uniformly distributed in the middle latitudes of the entire Northern Hemisphere and was for the first time also measured in the Southern Hemisphere (Darwin station, Australia). In general, simulated and observed concentrations of 133Xe and 137Cs both at Japanese as well as at remote sites were in good quantitative agreement. Altogether, we estimate that 6.4 PBq of 137Cs, or 18% of the total fallout until 20 April, were deposited over Japanese land areas, while most of the rest fell over the North Pacific Ocean. Only 0.7 PBq, or 1.9% of the total fallout were deposited on land areas other than Japan.

Xenon-133 and caesium-137 releases into the atmosphere from the Fukushima Dai-ichi nuclear power plant: determination of the source term, atmospheric dispersion, and deposition

Accident tolerant fuel cladding development: Promise, status, and challenges

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

Plasma-facing materials and components in a fusion reactor are the interface between the 
plasma and the material part. The operational conditions in this environment are probably 
the most challenging parameters for any material: high power loads and large particle and 
neutron fluxes are simultaneously impinging at their surfaces. To realize fusion in a tokamak 
or stellarator reactor, given the proven geometries and technological solutions, requires an 
improvement of the thermo-mechanical capabilities of currently available materials. In its 
first part this article describes the requirements and needs for new, advanced materials for the 
plasma-facing components. Starting points are capabilities and limitations of tungsten-based 
alloys and structurally stabilized materials. Furthermore, material requirements from the 
fusion-specific loading scenarios of a divertor in a water-cooled configuration are described, 
defining directions for the material development. Finally, safety requirements for a fusion 
reactor with its specific accident scenarios and their potential environmental impact lead to 
the definition of inherently passive materials, avoiding release of radioactive material through 
intrinsic material properties. The second part of this article demonstrates current material 
development lines answering the fusion-specific requirements for high heat flux materials. 
New composite materials, in particular fiber-reinforced and laminated structures, as well 
as mechanically alloyed tungsten materials, allow the extension of the thermo-mechanical 
operation space towards regions of extreme steady-state and transient loads. Self-passivating

/pdf/development-of-advanced-high-heat-flux-and-plasma-facing-4cdcf31umx.pdf

Development of advanced high heat flux and plasma-facing materials

Current status of materials development of nuclear fuel cladding tubes for light water reactors

Adhesion property and high-temperature oxidation behavior of Cr-coated Zircaloy-4 cladding tube prepared by 3D laser coating

Ever since the Fukushima accident, accident tolerant fuel (ATF) has been widely studied, and a coating technology for a fuel cladding surface has been considered to decrease the high-temperature oxidation rate of zirconium-based alloy In this study, arc ion plating technology was applied to coat a Cr layer on Zircaloy-4 as the oxidation-resistant coating layer for ATF cladding A pure Cr phase was confirmed through an XRD analysis, and Cr film was also found to be homogeneous, dense, and uniform in the cross-sectional observation by SEM Cr-coated Zircaloy-4 was tested in a high-temperature steam environment at 1473 K for 2000 s The high-temperature oxidation tests showed that the Cr layer protected Zircaloy-4 from high-temperature oxidation in the steam The stability of the Cr-coated layer deposited was maintained without peeling after a high-temperature oxidation test, and a Cr2O3 thin oxide layer was observed on the Cr coating layer Furthermore, the TEM results showed that a Cr-Zr mixed layer was formed during the high-temperature oxidation tests Based on the results obtained, the Cr film on Zircaloy-4 successfully acted as a corrosion protective layer, and arc ion plating was shown to be a corrosion protective coating system for ATF claddings

High temperature steam-oxidation behavior of arc ion plated Cr coatings for accident tolerant fuel claddings

The Fukushima accident has had a tremendous impact on Japan and the rest of the world in the areas of public health, economy, and nuclear energy policy. Thus, international consensus has been reach...

Jeong-Yong Park

Papers

Adhesion property and high-temperature oxidation behavior of Cr-coated Zircaloy-4 cladding tube prepared by 3D laser coating

High temperature steam-oxidation behavior of arc ion plated Cr coatings for accident tolerant fuel claddings

KAERI’s Development of LWR Accident-Tolerant Fuel

Ex-reactor corrosion and oxide characteristics of Zr–Nb–Fe alloys with the Nb/Fe ratio

Corrosion behavior and oxide properties of Zr-1.1 wt%Nb-0.05 wt%Cu alloy