Hafnium Oxidation at High Temperature in Steam

To assess the potential impact of using hafnium as absorber material in LWRs in high temperature accidental situations, the oxidation behavior of hafnium was studied up to 1400 °C, i.e. at temperature conditions relevant to severe accidents. Different sample geometries were tested and oxidized in steam/argon mixtures, either in a furnace or in a thermogravimetric analyzer. Metallographic examinations, hydrogen measurements and EPMA oxygen profiles were then performed. For hafnium rods/discs, metallographic examinations showed the presence of a dense and protective oxide film after steam oxidation. No or little hydrogen was detected in the metallic part of the rod/disc specimens. The reaction rate can be described by a parabolic law in the tested temperature range in the mid-to-long term, and the value of the effective activation energy determined from the experimental data in steam is in good agreement with the ones published in the literature. The diffusion coefficient of oxygen in hafnium was estimated at each temperature by fitting the experimental oxygen profile obtained on hafnium rods and its temperature dependence is derived in the temperature range 700-1400 °C. The hafnium claddings produced for the application in integral bundle tests exhibited a lower resistance to steam oxidation than hafnium rods/discs. Metallographic examinations showed a non-protective layer and a significant hydrogen amount was picked up by hafnium claddings. Above 800 °C, the oxidation rate for hafnium claddings follows a cubic to quartic law and the effective activation energy was determined in the temperature range 800-1100 °C. These tests highlighted the influence of the surface conditions on the oxidation rate of hafnium in steam. However, hafnium oxidation rate remains well below the oxidation rate of zirconium alloys in the same temperature range

Review of Fukushima Daiichi Nuclear Power Station debris endstate location in OECD/NEA preparatory study on analysis of fuel debris (PreADES) project

Much is still not known about the end-state of core materials in each of the units at Fukushima Daiichi Nuclear Power Station (Daiichi) that were operating on March 11, 2011. The Nuclear Energy Agency of the Organization for Economic Development has launched the Preparatory Study on Analysis of Fuel Debris (PreADES) project as a first step to reduce some of these uncertainties. As part of the PreADES Task 1, relevant information was reviewed to confirm the accuracy of graphical depictions of the debris endstates at the damaged Daiichi units, which provides a basis for suggesting future debris examinations. Two activities have been completed within the PreADES Task 1. First, relevant knowledge from severe accidents at the Three Mile Island Unit 2 and the Chernobyl Nuclear Power Plant Unit 4 was reviewed, along with results from prototypic tests and hot cell examinations, to glean insights that may inform future decommissioning activities at Daiichi. Second, the current debris endstate diagrams for the damaged reactors at Daiichi were reviewed to confirm that they incorporate relevant knowledge from plant observations and from severe accident code analyses of the BSAF (Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station) 1 and 2 projects. This paper highlights Task 1 insights, which have the potential to not only inform future Decontamination and Decommissioning activities at Daiichi but also provide important perspectives for severe accident analyses and management, particularly regarding the long-term management of a damaged nuclear site following a severe accident

CoreSOAR Core Degradation State-of-the Art Report Update: Conclusions [in press]

In 1991 the CSNI published the first State-of-the-Art Report on In-Vessel Core Degradation, which was updated to 1995 under the EC 3rd Framework programme. These covered phenomena, experimental programmes, material data, main modelling codes, code assessments, identification of modelling needs, and conclusions including the needs for further research. This knowledge was fundamental to such safety issues as in-vessel melt retention of the core, recovery of the core by water reflood, hydrogen generation and fission product release. In the last 20 years, there has been much progress in understanding, with major experimental series finished, e.g. the integral in-reactor Phébus FP tests, while others have many tests completed, e.g. the electrically-heated QUENCH series on reflooding degraded rod bundles, and one test using a debris bed. The small-scale PRELUDE/PEARL experiments study debris bed quench, while LIVE examines melt pool behaviour in the lower head using simulant materials. The integral severe accident modelling codes, such as MELCOR and MAAP (USA) and ASTEC (Europe), encapsulate current knowledge in a quantitative way. After two EC-funded projects on the SARNET network of excellence, continued in NUGENIA, it is timely to take stock of the vast range of knowledge and technical improvements gained in the experimental and modelling areas. The CoreSOAR project, in NUGENIA/SARNET, drew together the experience of 11 European partners to update the state of the art in core degradation, finishing at the end of 2018. The review covered knowledge of phenomena, available integral experiments, separate-effects data, modelling codes and code validation, then drawing overall conclusions and identifying needs for further research. The final report serves as a reference for current and future research programmes concerning core degradation in NUGENIA, in other EC research projects such as in Horizon2020 and for projects under the auspices of OECD/NEA/CSNI

Source term evaluation for accident transients in the experimental fusion facility ITER

International audienceWe have studied the transport and chemical speciation of radiotoxic and toxic species for an event of water ingress in the vacuum vessel of experimental fusion facility ITER with the ASTEC code. In particular our evaluation takes into account an assessed thermodynamic data for the beryllium gaseous species. This study shows that deposited beryllium dusts of atomic Be and Be(OH)2 are formed. It also shows that Be(OT)2 could exist in some conditions in the drain tank. © 2015, American Nuclear Society. All rights reserved

Core melt composition at Fukushima Daiichi Results of transient simulations with ASTEC

International audienceThe Accident Source Term Evaluation Code (ASTEC) is used to perform numerical simulations of the accidents at the Fukushima Daiichi nuclear power station in the frame of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) project. At present, simulations are available for Units 1, 2, and 3 of Fukushima Daiichi and for 6 days from the earthquake. A clear lesson from phase 1 of the project was that the uncertainties in the functioning of the safety systems and in accident progression are still large and there are many explanations for the measured thermohydraulic behavior. Rather than focusing on the thermohydraulic key parameters for which comparisons with measurements are available, this paper will address melt composition computation results that may provide insights relevant for the decommissioning process. When molten corium relocates from the core down to the vessel lower head, the melt jets interact with water and may be totally or partially fragmented depending on the level of water. A U-Zr-O-Fe molten pool may form in the lower head, and because of chemical reactions, separation between nonmiscible metallic and oxide phases may occur. The models implemented in ASTEC enable the simulation of these phenomena. Up to five different axisymmetric corium layers in the vessel bottom head can be formed, which are, from bottom to top, a debris layer, a heavy metallic layer, an oxide layer, a light metallic layer, and another debris layer. An important process is the UO2 fuel reduction to metallic uranium by nonoxidized zirconium, which results in uranium transport to the dense metallic layer as demonstrated in the MAterial SCAling (MASCA) program. Because of the large consensus on the accident progression of Fukushima Daiichi Unit 1, in this paper we present complex melt compositions before vessel failure for the current best-estimate cases for Unit 1. We do not present similar work performed for Units 2 and 3. It should be underlined that in the case of vessel bottom failure, a part of this complex melt will relocate to the pedestal and molten core-concrete interaction will take place enhancing other complex physical phenomena with possible large consequences on the melt chemical composition and behavior. © 2016, American Nuclear Society. All rights reserved

Progress on source term evaluation of accidental events in the experimental fusion installation ITER

International audienceThe French "Institut de Radioprotection et de Sûreté Nucléaire" (IRSN) in support to the French nuclear safety authority performs the safety analyses of the ITER experimental installation. We present the progress in the RandD activities related to a better evaluation of the source term in the event of an accident in this installation. These improvements are illustrated by an evaluation of the source term of a LOCA transient with the dedicated ASTEC code. © 2015 Elsevier B.V. All rights reserved

Fuel and fission product behaviour in early phases of a severe accident. Part I Experimental results of the PHEBUS FPT2 test

International audienceOne objective of the FPT2 test of the PHEBUS FP Program was to study the degradation of an irradiated UO2 fuel test section and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 and 900 mm) of the 1-m long test section are presented in this paper. Material interactions leading to local corium formation were identified firstly between fuel and Zircaloy-4 cladding, notably at 823 mm, where the cladding melting temperature was reached, and secondly between fuel and stainless steel oxides. Regarding fission products, molybdenum left so-called metallic precipitates mainly composed of ruthenium. Xenon and caesium behave similarly whereas barium and molybdenum often seems to be associated in precipitates. © 2014 Elsevier B.V. All rights reserved

Progress in understanding fission-product behaviour in coated uranium-dioxide fuel particles

Supported by results of calculations performed with two analytical tools (MFPR, which takes account of physical and chemical mechanisms in calculating the chemical forms and physical locations of fission products in UO2, and MEPHISTA, a thermodynamic database), this paper presents an investigation of some important aspects of the fuel microstructure and chemical evolutions of irradiated TRISO particles. The following main conclusions can be identified with respect to irradiated TRISO fuel: first, the relatively low oxygen potential within the fuel particles with respect to PWR fuel leads to chemical speciation that is not typical of PWR fuels, e.g., the relatively volatile behaviour of barium; secondly, the safety-critical fission-product caesium is released from the urania kernel but the buffer and pyrolytic-carbon coatings could form an important chemical barrier to further migration (i.e., formation of carbides). Finally, significant releases of fission gases from the urania kernel are expected even in nominal conditions. © 2008 Elsevier B.V. All rights reserved