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

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

Investigation of the diffusion of atomic fission products in UC by density functional calculations

Activation energies of U and C atoms self-diffusion in UC, as well as activation energies of hetero-diffusion of fission products (FPs) are investigated by first-principles calculations. According to a previous study which showed a likely U site occupation was favoured for all the FPs, their diffusion is restricted to the uranium sublattice of UC in the present study. In this framework, long-range displacements are only possible through a concerted mechanism with a surrounding uranium vacancy. Using the apparent formation energies of the uranium vacancy defect calculated in our previous study and the classical approach used in UO2 by Andersson et al., the activation energies of the main fission products in the various stoichiometric domains have been calculated. The results are compared to those obtained with the five frequency model applied to two representative fission products, Xe and Zr. Interestingly, despite strong differences of formalism, both models provided similar activation energies. © 2012 Elsevier B.V. All rights reserved

First-principles study of the stability of fission products in uranium monocarbide

International audienceThe incorporation and stability of fission products in uranium monocarbide are studied by means of Density Functional Theory using the generalized gradient approximation and projector-augmented waves method. The computations are performed considering incorporation sites of UC, such as the U, C and interstitial sites, and Schottky defects. The computed incorporation energies are discussed on the basis of the atomic size of the fission products, their chemical environment and the electronic structure. These energies show that all the studied fission products would preferentially occupy the U site. However, incorporation energies do not provide any further information on the fission product location in the case of unavailability of the sites which is why the concept of solution energies is also used. The solution energies obtained confirm that all the fission products are expected to be more stable on a U site of a single uranium vacancy or within a non-bound Schottky defect in equilibrium conditions. © 2012 Elsevier B.V. All rights reserved

A comparison of core degradation phenomena in the CORA, QUENCH, Phébus SFD and Phébus FP experiments

International audienceOver the past 20 years, integral fuel bundle experiments performed at IRSN Cadarache, France (Phébus-SFD and Phébus FP - fission heated) and at Karlsruhe Institute of Technology, Germany (CORA and QUENCH - electrically heated), accompanied by separate-effect tests, have provided a wealth of detailed information on core degradation phenomena that occur under severe accident conditions, relevant to such safety issues as in-vessel retention of the core, recovery of the core by water reflood, hydrogen generation and fission product release. These data form an important basis for development and validation of severe accident analysis codes such as ASTEC (IRSN/GRS, EC) and MELCOR (USNRC/SNL, USA) that are used to assess the safety of current and future reactor designs, so helping to reduce the uncertainty associated with such code predictions. Following the recent end of the Phébus FP project, it is appropriate now to compare the core degradation phenomena observed in these four major experimental series, indicating the main conclusions that have been drawn. This covers subjects such as early phase degradation up to loss of rod-like geometry (all the series), late phase degradation and the link between fission product release and core degradation (Phébus FP), oxidation phenomena (all the series), reflood behaviour (CORA and QUENCH), as well as particular topics such as the effects of control rod material and fuel burn-up on core degradation. It also outlines the separate-effects experiments performed to elucidate specific phenomena such as the impact of chemical reactions involving boron carbide absorber material. Finally, it indicates the remaining topics for which further investigation is still required and/or is under way. © 2014 Elsevier B.V. All rights reserved