Synthesis and characterization of thorium-bearing britholites

RADIOCHIn the field of the immobilization of tri- and tetravalent minor actinides, apatites and especially britholites were already proposed as good candidates. In order to simulate tetravalent minor actinides, the incorporation of thorium, through dry chemical routes, was studied in britholite samples of general formula Ca9Nd1−xThx(PO4)5−x(SiO4)1+xF2. The study showed that the incorporation of thorium was effective whatever the thorium reagent used or the grinding conditions considered. Nevertheless, it appeared necessary to use mechanical grinding (30 Hz, 15 min) before heating treatment (T = 1400 °C, 6 h) to improve the reactivity of powders and the sample homogeneity. In these conditions, the incorporation of thorium in the britholite structure occurred above 1100 °C. The heating treatment at 1400 °C led to single phase and homogeneous compounds. This work also underlined the necessity to prefer the coupled substitution Click to view the MathML source instead of (Nd3+, F−) left right double arrow (Th4+, O2−) in order to prepare pure and single phase samples in all the range of composition examined

Oxide phase characterization in simulated high burn-up UO2 fuels in the early stages of a nuclear severe accident

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Determination of Ba, Cs, Mo, Zr and U in SIMFuel samples by ICP-AES and ICP-MS for the study of fission products behavior during a nuclear severe accident.

International audienceThe development of realistic models for fission products behavior during a nuclear severe accident requires experimental data on fission products speciation into the fuel. In this context, several batches of dense UO$_2$ samples containing fission products surrogates under different chemical forms have been prepared and sintered, to be further submitted to thermal treatments in order to characterize fission products speciation under controlled temperature and oxygen potential conditions.To that end, a large number of as-fabricated samples from these experiments were analyzed by ICP-AES and ICP-MS after dissolution. A separation step by liquid chromatography on UTEVA resin was essential before the ICP-AES measurements to overcome the problems of spectral interferences and matrix effects caused by uranium.This study emphasizes the complementarity of these two techniques in nuclear fuel characterization. The advantage of the ICP-AES analysis on simultaneous device is explained in details. The importance of the integrated collision reaction cell in ICP-MS to avoid many problems of polyatomic interferences for the quantification of Cs and Ba is highlighted

Synthesis of the safety studies carried out on the GFR2400

International audienceThe present paper is dedicated to the synthesis of the safety studies carried out on the 2400 MWthgas-cooled fast reactor (GFR2400) concept developed at CEA. The analysis of the reference design basisaccidents investigated up to now, has shown margins up to the acceptance criteria, equal at least to300 ◦C for the category 3 situations and larger than 100 ◦C for the category 4 situations. The dimensioningof the decay heat removal (DHR) loops and of the power conversion system (PCS) loops has beenshown adequate even for bounding degraded situations including multiple failures. Furthermore, in thefollowing part of the paper, it is shown how the main insights provided by a level 1 probabilistic safetyassessment (PSA) carried out at an early stage of the design, have led to reinforce the reliability of theDHR function in high pressure conditions by using the PCS as the first mean to cool the core; in the sametime, on the basis of a combination of deterministic augments and of PSA results, a design simplificationprocess has led to add a low pressure DHR loop to replace a high pressure DHR loop. The last sectionis dedicated to prevention and preliminary study of severe accidents (SA). Four SA families have beenidentified depending on the dynamics and on the scale of the considered accident. The possibility to preventcore degradation by using an adapted accident management (nitrogen injection, use of PCS loops)has been preliminarily shown in several particularly challenging situations (loss of active means, unprotectedtransients, full depressurization). Finally, preliminary results regarding analytical studies carriedout on phenomena involved in GFR2400 core degradation (physico-chemistry and neutron physics) arepresented. Then, the application of the separate results aforementioned by considering results of analyticalsimplified thermalhydraulic calculations and of system calculations (carried out with the CATHARE2code) have enabled a preliminary assessment of GFR2400 behaviour in case of core degradation. For som

Immobilisation of actinides in phosphate matrices

RADIOCHIn the field of the immobilisation of high-activity-level and long-life radwaste (HAVL) for a deep underground repository, several phosphate matrices were already proposed as good candidates to delay the release of actinides in the near-field of such disposal. Among them, thorium phosphate–diphosphate (TPD), monazites/brabantites, britholites, and TPD/monazite compositeswere extensively studied. The synthesis of samples doped with actinides (Th, U...) through wet and dry chemistry methods then their complete characterisation are reported. Their chemical durability is also examined. These materials appear as promising matrices to immobilise tetravalent and/or trivalent actinide

Severe Accident Research activities at the CEA Methodology and Main Insights Related to Source Term Quantification and Fuel Behavior.

International audienceDespite the very high level of control and safety in Nuclear Power Plant, they are not exempt from device malfunction, human errors or even natural disasters, which may lead to nuclear incidents or accidents. When consequences involve a high degradation of the reactor core (melting), associated with the release of Fission Products (FPs) and other radioactive materials from the reactor core, it is classified as a Severe Accident (SA, TMI-2, Chernobyl and Fukushima for instance). These released FPs may be transported by air masses or water and thus cover extensive areas and affect all living beings due to their radiological effect and may as well act as poisons. The amount and isotopic composition of the radioactive material released from the core is called Source Term, and its assessment has been the main objective of several international research programs for more than thirty years. These research programs are commonly classified into two main groups, according to their approach (1) Integral programs, such as PHEBUS FP, studying the response of a whole nuclear core during a severe accident, in a reduced scale; (2) On the other hand, analytical programs studying the fuel and FPs when submitted to accidental conditions, by means of Separate-Effect Tests (SET). Examples of the latter are the HI/VI, VEGA, VERCORS and VERDON programs. As result from all the research programs an extensive experimental database has been generated. However, up to now, predicting correctly the FPs release from UO2 and/or MOX fuels in SA conditions is still a significant and very important challenge since there are many remaining uncertainties. In order to improve these estimations, the global fuel and FPs behavior during the accidental sequence must be better understood, and specific emphasis has to be put on mechanisms which promote FPs release and fuel relocation. One of the most useful ways to do that is to perform appropriated annealing treatments with representative thermal transients in order to measure the absolute level and kinetics of the released FPs. To understand the promoting mechanisms, theses FPs release measurements have to be coupled with the corresponding fuel micro-structural changes resulting from these thermal transient.To this end, since the last decade, CEA has set up two complementary research axes, aiming at reproducing conditions representative of nuclear severe accidents, using both high burn-up irradiated fuel samples and model materials. The first axis corresponds to the VERDON program and deals with commercial UO2 and MOX fuels irradiated in French PWR. Model materials (often called SIMFUELS) consist in natural UO2 doped with stable isotopes of FP in concentrations that match a targeted burn-up. Therefore, SIMFUELS are representative of irradiated nuclear fuels but without their radioactivity. The importance of such materials lies in the possibility of using powerful characterization techniques, such as X-ray Absorption Spectroscopy, which today are unavailable for large samples of irradiated nuclear fuels.The present paper, organized in four main parts, presents successively the experimental facilities available at the CEA Cadarache and Marcoule centers together with the corresponding RetD axes SA experimental VERDON laboratory and associated annealing test device as MERARG, Analytical and micro analysis laboratories by which all the pre- and post-test fuels examinations, supported by analytical development on simulated corium samples, are performedUse of SIMFUELS methodology.The last part of the paper focuses on results obtained with this general approach, with special emphasis on VERDON-1 test

Fission products and nuclear fuel behaviour under severe accident conditionsSpeciation of fission productsin the Verdon 3 and -4 samples

International audienceDuring the irradiation of nuclear fuel in a light water reactor (LWR), fission products (FP) are created and retained in the fuel, modifying its microstructure and properties (thermal conductivity, creep behavior, etc.). In the case of failure of all the three containment barriers, as it could the case during a severe accident, FPs could be released from the fuel to the environment. The chemical nature of FP released from the fuel strongly depends on the scenario and the accident conditions involved (temperature, type of fuel, burn-up, oxygen potential, etc.). Moreover, chemical nature greatly affects FP behavior during their transport in the reactor and hence the source term. FP speciation in the fuel was studied in order to provide experimental data needed to support the speciation mechanisms currently proposed in the literature. Within the framework of the International Source Term Program, the VERDON-3 and-4 tests were devoted to high burn-up MOX fuel behavior and FP releases respectively under oxidizing and reducing conditions at very high temperature (up to 2570 K and 2800 K respectively). Qualitative and quantitative analyses on these fuel samples made it possible to obtain valuable information on fission product behavior in the fuel during the test as a function of the atmosphere. This presentation will provide a series of detailed characterizations including scanning electron microscopy (SEM) and optical microscopic (OM) observations, as well as electron probe microanalyses (EPMA) and Secondary Ion Mass Spectrometry (SIMS) of the VERDON-3 and -4 fuel samples before and after the test