[en] Fuel performance is a key factor for minimizing operating costs in nuclear plants. One of the important aspects of fuel performance is fuel rod design, based upon reliable tools able to verify the safety of current fuel solutions, prevent potential issues in new core managements and guide the invention of tomorrow's fuels. AREVA is developing its future global fuel rod code COPERNIC3, which is able to calculate the thermal-mechanical behavior of advanced fuel rods in nuclear plants. Some of the best practices to achieve this goal are described, by reviewing the three pillars of a fuel rod code: the database, the modelling and the computer and numerical aspects. At first, the COPERNIC3 database content is described, accompanied by the tools developed to effectively exploit the data. Then is given an overview of the main modelling aspects, by emphasizing the thermal, fission gas release and mechanical sub-models. In the last part, numerical solutions are detailed in order to increase the computational performance of the code, with a presentation of software configuration management solutions. (authors)

[en] Proper design of fuel rods for demanding operation regarding BU, power history and materials is the prerequisite for safe and economic fuel usage. The availability of reliable tools able to accurately predict the fuel behavior in such severe conditions is therefore of great importance. COPERNIC3 is an advanced fuel rod performance code developed by AREVA. Inspired from the COPERNIC2, CARO-E3 and RODEX4 codes, it incorporates state-of-the-art models largely designed or further developed at AREVA NP. Our goal is to accurately predict fuel behavior with special emphasis on the high duty and high burnup region for PWR and BWR reactors and modern fuel. Demanding fuel operation has relevant effects on fuel behavior which must be adequately modeled. For example the microstructure and the composition of nuclear fuel at high burnup are different from those of the as-fabricated fuel. The changes significantly affect fuel thermal conductivity and porosity, particularly in the pellet rim region which is subjected to restructuring. Also, the migration of the fission gas atoms in the fuel matrix is affected at high burnup. The result is that the fission gas release (FGR) is enhanced at high burnup and the distribution of the fission products between matrix, grain boundaries and rod void volume is drastically changed. Other examples are: - Evolution of radial power and burnup profiles. - Gaseous swelling and increased PCMI. - Fuel densification and swelling. - Helium production and release, especially for commercial-grade MOX. - And other clad specific issues like clad oxidation, fuel rod growth, etc. The paper will present in a first step the main advanced models implemented in COPERNIC3 and their associated databases. Then the code will be appraised by comparing the measurements to the calculated parameters, in the whole range of qualification of the code. Modeling solutions for the high burnup features are presented in the paper. In addition, all other relevant thermal mechanisms that must be properly taken into account and their modeling in COPERNIC3 are described. The mechanical modeling is explained with emphasis on new pellet features including creep, cracking, gaseous swelling and dish filling trends, occurring especially during high duty conditions. The global UO₂ or MOX fuel modeling at high burnup is supported by a huge database including irradiations campaigns conducted in various commercial and experimental reactors. We utilize PWR and BWR commercial fuel data (long rods) irradiated mainly in France, US, Germany, Switzerland, Belgium and Sweden at low and high duty up to maximum rod average burnup of 100 MWd/kgM and segmented rod PWR and BWR fuel with further transient operations in Experimental Reactors (e.g. Studsvik, Osiris, Petten, Mol) and burnups up to 73 MWd/kgM. The thermal modeling is validated through high burnup instrumented irradiation tests (IFA-562 and 597, Extrafort..) carried out in the Halden and Osiris experimental reactors. Various examples of code benchmarking for representative experimental cases will be plotted for relevant key parameters: centerline temperature, FGR, dimensional changes etc. In conclusion, the paper will show that the observed fuel performance behavior is adequately predicted by COPERNIC3 in a wide range of irradiation conditions, for UO₂, Gad fuel and MOX at high burnups or powers, for both PWR and BWR reactors. (authors)

[en] Measurement of the thermophysical properties of liquid metals is challenging because of their high chemical activity and high temperatures. The electromagnetic levitation allows one to hold the electrically conductive liquid sample containerless in an inert atmosphere in thermal equilibrium while measurements on the sample can be taken in a non-contact way followed by extraction of some thermophysical properties. Yet, the electromagnetic forces within the skin layer inside the sample cause convective flow of the liquid thus disabling the data extraction. A static magnetic field imposed over a sample is known to damp the convective flow. With these ideas, an experimental set-up with a DC magnetic field directed perpendicular to the gravity vector was constructed and first experiments were performed with liquid Ni and some other materials. In most of the experiments the instability of the levitated sample during a slow variation of the DC magnetic field was observed which is reported in the present article. (paper)

[en] The ASTRID prototype (Advanced Sodium Technological Reactor for Industrial Demonstration), is foreseen in operation by the 20’s. It will have to demonstrate progresses in Sodium Fast Reactor (SFR) technology at industrial scale, by qualifying innovative options. Some of these options still require development especially in the field of operability and safety. Thus, two power conversion systems (PCS) are investigated in parallel: the steam water Rankine PCS and the Gas Brayton PCS. This paper is placed in the context of a gas PCS with pure nitrogen which is mainly motivated by an enhanced safety and acceptability considerations with the “de facto” elimination of the sodium / water reaction risk. In this gas PCS, the main critical and innovative component is the sodium gas heat exchanger (SGHE) which transfers heat from the secondary sodium loops to the tertiary gas loop. This paper presents the development status of this component through the different topics, i.e. design, material studies, manufacturing process development, thermo-hydraulic and thermo-mechanical program, and a preliminary qualification plan as a conclusion. (author)