This paper presents the results of a research project aiming to optimise the scheduling of activities within a research laboratory of the ‘Commissariat à l'Energie Atomique et aux Energies Alternatives (CEA)’. To tackle this problem, we decompose every activity into a set of elementary tasks to apply standard scheduling methods. We model the problem as an extended version of the Multi-Skill Project Scheduling Problem (MSPSP). As a first approach, we propose a Multi-Skill Project Scheduling Problem with penalty for preemption, along with its mixed-integer/linear programming (MILP) formulation, where the preemption is allowed applying a penalty every time an activity is interrupted. However, the previous approach does not take into account all safety constraints at the facility, and a more accurate variant of the problem is needed. We propose then to integrate the concept of partial preemption to the MSPSP. This concept, that has not been yet studied in the scientific literature, implies that only a subset of resources is released during preemption periods. The resulting MSPSP with partial preemption (MSPSP-PP) is modelled using two methodologies: MILP and constraint programming. Regarding the industrial need of having good solutions in a short time, we also present a greedy algorithm for the MSPSP-PP.

The SPUA (Plutonium Uranium and minor Actinides Service), located in CEA (French Atomic Energy Commission) at Cadarache, is in charge of the elaboration and characterization studies of as-fabricated nuclear fuels, present or future, as well as actinides (Am, Cm) transmutation targets. Conventional techniques (in glove-boxes environment) are used to make uranium and/or plutonium oxide fuels and different equipments are dedicated to their microstructural and thermo-mechanical characterization. GCR (Gas Cooled Reactor) type needs innovating concepts such as new fissile materials (nitrides or carbides) or different fuel geometries (particles or fibers most likely inserted in spherical or hexagonal elements). Consequently, current elaboration and characterization techniques will evolve to fulfill these new requirements. The SPUA is already working on HTR (High Temperature Reactor) particles elaboration and is contributing to new GCR’s fuels choice by studying innovative elaboration processes. At the same time, we are investigating new microstructural characterization methods and different thermal properties determination devices. Thanks to that, we will be able to reach the intrinsic properties of fuel materials before irradiation. These techniques are developed with keeping in mind to be compatible with hot cell’s post irradiative examination (PIE). In the presentation, we shall provide a description of the currently investigated techniques and devices for unirradiated future fuels and their use in hot cells will be discussed.

The LCMl (Laboratorv for characterization of irradiated materials), located in CEA from Saclav. is in charge oi the mechanical tests on iriaiated materials. The dynamic tensiie testing machine, in a hot celi equipped with two remote handlings. has been first improved in 1995. to fulfiil the French safety programs on Reactivity Initiated Accident C A 0 1 .::U SCI .5 <l :11 S ~ 2 . n rr i to oni.3 11 rwt;ua,~cs cropcn, cnla on r..rrrrl L ,:a", c m " ng $,p4 ' l r U e u l i q . m t t , in,: -;u~ o j s ieapmsc .noer H A or .OCA lrios ert cscnq 3n.1 lnc!rnn coroton; Fc' lnt HIA. 1.1 2 illem, 1 2 i l ?g 1 wdr8 r.,'e, .p 1, 5 i' a m nsal ng rd!es .p !o 210.C S . ;n F 1% _OS? 01 CS dnl ACC odi i s . OCA led g ;: sra r i.?lr, U. 10 an0 n c x r g rales u. ?WC s KO. o oi. a00102r .?h: The tensile samples are machined with a spark erosion machine. directiy from pieces of cladding previously defueied. Two kinds of samples can be machined in the ciadding: Axial samples in order to test axial mechanical characteristics Rina samoles in order to test transverse mechanical characteristics, more reoresentative of HIA On one hand. the axial tensile tests were performed using the Jouie effect, and heating rates up to about 500°C.S' were obtained. This enabled us to perform the axiai tests in a satisfactory manner. On the other hand, the tensiie ring tests were first performed in a vertical furnace with a heating rate about 09T .s ' and a thermal stability about 1°C. For temperatures above 480°C. the mechanical characteristics showed a sharp drop which could be attributed to irradiation defect annealing. Therefore we have recently deveiaped an Induction heating system to reach heating rates high enough (200'c.s~') to prevent any significant anneaiing before performing the ring tensile tests. To apply a uniaxiai tangential tension, two matching half-cylinders are inserted inside the ring and are pulled apait. The main objective of this paper is to present this system that can be telemanipulated and achieve heatlng rates up to ~ O O ~ C . ~ . ' while taking into account the requirement forair-cooled coils in the hat celi. The same interface is 40 Aver@ ei ar /lnduclion heating on dynamic lensile rests in CEA Saclay used for mduction heating equipment and Joule effect (current system), in order to control the specimen temperature synchronously with the Iaadlextension, colleCting data from all transducers connected including load/displacement.

It is already 10 years since the (European) High Temperature Reactor Technology Network (HTR-TN) launched a program for development of HTR technology, which expanded through three successive Euratom framework programs, with many projects in line with the network strategy. Widely relying in the beginning on the legacy of the former European HTR developments (DRAGON, AVR, THTR, etc.) that it contributed to safeguard, this program led to advances in HTR/VHTR technologies and produced significant results, which can contribute to the international cooperation through Euratom involvement in the Generation IV International Forum (GIF). the main achievements of the European program, performed in complement to efforts made in several European countries and other GIF partners, are presented: they concern the validation of computer codes (reactor physics, as well as system transient analysis from normal operation to air ingress accident and fuel performance in normal and accident conditions), materials (metallic materials for vessel, direct cycle turbines and intermediate heat exchanger, graphite, etc.), component development, fuel manufacturing and irradiation behavior, and specific HTR waste management (fuel and graphite). Key experiments have been performed or are still ongoing, like irradiation of graphite and of fuel material (PYCASSO experiment), high burn-up fuel PIE, safety test and isotopic analysis, IHX mock-up thermohydraulic test in helium atmosphere, air ingress experiment for a block type core, etc. Now HTR-TN partners consider that it is time for Europe to go a step forward toward industrial demonstration. In line with the orientations of the “Strategic Energy Technology Plan (SET-Plan)” recently issued by the European Commission that promotes a strategy for development of low-carbon energy technologies and mentions Generation IV nuclear systems as part of key technologies, HTR-TN proposes to launch a program for extending the contribution of nuclear energy to industrial process heat applications addressing (1) the development of a flexible HTR that can be coupled to many different process heat and cogeneration applications with very versatile requirements, (2) the development of coupling technologies for such coupling, (3) the possible adaptations of process heat applications required for nuclear coupling, and (4) the integration and optimization of the whole coupled system. As a preliminary step for this ambitious program, HTR-TN endeavors to create a strategic partnership between nuclear industry and R&D and process heat user industries.

The "analytical" PYCASSO (PYrocarbon irradiation for Creep and Swelling/Shrinkage of Objects) irradiations focus on determining the effects of neutron irradiation in the temperature range of 900-1100 degrees C, excluding effects due to the presence of fuel, such as pressurization or chemical attack by fission products. These irradiations can therefore be considered separate effect tests, where only the influence of neutron fluence and temperature on coatings and coating combinations is investigated.For this purpose dedicated particles have been manufactured consisting of surrogate kernels (ZrO(2) and Al(2)O(3)) with different types of PyC/SiC/ZrC coatings and coating combinations. All specimens delivered have been extensively characterized, such that even potentially small changes due to the irradiation in dimensions, microstructure and density can be determined accurately after irradiation.Partners involved in this irradiation are CEA (France), JAEA (Japan) and KAERI (South Korea). The PYCASSO irradiations take place in the High Flux Reactor (HFR) in Petten, and are coordinated by NRG (The Netherlands). The partnership for PYCASSO was initiated by the RAPHAEL (V)HTR European 6th Framework Program and is integrated in the Generation IV International Forum VHTR Fuel and Fuel Cycle project. (C) 2010 Elsevier B.V. All rights reserved.

Within the Raphael (V)HTR 6th framework EU-program, the PYCASSO experiments have been devised to investigate coating behaviour under irradiation. Samples have been included from CEA (France), JAEA (Japan) and KAERI (Republic of Korea), which makes this irradiation a real Generation IV effort. The experiment is a separate effect test, where the influence of fuel (coating corrosion or micro structural change due to fission products), thermal gradients, and variation in coating microstructure and dimensions have been minimized by the use of dummy kernels (Al2O3 and ZrO2), high conductivity particle holder material combined with low energy production of the kernels, and strict (fabrication) quality control and selection procedures respectively. The purpose of the experiment is threefold for the partners involved: - for CEA to determine the behaviour of pyrocarbon under irradiation, especially the interaction of pyrocarbon swelling and creep with SiC coating layers. The results will be used to validate and improve HTR fuel performance modelling. - for JAEA to investigate the behaviour of ZrC coatings, which have been successfully manufactured, but require post-irradiation investigation and characterization. - for KAERI to determine the influence of fabrication of pyrocarbon layers with different densities on the behaviour under irradiation. The paper will go into more detail on the goals to be achieved by the different partners. The PYCASSO-I irradiation is performed in the High Flux Reactor (HFR) in Petten, The Netherlands. The experiment accommodates temperature regions of 900, 1000 and 1100°C, and contains 76 separate particle sample holders. The PYCASSO-I irradiation is a completely new design and will be described in detail, including the route from the concept definition via feasibility studies, fabrication and assembly, up to the irradiation, which took only 1, 5 year. At the time of the conference, the PYCASSO-I irradiation will be finished and a full evaluation of the irradiation will be presented. Additionally, the future post irradiation examination planned for the PYCASSO-I samples and the details of the PYCASSO-II irradiation will be outlined.

Silicon Carbide (SiC) exhibits good thermomechanical resistance and high thermal conductivity at high temperature As it is also compatible with fast neutron spectrum SiC is then a candidate for structural applications in some nuclear reactors of the next generation such as Gas Cooled Reactors However at high temperatures (T>1000 degrees C) the simultaneous effect of stress and irradiation emphasizes the creep of SIC and induces changes in mechanical properties depending on the temperature and the fluence In the scope of a wide program aiming to select materials for these applications ton irradiations have been performed to simulate neutron-induced damage in ceramics and evaluate its consequences on microstructure and mechanical properties The changes in microstructure and mechanical properties of commercial grade of alpha SiC induced by surface irradiations with 95 MeV Xe ions at 400 degrees C are investigated Irradiations resulted in damages affecting a thickness of about 10 mu m for fluences ranging from 30 10(14) to 36 10(15) ions/cm(2) Raman spectroscopy analysis revealed that irradiations produced homonuclear Si-Si bonds and disordered phase of crystalline SiC between 3 10(14) and 36 10(15) ions/cm(2) fluence These microstructural modifications contributed to a macroscopic swelling estimated by measuring the step height between the irradiated and virgin areas Between 3 10(14) and 1 2 10(15) ions/cm(2) of fluence the step height increases from 47 nm to 83 nm then stabilizes with increasing fluence Elastic modulus of alpha SiC did not seem to be significantly affected by irradiation Hardness of alpha SiC exhibited an increase of 15% in the near-surface region of the samples (up to 6 mu m depth at fluence of 36 10(15) ions/cm(2)) then a decrease of 60% at the damage peak range that was attributed to a continuous buried disordered layer Further mechanical tests such as creep and bending tests are planned on irradiated SiC thanks to a dedicated device allowing characterisation of very thin beams

Within the scope of the 5th EURATOM Framework Programme (FP) for the HTR-F and HTR-F1 projects, a new 4-year integrated project on very high temperature reactors (RAPHAEL: ReActor for Process Heat And Electricity) was started in April 2006 as part of the 6th Framework Programme. The Sub-Project on Fuel Technology (SP-FT) is one of eight sub-projects constituting the RAPHAEL project. R&D conducted in this sub-project focuses on understanding fuel behaviour, determining the limits of state-of-the-art fuel, and developing potential performance improvements. Fabrication processes were worked out for alternative fuel kernel composition (UCO instead of UO2) and coating (ZrC instead of SiC): i) UCO microstructure reduces fission product migration and is thus considered superior to UO2 under high burn-ups and high temperature gradients. For this reason, the manufacturing feasibility of UCO kernels using modified external sol-gel routes was addressed. The calcining and sintering steps were particularly studied. ii) For its better high temperature performance, ZrC is a candidate coating material for replacing SiC in TRISO (TRistructural ISOtropic) particles. One of the objectives was therefore to deposit a stoichiometric ZrC layer without impurities. An “analytical irradiation” experiment currently performed in the HFR — named PYCASSO for PYrocarbon irradiation for Creep And Swelling/Shrinkage of Objects — was set up to measure the changes in coating material properties as a function of neutron fluence, with samples coming from the new fabrication process. This experiment was started in April 2008 and will provide data on particle component behaviour under irradiation. This data is required to upgrade material models implemented in the ATLAS fuel simulation code. The PYCASSO irradiation experiment is a true Generation IV VHTR effort, with Korean and Japanese samples included in the irradiation. Further RAPHAEL results will be made available to the GIF VHTR Fuel and Fuel Cycle project partners in the future. Post-irradiation examinations and heat-up tests performed on fuel irradiated in an earlier project are being performed to investigate the behaviour of state-of-the-art fuel in VHTR normal and accident conditions. Very interesting results from destructive examinations performed on the HFR-EU1bis pebbles were obtained, showing a clear temperature (and high burn-up) influence on both kernel changes (including fission product behaviour) and the coating layers. Based on fuel particle models established earlier, the fuel modelling capabilities could be further improved: i) Modelling of fuel elements containing thousands of particles is expected to enable a statistical approach to mechanical particle behaviour and fission product release. ii) A database on historical and new fuel properties was built to enable validation of models. This paper reports on recent progress and main results of the RAPHAEL sub-project on fuel technology.

The thermophysical properties of plutonium–zirconium nitride (0–25at.% Pu), namely heat capacity, thermal conductivity and thermal expansion were measured on pellets produced thanks to the well established pellet pressing route. The experimental results obtained on ZrN were consistent with the existing literature data. The thermal properties measurements reported for (Pu0.25Zr0.75)N were in rather good agreement with predictive values calculated from ZrN and PuN available data.

In the framework of the ANTARES (AREVA New Technology for Advanced Reactor Energy Supply) programme, the Commissariat à l'Energie Atomique (CEA) and AREVA conduct R&D projects covering the mastering of fuel compact fabrication technology with the CAPRI (CEA AREVA PRoduction Integrated) line. Fuel compacts consist of coated particles bonded in a close- packed array by a carbonaceous matrix. This matrix is made by treating a fully graphitized powder (natural graphite and/or needle or pitch coke origin) with a binding agent which is a phenol-formaldehyde resin. The resin bonded matrix has to fulfill a number of essential tasks in the fuel element. Theses tasks are directly influenced by the graphite matrix structure and by the graphite sourcing. A bibliographic review of past irradiation tests performed in the 70's has been done to assess the relationship between graphite matrix materials used and the in pile compact behavior. The goal was to specify compact properties required for the definition of coming SIROCCO irradiation tests in the OSIRIS reactor. This paper presents the results of this bibliographic review. Compact properties leading to acceptable in pile behavior (i.e. dimensional changes, thermal conductivity and coefficient of thermal expansion) are highlighted and the influence of the matrix graphite structure is presented.