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This PhD dissertation focuses on the improvement of very high temperature thermometry (1100 deg. C to 2480 deg. C), with special emphasis on the application to the field of nuclear reactor safety and severe accident research. Two main projects were undertaken to achieve this objective: - The development, testing and transposition of high-temperature fixed point (HTFP) metal-carbon eutectic cells, from metrology laboratory precision (±0.001 deg. C) to applied research with a reasonable degradation of uncertainties (±3-5 deg. C). - The corrosion study and metallurgical characterization of Type-C thermocouple (service temp. 2300 deg. C) prospective sheath material was undertaken to extend the survivability of TCs used for molten metallic/oxide corium thermometry (below 2000 deg. C)
Linac4 accelerator of Centre Europeen de Recherches Nucleaires is under construction and a RF-driven H{sup -} ion source is being developed. The beam current requirement for Linac4 is very challenging: 80 mA must be provided. Cesiated plasma discharge ion sources such as Penning or magnetron sources are also potential candidates. Accelerator ion sources must achieve typical reliability figures of 95% and above. Investigating and understanding the underlying mechanisms involved with source failure or ageing is critical when selecting the ion source technology. Plasma discharge driven surface ion sources rely on molybdenum cathodes. Deformation of the cathode surfaces is visible after extended operation periods. A metallurgical investigation of an ISIS ion source is presented. The origin of the deformation is twofold: Molybdenum sputtering by cesium ions digs few tenths of mm cavities while a growth of molybdenum is observed in the immediate vicinity. The molybdenum growth under hydrogen atmosphere is hard and loosely bound to the bulk. It is, therefore, likely to peel off and be transported within the plasma volume. The observation of the cathode, anode, and extraction electrodes of the magnetron source operated at BNL for two years are presented. A beam simulation of H{sup -}, electrons, and Cs{sup -} ions was performed with the IBSimu code package to qualitatively explain the observations. This paper describes the operation conditions of the ion sources and discusses the metallurgical analysis and beam simulation results.
The CABRI experimental reactor is located at the Cadarache nuclear research center, southern France. It is operated by the Atomic Energy Commission (CEA) and devoted to IRSN (Institut de Radioprotection et de Surete Nucleaire) safety programmes. It has been successfully operated during the last 30 years, enlightening the knowledge of FBR and LWR fuel behaviour during Reactivity Insertion Accident (RIA) and Loss Of Coolant Accident (LOCA) transients in the frame of IPSN (Institut de Protection et de Surete Nucleaire) and now IRSN programmes devoted to reactor safety. This operation was interrupted in 2003 to allow for a whole facility renewal programme for the need of the CABRI International Programme (CIP) carried out by IRSN under the OECD umbrella. The principle of operation of the facility is based on the control of {sup 3}He, a major gaseous neutron absorber, in the core geometry. The purpose of this paper is to illustrate how several dosimetric devices have been set up to better characterize the core during the upcoming commissioning campaign. It presents the schemes and tools dedicated to core characterization. (authors)
We have measured the responses of Fuji MS, SR, and TR imaging plates (IPs) to protons with energies ranging from 0.6 to 3.2 MeV. Monoenergetic protons were produced with the 3.5 MV AIFIRA (Applications Interdisciplinaires de Faisceaux d'Ions en Region Aquitaine) accelerator at the Centre d'Etudes Nucleaires de Bordeaux Gradignan (CENBG). The IPs were irradiated with protons backscattered off a tantalum target. We present the photo-stimulated luminescence response of the IPs together with the fading measurements for these IPs. A method is applied to allow correction of fading effects for variable proton irradiation duration. Using the IP fading corrections, a model of the IP response function to protons was developed. The model enables extrapolation of the IP response to protons up to proton energies of 10 MeV. Our work is finally compared to previous works conducted on Fuji TR IP response to protons.
The High Luminosity (HiLumi) Large Hadron Collider (LHC) project has, as the main objective, to increase the LHC peak luminosity by a factor five and the integrated luminosity by a factor ten. This goal will be achieved mainly with a new interaction region layout, which will allow a stronger focusing of the colliding beams. The target will be to reduce the beam size in the interaction points by a factor of two, which requires doubling the aperture of the low-β (or inner triplet) quadrupole magnets. The use of Nb3Sn superconducting material and, as a result, the possibility of operating at magnetic field levels in the windings higher than 11 T will limit the increase in length of these quadrupoles, called MQXF, to acceptable levels. After the initial design phase, where the key parameters were chosen and the magnet's conceptual design finalized, the MQXF project, a joint effort between the U.S. LHC Accelerator Research Program and the Conseil Europeen pour la Recherche Nucleaire (CERN), has now entered the construction and test phase of the short models. Concurrently, the preparation for the development of the full-length prototypes has been initiated. Lastly, this paper will provide an overview of the project status, describing and reporting on the performance of the superconducting material, the lessons learnt during the fabrication of superconducting coils and support structure, and the fine tuning of the magnet design in view of the start of the prototyping phase.
This paper presents the Material Report dedicated to stainless steels parts and products issued by AFCEN (Association Francaise pour les regles de Conception et de Construction des Materiels des Chaudieres Electro-Nucleaires) in support to RCC-MRx 2010 Code. The RCC-MRx Code is the result of the merger of the RCC-MX 2008, developed in the context of the research reactor Jules Horowitz Reactor project, in the RCC-MR 2007, which set up rules applicable to the design of components operating at high temperature and to the Vacuum Vessel of ITER (a presentation of RCC-MRx 2010 Code is the subject of another paper proposed in this Congress; it explains in particular the status of this Code). This Material Report is part of a set of Criteria of RCC-MRx (this set of Criteria is under construction). The Criteria aim at explaining the design and construction rules of the Code. They cover analyses rules as well as part procurement, welding, methods of tests and examination and fabrication rules. The Material Report particularly provides justifications and explanations on requirements and features dealing with parts and products proposed in the Code. The Material Report contains the following information: Introduction of the grade(s): codes and standards and Reference Procurement Specifications covering parts and products, applications and experience gained, - Physical properties, - Mechanical properties used for design calculations (base metal and welds): basic mechanical properties, creep mechanical properties, irradiated mechanical properties, - Fabrication: experience gained, metallurgy, - Welding: weldability, experience gained during welding and repair procedure qualifications, - Non-destructive examination, - In-service behaviour. In the article, examples of data supplied in the Material Report dedicated to stainless steels will be exposed. (authors)
In this first inaugural joint ESTRO-AAPM session we will attempt to provide some answers to the problems encountered in the clinical application of particle therapy. Indeed the main advantage is that the physical properties of ion beams offer high ballistic accuracy for tightly conformal irradiation of the tumour volume, with excellent sparing of surrounding healthy tissue and critical organs, This also its Achilles' heel calling for an increasing role of imaging to ensure safe application of the intended dose to the targeted area during the entire course of fractionated therapy. We have three distinguished speakers addressing possible solutions. Katia Parodi (Ludwig Maximilians University, Munich, Germany) To date, Positron Emission Tomography (PET) is the only technique which has been already clinically investigated for in-vivo visualization of the beam range during or shortly after ion beam delivery. The method exploits the transient amount of β{sup 2}-activity induced in nuclear interactions between the primary beam and the irradiated tissue, depending on the ion beam species, the tissue elemental composition and physiological properties (in terms of biological clearance), as well as the time course of irradiation and imaging. This contribution will review initial results, ongoing methodological developments and remaining challenges related to the clinical usage of viable but often suboptimal instrumentation and workflows of PET-based treatment verification. Moreover, it will present and discuss promising new detector developments towards next-generation dedicated PET scanners relying on full-ring or dual-head designs for in-beam quasi real-time imaging. Denis Dauvergne (Institut de Physique Nucleaire de Lyon, Lyon, France) Prompt gamma radiation monitoring of hadron therapy presents the advantage of real time capability to measure the ion range. Both simulations and experiments show that millimetric verification of the range can be achieved at the pencil beam scale for active proton beam delivery in homogenous targets. The development of gamma cameras, that has been studied by several groups worldwide over the last years, now reaches - for some of them - the stage of being applicable in clinical conditions, with real size prototypes and count rate capability matching the therapeutic beam intensities. We will review the different concepts of gamma cameras, the advantages and limitations of this method, and the main challenges that should still be overcome before the widespread of prompt gamma quality assurance for proton and hadrontherapy. Jon Kruse (Mayo Clinic, Rochester, MN, USA) Treatment simulation images for proton therapy are used to determine proton stopping power and range in the patient. This talk will discuss the careful control of CT numbers and conversion of CT number to stopping power required in proton therapy. Imaging for treatment guidance of proton therapy also presents unique challenges which will be addressed. Among them are the enhanced relationship between internal anatomy changes and dosimetry, the need for imaging to support adaptive planning protocols, and high operational efficiency. Learning Objectives: To learn about the possibilities of using activation products to determine the range of particle beams in a patient treatment setting To be informed on an alternative methodology using prompt gamma detectors To understand the impact of the accuracy of the knowledge of the patient information with respect to the delivered treatment.
Tokamak Cooling Water System (TCWS) drain tanks (DTs) serve two functions: normal operation and safety operation. Normal DTs are used for regular maintenance operations when draining is necessary. Safety DTs are used to receive the water leaked into the Vacuum Vessel (VV) after an in-vessel loss of coolant accident (LOCA) event. The preliminary design of the DTs shall be based on the information provided by this document. The capacity of the normal DTs is estimated based on the internal volume of in-vessel components [e.g., First Wall/Blanket (FW/BLK) and Divertor (DIV)]; Neutral Beam Injector (NBI) components; and TCWS piping, heat exchangers, electric heaters, pump casing, pressurizers, and valves. Water volumes have been updated based on 2004-design information, changes adopted because of approved Project Change Requests (PCRs), and data verification by US ITER and AREVA Federal Services, the US ITER A and E Company. Two tanks will store water from normal draining operations of the FW/BLK and DIV Primary Heat Transfer Systems (PHTSs). One tank will store water from normal draining operations of the NBI PHTS. The capacity of the safety DTs is based on analysis of a design basis accident: a large leak from in-vessel components. There are two safety DTs that will receive water from a VV LOCA event and drainage from the VV, as needed. In addition, there is one sump tank for the DIV that will be used for collecting drain water from the draining and drying processes and specifically for draining the DIV system as the DIV cassette lines are at a lower elevation than the DT connection point. Information documented in this report must be refined and verified during the preliminary design of the DTs, and there are several aspects to be considered to complete the preliminary design. Input to these design considerations is discussed in this report and includes, but is not limited to, water inventory; operating procedures/maintenance; Failure Modes and Effects Analysis (FMEA); tank layout and dimensions, including design margin; classification under French Nuclear Pressure Directives, Equipements Sous Pression Nucleaires (ESPN); and adaptations for construction.
Uranium electrorefining operations at Y-12 require validation for chlorine absorption and reflection. Plutonium chloride solution operations at Los Alamos National Laboratory require chlorine absorption validation. The growing need for chlorine validation is evident to the wider criticality safety community, with multiple attendees at the recent TEX 2.0 meeting at Lawrence Livermore National Laboratory (May 2023) requesting validation for chlorine (Idaho National Laboratory (INL), Institute de radioprotection et de surete nucleaire (IRSN), Savannah River Nuclear Site (SRNS), LANL, and Y-12). Los Alamos National Laboratory has recently performed, and benchmarked, an experiment titled Chlorine Worth Study (CWS) for the internal operations at the lab, but due to the difficulties in precisely characterizing the material compositions of the chlorine absorbers it is advantageous to perform a complimentary study with a different chlorine-based absorber material. Furthermore, having a uranium-based vs plutonium-based experiment provides a separate and important validation basis for criticality safety and nuclear data evaluation. The original final design report for the study of chlorine absorption using the TEX-HEU experimental base was presented in 2022, but used the same chlorine-bearing materials that were found to be difficult to characterize in the LANL benchmark. A complete redesign of the experiment has been performed looking at alternative absorber materials in various forms to produce an experiment that is fully characterizable. This report presents five novel chlorine experiments using the TEX-HEU test bed which provides direct comparison to the Y-12 and INL/Terrapower application needs, utilizing sodium chloride (NaCl) absorber plates. The absorber plates will consist of granulated NaCl (≥99.5% pure), which will be fully encapsulated in aluminum tins, providing a simple but effective chlorine-based absorber material that can be completely characterized. Of the five configurations presented in this report, it is expected that two or three configurations will be down selected for the actual experiment, with the other configurations being alternates. Three of the configurations are in the standard configuration, where the absorber is placed directly on the HEU fuel plates, and two in the sandwich configuration, where the absorber is surrounded by polyethylene moderators to force additional neutron thermalization prior to reaching the absorber. There are two thicknesses of NaCl absorber plates: 3/16” and 1/4” active thicknesses (i.e. not including the encapsulation). Both variations of the absorbers have an active absorber radius of 6” and a total radius of 7.5” to match the diameter of the HEU plates, with the outer 1.5” being aluminum encapsulation. The high-density polyethylene (HDPE) moderators are of the thicknesses: 27/16”, 7/4”, 1/8”, 11/16”, and 3/4”. The final configurations have six (one sandwich and one standard configuration), eight (one sandwich and one standard configuration), and 18 total fuel layers (standard configuration). The standard and sandwich configurations were designed such that the differences in moderator thicknesses are 1” HDPE, which were already procured for the original CED-2. The proposed configurations were precisely tuned to closely match the sensitivity profiles and neutron spectra of the Y-12 upset cases and were also compared to the INL/Terrapower upset cases. The assessment of experimental uncertainties of the non-absorber components was predicted to be 0.00114 Δk eff . The assessment of uncertainties resulting from the absorbers was predicted to be 0.00029 Δk eff . This results in a total uncertainty of 0.00118 Δk eff . Many of the largest uncertainties, namely the moderator densities, may be reduced with precision dimensional inspection of the components. The 1” HDPE moderators as well as the HDPE reflectors from the original CED-2 were incorporated in the final designs presented here. Additional HDPE moderator plates must be fabricated to complete the configurations. NaCl absorber plates will by fabricated at LLNL. The total additional cost is expected to be $\$$54,250 for the remaining components. Precise inspection, including dimensional, mass, density, and impurity, is recommended for all components. LLNL estimates that these costs are around $\$$12,000. It is expected, based on previous TEX-HEU experiments, that three weeks of experimental facility time is needed to complete the experiments.
The scope of this paper is the utilization of inheritance for requirements specification, i.e., the tasks of analyzing and modeling the domain, as well as forming and defining requirements. Our approach and the tool supporting it are named RETH (Requirements Engineering Through Hypertext). Actually, RETH uses a combination of various technologies, including object-oriented approaches and artificial intelligence (in particular frames). We do not attempt to exclude or replace formal representations, but try to complement and provide means for gradually developing them. Among others, RETH has been applied in the CERN (Conseil Europeen pour la Rechereche Nucleaire) Cortex project. While it would be impossible to explain this project in detail here, it should be sufficient to know that it deals with a generic distributed control system. Since this project is not finished yet, it is difficult to state its size precisely. In order to give an idea, its final goal is to substitute the many existing similar control systems at CERN by this generic approach. Currently, RETH is also tested using real-world requirements for the Pastel Mission Planning System at ESOC in Darmstadt. First, we outline how hypertext is integrated into a frame system in our approach. Moreover, the usefulness of inheritance is demonstrated as performed by the tool RETH. We then summarize our experiences of utilizing inheritance in the Cortex project. Lastly, RETH will be related to existing work.
An experiment designed to compare neutron noise analysis techniques on subcritical assemblies using differing neutron detector designs was developed as a collaborative effort between the Institut de radioprotection et de surete nucleaire (IRSN), Los Alamos (LANL), Lawrence Livermore (LLNL), and Sandia National Laboratories. This experiment has the final goal of characterizing neutron noise analysis accuracy in determining multiplication in light water systems for various neutron detection systems. This experiment utilized the Seven Percent Critical Experiment (7uPCX) facility at Sandia National Laboratories. The 7uPCX is an assembly designed to investigate light water systems with fuel enrichment above 5% and is able to use square or hex grid plates with varying pitches.
The CEA is involved in a substantial effort on the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) pre-conceptual design in cooperation with EDF, as experienced Sodium-cooled Fast Reactor (SFR) operator, AREVA, as experienced SFR Nuclear Island engineering company and components designer, ALSTOM POWER as energy conversion system designer and COMEX NUCLEAIRE as mechanical systems designer. The CEA is looking for other partnerships, in France and abroad. The ASTRID preliminary design is based on a sodium-cooled pool reactor of 1500 MWth generating about 600 MWe, which is required to guarantee the representativeness of the reactor core and the main components with regard to future commercial reactors. ASTRID lifetime target is 60 years. Two Energy Conversion Systems are studied in parallel until the end of 2012: Rankine steam cycle or Brayton gas based energy conversion cycle. ASTRID design is guided by the following major objectives: improved safety, simplification of structures, improved In Service Inspection and Repair (ISIR), improved manufacturing conditions for cost reduction and increased quality, reduction of risks related to sodium fires and water/sodium reaction, and improved robustness against external hazards. The core is supported by a diagrid, which lay on a strong back to transfer the weight to the main vessel. AREVA is involved in a substantial effort in order to improve the core support structure in particular regarding the ISIR and the connection to primary pump. In the preliminary design, the primary system is formed by the main vessel and the upper closure comprising the reactor roof, two rotating plugs - used for fuel handling - and the components plugs located in the roof penetrations. The Above Core Structure deflects the sodium flow in the hot pool and provides support to core instrumentation and guidance of the control rod drive mechanisms. The number of the major components in the main vessel, primary pumps, Intermediate Heat Exchangers, and Decay Heat Exchangers are now under consideration. Under normal conditions, power release is achieved using the steam/water plant (in case of Rankine steam cycle) or the gas plant (in case of Brayton gas cycle). The diverse design and operating modes of Decay Heat Removal systems provide protection against common cause failures. A Decay Heat Removal system through the reactor vault is in particular studied with the objective to complement Direct Reactor Cooling systems. At this stage of the studies, the secondary system comprises four independent sodium loops (two and three sodium loops configurations are also investigated). Each loop includes one mechanical pump (or a large capacity Annular Linear Induction Electromagnetic Pump), and three modular Steam Generator Units characterized by once through straight tube units with a ferritic tube bundle; nevertheless, helical coil steam generator with tubes made of Alloy 800, and inverted type steam generator with a ferritic tube bundle are also investigated. The limited power of each modular Steam Generator Unit allows the whole secondary loop to withstand a large water/sodium reaction consecutive to the postulated simultaneous rupture of all the heat exchange tubes of one module. The arrangement of the components is based on the 'Regain' concept, in which the secondary pump is situated at a low level in the circuit; conventional arrangement, as SUPERPHENIX type, is a back-up option. Alternative arrangements based on gas cycles are also studied together with Na-gas heat exchanger design. This paper presents a status of the ASTRID pre-conceptual design. The most promising options are highlighted as well as less risky and back-up options. (authors)
In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. The reflooding (injection of water into core) may be applied if the availability of safety injection is recovered during accident. If the injection becomes available only in the late phase of accident, water will enter a core configuration that will differ significantly from original rod-bundle geometry. Any attempt to inject water after significant core degradation can lead to further fragmentation of core material. The fragmentation of fuel rods may result in the formation of a 'debris bed'. The typical particle size in a debris bed might reach few millimeters (characteristic length-scale: 1 to 5 mm), i.e., a high permeability porous medium. The French 'Institut de Radioprotection et de Surete Nucleaire' is developing experimental programs (PEARL and PRELUDE) and simulation tools (ICARE-CATHARE and ASTEC) to study and optimize the severe accident management strategy and to assess the probabilities to stop the progress of in-vessel core degradation. It is shown that the quench front exhibits either a ID behaviour or a 2D one, depending on injection rate or bed characteristics. The PRELUDE experiment covers a rather large range of variation of parameters, for which the developed model appears to be quite predictive. (authors)
The spectrometer ATHENA (Accurate Telescope for High Energy Neutron metrology Applications), is being developed at the IRSN / LMDN (Institut de Radioprotection et de Surete nucleaire / Laboratoire de Metrologie et de dosimetrie des neutrons) and aims at characterizing energy and fluence of fast neutron fields. The detector is a Recoil Proton Telescope and measures neutron fields in the range of 5 to 20 MeV. This telescope is intended to become a primary standard for both energy and fluence measurements. The neutron detection is achieved by a polyethylene radiator for n-p conversion, three 50{sub m} thick silicon sensors that use CMOS technology for the proton tracking and a 3 mm thick silicon diode to measure the residual proton energy. This first prototype used CMOS sensors called MIMOSTAR, initially developed for heavy ion physics. The use of CMOS sensors and silicon diode increases the intrinsic efficiency of the detector by a factor of ten compared with conventional designs. The first prototype has already been done and was a successful study giving the results it offered in terms of energy and fluence measurements. For mono energetic beams going from 5 to 19 MeV, the telescope offered an energy resolution between 5 and 11% and fluence difference going from 5 to 7% compared to other home standards. A second and final prototype of the detector is being designed. It will hold upgraded CMOS sensors called FastPixN. These CMOS sensors are supposed to run 400 times faster than the older version and therefore give the telescope the ability to support neutron flux in the order of 107 to 108cm{sup 2}:s{sup 1}. The first prototypes results showed that a 50 m pixel size is enough for a precise scattering angle reconstruction. Simulations using MCNPX and GEANT4 are already in place for further improvements. A DeltaE diode will replace the third CMOS sensor and will be installed right before the silicon diode for a better recoil proton selection. The final prototype with its new geometry will increase the telescopes efficiency by a factor of 1.5. It will also cover some of the most important points in metrology; repeatability, reproducibility and sustainability. (authors)
This present work deals with a collaborative program called GAMMA-MAJOR 'Development and qualification of a deterministic scheme for the evaluation of GAMMA heating in MTR reactors with exploitation as example MARIA reactor and Jules Horowitz Reactor' between the National Centre for Nuclear Research of Poland, the French Atomic Energy and Alternative Energies Commission and Aix Marseille University. One of main objectives of this program is to optimize the nuclear heating quantification thanks to calculation validated from experimental measurements of radiation energy deposition carried out in irradiation reactors. The quantification of the nuclear heating is a key data especially for the thermal, mechanical design and sizing of irradiation experimental devices in specific irradiated conditions and locations. The determination of this data is usually performed by differential calorimeters and gamma thermometers such as used in the experimental multi-sensors device called CARMEN 'Calorimetric en Reacteur et Mesures des Emissions Nucleaires'. In the framework of the GAMMA-MAJOR program a new calorimeter was designed for the nuclear energy deposition quantification. It corresponds to a single-cell calorimeter and it is called KAROLINA. This calorimeter was recently tested during an irradiation campaign inside MARIA reactor in Poland. This new single-cell calorimeter differs from previous CALMOS or CARMEN type differential calorimeters according to three main points: its geometry, its preliminary out-of-pile calibration, and its in-pile measurement method. The differential calorimeter, which is made of two identical cells containing heaters, has a calibration method based on the use of steady thermal states reached by simulating the nuclear energy deposition into the calorimeter sample by Joule effect; whereas the single-cell calorimeter, which has no heater, is calibrated by using the transient thermal response of the sensor (heating and cooling steps). The paper will concern these two kinds of calorimetric sensors. It will focus in particular on studies on their out-of-pile calibrations. Firstly, the characteristics of the sensor designs will be detailed (such as geometry, dimension, material sample, assembly, instrumentation). Then the out-of-pile calibration methods will be described. Furthermore numerical results obtained thanks to 2D axisymmetrical thermal simulations (Finite Element Method, CAST3M) and experimental results will be presented for each sensor. A comparison of the two different thermal sensor behaviours will be realized. To conclude a discussion of the advantages and the drawbacks of each sensor will be performed especially regarding measurement methods. (authors)
The experience feedback shows that the complete dismantling of one Light Water Reactor (LWR) unit is an activity now globally well under control (reactor pressure vessel and its internals removal and cutting included). However some specific industrial difficulties may arise in view of the dismantling of numerous LWR units simultaneously. Indeed more than 300 LWR units are currently operated worldwide, which have been commissioned these last 50 years. So in some countries, many units may be permanently shut down over a period of few years in the next decades. Such a situation addresses the issue of the overall management of large quantities of spent nuclear fuel (SNF) and decommissioning waste for the concerned units. For example, in France, the legislative and regulatory framework for the nuclear facilities favors their dismantling 'as soon as possible' after their permanent shutdown which implies as well to limit the duration of the transition period from operation to decommissioning. Furthermore 58 LWR units have been commissioned between 1977 and 1999 in France - on average more than 2 units per year. In this context, the operator (Electricite de France - EDF) plans to remove quickly the SNF then to perform dismantling actions immediately after the permanent shutdown of the LWR units. One issue is to remove the SNF from all the relevant units, even if this removal is simultaneous in many units (permanently shut down or still under operation). Similar issue has to be taken into account regarding the management of the radioactive waste (RW) produced by the dismantling and clean-up actions, notably the RW that cannot be disposed of in a near surface repository. One method followed by the French technical support organization (Institut de Radioprotection et de Surete Nucleaire - IRSN) to analyze these issues is the use of estimates of flows of SNF and RW. These estimates are based notably on a phasing-scenario and a planning template defined for the decommissioning of the units of one nuclear power plant (NPP) and coupled to an overall schedule for phase out all the units of the fleet. The estimates relative to the next decades can be compared to the current experience feedback of flows of SNF and RW for units under operation, in order to identify risks when facing decommissioning. The risks highlighted are driven by key parameters (as duration of the main dismantling actions) of the estimates which can be adapted to minimize their impact. On this basis, it is possible to identify the key-factors to dismantle each unit of NPPs and phase out the fleet regarding SNF and RW management. It is noteworthy that this work needs to be done in any case upstream the studies and the implementation of dismantling actions. Finally it can be underlined that another issue is the human resources (staff, skills and knowledge) necessary to perform all the decommissioning actions, but this aspect is not addressed in the present paper. (authors)
Model improvements and validation efforts for SAS4A code, which is the most advanced computer code for simulating the the primary phase of the Core Disruptive Accident (CDA) of MOX-fueled Sodium-cooled Fast Reactors, have been performed utilizing the experimental data from the CABRI programs by JAEA (Japan Atomic Energy Agency), IRSN (Institut de Radioprotection et de Surete Nucleaire), CEA (Commissariat a l'Energie Atomique et aux Energies Alternatives) and KIT (Karlsruhe Institute of Technology). In order to confirm validity of improved models of SAS4A, a systematic and comprehensive validation plan was defined. It was based on the PIRT (Phenomena Identification and Ranking Table) result for CDA of MOX-fueled fast reactors. The fuel failure position and time are the most important parameters, which dominate the analytical results in a reactor calculation. The failure position will affect the reactivity feedback due to the fuel distribution after failure and consequently would have an effect on the energy release in the initiating phase. This paper describes the preliminary result of validation study for SAS4A code in simulated Transient Over-Power (TOP) condition. (authors)