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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

A 12 GHz RF Power Source for the CLIC Study

The CLIC RF frequency has been changed in 2008 from the initial 30 GHz to the European X-band 11.9942 GHz permitting beam independent power production using klystrons for CLIC accelerating structure testing. A design and fabrication contract for five klystrons at that frequency has been signed by different parties with SLAC. France (IRFU, CEA Saclay) is contributing a solid state modulator purchased in industry and specific 12 GHz RF network components to the CLIC study. RF pulses over 120 MW peak at 230 ns length will be obtained by using a novel SLED-I type pulse compression scheme designed and fabricated by IAP, Nizhny Novgorod, Russia. The X-band power test stand is being installed in the CLIC Test Facility CTF3 for independent structure and component testing in a bunker, but allowing, in a later stage, for powering RF components in the CTF3 beam lines. The design of the facility, results from commissioning of the RF power source and the expected performance of the Test Facility are reported.

43 PARTICLE ACCELERATORS↗
DOE OSTI2021

Mapping the scattered family tree of fission neutrons

A statistically predicted tendency for neutrons produced inside fission reactors to form in clusters can cause asymmetrical energy production that is counterbalanced, at least in part, by the spontaneous fission of radioactive material in the reactor. This neutron-clustering effect theory has been demonstrated in a nuclear reactor for the first time and was described in an article published by Los Alamos National Laboratory. The findings of a study supported by the Department of Energy’s Nuclear Criticality Safety Program, funded through the National Nuclear Security Administration, and carried out in collaboration with two French nuclear agencies—the Institute for Radiological Protection and Nuclear Safety (IRSN) and the Atomic Energy Commission (CEA)—could improve reactor safety and lead to more accurate simulations, according to LANL. The team found that a complete die-off was avoided in the small reactor because spontaneous fission—nuclear splitting of radioactive material inside reactors that is not caused by direct impact from a fission neutron—creates more neutrons. The balance of fission chain reactions and spontaneous fission lessens the impact of energy bursts created by clustering neutrons.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2011

Development and experimental validation of a calculation scheme for nuclear heating evaluation in the core of the OSIRIS material testing reactor

The control of the temperature in material samples irradiated in a material testing reactor requires the knowledge of the nuclear heating caused by the energy deposition by neutrons and photons interacting in the irradiation device structures. Thus, a neutron-photonic three-dimensional calculation scheme has been developed to evaluate the nuclear heating in experimental devices irradiated in the core of the OSIRIS MTR reactor (CEA/Saclay Center). The aim is to obtain a predictive tool for the nuclear heating estimation in irradiation devices. This calculation scheme is mainly based on the TRIPOLI-4 three-dimensional continuous-energy Monte Carlo transport code, developed by CEA (Saclay Center). An experimental validation has been carried out on the basis of nuclear heating measurements performed in the OSIRIS core. After an overview of the experimental devices irradiated in the OSIRIS reactor, we present the calculation scheme and the first results of the experimental validation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

CALMOS: Innovative device for the measurement of nuclear heating in material testing reactors

An R and D program has been carried out since 2002 in order to improve gamma heating measurements in the 70 MWth OSIRIS Material Testing Reactor operated by CEA's Nuclear Energy Div. at the Saclay research center. Throughout this program an innovative calorimetric probe associated to a specific handling system has been designed in order to make measurements both along the fissile height and on the upper part of the core, where nuclear heating rates still remain high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for the process validation, while a displacement system has been especially designed to move the probe axially. A final probe has been designed thanks to modeling results and to preliminary measurements obtained with mock-ups irradiated to a heating level of 2W/g, This paper gives an overview of the development, describes the calorimetric probe, and expected advantages such as the possibility to use complementary methods to get the nuclear heating measurement. Results obtained with mock-ups irradiated in ex-core area of the reactor are presented and discussed. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

Jules Horowitz Reactor, a new irradiation facility: Improving dosimetry for the future of nuclear experimentation

Document available in abstract form only, full text of document follows: The Jules Horowitz Reactor (JHR) is an experimental reactor under construction at the French Nuclear Energy and Alternative Energies Commission (CEA) facility at Cadarache. It will achieve its first criticality by the end of 2014. Experiments that will be conducted at JHR will deal with fuel, cladding, and material behavior. The JHR will also produce medical radio-isotopes and doped silicon for the electronic industry. As a new irradiation facility, its instrumentation will benefit from recent improvements. Nuclear instrumentation will include reactor dosimetry, as it is a reference technique to determine neutron fluence in experimental devices or characterize irradiation locations. Reactor dosimetry has been improved with the progress of simulation tools and nuclear data, but at the same time the customer needs have increased: Experimental results must have reduced and assessed uncertainties. This is now a necessary condition to perform an experimental irradiation in a test reactor. Items improved, in the framework of a general upgrading of the dosimetry process based on uncertainty minimization, will include dosimeter, nuclear data, and modelling scheme. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Measurements of actinide-fission product yields in Caliban and Prospero metallic core reactor fission neutron fields

In the 1970's and early 1980's, an experimental program was performed on the facilities of the CEA Valduc Research Center to measure several actinide-fission product yields. Experiments were, in particular, completed on the Caliban and Prospero metallic core reactors to study fission-neutron-induced reactions on {sup 233}U, {sup 235}U, and {sup 239}Pu. Thick actinide samples were irradiated and the number of nuclei of each fission product was determined by gamma spectrometry. Fission chambers were irradiated simultaneously to measure the numbers of fissions in thin deposits of the same actinides. The masses of the thick samples and the thin deposits were determined by mass spectrometry and alpha spectrometry. The results of these experiments will be fully presented in this paper for the first time. A description of the Caliban and Prospero reactors, their characteristics and performances, and explanations about the experimental approach will also be given in the article. A recent work has been completed to analyze and reinterpret these measurements and particularly to evaluate the associated uncertainties. In this context, calculations have also been carried out with the Monte Carlo transport code Tripoli-4, using the published benchmarked Caliban description and a three-dimensional model of Prospero, to determine the average neutron energy causing fission. Simulation results will be discussed in this paper. Finally, new fission yield measurements will be proposed on Caliban and Prospero reactors to strengthen the results of the first experiments. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Sensitivity studies associated with dosimetry experiment interpretation

Document available in abstract form only, full text of document follows: Interpretation of reactor dosimetry experiments with C/E comparison requires precise knowledge of parameters involved in modeling. Some parameters have more weight than others on the calculated values. So, sensitivity studies should be conducted to verify the importance of these parameters. The conclusions of these studies are used to refine the experiment modeling, or to correct uncertainty calculations. The results of these sensitivity studies allow a post-irradiation analysis, which can justify the discarding of some atypical C/M values. Derived uncertainties may be improved by the sensitivity analyses. Beyond classical parameters as geometry or composition, this paper describes some specific sensitivity studies conducted for dosimetry irradiation in reactor, and presents conclusions. These studies are based on dosimeters irradiated in the EOLE reactor facility at Cadarache CEA center. Conclusions drawn from these studies are generic and can be applied to any dosimetry study. Calculations performed for these studies were realized using TRIPOLI-4 Monte Carlo code. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Core characterization of the new CABRI Water Loop Facility

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)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Monte Carlo interpretation of the AMMON/REF experiment in EOLE for the JHR reactor calculations

A new experiment, named AMMON and dedicated to the analysis of the JHR neutron physics and to the qualification of the associated neutron calculation tools, is currently in progress in the EOLE zero-power experimental reactor at CEA Cadarache. The first core configuration, so called 'reference configuration' (AMMON/REF) has already been loaded: it consists of an experimental zone of 7 JHR assemblies with U{sub 3}Si{sub 2} 27% {sup 235}U enriched fuel curved plates surrounded by a driver zone with 623 standard PWR UOx fuel pins. The analysis of the AMMON/REF measurements is presented in this paper. It is based on calculations performed with the 3-dimensional reference Monte Carlo TRIPOLI4.7 code and the JEFF3.1.1 European library: it highlights a good agreement between calculation and experiment concerning reactivity and power distribution in the experimental zone. Reactivity prediction is very satisfactory, despite the presence of a large aluminum quantity in the core: C-E=+ 376 pcm {+-} 290. The radial and axial plate power profiles are also in good agreement. Assembly power distribution is predicted within the experimental uncertainties (1{sigma}). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Sensitivity and uncertainty analysis applied to the JHR reactivity prediction

The on-going AMMON program in EOLE reactor at CEA Cadarache (France) provides experimental results to qualify the HORUS-3D/N neutronics calculation scheme used for the design and safety studies of the new Material Testing Jules Horowitz Reactor (JHR). This paper presents the determination of technological and nuclear data uncertainties on the core reactivity and the propagation of the latter from the AMMON experiment to JHR. The technological uncertainty propagation was performed with a direct perturbation methodology using the 3D French stochastic code TRIPOLI4 and a statistical methodology using the 2D French deterministic code APOLLO2-MOC which leads to a value of 289 pcm (1{sigma}). The Nuclear Data uncertainty propagation relies on a sensitivity study on the main isotopes and the use of a retroactive marginalization method applied to the JEFF 3.1.1 {sup 27}Al evaluation in order to obtain a realistic multi-group covariance matrix associated with the considered evaluation. This nuclear data uncertainty propagation leads to a K{sub eff} uncertainty of 624 pcm for the JHR core and 684 pcm for the AMMON reference configuration core. Finally, transposition and reduction of the prior uncertainty were made using the Representativity method which demonstrates the similarity of the AMMON experiment with JHR (the representativity factor is 0.95). The final impact of JEFF 3.1.1 nuclear data on the Begin Of Life (BOL) JHR reactivity calculated by the HORUS-3D/N V4.0 is a bias of +216 pcm with an associated posterior uncertainty of 304 pcm (1{sigma}). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Innovative power conversion system for the French SFR prototype, ASTRID

In the framework of the French Act of 28 June 2006 about nuclear materials and waste management, the prototype ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), foreseen in operation by the 20's, will have to demonstrate not only the minor actinide transmutation capability, but also the progress made in Sodium Fast Reactor (SFR) technology on an industrial scale, by qualifying innovative options. Some of these options still require improvements, especially in the field of operability and safety. In fact, one of the main issues with the standard steam/water Power Conversion System (PCS) of SFR is the fast and energetic chemical reaction between water and sodium, which could occur in steam generators in case of tube failure. To manage the sodium/water reaction, one way consists in minimizing the impact of such event: hence studies are carried out on steam generator design, improvement of the physical knowledge of this phenomenon, development of numerical simulation to predict the reaction onset and consequences, and associated detection improvement. On the other hand, the other way consists in eliminating sodium/water reaction. In this frame, the CEA contribution to the feasibility evaluation of an alternative innovative PCS (replacing steam/water by 180 bar pressurised nitrogen) is focused on the following main topics: - The parametric study leading to nitrogen selection: the thermodynamic cycle efficiency optimisation on Brayton cycles is performed with several gases at different pressures. - The design of innovative compact heat exchangers for the gas loop: here the key points are the nuclear codification associated with inspection capability, the innovative welding process and the thermal-hydraulic and thermal-mechanic optimisations. After a general introduction of the ASTRID project, this paper presents in detail these different feasibility studies being led on the innovative gas PCS for an SFR. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Use of simplified PSA studies in support of the ASTRID design process

In the framework of the French Act of 28 June 2006 about nuclear materials and waste management, a GEN IV and actinides incineration demonstration prototype is to be commissioned in the 2020 decade. In this objective a prototype called ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) is proposed to demonstrate the progress made in SFR technology at an industrial scale by qualifying innovative options, some of which still remain open in the areas requiring improvements, especially safety and operability. More specifically, we aim for a level of safety that is at least equivalent to that of the EPR (third generation), with improvements made in SFR-specific fields. The integration of safety issues in the early phase of the design of ASTRID is necessarily expected. For this purpose, CEA and its partners AREVA and EDF have planned to perform a level-1 PSA to support and orientate the preliminary design of ASTRID reactor. This paper presents the PSA approach and current studies for the assessment of safety systems and the future work to be done for the 2012-2014 period. The preliminary preparation of PSA studies is presented: objectives and scope of the early design phase PSA, definition of core damage states, selection and grouping of initiating events, assessment of safety functions and related systems. Work under progress is also presented: modelling of event trees, construction of fault trees of safety systems, transient calculations of accident sequences with the CATHARE2 code and reliability data assessment. Main objectives of a level-1 PSA performed at conceptual design stage are an early assessment of the safety architecture of the reactor and findings about the most effective areas for improvement, but also the identification of dominant accident sequences and comparison with alternative designs. After the elaboration of a simplified level-1 PSA model for nominal state and main internal initiators, various design alternatives will be evaluated from the viewpoint of PSA in order to support the design choices of the ASTRID reactor. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

CATHARE calculations of Phenix ultimate natural convection test

The Phenix Sodium cooled Fast Reactor (SFR) started operation in 1973 and it was stopped in 2009. Before the reactor was definitively shutdown, in order to collect experimental data for code assessments in the frame of Generation IV an intensive program of ultimate tests was set up. Among these ultimate experiments, two thermal hydraulic tests were performed: an asymmetrical test consisting in a trip on one secondary pump and a natural convection test in the primary circuit. The natural convection test has been used for an international benchmark on system codes in the frame of the IAEA. The CATHARE code - initially developed for water cooled reactors and now extended for safety analyses for other kinds of reactors, including Sodium Fast Reactor - was used by CEA for this benchmark. The paper reminds briefly the Phenix reactor with the main physical parameters and the instrumentation used during the natural convection test. Main test results are also briefly reminded including the evolution of the core and the heat exchangers inlet and outlet temperatures, and some local temperature measurements. The main developments to perform CATHARE SFR computations and the strategy of system code assessment are presented. Then the CATHARE modelling of Phenix reactor is depicted and the various assumptions are pointed out. CATHARE encountered no problem to predict the initial nominal state. Afterwards, the whole transient scenario is calculated and CATHARE calculations are compared to the Phenix measurements. The global trend is rather well predicted by the CATHARE code. Nevertheless, due to complex flow phenomena occurring in large plena and components, the system code encountered physical limitations, leading to remaining discrepancies between code prediction and plant data. Various sensitivity calculations are presented and they bring partial answers. Additional analyses are in progress to understand more deeply the complex 3D phenomena involved during the different phases of the natural convection test. Additional work for coupling CATHARE system code and TRIO-U CFD code is in progress and will bring useful information to better understand the physical phenomena involved during the natural convection test and to improve system modeling for future SFR safety analysis. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Decontamination Process of Internal Part Pipes - 13442

The Marcoule Site, created in 1955 is one of the first nuclear sites in France. It combines the activities of the Research Centre of the French Atomic Energy Commission (CEA) and AREVA industrial operations. Today, a large part of the operations on this site consists of the cleaning and the dismantling of nuclear Installations, once the end of their life cycle has been reached. An example can be the reprocessing plant UP1. This unit, started in 1958 has been stopped in 1997 and its dismantling started quickly thereafter. Technical challenges of the UP1 dismantling are mainly linked to a very high risk of exposure due to a large variety of contaminated equipments and residuals of fission products, potential sources of irradiation. The dismantling of Hall 71 is a typical example of such challenge. This paper will present a solution developed by AREVA Clean-Up business unit, in collaboration with COFIM Industry, to remove contamination incrusted inside the pipes before starting the cutting operations, thus reducing irradiation risk. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Dismantling of the 904 Cell at the HAO/Sud Facility - 13466

La Hague facility, in France, is the spent fuel recycling plant wherein a part of the fuel coming from some of the French, German, Belgian, Swiss, Dutch and Japanese nuclear reactors is reprocessed before being recycled in order to separate certain radioactive elements. The facility has been successively handled by the CEA (1962-1978), Cogema (1978-2006), and AREVA NC (since 2006). La Hague facility is composed of 3 production units: The UP2-400 production unit started to be operated in 1966 for the reprocessing of UNGG metal fuel. In 1976, following the dropout of the graphite-gas technology by EDF, an HAO workshop to reprocess the fuel from the light water reactors is affiliated and then stopped in 2003. - UP2-400 is partially stopped in 2002 and then definitely the 1 January 2004 and is being dismantled - UP2-800, with the same capacity than UP3, started to be operated in 1994 and is still in operation. And UP3 - UP3 was implemented in 1990 with an annual reprocessing capacity of 800 tons of fuel and is still in operation The combined licensed capacity of UP2-800 and UP3 is 1,700 tons of used fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

A Rinsing Effluent Evaporator for Dismantling Operations - 13271

Between 1958 and 1997, the UP1 plant at Marcoule - located in the south of France - reprocessed and recycled nearly 20,000 MT of used fuel from special defense applications reactors, as well as fuel from the first generation of electricity generating reactors in France (natural uranium fuel, CO{sub 2}-cooled, graphite-moderated). Decommissioning and Dismantling of the UP1 plant and its associated units started in 1998. Since 2005, the UP1 facility has been operated by AREVA as the Marcoule Management and Operation contractor for French Atomic Energy Commission (CEA). An important part of this decommissioning program deals with the vitrification facility of Marcoule. This facility includes 20 tanks devoted to interim storage of highly active solutions, prior to vitrification. In 2006, a rinsing program was defined as part of the tank cleanup strategy. The main objective of the rinsing phases was to decrease activity in order to limit the volume of 'long-life active' waste produced during the decommissioning operations, so the tanks can be dismantled without the need of remote operations. To enable this rinsing program, and anticipating large volumes of generated effluent, the construction of an evaporation unit proved to be essential. The main objective of this unit was to concentrate the effluent produced during tank rinsing operations by a factor of approximately 10, prior to it being treated by vitrification. The evaporator design phase was launched in September 2006. The main challenge for the Project team was the installation of this new unit within a nuclear facility still in operation and in existing compartments not initially designed for this purpose. Cold operating tests were completed in 2008, and in May 2009, the final connections to the process were activated to start the hot test phase. During the first hot test operations performed on the first batches of clean-up effluent, the evaporator had a major operating problem. Extremely large quantities of foam were produced, affecting the evaporator operation, and creating the risk of a reduction in its capacity and throughput performance. A task force of AREVA process, operations, and safety experts from Marcoule and the La Hague reprocessing complex was assembled. New operating parameters were defined and tested to improve the process. Since then, the evaporator has performed very satisfactorily. The foam buildup phenomenon has been brought under complete control. All the different types of effluents produced during cleanup operations have been concentrated, and the results obtained in terms of quality and throughput, have ensured a consistent supply to the vitrification unit. The evaporator was operated until the end of April 2012, and enabled the production of 500 cubic meters of very high activity effluent, concentrating the fission products rinsed from the storage tanks. The evaporator will now be deactivated and decommissioned, with the first rinsing and cleanup operations scheduled to begin in 2014. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

An assessment of the effect on Olkiluoto repository capacity achievable with advanced fuel cycles

Previously a few scenarios have been simulated for transition from thermal to fast reactor fleet in Finland in order to determine how much the transuranic inventory could be reduced with the partitioning and transmutation (P-T) technologies. Those calculations, performed with COSI6 code developed by CEA, are extended in the present study, in which the effect of P-T on the capacity of the planned final disposal repository at Olkiluoto (Finland) is evaluated by taking into account the created fission products and transuranic residuals from the reprocessing operations. The decay heat is assumed to be the most restrictive factor in defining the waste disposal packing density. The repository capacity evaluation of this study is based on the comparison of the decay heats produced by the deposited waste in various scenarios. The reference scenario of this article involves only Light Water Reactors (LWR) in an open fuel cycle. The capacity requirement of the geological repository is estimated in a few closed fuel cycle scenarios, all including actinide transmutation with Fast Reactors (FR). The comparison between the P-T scenarios and reference is based on the decay heat production of the deposited waste. The COSI6 code is used for simulations to provide the repository decay heat curves. Applying the closed fuel cycle would change the disposal concept and schedule, because of which it is not quite straightforward to assess the impact of P-T on the capacity. However, it can be concluded that recycling the transuranic nuclides probably decreases the required volume for the disposal, but thermal dimensioning analysis is needed for more specific conclusions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

New developments and prospects on COSI, the simulation software for fuel cycle analysis

COSI, software developed by the Nuclear Energy Direction of the CEA, is a code simulating a pool of nuclear power plants with its associated fuel cycle facilities. This code has been designed to study various short, medium and long term options for the introduction of various types of nuclear reactors and for the use of associated nuclear materials. In the frame of the French Act for waste management, scenario studies are carried out with COSI, to compare different options of evolution of the French reactor fleet and options of partitioning and transmutation of plutonium and minor actinides. Those studies aim in particular at evaluating the sustainability of Sodium cooled Fast Reactors (SFR) deployment and the possibility to transmute minor actinides. The COSI6 version is a completely renewed software released in 2006. COSI6 is now coupled with the last version of CESAR (CESAR5.3 based on JEFF3.1.1 nuclear data) allowing the calculations on irradiated fuel with 200 fission products and 100 heavy nuclides. A new release is planned in 2013, including in particular the coupling with a recommended database of reactors. An exercise of validation of COSI6, carried out on the French PWR historic nuclear fleet, has been performed. During this exercise quantities like cumulative natural uranium consumption, or cumulative depleted uranium, or UOX/MOX spent fuel storage, or stocks of reprocessed uranium, or plutonium content in fresh MOX fuel, or the annual production of high level waste, have been computed by COSI6 and compared to industrial data. The results have allowed us to validate the essential phases of the fuel cycle computation, and reinforces the credibility of the results provided by the code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗