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At least 505 records · Page 28
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Development of Cr-coated zirconium alloy cladding for enhanced accident tolerance

In the context of the development of Enhanced Accident Tolerant Fuels (EATF), AREVA NP, along with CEA and EDF, is developing, among others, a chromium (Cr) coating (5- 20 μm thick) on zirconium alloy cladding. The coating is deposited on the cladding surface by Physical Vapor Deposition (PVD) and forms a dense and adherent protective layer. It preserves the good mechanical behavior of the current cladding in normal operating conditions while significantly improving the oxidation behavior in accidental conditions. This behavior is being evaluated through out-of-pile corrosion and mechanical tests at the AREVA NP Technical and Research Centers. The corrosion behavior was evaluated through autoclave tests in both PWR water chemistry at 360 and in 415 Celsius degrees steam to increase the corrosion kinetics. In both cases very small weight gain was measured on the order of 1 mg/dm{sup 2} and remains stable throughout the length of the tests. The mechanical behavior of Cr-coated samples was evaluated through tensile tests on Cr-coated and uncoated M5 tube samples at both room temperature and 400 Celsius degrees. The curves demonstrate that the Cr-coated samples exhibit similar mechanical behavior as uncoated samples, and for all the tests performed, the mechanical properties (yield strength 0.2%, ultimate tensile strength, and total elongation) of the Cr-coated samples all fell within the range of values for uncoated samples. In order to obtain in-pile data as early as possible on this solution, an irradiation program (called IMAGO) in the Goesgen reactor will be performed in the years to come. This will be the first irradiation of EATF solutions in a commercial reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Benefits from developments in the field of Decommissioning for Fukushima Daiichi fuel debris retrieval: Remote-Controlled Laser Cutting Process - 17110

Among the numerous challenges towards the decommissioning of Fukushima Dai-ichi reactors, removing fuel debris is certainly one of the tougher ones, requiring several technical developments. One of them, related to laser cutting of fuel debris is being developed by a French consortium 'ONET Technologies, CEA and IRSN'. Remote controlled laser cutting was implemented for the first time in a real decommissioning workshop in 2015, in UP1 reprocessing plant. In addition to laser cutting development in air and underwater, key aspects of fuel debris cutting of representative of corium debris simulants was performed to test the quantities of aerosols, gas and dusts released during the debris laser cutting. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Using FENOSOL{sup TM} Phenolic Foam as a Shock Absorber for Type B Packaging - 17256

Type B and fissile material transportation packages are designed to withstand deformation from dynamic forces experienced during impacts in hypothetical accident conditions. A critical aspect is to ensure the containment system is not breached. The package design normally includes impact limiters made up of a shell filled with energy-absorbing material. Various types of materials are used as energy absorbers, including wood, foam, and honeycomb structures. Synthetic foam is an ideal material for impact limiters; it has isotropic material properties, it acts as a heat insulator during fire conditions, and it maintains structural integrity at high temperatures. In addition, the density of the foam may be controlled during the manufacturing process, allowing the cask designer flexibility to select the right density for each application. The foam properties can be used as inputs for numerical simulations of dynamic impacts. These simulations are a key component of the safety analysis report for Type B transportation casks. Robatel has recently invested in equipment to produce FENOSOL{sup TM} phenolic foam in our factory for use in our cask designs. This acquisition was made jointly between Robatel and the CEA, with Robatel ensuring the industrialization of the product. In the past decade, this foam was successfully integrated into more than 10 European cask designs, and there has been extensive in-laboratory characterization of its properties. FENOSOL{sup TM} has already been accepted by the French Nuclear Safety Authority (ASN) as a shock absorbing material for use in cask designs. Since the acquisition of the phenolic foam production process by Robatel, a new research and development program has been launched to provide more accurate information to our clients, extend the range of application of this foam, and to improve its properties. Robatel has built an extensive database of material characteristics through the use of a universal mechanical tester, a scanning electron microscope, and a nuclear magnetic resonance spectrometer. In addition, Robatel has successfully implemented FENOSOL{sup TM} in two recent cask designs. This included scale model drop testing of a 9 meter drop in three orientations, and a pin puncture drop test. Robatel is in the process of producing FENOSOL{sup TM} samples for testing according to ASTM standards. Various densities will be included in the study. The testing will include duro-meter hardness, specific heat, thermal conductivity, compressive strength and modulus, and tensile strength. Once the testing is complete, Robatel will compile the data in the form of product specification sheets for each density. These information sheets and test reports will be used to validate FENOSOL{sup TM} for use in future impact limiters for the North American Type B cask market. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Irradiation in the Swiss Spallation Neutron Source and its Applications

In high power (∼ 1 MW or higher) spallation targets being operated or developed for different attractive applications such as neutron scattering science and technology or accelerator driven systems (ADS) for nuclear waste transmutation, both target and structural materials are exposed to intensive irradiation of high energy protons and spallation neutrons. Since spallation reactions induced by high energy protons can produce helium (He) and hydrogen (H) and other transmutation elements at very high rates, the behavior of materials used in a spallation irradiation environment can be much different from those observed from materials irradiated with neutrons in fission reactors. Therefore, the damage induced by spallation irradiation cannot be simply simulated by fission neutron irradiation experiments. Seven irradiation experiments have been conducted in the targets of the Swiss Spallation Neutron Source (SINQ) at the Paul Scherrer Institute (PSI) since 1998 within a broad international collaboration with Commissariat a L'energie Atomique Centre d'Etudes de Saclay (CEA), Forschungszentrum Juelich (FZJ), Japan Atomic Energy Agency (JAEA), Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL), University California Santa Barbara (UCSB) and several institutes of Chinese Academy of Sciences. More than 7000 samples from various Al-, Ti-, Fe-, Ni-, Zr-, and W-alloys and SiC/SiC, C/SiC and MAX-phase ceramics composites have been irradiated to doses up to ∼25 dpa (in Fe) and ∼2000 appm He at temperatures up to ∼800 deg. C. In addition, quite many specimens were irradiated in contact with stagnant liquid metals such as Hg, PbBi, PbLi etc. Dedicated post-irradiation examinations (PIE) have been performed, together with the international collaborators. Various mechanical tests including tensile, bend, Charpy-impact, small punch and hardness have been carried out to obtain the mechanical properties of the materials after irradiation. The microstructural changes in the irradiated specimens have been investigated by means of transmission electron microscopy, positron annihilation spectroscopy, atom probe tomography etc. Many novel results have been obtained from the PIE conducted by PSI and other STIP partner laboratories. The DBTT-shift of different FM steels increases more or less linearly with dose after irradiation in the SINQ target, while it saturates at a level of few dpa in fission neutron irradiation cases. This phenomenon can be mostly attributed to helium effects. In addition to the strong embrittlement effect, pronounced hardening effect was also detected in high dose specimens, which was contributed by high density helium bubbles.

36 MATERIALS SCIENCE↗
DOE OSTI2019

French Innovative Thermal Treatment Processes for The Management Radioactive Organic Liquid Waste - 19017

The nuclear industry as well as hospitals or some research laboratories produce Radioactive Organic Liquid Waste (ROLW) such as scintillation cocktails for medical imaging, labeled molecules for medical research or liquids from former activities. In order to be compatible with disposal facilities requirements, this waste must be inerted and stabilized in solid matrix. The route generally used is incineration in the Centraco facilities at the Marcoule site (Gard, France), ashes produced being cemented downstream. However, liquids whose chemical or radiological characteristics are not compatible with the incinerator operating rules (high level of {sup 14}C or halogen content, corrosive gas production etc.) cannot follow this path. The purpose of the MILOR project is then to propose alternative paths to treat all kinds of ROLW. Initiated in September 2017, this project has the technical objective of developing, over a four-year period, two pilot-scale mineralization processes. Two variants of complementary plasma technologies are evaluated: aerial plasma with the 'IDOHL' process and submerged plasma with the 'ELIPSE' process. Compared with other methods, thermal plasmas have unique characteristics such as rapid decomposition with high throughput, fast startup and shutdown, and high-energy delivery. The process called IDOHL consists of a process involving an induction plasma in which the liquids are directly introduced. It has been designed for organic liquids without mineral load and for treatment rates less than 0.5 l/h. This type of plasma allows the destruction of liquid organohalogens without any prior evaporation or treatment. Specific gas treatment systems have been developed to direct the gas produced to a dry or wet gas treatment system. One aim of the project is to subsequently install the process in a radioactive zone at Saclay DRF. The second called ELIPSE consists of a breakthrough process involving a submerged plasma at the heart of which organic liquids are also introduced. Current research has demonstrated that such a technique may enable the instantaneous and complete destruction of liquids with a wide variety of constituents, such as chlorine, fluorine, or phosphorus. The submersion solution offers many advantages: quenching and cleaning of combustion gases; filtering of the particles they contain; and cooling maintained for the entire process, which guarantees excellent corrosion control. An advantage of this type of design is that the gas treatment system can be reduced to a demister-condenser followed by a simple safety filter, thereby offering the additional advantage of an extremely compact treatment system. This design also allows the ELIPSE process to become by this way an embeddable process if required. The MILOR project, financed by the French State through investment program is a tripartite project bringing together Andra, in charge of radioactive waste management, the CEA in charge of leading scientific and technological researches to propose effective ways of treatment and the company A3i whose role is to ensure the subsequent industrial development. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

An IRSN Project For Experimental Data Acquisition In The Frame Of Criticality Safety And Reactor Physics

In the frame of its research missions for criticality safety assessment support, IRSN performed various experimental programs related to criticality safety in the CEA Valduc facilities. Following the shutdown of these facilities and the cancellation of the MIDAS project, which aimed at refurbishing these facilities, IRSN has initiated a new project named PRINCESS for 'Project for IRSN Neutron physics and Criticality Experimental data Supporting Safety'. PRINCESS is the practical implementation of IRSN's international partnership strategy to build relations with organizations that are equipped with experimental facilities for criticality safety and reactor physics research. The aim is to analyze experimental data from programs already performed and to conduct joint research programs on topics of interest of both parties. This paper presents an overview of the status of the PRINCESS project. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Validation Study in SAS4A Code in Simulated Mild TOP Condition

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)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

A burnup credit approach irradiated fast-neutron reactor fuels/ nuclides of interest and fuel storage application - 14248

The concept of taking into account the reduction of the reactivity of nuclear used fuel due to their burnup is referred to as 'Burnup Credit' (BUC). Economic incentives offered by the implementation of a BUC methodology in criticality analyses of Light Water Reactor (LWR) used fuel facilities are nowadays well-demonstrated. The current prospect of reprocessing the fissile and fertile subassemblies of the PHENIX reactor in France, and the development of the 4.-Generation Fast-neutron Reactors (FR) enable research projects in relevant fuel cycle operations. Under these circumstances, the CEA and AREVA-NC have decided to study a burnup credit approach for irradiated fuels of the PHENIX reactor. This paper focuses on a preliminary investigation on the use of burnup credit for used FR fuel operations. The first step consists in analyzing and selecting the nuclides of interest for burnup calculation, on the basis of the 27 BUC nuclides (12 actinides and 15 fissions products) chosen for LWR-MOx fuels [1]. These nuclides are then involved in a criticality calculation (a used fuel pool) to highlight the interest of a BUC approach. This study highlights the interest of a burnup credit approach for FR fuels and shows that the highest contribution to reactivity loss is due to only one fission product: the {sup 149}Sm. The opportunity to take account of only one fission product is highly promising and will considerably simplify the way to determine a penalizing fuel inventory. Moreover, it is to notice that results for different configurations may differ from those obtained for a fuel storage application. Thus, specific analyses have to be carried out in order to study the applicability of the conclusions of this paper. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Joint neutron measurements on the flat-top HEU critical assembly machine - 14388

Flat-Top is a critical assembly with a spherical core (either ∼17 kg of HEU or ∼6 kg of δ-phase Pu) surrounded by a spherical 1000 kg natural uranium reflector. In 2013, joint LANL and CEA measurements were conducted to measure Flat-Top configurations with the HEU core using He-3 detector systems. Similar experiments were previously documented in which the Caliban and Godiva critical assemblies were measured. The core and reflector of Flat-Top have a glory hole in which nuclear material or non-nuclear material may be placed. Five configurations were measured and included subcritical configurations at various reactivity states as well as a configuration at delayed critical. The various reactivity states were achieved by varying the placement of the safety blocks and control rods (composed of natural uranium). A description of the assembly and the specific configurations measured will be presented. The subcritical and delayed critical measurements were analyzed using the Hage-Cifarelli formalism of the Feynman Variance-to-Mean method. This analysis can be used to calculate the multiplication of each configuration, from which the multiplication factor and reactivity are determined. These results are compared to the reactivity states based upon control rod worth curves and MCNP simulations. For this work, a new MCNP model was made to accurately represent the measured configurations. The ICSBEP handbook contains a benchmark of Flat-Top but all models are very simple and do not have the required fidelity to accurately represent the measured configurations. This work also examines the number of neutrons emitted per fission for this assembly (which has a significant contribution of induced fissions in U-238 due to the large natural uranium reflector). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Atalante research facility new mass limits in a unit containing plutonium and minor actinides - 14402

ATALANTE, located in Marcoule, is one of the main Nuclear Facilities of the French Alternative Energies and Atomic Energy Commission (CEA). This facility is mainly dedicated to research and development for nuclear energy related to the back end of the fuel cycle: spent fuel and final waste management. It is a large research complex housing chemistry and analytical laboratories and shielded cells (240 glove boxes and 11 shielded cells), equipped for basic studies through to demonstration experiments. Within the facility, one unit can contain plutonium and curium at the same time. In this case, nuclear criticality safety is achieved by exercising control over mass. A subcritical mass limit is defined for plutonium (280 g for {sup 239}Pu) and an exemption mass limit (3 g for {sup 245}Cm) is used for the actinide. In order to increase the maximum mass (in particular {sup 245}Cm) which can be handled in the unit, new mass control limits were studied, based on control over the sum of the mass ratio to the upper subcritical mass, noted m{sub lim}, for each nuclide (350 g for {sup 239}Pu; 34.8 g or 23 g for {sup 245}Cm) as shown in equation (1) below. m({sup 239}Pu)/ m{sub lim}({sup 239}Pu)+ m({sup 245}Cm)/m{sub lim}({sup 245}Cm)≤1 (1) To validate this criticality safety control method, a study was conducted to demonstrate that, for different masses of {sup 239}Pu and {sup 245}Cm meeting equation (1), the calculated effective multiplication factor is always lower than the effective multiplication factor corresponding to a sphere of 350 g of pure {sup 239}Pu, moderated, with 20 cm water reflection. The two pseudo fissile media are: - metallic plutonium (100% {sup 239}Pu, density 19.86), moderated with water, - metallic curium (100% {sup 245}Cm, density 13.566), moderated with water. The calculation model is a sphere of {sup 245}Cm metal positioned in a sphere of {sup 239}Pu metal, with optimum moderation for each fissile material, and a full water reflection (20 cm). The results show that, under certain conditions, the effective multiplication factor of the mixture can be higher than the effective multiplication factor of each pure fissile material when using the upper subcritical mass of 34.8 g for {sup 245}Cm. When using 23 g of {sup 245}Cm, the k{sub eff} of the mixture is always lower than the k{sub eff} of pure {sup 239}Pu. These calculations enable the safety documentation to be changed by introducing a new process operational procedure, based on equation (1). This change in procedure allows the operator to work with a larger quantity of curium in the unit. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Storage and concentration of plutonium and uranium solutions in LES 401 hot cell (ATALANTE) - criticality safety considerations - 14414

ATALANTE, located in Marcoule, is one of the main nuclear facilities of the French CEA (Atomic Energy and Alternative Energies Commission). The LES 401 hot cell was built in the 1980's and, since then, has been operating only with low masses of fissile material. As part of the management optimization for purified Uranium and Plutonium solutions ('Fin U Pu' Project), the new functions provided in LES 401 are now expected to be the following: - Reception of Uranium and Plutonium solutions, - Concentration of these solutions, - Storage of the concentrates and distillates. From the criticality safety point of view, the equipment was originally designed in safe geometry for Plutonium Nitrate, considering only a 2.5 cm -thick water reflector. However, the new criticality safety case had to take into account the following points: - Reception of not only Plutonium solutions, but also highly enriched Uranium solutions, - Considering flood risk as a possible contingency, - Considering precipitation as a possible contingency. These more conservative assumptions have meant introducing a mass limit for the cell, in addition to the geometry control mode. For the equipment containing very low concentrations of fissile material in normal operation, it was necessary, using the double contingency principle, to take a concentration limit into account in the specific situation of water flooding the cell. Finally, given the mass and geometry control mode, it was necessary to perform all the calculations with both fissile media (Plutonium and Uranium), as it was found that depending on the selected mass, geometry, and reflection conditions, the most penalizing media is not always the same. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

DARWIN2 Low burnup validation for burnup credit using SFCOMPO PIE database - Verification of ICF values - 14433

The DARWIN package, developed by the CEA and its French partners (AREVA and EDF), provides the parameters required for fuel cycle applications, especially fuel inventory. It is widely used in the French Burnup Credit (BUC) calculation route, based on the connection of the depletion code DARWIN and the Criticality-Safety Package CRISTAL. The experimental validation process of BUC nuclides provided by DARWIN2.3 package consists in the comparison between JEFF-3.1.1-based calculated values of isotopic concentrations and experimental values obtained from the chemical analysis of PWR fuel rod cuts. As a complement to the extensive validation using the French PIE database, specific PIE data chosen for their quality were selected and analyzed in the SFCOMPO database of NEA. A set of 56 samples chosen in PWR-UO{sub 2} experiments was selected, with initial {sup 235}U enrichment between 2.7% and 4.1% and local burnup estimated from {sup 148}Nd measurement varying between 7 and 35 GWd/tHM. This paper provides an overview of calculation - experiment comparison obtained in Takahama-3 17x17, Trino-Vercellese 15x15 and Obrigheim 14x1 for uranium, plutonium isotopes, as well as available fission products contributing to the assembly reactivity. Consistent biases compared to SCALE5 results were obtained. From these results, an analysis is driven using 15 selected samples with low burnup, selected among asymptotic rods of irradiated fuels at 1 or 2 cycles. This paper shows the consistency of the calculation - experiment biases obtained with DARWIN2.3 towards previously defined penalized biases used for the Isotopic Correction Factors (ICF) in BUC applications. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

CIRCEE v1 - a versatile tool for criticality accident dose assessment - 14527

CIRCEE is a calculation code developed by the French Atomic Energy Commission (CEA), allowing fast assessment of operational dose over complex geometries. It is used to assess dosimetric consequences of a criticality accident. More specifically, it can be used both during the preliminary studies phase (isodoses curves for safety report preparation or emergency preparedness, CAAS implantation and validation...) and during crisis management (effective dose estimation, emergency response). An early development stage of CIRCEE has been presented in a talk at the 2009 NCSD edition [1]. Since then, an extensive code rewriting increased its reliability and maintainability while keeping its performance at an excellent level. Now in a stable production version, and following the grant of several international patents, it should be available under commercial license terms in the near future. (author)

42 ENGINEERING↗
DOE OSTI2015

2014 CALIBAN and PROSPERO experiments for the criticality accident dosimetry intercomparison - 14529

Nuclear criticality accident dosimetry is very specific, due to the mixed fields and to energy and dose ranges which could be out of the range of classical individual dosimetry. The maximal neutron and gamma dose must be determined as well as other information such as the orientation of the victim in the radiation field. It is exceptionally challenging to establish doses in the event of partial body exposures. Therefore, frequent exercises are organized to maintain competencies and to train staff to perform such dosimetry. Experiments using the CALIBAN and PROSPERO reactors were performed in June and September 2014 in order to test various criticality accident dosimeters (CAD). These two metallic reactors are located at the CEA research center at Valduc (Cote-d'Or, France). They are composed of a metallic alloy of uranium with a high enrichment in the isotope {sup 235}U. These reactors are used for the production of mixed radiation fields primarily comprised of fission neutrons and gammas. CADs from three different laboratories (AWE, UK; IRSN, FR; LLNL, USA) were irradiated on water phantoms with three different orientations (0 deg. and 45 deg., front and rear faces) and free in air at both CALIBAN and PROSPERO reactors, in pulse and steady state modes. This paper will present the comparison of some results. These experiments will be useful preparation for other exercises planned in 2016 at the GODIVA-IV and FLATTOP reactors located at the Nevada National Security Site, USA. (authors)

42 ENGINEERING↗
DOE OSTI2015

Recent activities regarding the use of burnup credit in France - 14539

Burnup credit refers to the concept of taking credit for the reduction in reactivity due to irradiation of nuclear fuel. A working group, bringing together several French nuclear companies and institutions (AREVA, CEA, EDF and IRSN), was created in 1997 to analyze different propositions to introduce a limited number of fission products together with a more realistic axial burnup profile into the criticality safety studies for the nuclear fuel cycle. Initially, the discussions of the working group focused on PWR UOX fuel. The work and accomplishments performed by this working group on burnup credit regarding PWR UOX fuel assemblies has been presented in various publications. This paper gives an outlook of the French working group on burnup credit activities afterwards. These activities relate to industrial applications of burnup credit for PWR UOX fuel, burnup credit implementation for PWR MOX fuel allowing the plutonium recycling in French PWRs, as well as applications for BWR UOX fuel since it may be transported, stored and reprocessed in France. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Status of French regulation regulations concerning nuclear criticality safety - 14585

Hitherto, in France, licensees used to demonstrate compliance with Fundamental Safety Rule I.3.c. This Rule sets out the principles for demonstrating Nuclear Criticality Safety in all nuclear facilities (excluding reactors). Although there is a good level of consensus between the licensees and the Regulator, this Rule is now somewhat dated (it came into force in 1984) and does not cover all aspects. The French Nuclear Safety Authority (ASN) used the 2006 Law on Transparency and Nuclear Safety as the catalyst for a thorough revision and updating of its regulatory documentation. Thus, in 2010, ASN decided to set up a 'criticality working group' involving ASN, the licensees (AREVA, CEA, EDF...) and the Technical Safety Organization (IRSN). This group was tasked with drafting a regulatory text (a 'resolution') and a criticality guide. The objectives were to update Rule I.3.c, clearly defining the scope for reactors and transport packages of fissile material, to further develop areas not covered in the existing rule and to take account of the feedback and lessons learned from past events. Thus, this paper presents the status of French regulations concerning Nuclear Criticality Safety (NCS). Firstly, it focuses on the existing Rule I.3.c and then explains how this will be modified by the forthcoming improvements to the resolution and the criticality guide. It in particular addresses the scope of the new regulations (concerning reactors and the transport of fissile material), how the principle of defence in depth is to be applied to criticality safety, the double contingency principle, control modes, the reference fissile medium, the organization, the role of the Criticality Engineer, admissibility criteria, validation, etc. (authors)

42 ENGINEERING↗
DOE OSTI2015

Experimental validation of the Cristal V2.0 French criticality calculation package - 14587

A new version of the CRISTAL criticality calculation package, named CRISTAL V2.0 [1], has been developed by IRSN and CEA, in collaboration with AREVA and EDF. Four calculation routes are now available, using nuclear data issued from the JEFF-3.1.1 evaluation: two multigroup routes based on 281 groups cross-sections (APOLLO2 - MORET 5 or APOLLO2-Sn calculations), a pointwise Monte Carlo route (TRIPOLI-4{sup R}) and a criticality standard calculation route. The CRISTAL V2.0 package, which will be delivered soon to OECD/NEA data bank, benefits from a broad validation database covering almost all areas of criticality-safety applications, from fuel fabrication to reprocessing, including transportation. 3,127 critical experiments, corresponding to 342 series, either issued from the OECD/ICSBEP Handbook or performed in French facilities (confidential data) were selected for the CRISTAL V2.0 validation database. Currently, more than 2,300 experiments are already part of the validation database. This validation work highlights an overall consistency to experimental data and inter-comparisons between the different calculation routes mainly show good agreements on a large set of common experiments. (authors)

42 ENGINEERING↗
DOE OSTI2022

General overview of multiphysics modeling and simulation at CEA/DES/IRESNE supporting safety, operating and development of nuclear reactors and facilities

This paper presents a general overview of the innovative multiphysics research/development activities carried on by the CEA/DES/IRESNE institute. Eight main different multiphysics fields addressed by IRESNE have been listed (among others: fuel fabrication, behavior of fuels, of reactors, of corium, radionuclide transfer to environment,...) in relation with the involved coupled physics: neutron transport, fluid mechanics, electro-magnetism, heat transfer, solid mechanics and (physico-)chemistry. The major mission of IRESNE institute is to study current and future nuclear systems integrated into a low carbon energy system. To fulfill this mission, IRESNE's activities rely on the development and the implementation of multiphysics modeling, advanced coupling scheme, data transfer techniques and dedicated multiphysics experimentations. Indeed, such multiphysics simulation approach of nuclear systems is today required to deal with the complex intrinsic features of studied phenomena. This approach aims at taking into account physical phenomena interactions, at each interest modeling scale, in order to reach an accurate and predictive behavior representation of a system component or of the whole system. The final objective is to contribute to the development of digital reactor, core and fuel at CEA, by progressing on applied mathematics, high-performance computing and physical mechanism understanding.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗