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[en] This publication gathers Power Point presentations and texts of contributions to a meeting. The contributions contains activity reports, presentations of works and researches, lists of publications related to different programs, experiments and projects. They more particularly address the issue of nuclear data within the IN2P3 (general objectives of studies on nuclear data; study framework within the IN2P3; presentation of topics such as production of major and minor actinides, core reactivity and regeneration performance, gamma emission, residual heat and radiotoxicity, benchmarks and data validation, transverse needs such as uncertainty management, modelling and assessment; required tools), accelerators for the ADSs or Accelerator Driven Systems (requirements regarding beams, concept of ADS accelerator, R and D activities on the injector and on the main linac), physics experiments and reactor technology (reactor physics for Accelerator Driven Systems with the MUSE program and the GUINEVERE and FREYA projects; the FFFER experiment for molten salt reactors), the n2EDM project for the measurement of the neutron dipole moment, the nEDM experiment (physics motivations, status of the PSI UCN source, status of the running EDM experiment in terms of systematics and statistical sensitivity), assessment and prospective for systems studies and scenarios (study of innovative systems for the uranium-plutonium cycle and transmutation and for the thorium-uranium cycle, development of methods, tools and computing codes for neutronics and fuel evolution, for safety and for uncertainty propagation, study of scenarios for the nuclear fuel, for the thorium cycle, for proliferation, for technical-economic aspects), an analysis of prompt decay experiments for ADS reactivity monitoring at Venus-F facility, reactivity monitoring using the area method for the subcritical venus-F core within the framework of the FREYA project, proposal for an IN2P3 contribution to the n2EDM project at PSI.
[en] In order to develop alternative fabrication method of fuel slugs for preventing the evaporation of volatile elements such as AM,U-10wt.%Zr-Mn fuel slugs for SFR (: Sodium-cooled Fast Reactor) have been fabricated by modified casting method and characterized to evaluate the feasibility of the alternative fabrication. Metal fuels such as U-Zr/U-Pu-Zr alloys have been considered as a nuclear fuel for a sodium-cooled fast reactor (SFR) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980's In order to develop innovative fabrication method of metal fuel slugs for preventing the evaporation of volatile elements such as Am, modified casting under inert atmosphere has been applied for metal fuel slugs for SFR. Alternative fabrication method of fuel slugs has been introduced to develop an improved fabrication process of metal fuel for preventing the evaporation of volatile elements. In this study, U- 10wt.%Zr-Mn fuel slugs for SFR have been fabricated by modified casting method and characterized to evaluate the feasibility of the alternative fabrication method. Alternative casting such as modified casting has been applied to develop fabrication method of fuel slugs for preventing the evaporation of volatile elements such as Am. U-10wt.%Zr-Mn containing a volatile surrogate Mn fuel slugs was soundly fabricated under inert atmosphere with dimensions of Φ5 - L 250 mm. Mass fraction of fuel loss was so low, up to 0.2%. Mn element was most recovered with prevention in evaporation of Mn. It was seen that the losses of volatile Am can be effectively controlled to below detectable levels using modest pressure. (authors)
[en] There are several options of nuclear fuel utilisation in the HTGR-based Experimental Power Reactor (Reaktor Daya Eksperimental / RDE). Although mainly RDE utilises low enriched uranium (LEU)-based fuel, which is the most viable option at the moment, it is possible for RDE to utilise other fuel, for example thorium-based and possibly even plutonium-based fuel. Different fuel yields different spent fuel characteristics, so it is necessary to identify the characteristics to understand and evaluate their handling and interim storage. This paper provides the study on the characteristics of thorium-fuelled RDE spent fuel, assuming typical operational cycle. ORIGEN2.1 code is employed to determine the spent fuel characteristics. The result showed that at the end of the calculation cycle, each thorium-based spent fuel pebble generates around 0,627 Watts of heat, 28 neutrons/s, 8.28 x 1012 photons/s and yield 192.53 curies of radioactivity. These higher radioactivity and photon emission possibly necessitate different measures in spent fuel management, if RDE were to use thorium-based fuel. Tl-208 activity, which found to be emitting potentially non-negligible strong gamma emission, magnified the requirement of proper spent fuel handling especially radiation shielding in spent fuel cask. (author)
[en] The paper describes studies that have been undertaken to analyse the mass inventory of the thorium-uranium mixed oxide fuel removed from the standard CANDU 6 reactor. The analysed T37 fuel bundle configuration is based on thorium and low enriched uranium (LEU) oxide, considering the ThO2 ratio of 100% in the central pin, 60% in the inner and intermediate rings and 80% in the outer ring. The 235U enrichment has been varied in the radial direction from 6% to 10%. The study focuses on identifying optimal thorium - uranium combination from the viewpoint of reducing the content of plutonium and highly radiotoxic minor actinides content the spent fuel accumulated in 30 years operating reactor. The lattice simulations have been performed using the Continuous-energy Monte Carlo Reactor Physics Burn-up Calculation Code - SERPENT 2, version 2.1.30. In order to perform the depletion calculations, the ACE format cross section library JEFF-3.1.1 have been used. (author).
[en] Automatic translation: 1. Summary: The existing uncertainty of supply with highly enriched uranium requires a back-up solution for the U-Th fuel cycle (HEU). A medium-enriched cycle in the burn-up-brood system has the most advantages for this on, deviates the least from the existing concept and uses the HTR potential not by switching to this cycle in the meantime. The at Fuel element manufacturers allow existing experience, the XVI. Manufacture AVR refill batch with fuel elements from the back-up cycle. Nuclear bills for the AVR showed that the desired fuel element design can be tested with 6000 elements in the AVR. The work for creating a target specification and initiation of an approval process have started.
[de]1. Zusammenfassung: Die bestehende Versorgungsunsicherheit bei hochangereichertem Uran erfordert eine back-up-Lösung für den U-Th-Brennstoffkreislauf (HEU). Ein mittelangereicherter Zyklus im Abbrand-Brut-System weist hierfür die meisten Vorteile auf, weicht am wenigsten vom bestehenden Konzept ab und verbaut das HTRPotential nicht durch zwischenzeitliche Umstellung auf diesen Zyklus. Die beim Brennelementhersteller vorliegenden Erfahrungen erlauben, die XVI. AVR-Nachfüll-Charge mit Brennelementen des back-up-Zyklus herzustellen. Nukleare Rechnungen für den AVR ergaben, daß die angestrebte Brennelementauslegung mit 6000 Elementen im AVR getestet werden kann. Die Arbeiten für Erstellung einer Zielspezifikation und Einleitung eines Genehmigungsverfahrens sind angelaufen.
[en] The present study represents a brief analysis of the neutronic parameters (infinite multiplication factor and power distribution) in the CANDU 6 reactor estimated by both Monte Carlo and Collision Probability methods. The simulations were performed using the Continuous-energy Monte Carlo Reactor Physics Burn-up Calculation Code SERPENT 2 along with JEFF-3.1.1 library and the Open-Source Deterministic Transport Code DRAGON 5 with the IAEA 69 energy groups library. The fuel composition chosen for this study has different fissile concentration placed in the bundle rings. The central element contains only ThO2, while the inner and intermediate rings contain 60% ThO2, and the outer ring 80% ThO2. Different bundle fuel charges were investigated by varying the 235U enrichment in radial direction and keeping constant the Th/U ratio. The results highlight that using of (Th,U)O2 in CANDU 6 reactors is both technically feasible and economically useful. (author).
[en] Automatic translation: 1. Introduction and objective: For the HTR in the open fuel cycle - or during one Phase in which the cycle has not yet been closed is - there is a need to store the spent fuel elements for longer periods of time. A prerequisite for this long-term storage is knowledge of the properties of the stored goods and the storage concepts. They should be summarized in a database. The aim of the present data compilation is the listing the properties of the stored goods, insofar as they are directly relevant for long-term storage. In addition, the data included, from which important additional information result or can be derived. The data compilation should be in the genlante in general Reference work ''Data and technologies of the HTR fuel Flow into the cycle''.
[de]1. Einleitung und Zielsetzung: Für den HTR im offenen BrennstoffKreislauf - oder während einer Phase, in der eine Schließung der Kreislaufs noch nicht erfolgt ist - ergibt sich die Notwendigkeit, die abgebrannten Brennelemente über längere Zeiträume zu lagern. Vorraussetzung für diese Langzeitlagerung sind Kenntnisse über die Eigenschaften des Lagergutes und der Lagerkonzepte. Sie sollen in einer Datenbank zusammengefaßt werden. Ziel der vorliegenden Datenzusammenstellung ist die Auflistung der Eigenschaften des Lagergutes, soweit sie für die Langzeitlagerung direkt relevant sind. Darüberhinaus sind die Daten aufgenommen, aus denen sich wichtige Zusatzinformationen ergeben oder abgeleitet werden können. Die Datenzusammenstellunq soll in das genlante allaemeine Nachschlagwerk ''Daten und Technologien des HTR-Brennstoff- Kreislaufs'' einfließen.
[en] Outline: - The Lead cooled Fast Reactors in GIF: SSTAR(USA) Small-sized, battery type reactor with long core life; BREST-OD-300(Russia) Medium-sized, 'pools-in-loop' type reactor with associated closed fuel cycle facilities; ELFR(Europe) Large-sized, integral type reactor for closing of the fuel cyle. - Activities of the GIF LFR provisional SSC (pSSC). - Status of LFR R&D activities in MoU Countries/Entities: Japan, Russian Federation, Republic of Korea, USA, China and Euratom. - Development of the GIF LFR Safety Design Criteria. Outlook on LFR SDC: Work started during 2014, and the present report is the result of discussions among members of the LFR pSSC, benefiting greatly from review and consultations with the GIF RSWG, ANL, IRSN and other partners of the Euratom collaborative project ARCADIA; LFR SDC Report has been updated following the IAEA SSR 2/1 (rev. 1) as well as the IAEA Safety Glossary (2018); The report endorsed by RSWG in February 2021 and approved by the GIF Experts Group in March 2021; Planned to be followed by reviews by external partners (IAEA, WGSAR); Further steps will include the development of detailed Safety Design Guidelines for selected topics.
[en] The molten salt reactor (MSR), as one of the Generation Ⅳ advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistance. The aim of this work is to provide and evaluate possible solutions to fissile 233U production and further the fuel transition to thorium fuel cycle in a thermal MSR by using plutonium partitioned from light water reactors spent fuel. By using an in-house developed tool, a breeding and burning (B&B) scenario is first introduced and analyzed from the aspects of the evolution of main nuclides, net 233U production, spectrum shift, and temperature feedback coefficient. It can be concluded that such a Th/Pu to Th/233U transition can be accomplished by employing a relatively fast fuel reprocessing with a cycle time less than 60 days. At the equilibrium state, the reactor can achieve a conversion ratio of about 0.996 for the 60-day reprocessing period (RP) case and about 1.047 for the 10-day RP case.The results also show that it is difficult to accomplish such a fuel transition with limited reprocessing (RP is 180 days), and the reactor operates as a converter and burns the plutonium with the help of thorium. Meanwhile, a prebreeding and burning (PB&B) scenario is also analyzed briefly with respect to the net 233U production and evolution of main nuclides. One can find that it is more efficient to produce 233U under this scenario, resulting in a double time varying from about 1.96 years for the 10-day RP case to about 6.15 years for the 180-day RP case. (authors)
[en] Compilation of two papers presented at the 89. Meeting of the American Ceramic Society held in Pittsburgh in 1987: ''IRRADIATION BEHAVIOR OF ADVANCED FUEL ELEMENTS FOR THE HIGH TEMPERATURE REACTOR'' - Following the development, testing and utilization in AVR and THTR of mixed oxide Thorium/High Enriched Uranium Fuels, the German HTR program has concentrated on Low Enriched UO2 (LEU) fuels. A series of irradiation tests have been completed with these fuels. These experiments have shown in-pile failure fractions below 2x10-5 and fission product release orders of magnitude lower than in previous HTR fuels. ''FAILURE OF TRISO FUEL PARTICLES AT VERY HIGH TEMPERATURES'' - The fuel particle for the High-Temperature Gas-Cooled Reactor, which is coated with multiple layers of pyrolytic carbon and silicon carbide, fulfills the functional requirements of a fuel element in retaining fission products. Irradiated and unirradiated fuel particles have been evaluated for their high temperature (2000- 2500°C) performance and failure mechanisms in multi-national laboratory heating tests. A phenomenalistic failure model, which incorporates the SiC failure mechanisms of corrosion by fission products and SiC thermal decomposition, as well as diffusion of fission products in the pyrocarbon, is fitted to the data.