The XT-ADS Core Design (original) (raw)
Related papers
XT-ADS: Neutronics, Shielding, and Radiation Damage Calculations
Nuclear Technology, 2009
This work is related to the design of the core of the eXperimental demonstration of the technological feasibility of Transmutation in an Accelerator-Driven System (XT-ADS) facility in the framework of the EUROpean Research Programme for the TRANSmutation of High Level Nuclear Waste in an Accelerator Driven System (EUROTRANS) project. The design specifications for the proton accelerator of the XT-ADS are 600 MeV and up to 3.5 mA for the beam energy and current, respectively. The proton beam impinges on a liquid target consisting of a lead-bismuth-eutectic mixture. The state-of-the-art Monte Carlo code MCNPX was used to assess the neutronics performance and shielding properties of the system. The nuclear dataprocessing system NJOY 99 was also used. The work consisted of the optimization of the core configuration (geometry, number, and location of the fuel and absorber assemblies) and the appropriate fuel composition in order to reduce radiation damage (namely, the displacement per atom values) on the core barrel and top grid plate, while maintaining the high neutron fluxes (10 15 n{cm Ϫ2 {s Ϫ1) and the k eff of the system of ;0.95. The assessment of the core configuration and fuel composition was performed, resulting from the interplay among parameters such as the desired high neutron fluxes, the k eff value wanted for safety and core performance reasons, the as-lowas-possible radiation damage of the core barrel and top grid plate, and the fuel composition, among others.
A forward step towards the Pu, MA and LLFP transmuter Accelerator Driven System (ADS) is the realisation of a 80 MWt ADS-demo (XADS) whose basic objective is the system feasibility demonstration. The XADS is forecasted to adopt the UO 2 -PuO 2 mixed-oxides (MOX) fuel already experimented in the French SPX-1 sodium cooled fast reactor. The present analysis, performed by using Transuranus Code, was carried out for the Normal Operation (nominal reactor power and 120% nominal reactor power), aimed at verifying that the fuel system allows for an ample margin with respect to the design limits (i.e., centreline fuel temperature, cladding temperature and damage) during all the inreactor lifetime. Most relevant assumptions in the present calculations were the AISI-316 as cladding material and no consideration of any LBE (Lead-Bismuth Eutectic) corrosion effect. It is there confidence that this ample margin would help to cover also this limitation. Furthermore, it is shown that some modifications on fuel rod specifications, such as increasing the fabrication filling gas pressure, will reduce significantly the FGR (Fission Gas Release). The analysis was performed in the context of a cooperation between POLIMI (Politecnico di Milano), ENEA and ITU (Institute for Transuranium Elements, Karlsruhe, Germany).
Neutronics and Shielding Issues of ADS
Accelerator Driven Systems (ADSs) are hybrid systems consisting of a high-intensity proton accelerator with beam energy in the hundreds of MeV range impinging on a target of a heavy element and coupled to a sub-critical core. The intense (of the order of 10 15 n/cm 2 /s) and fast neutron fluxes produced by the spallation reactions triggered by the impinging protons in the target can be used to induce fission reactions in the actinides and capture reactions in the longlived fission products in the fuel assemblies in the core of the system. ADSs have been considered during the last fifteen years as one of the promising technological solutions for the transmutation of nuclear waste, reducing the radiotoxicity of the high-level nuclear waste and hence reducing the burden to the geological repositories. The European Commission´s Green Paper entitled "Towards a European Strategy for the Security of Energy Supply" clearly pointed out the importance of nuclear energy in Europe. With 145 operating reactors producing a total power of 125 GW e , the resulting energy generation of 850 TWh per year provides 35% of the electricity consumption of the European Union. The Green Paper also points out that the nuclear industry has mastered the entire nuclear fuel cycle with the exception of waste management and for this reason, "focusing on waste management has to be continued". Amongst the several solutions being studied in recent years, MYRRHA (concept developed at SCK-CEN, Belgium), XADS (design studies co-funded by the European Union in the framework of the 5 th Framework Programme) and XT-ADS and EFIT (acronyms standing for an experimental machine and for the long term transmuter to be deployed on an industrial scale, both in the EUROTRANS project of the 6 th Framework Programme) have deserved the attention of different communities of specialists in the field of Nuclear Technology and Radioactive Waste Management. Although these machines have been designed with different parameters, their implementation and deployment have in common the fact that they raise cutting edge scientific and technological issues, associated to the operation of the high-intensity proton accelerator, the high-power (in the multi-MegaWatt range) delivered to the target and the material damage in the target and surrounding structures. The thermal power in the core, the thermal-hydraulic aspects associated to the heat removal in steady state and also in transient mode, the subcriticality level of the system and the efficiency of the transmutation process, are particularly sensitive to the core design (geometry, number of subassemblies, fuel composition, among many other aspects). Neutronics and shielding issues and the computation and mapping of neutron fluxes and doses are important throughout all stages of design of these systems. In this paper, i) the main characteristics and parameters of the ADS systems previously alluded to will be reviewed ii) the neutronics and shielding calculations of relevance for the design of the ADS systems, for radiation damage and for radiation protection purposes will be extensively described.
1999
The laser glass for the National Ignition Facility (NIF) Main Amplifier system is pumped by a system of 192 pulsed power/flash lamp assemblies. Each of these 192 assemblies consists of a 1.6 MJ (nominal) capacitor bank working with a Pre-Ionization/Lamp Check (PILC) pulser to drive an array of40 flash lamps. This paper describes the predicted performance of these Power Conditioning System (PCS) modules in concert with flashlamp assemblies in NIF. Each flashlamp assembly consists of 20 parallel sets of lamps in series pairs. The sensitivity of system performance to various design parameters of the PILC pulser and the main capacitor bank is described, Results of circuitmodels are compared to sub-scaleflashlamptests and to measurementstaken in. tests of a PCS module driving a flashlamp assembly in the First Article NIF Test Module facility at Sandia National Laboratories. Also included are predictions from a physics-based, semi-empirical amplifier gain code.
ADS-Demo Fuel Rod Performance Analysis
A forward step towards the Pu, MA and LLFP transmuter Accelerator Driven System (ADS) is the realisation of a 80 MWt ADS-demo (XADS) whose basic objective is the system feasibility demonstration. The XADS is forecasted to adopt the UO2-PuO2 mixed-oxides (MOX) fuel already experimented in the French SPX-1 sodium cooled fast reactor. The present analysis, performed by using Transuranus Code, was carried out for the Normal Operation (nominal reactor power and 120% nominal reactor power), aimed at verifying that the fuel system allows for an ample margin with respect to the design limits (i.e., centreline fuel temperature, cladding temperature and damage) during all the inreactor lifetime. Most relevant assumptions in the present calculations were the AISI-316 as cladding material and no consideration of any LBE (Lead-Bismuth Eutectic) corrosion effect. It is there confidence that this ample margin would help to cover also this limitation. Furthermore, it is shown that some modification...
High Flux Isotope Reactor Low Enriched Uranium Low Density Silicide Fuel Design Parameters
2021
High Flux Isotope Reactor (HFIR) highly enriched uranium (HEU) to low-enriched uranium (LEU) conversion activities are ongoing as part of the Department of Energy (DOE) National Nuclear Security Administration (NNSA)'s nuclear nonproliferation mission. Design activities studying the conversion of HFIR from HEU to LEU fuel explored different fuel design features and shapes with a low density uranium-silicide dispersion (U 3 Si 2-Al) fuel, which has a uranium density of 4.8 gU/cm 3. The goal of these studies is to generate several HFIR LEU fuel designs of varying fuel fabrication complexity that meet the current HEU performance metrics and safety requirements. The documented designs will serve as references for fuel fabrication and qualification activities. Recent advancements in modeling and simulation tools enable quick prototyping of fuel designs. Shift, a Monte Carlo neutron transport and depletion tool optimized for high-performance computing (HPC) architectures, is used for efficient fuel cycle and performance metrics calculations. The HFIR Steady State Heat Transfer Code (HSSHTC) is used to vet the thermal safety margin. Also, a new automation tool that connects all fuel design analysis steps, named Python HFIR Analysis and Measurement Engine (PHAME), has been developed to expedite the design study in an efficient and reproducible manner. Leveraging these tools, several candidate fuel designs were selected for varying fabrication complexity. This report provides design feature details for four selected HFIR LEU low density U 3 Si 2-Al fuel designs and their corresponding performance and safety metrics. Nominal, best-estimate design parameters and irradiation conditions, including fission rate densities, power densities, heat fluxes, and cumulative fission densities are provided for candidate fuel designs relevant to framing irradiation experiments to support fuel qualification efforts. Simulations show that the low density U 3 Si 2-Al, with design features to enhance safety, can meet HEU core performance metrics and safety requirements if the reactor power is increased from 85 MW (HEU) to 95 MW (LEU) and if the active fuel length is increased from 50.80 cm (HEU) to 55.88 cm (LEU). * This Q value of 200.7 Mev/fission is a 'typical' Q value used in HFIR analyses close to the approximate cycle-averaged value that was recently calculated for the optimized silicide design. The BOC and EOC Q values were estimated to be 200.49 and 201.46 MeV/fission, respectively, givin the average Q value of 200.98 MeV/fission [9].