Next generation fuel irradiation capability in the High Flux Reactor Petten

This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.     

Design concepts for reactor assembly components of 500 MWe future SFRs

500 MWe Prototype Fast Breeder Reactor (PFBR) is under construction in India. Beyond PFBR, it is planned to construct 3 twin units; each one is 2 MWe × 500 MWe capacity reactors with improved economy and enhanced safety. Significant capital cost reduction is targeted for the reactor assembly, by way of introducing new concepts for the grid plate, primary pipes, top shield and fuel handling system and optimizing the main vessel diameter and bottom dished head shape. The capital cost reduction of the reactor assembly components that could be achieved through these improved concepts is estimated to be about 25%. To validate these concepts, preliminary analysis has been completed, R&D areas have been identified and strategy to execute the R&D has been defined clearly. The basis of each concept is highlighted to depict the Indian approach and strategy to make the fast reactor economically competitive.     

CAST3M/ARCTURUS: A coupled heat transfer CFD code for thermal–hydraulic analyzes of gas cooled reactors

The safety of gas cooled reactors (High Temperature Reactors (HTR), Very High Temperature Reactors (VHTR) or Gas Cooled Fast Reactors (GFR)) must be ensured by systems (active or passive) which maintain loads on component (fuel) and structures (vessel, containment) within acceptable limits under accidental conditions. To achieve this objective, thermal–hydraulics computer codes are necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Some key safety questions are related to the evaluation of decay heat removal and containment pressure and thermal loads. This requires accurate simulations of conduction, convection, thermal radiation transfers and energy storage. Coupling with neutronics is also an important modeling aspect for the determination of representative parameters such as neutronics coefficient (Doppler coefficient, Moderator coeffcient, …), critical position of control rods, reactivity insertion aspects, …. For GFR, the high power density of the core and its necessary reduced dimension cannot rely only on passive systems for decay heat removal. Therefore, forced convection using active safety systems (gas blowers, heat exchangers, …) are highly recommended. Nevertheless, in case of station black-out, the safety demonstration of the concept should be guaranteed by natural circulation heat removal. This could be performed by keeping a relatively high back-up pressure for pure helium convection and also by heavy gas injection. So, it is also necessary to model mixing of different gases, the on-set of natural convection and the pressure and thermal loads onto the proximate or guard containment. In this paper, we report on the developments of the CAST3M/ARCTURUS thermal–hydraulics (Lumped Parameter and CFD) code developed at CEA, including its coupling to the neutronics code CRONOS2 and the system code CATHARE. Elementary validation cases are detailed, as well as application of the code to benchmark problems such as the HTR-10 thermal–hydraulic exercise. Examples of containment thermal–hydraulics calculations for fast reactor design (GFR) are also detailed.     

Meeting EU's energy needs through nuclear fission: Synergy of public and private research in an international context

The availability of an affordable and sustainable energy supply is becoming a major and growing concern for world's future. It is very likely that there is not one single solution to the problem but that it is necessary to call upon a whole set of means such as energy efficiency improvement, deployment of renewable energies, clean coal technologies including CO₂ capture and storage, nuclear development. Indeed, it is more and more recognized that nuclear energy offers a very effective way to contribute to this worldwide challenge. It can be a safe, clean, reliable and cost-effective source of energy, the price of which remaining quite stable. Although the “generations” of nuclear systems are at different degrees of maturity, the scientific, technological and industrial gaps are quite well identified and assessed so that it is possible to describe a detailed roadmap of their development, including R&D needs.A significant part of these R&D needs should be addressed through cooperation involving public and private sector. It is the case for programs relating to safety, radiation protection, PRA (probabilistic risk assessment) methodology, background knowledge about ageing, fuel and fuel cycle for future light water reactors (Gen 3), pre-normative research for the purpose of harmonizing safety demonstration methodologies, Gen 4 systems with an emphasis on sodium-cooled fast breeder, large R&D infrastructures like test reactors and more generally, all obstacles to a consensual development of nuclear energy. R&D program should also be helpful in maintaining appropriate expertise and competencies.Strong cooperation between countries and between stakeholders is necessary to face all these challenges.     

Comparative study of fast critical burner reactors and subcritical accelerator driven systems and the impact on transuranics inventory in a regional fuel cycle

In the frame of Partitioning and Transmutation (P&T) strategies, many solutions have been proposed in order to burn transuranics (TRU) discharged from conventional thermal reactors in fast reactor systems. This is due to the favourable feature of neutron fission to capture cross section ratio in a fast neutron spectrum for most TRU. However the majority of studies performed use the Accelerator Driven Systems (ADS), due to their potential flexibility to utilize various fuel types, loaded with significant amounts of TRU having very different Minor Actinides (MA) over Pu ratios. Recently the potential of low conversion ratio critical fast reactors has been rediscovered, with very attractive burning capabilities. In the present paper the burning performances of two systems are directly compared: a sodium cooled critical fast reactor with a low conversion ratio, and the European lead cooled subcritical ADS-EFIT reactor loaded with fertile-free fuel. Comparison is done for characteristics of both the intrinsic core and the regional fuel cycle within a European double-strata scenario. Results of the simulations, obtained by use of French COSI6 code, show comparable performance and confirm that in a double strata fuel cycle the same goals could be achieved by deploying dedicated fast critical or ADS-EFIT type reactors. However the critical fast burner reactor fleet requires ∼30-40% higher installed power then the ADS-EFIT one. Therefore full comparative assessment and ranking can be done only by a parametric sensitivity study of both the fuel cycle and the electricity generating costs.     

Waste disposal aspects of the long term cooling characteristics of irradiated nuclear fuels

A study has been made of the long term cooling characteristics of nuclear fuels irradiated in commercial reactor designs of interest within the U.K. In the case of thermal reactors, Magnox, AGR, SGHWR, LWR and HTR systems fuelled with either natural or enriched uranium are considered, together with a fast reactor fuelled with plutonium derived either from the Magnox or the AGR programmes. Alternative uses for Magnox plutonium are considered by simulating a plutonium fuelled HTR thermal system and the development of a $\text{Th}{}^{233}\text{U}$ fuel cycle has been anticipated for both a fast reactor and an HTR.For each system the activities as a function of cooling time are considered on the assumption of U/Pu recovery from the fuel during reprocessing within a year of discharge from the reactor and for the alternative case of no U/Pu extraction. The reprocessing waste products associated with the various fuel cycles have then been compared both on the basis of decay heating and radiological hazard per GW(e) yr. Finally, recycling of transplutonium elements is also considered with a view to reducing the long term heating commitment from the higher actinides.     

Role of Nuclear Power in Large-Scale Power Production in Russia in the 21st Century

The results are summarized of investigations performed over the last 20 years which culminated in the current fundamental formulation of the problem of nuclear power based on fast reactors and the basic criteria for such nuclear power in the strategy for growth and which served as a scientific basis for the preparation of the initiative of the President of the Russian Federation. These investigations have shown that the experience gained over half a century and the technologies developed in Russia permit developing and demonstrating at the start of the century a fast reactor which meets the requirements of large-scale nuclear power. An experimental prototype of such a reactor (BREST-300) has been designed and R&D work substantiating the design is continuing.     

Role of ADS in the back-end of the fuel cycle strategies and associated design activities: The case of Japan

Reduction of burden caused by radioactive waste management is one of the most critical issues for the sustainable utilization of nuclear power. The Partitioning and Transmutation (P&T) technology provides the possibility to reduce the amount of the radiotoxic inventory of the high-level radioactive waste (HLW) dramatically and to extend the repository capacity. The accelerator-driven system (ADS) is regarded as a powerful tool to effectively transmute minor actinides (MAs) in the “double-strata” fuel cycle strategy. The ADS has a potential to flexibly manage MA in the transient phase from light water reactors (LWRs) to fast breeder reactors (FBRs), and can co-exist with FBR symbiotically and complementarily to enhance the reliability and the safety of the commercial FBR cycle. The concept of ADS in JAEA is a lead-bismuth eutectic (LBE) cooled, tank-type subcritical reactor with the power of 800 MWth driven by a 30 MW superconducting LINAC. By such an ADS, 250 kg of MA can be transmuted annually, which corresponds to the amount of MA produced in 10 units of LWR with 1 GWe. The design study was performed mainly for the subcritical reactor and the spallation target with a beam window. In Japan, Atomic Energy Commission (AEC) has implemented the check and review (C&R) on P&T technology from 2008 to 2009. In the C&R, the benefit of P&T technology, the current status of the R&D, and the way forward to promote it were discussed.     

Improving SFR economics through innovations from thermal design and analysis aspects

Achieving economic competitiveness as compared to LWRs and other Generation IV (Gen-IV) reactors is one of the major requirements to attract large-scale investment in commercial sodium cooled fast reactor (SFR) power plants. Advances in R&D for advanced SFR fuel and structural materials provide key long-term opportunities to improve SFR economics. In addition, other new opportunities are emerging to further improve SFR economics. This paper provides an overview on potential ideas from the perspective of thermal hydraulics to improve SFR economics. These include: (1) a new hybrid loop-pool reactor design to further optimize economics, safety, and reliability of SFRs with more flexibility, (2) a multiple-reheat and intercooling helium Brayton cycle to improve plant thermal efficiency and to reduce safety related overnight and operation costs, and (3) modern multi-physics thermal analysis methods to reduce analysis uncertainties and associated requirements for over-conservatism in reactor design. This paper reviews advances in all three areas and their potential beneficial impacts on SFR economics.     

Overview of past and current activities on fuels for fast reactors at the Institute for Transuranium Elements

Nuclear energy has the potential to provide a secure and sustainable electricity supply at a competitive price and to make a significant contribution to the reduction of greenhouse gas emissions. The renewal of interest in fast neutron spectra reactors to meet more ambitious sustainable development criteria (i.e., resource maximisation and waste minimisation), opens a favourable framework for R&D activities in this area. The Institute for Transuranium Elements has extensive experience in the fabrication, characterization and irradiation testing (Phénix, Dounreay, Rapsodie) of fast reactor fuels, in oxide, nitride and carbide forms. An overview of these past and current activities on fast reactor fuels is presented.     

Partitioning and transmutation studies at JAERI both under OMEGA program and High-Intensity Proton Accelerator Project

At the Japan Atomic Energy Research Institute (JAERI), active and comprehensive studies on partitioning and transmutation (P&T) of long-lived nuclear waste from the reprocessing processes of spent fuel has been carried out under the OMEGA program. Studies at JAERI include a design study of dedicated transmutation systems both of an MA burner fast reactor (ABR) and an accelerator-driven subcritical system (ADS), and the development of a high intensity proton accelerator as well as the development of partitioning process, nitride fuel fabrication/dry separation process technologies and nuclear data studies.

During the course of studies, JAERI developed the concept of the double-strata fuel cycle, in which a dedicated system is used for transmutation. Comparing the various transmutation systems, such as thermal neutron spectrum or fast neutron spectrum systems, power reactors or dedicated systems, from the viewpoints of reactor physics, nuclear fuel cycle and socio-technical issues, it was concluded that the ADS is the best option for transmutation of minor actinide(MA). JAERI, therefore, decided to concentrate its R&D efforts on the development of ADS and related technologies.

One of the goals of R&D is to provide a basis for designing demonstration facilities of ADS, aqueous partitioning process and nitride fuel fabrication and dry separation technologies. As the initial step toward this purpose, the construction of an ADS experimental facility is planned under the High-Intensity Proton Accelerator Project which JAERI and the High Energy Accelerator Research Organization (KEK) are jointly proposing since 1998.

The paper discusses the some of the results of P&T studies and the outline of the High-Intensity Proton Accelerator Project under which ADS experimental facility will be constructed.     

Advanced technologies for CANDU reactors

The inherent features of a reactor based on multiple pressure tubes, rather than a single pressure vessel, provide CANDU with considerable flexibility for continuous design improvements. Of particular significance is the presence of a large volume of cool heavy-water moderator within the core. This can act as a heat sink for passive heat rejection during high-temperature accidents. We describe elements of our R&D program that are aimed to exploit this benefit. Similarly, safety, reliability and economy are the focus for our exploration of advanced technologies in the areas of advanced fuel and fuel channels, improved heavy-water production and computer-based systems that facilitate reactor design, construction and operation.     

Requirements for helium cooled pebble bed blanket and R&D activities

This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R&D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM.The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R&D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R&D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine.     

Thermal response of a modular high temperature reactor during passive cooldown under pressurized and depressurized conditions

The concept of inherent safety features of the modular HTR design with respect to passive decay heat removal through conduction, radiation and natural convection was first introduced in the German HTR-module (pebble fuel) design and subsequently extended to other modular HTR design in recent years, e.g. PBMR (pebble fuel), GT-MHR (prismatic fuel) and the new generation reactor V/HTR (prismatic fuel).This paper presents the numerical simulations of the V/HTR using the thermal-hydraulic code THERMIX which was initially developed for the analysis of HTRs with pebble fuels, verified by experiments, subsequently adopted for applications in the HTRs with prismatic fuels and checked against the results of CRP-3 benchmark problem analyzed by various countries with diverse codes.In this paper, the thermal response of the V/HTR (operating inlet/outlet temperatures 490/1000 °C) during post shutdown passive cooling under pressurized and depressurized primary system conditions has been investigated. Additional investigations have also been carried out to determine the influence of other inlet/outlet operating temperatures (e.g. 490/850, 350/850 or 350/1000 °C) on the maximum fuel and pressure vessel temperature during depressurized cooldown condition. In addition, some sensitivity analyses have also been performed to evaluate the effect of varying the parameters, i.e. decay heat, graphite conductivity, surface emissivity, etc., on the maximum fuel and pressure vessel temperature. The results show that the nominal peak fuel temperatures remain below 1600 °C for all these cases, which is the limiting temperature relating to radioactivity release from the fuel. The analyses presented in this paper demonstrate that the code THERMIX can be successfully applied for the thermal calculation of HTRs with prismatic fuel. The results also provide some fundamental information for the design optimization of V/HTR with respect to its maximum thermal power, operating temperatures, etc.     

Plutonium management and Generation IV systems

This paper discusses the potential role of Generation IV nuclear energy systems in managing plutonium. It briefly reviews the Generation IV goals and their relevance to plutonium management. Each of the six selected Generation IV systems [very high temperature reactor (VHTR), gas-cooled fast reactor (GFR), sodium-cooled fast reactor (SFR), super-critical-water-cooled reactor (SCWR), lead-cooled fast reactor (LFR), molten salt reactor (MSR)] is briefly discussed. The main characteristics of each system are summarised and the capability for plutonium management indicated. The potential for the management of plutonium using Generation IV systems is briefly reviewed from a complete fuel cycle perspective to illustrate the issues in the context of a fleet of reactor and fuel cycle facilities.     

The advanced high-temperature reactor: High-temperature fuel, liquid salt coolant, liquid-metal-reactor plant

The Advanced High-Temperature Reactor is a new reactor concept that combines four existing technologies in a new way: (1) coated-particle graphite-matrix nuclear fuels (traditionally used for helium-cooled reactors), (2) Brayton power cycles, (3) passive safety systems and plant designs from liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants with boiling points far above the maximum coolant temperature. The new combination of technologies enables the design of a large [2400- to 4000-MW(t)] high-temperature reactor, with reactor-coolant exit temperatures between 700 and 1000°C (depending upon goals) and passive safety systems for economic production of electricity or hydrogen. The AHTR [2400-MW(t)] capital costs have been estimated to be 49 to 61% per kilowatt (electric) relative to modular gas-cooled [600-MW(t)] and modular liquid-metal-cooled reactors [1000-MW(t)], assuming a single AHTR and multiple modular units with the same total electrical output. Because of the similar fuel, core design, and power cycles, about 70% of the required research is shared with that for high-temperature gas-cooled reactors.     

The HTR and nuclear process heat applications

The High Temperature Reactor HTR offers beside the production of electricity the potential of the production of secondary energy carriers for the fuel and heat market. Therefore the HTR can considerably contribute to solutions of future problems in the energy supply of the Federal Republic of Germany as well as of the world. On the basis of the experiences with the power plants AVR, Fort St. Vrain and THTR-300 new concepts of reactors have been proposed: the medium size reactor HTR 500 and the Modular HTR concept. The high temperature heat application is directed towards the refinement of fossil fuels, the long distance energy system and other applications as e.g. process steam for chemical industry, enhanced oil recovery and water splitting. The research and development program in the projects Prototype Plant Nuclear Process Heat (PNP) and Nuclear Long Distance Energy (NFE) has shown very promising results. These results show that nuclear process heat is technically feasibly and that it is possible to reach a commercial application in the next decades.     

The reactor physics characteristics of a transuranic mixed oxide fuel in a heavy water moderated reactor

The reprocessing actinide materials extracted from spent fuel for use in mixed oxide fuels is a key component in maximizing the spent fuel repository utility. While fast spectrum reactor technologies are being considered in order to close the fuel cycle, and transmute these actinides, there is potential to utilize existing pressurized heavy water reactors such as the CANDU^®1 design to meet these goals. The use of current thermal reactors as an intermediary step which can burn actinide based fuels can significantly reduce the fast reactor infrastructure needed. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a typical CANDU nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 4.75% actinide MOX fuel. The WIMS-AECL model of the fuel lattice was created and the two neutron group properties were transferred to RFSP in order to create a 3 dimensional time average full core model. The model was created with typical CANDU limits on bundle and channel powers and a burnup target of 45 MWd/kgHE. The TRUMOX fuel design achieved its goals and performed well under normal operations simulations. This effort demonstrated the feasibility of using the current fleet of CANDU reactors as an intermediary step in burning reprocessed spent fuel and reducing actinide burdens within the end repository. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle using existing and proven reactor technologies.     

Advanced fast reactor development requirements: Is there any need for HEU?

The present paper reviews R&D needs foreseen for advanced fast reactor development. The focus is on physics, safety and fuel validation needs using experimental facilities, i.e. critical zero power facilities or irradiation reactors. This review underlines the crucial need for a continuous availability of such facilities in the next decades. As for the fuel to be used in the experiments the need is for plutonium fuelled experiments and no specific need for HEU seems to be foreseen. However, the already existing stocks of HEU at some critical facilities like MASURCA in France will be certainly needed and should be kept in order to allow the required flexibility to perform critical experiments of large size, representative of most future advanced systems.