Material effects: HTR irradiation examined3 October 2002
Results from the study of irradiation effects on advanced ceramic and metallic materials and on the development of in-core instrumentation for high-temperature reactors.
A recent meeting on Basic Studies in the Field of High Temperature Engineering covered research on materials relating to high temperature and neutron irradiation on advanced ceramic and metallic materials and on the development of in-core instrumentation.
The meeting (organised by the Nuclear Science Committee of the OECD Nuclear Energy Agency, and the Japan Atomic Energy Research Institute) began with papers giving overviews of high-temperature engineering research in each country.
Research under the 5th Euratom Framework Programme (FP5) considders the safety and efficiency of future reactor systems in order to investigate and evaluate new or revised concepts. Several projects in this programme address the high-temperature gas-cooled reactor (HTR or HTGR): fuel technology, fuel cycle, materials, power conversion systems and licensing.
Nine HTR-related projects selected by the European Commission form a structured cluster covering both fundamental research and technology (see Table).
Two projects (HTR-F and HTR-F1) look at HTR fuel technology. The objectives of these projects are to restore and improve fuel fabrication capability in Europe, to qualify fuel at high burnup with a high reliability and to study innovative fuels that can be used for new applications.
There are two projects (HTR-N and HTR-N1) both of which provide numerical nuclear physics tools to analyse and design innovative HTR cores; to investigate fuel cycles that can minimise the generation of long-lived actinides and optimise Pu-burning capabilities; and to analyse HTR-specific waste and the disposal behaviour of spent fuel.
There are two projects (HTR-M and HTR-M1) that provide materials data for key HTR components - the reactor pressure vessel, high temperature areas and graphite structures.
Project HTR-E addresses innovative key components, systems and equipment related to the direct cycle of modern HTRs. These include the turbine, heat exchanger, active and permanent magnetic bearings, rotating seals, sliding parts, and the helium purification system. The programme covers both design and experiments.
Project HTR-L proposes a safety approach for a licensing framework specific to modular high-temperature reactors and a classification for the design basis operating conditions and associated acceptance criteria.
Project HTR-C coordinates all the separate projects.
IAEA activities on high-temperature gas-cooled reactors are conducted primarily through the Technical Working Group on Gas-cooled Reactors. In particular, there has been significant work on the evaluation of HTGR performance.
Modular HTGRs rely largely on the integrity of ceramic fuel elements of the core to confine fission products. A combination of strongly negative temperature coefficient, a low power density, a large core thermal capacity and a passive decay heat removal path promise to assure fuel integrity under all postulated severe accidents.
The IAEA has defined benchmark problems for reactor physics and thermal hydraulics, and is supporting their investigation.
As is illustrated by the German programme, safety analysis of the HTR is dedicated to developing a system which will avoid the problems that are connected to existing reactors.
HTR reactors cannot melt even in severe loss of coolant accidents, and fuel cannot be destroyed by other severe accidents. Therefore, the release of radioactivity to the environment is very limited.
Kurt Kugeler of the Jülich Research Centre said the new type of technology can be designated "catastrophe-free nuclear technology". The allowed release of fission products in case of severe accidents is very limited. For the case of a reactor with a thermal power of 300MW, this amount should be less than 10-5 to 10-6 of the fission product inventory inside the fuel elements.
The range of catastrophe-free nuclear technology must be defined. There are three categories of accidents:
• Category 1. Internal reasons (totalloss of coolant, total failure of decay heat removal, ingress of water to primary system, ingress of air to primary system, extreme reactivity accidents, and massive failures of components).
• Category 2. External reasons (air crash, gas cloud explosion, earthquakes, fire, and hurricanes).
• Category 3. External reasons outside today's licensing procedures (sabotage, terrorist attack, war, extreme earthquakes, and meteorite strike).
The work to be done in the future is to identify systems in which there is no release of intolerable amounts of radioactivity to the environment through internal or external causes. This safety behaviour of the reactors must be improved by convincing experiments.
All new HTR concepts are designed to integrate inherent safety features as much as possible. The main aspects of this are the self-acting decay heat removal and the limitation of the temperature of fuel elements to allowed values (1600ºC for the Trisco-coated particles in pebble bed fuel).
The concept of self-acting decay heat removal exists - in principle - in all high-temperature reactors, including the old reactor systems. Depending on power, power density and core dimensions, the maximum fuel temperatures during loss of coolant and loss of active decay heat removal accidents is different for the reactors.
In large cores (THTR-300, GAC-1160, Fort St Vrain), high fuel temperatures above 2000ºC would have occurred in decay heat removal accidents, resulting in large fission product releases from the fuel elements. Even these large cores would not have melted, because graphite is stable up to 3600ºC.
Experiments to improve the concept of self-acting decay heat removal
Testing the inherent safety features of reactors is one of the most important tasks for further work. Some main aspects of fulfilling the principle self-acting decay heat removal have been tested very well during the development of the pebble-bed HTR.
Important experiments that have been performed are:
• The effective heat conductivity of the pebble bed is dependent on the temperature inside the reactor core. This parameter directly influences the maximum temperature and the temperature distribution inside the core during a loss of coolant accident. Heat transfer by conduction, thermal radiation and free convection in the core region have all been measured in several experiments.
• Experiments to measure the temperature distribution inside a large arrangement of fuel elements, to control the dynamic temperature profiles and to thereby validate computer codes for the calculation of the three-dimensional real temperature distribution in the core, were carried out.
• Experiments to measure the heat fluxes and the temperature distribution from the structures of core internals into the reactor vessel and from there to the surrounding concrete cell outside the reactor vessel have been performed.
• An experiment to measure the heat flux, including free convection, from the surface of the reactor pressure vessel has been carried out in Japan in the framework of the HTTR (high-temperature engineering test reactor) project.
• A real reactor experiment has been carried out at the AVR reactor in Jülich. The loss of shutdown and active cooling were simulated by switching off the blower power and measuring the dynamic behaviour of the reactor system.
• In the future, a full-scale experiment for a real reactor can been carried out, in which the decay heat is simulated by electric heating. All accidental situations can be simulated in this model, and measurements can indicate the system behaviour as well as the accuracy of codes.
At the Dutch high flux reactor (HFR) irradiation experiemeents are being carried out on fuel pebbles and fuel compacts.
The main objective of the fuel pebble irradiation test is to demonstrate the feasibility of high burn-up for German LEU fuel using Trisco coated particles. It will irradiate fuel to a burn-up of 20% FIMA (fissions per initial metal atom), to explore the real limits that have formerly been designed for operational conditions of the HTR module. It will also extend the existing database for metallic fission product release, particularly of Ag-110, for an improved assessment of the particle choice for the HTR concept.
Finally it will demonstrate the ability of the fuel to retain fission products in the event of an accident.
Other experiments examine irradiation of HTR structural materials. These are intended to help build a database on pressure vessel steel plate and weld materials. It will lead to recommendations for HTR pressure vessel feasibility, choice of material, defect assessment and leak-before-break applicability and evaluation factors for defect tolerance assessment for HTR relevant welds.
The centre is also carrying out experiments and studies into reactivity effects and start-up core physics.
CEA and Framatome ANP researches in high-temperature engineering are for the short term directly connected to the development of small- or medium-sized thermal HTRs with direct cycle and the necessary recovery at a national level of expertise on some associated crucial technology issues. For the longer term, the CEA's R&D effort is directed toward the study of technologies in support of HTTRs with hardened neutron spectra and refractory fuel, possibly compliant with on-site reprocessing.
In the medium term two variants are under consideration. One is a very high temperature GCR (>1000ºC) for very high thermodynamical efficiency and hydrogen production. The other is a robust GCR with maximised safety features and intrinsic resistance to proliferation, for deployment in countries which have little nuclear experience.
France wants to recover the capability of HTR particle fuel fabrication. It aims to characterise and qualify fuel behaviour under irradiation to show that it is reliable in nominal and high-temperature conditions.
The experimental programme is accompanied by modelling of the particle fuel behaviour to provide a fully qualified calculation tool for the thermomechanical behaviour of the particle and for fission product transport. Key dates are:
• 2003. Release of a first qualified version of the particle fuel code.
• 2004. Fabrication of Trisco fuel.
• 2004-2006. Start of irradiation tests in the Osiris reactor at CEA Saclay.
• 2006-2010. Characterisation of irradiated particles and qualification of the fabrication process.
The study of high-temperature structural materials, together with the technology of He circuit, clearly show a common body of development. There are also potential benefits from synergies with industry. Of course, requirements for a fast GCR introduce supplementary constraints for inner core structure materials.
The programme aims to select materials for a feasibility study of a fast GCR by 2006.
R&D work has two strands. The first is selecting and establishing specifications and determining the in-pile and out-of-pile behaviour of the structural materials. The second is establishing the rules for mechanical assessments and to predict lifetimes for the structures and the welded joints.
The oxidation of graphite by air is also being considered. Graphite characteristics and oxidation conditions vary and it is difficult to extrapolate the behaviour of a graphite grade to a different location.
The helium programme aims to move from bench experiments in 2001-2005 to a 15MW system loop in 2005-2015. Among the requirements are:
• Control of coolant quality.
• Management of gas inventories.
• Generic technologies such as tribology, leak tightness and insulation.
• Heat transfers and fluid flows in the core, circuits and heat exchangers.
• Dynamics of the circuit and structures.
There are three steps to achieve this. In 2002-2003, small dedicated benches will be used for tribology studies - the oxygen/impurities influence on the binding risk will have to be considered, as well as the evaluation of coating performances. Thermal insulation and leak tightness studies will consider wall diffusion, and leaks through valves, seals and shaft penetration in the case of an alternator located outside the turbine vessel.
In 2004-2005, dynamic loops will be used for purification studies to develop and validate methods and instrumentation to guarantee He quality; to monitor the level of chemical impurities, activated corrosion products and fission products; and to determine the efficiency and purification rates of the various designs. A 1MW multi-purpose loop designed to operate up to 1000ºC and at pressures of 10MPa will be used to qualify components such as recuperator, circulator, valves and to perform fuel element and sub-assembly characterisations.
Finally a multi-channel and system multi-purpose loop, still to be specified, will be used for testing in fully representative conditions large components and GCR systems. The objective will be to demonstrate the robustness of the selected technologies in both nominal and accident situations.
In Japan HTR work has been under way for some time. Japan Atomic Energy Research Institute's (JAERI's) 30MWt high-temperature engineering test reactor (HTTR) aims to establish and upgrade HTGR technologies as well as hosting basic research in high-temperature engineering.
JAERI is carrying out basic research in the following areas:
• New materials development of high-temperature oxide superconductors, high performance silicon carbide semiconductors, and heat-resistant ceramic composites.
• High-temperature radiation chemistry research on polysilane decomposition into silicon carbide fibres, and high-temperature radiolysis of heavy oils and plastics, as well as in situ measurement of property changes of solid tritium breeding materials under irradiation.
• High-temperature in-core instrumentation development: a heat- and radiation-resistant optical fibre system and devices for monitoring neutron and gamma-ray spectra.
JAERI is preparing for the first HTTR irradiation in 2003. It will then proceed to international collaboration on high-temperature irradiation tests of new materials.
TablesOngoing HTR-related research projects in Euratom FP5