[en] The OECD Halden Reactor Project is an international network dedicated to enhanced safety and reliability of nuclear power plants. The Project operates under the auspices of the OECD Nuclear Energy Agency and aims at addressing and resolving issues relevant to safety as they emerge in the nuclear community. This paper gives a concise presentation of the Project goals and of its technical infrastructure. The paper contains also a brief overview of results from the programme carried out in the time period 1997-1999 and of the main issues contemplated for the 3-year programme period 2000-2002

[en] The OECD Halden Reactor Project is an international organisation devoted to improved safety and reliability of nuclear power station through an user-oriented experimental programme. A significant part of this programme consists of studies addressing fuel performance issues in a range of conditions realised in specialised irradiation. The key element of the irradiation carried out in the Halden reactor is the ability to monitor fuel performance parameters by means of in-pile instrumentation. The paper reviews some of the irradiation rigs and the related instrumentation and provides examples of experimental results on selected fuel performance items. In particular, current irradiation conducted on high/very high burn-up fuels are reviewed in some detail

[en] One of the major achievements of the OECD Nuclear Energy Agency (NEA) is the knowledge it has helped to generate through the organisation of joint international research projects. Such projects, primarily in the areas of nuclear safety and radioactive waste management, enable interested countries, on a cost-sharing basis, to pursue research or the sharing of data with respect to particular areas or issues. Over the years, more than 30 joint projects have been conducted with wide participation of member countries. The purpose of this report is to describe the achievements of the OECD/NEA joint projects on nuclear safety research that have been carried out over the past three decades, with a particular focus on thermal-hydraulics, fuel behaviour and severe accidents. It shows that the resolution of specific safety issues in these areas has greatly benefited from the joint projects' activities and results. It also highlights the added value of international co-operation for maintaining unique experimental infrastructure, preserving skills and generating new knowledge

[en] The results of Reactivity-Initiated Accidents (RIA) experiments have been analysed and the main variables affecting the fuel failure propensity identified. Fuel burn-up aggravates the mechanical loading of the cladding, while corrosion, or better the hydrogen absorbed in the cladding as a consequence of corrosion, may under some conditions make the cladding brittle and more susceptible to failure. Experiments point out that corrosion impairs the fuel resistance for RIA transient occurring at cold conditions, whereas there is no evidence of important embrittlement effects at hot conditions, unless the cladding was degraded by oxide spalling. A fuel failure threshold correlation has been derived and compared with experimental data relevant for BWR and PWR fuel. The correlation can be applied to both cold and hot RIA transients, account taken for the lower ductility at cold conditions and for the different initial enthalpy. It can also be used for non-zero power transients, provided that a term accounting for the start-up power is incorporated. The proposed threshold is easy to use and reproduces the results obtained in the CABRI and NSRR tests in a rather satisfactory manner. The behaviour of advanced PWR alloys and of MOX fuel is discussed in light of the correlation predictions. Finally, a probabilistic approach has been developed in order to account for the small scatter of the failure predictions. This approach completes the RIA failure assessment in that after determining a best estimate failure threshold, a failure probability is inferred based on the spreading of data around the calculated best estimate value

[en] A correlation for RIA fuel failure threshold has been derived and compared with recent experimental data. The correlation can be used for UO₂ and MOX fuel and at hot zero power and cold zero power transients, account taken for the different initial enthalpy and for the lower ductility at cold conditions. It can also be used for non-zero power transients, provided that a term accounting for the start-up power is incorporated. A probabilistic approach has been attempted for the cold zero power case. For LOCA, a correlation has been derived expressing the LOCA limit as a function of the hydrogen accumulated in the cladding during base irradiation. However, LOCA experimental data exhibit a significant scatter, possibly because results depend on details in the conduct of the tests. Consequently, predictions of LOCA limits vs. burn-up are affected by appreciable uncertainty, and more data derived from testing of high burn-up cladding specimens are needed in order to reach firm conclusions. (author)

[en] An overview of the experimental work scope at the Halden reactor is provided, together with the most important achievements related to fuel and materials safety and reliability. The most important characteristics of the Halden tests consists in the in-reactor monitoring capability, which provides unique on-line data on specific performance items, such as for instance fuel temperature, fission gas release and pellet-cladding mechanical interaction for fuel rods and crack growth, creep and stress/strain relaxation for structural materials. The type of measurements, the techniques involved, the data and their relevance for the nuclear community are presented and discussed. In particular, licensing items related to high burn-up fuel, such as overpressure criteria and response to LOCA, are addressed. Considerations are also given to the long term availability of the Halden reactor as a key support tool for the nuclear industry in the decades ahead. (author)

[en] The development of the new accident-tolerant fuel for LWRs is a long-term endeavour, which will require a very extensive selection and validation process consisting of: 1 - laboratory assessments; 2 - out-of reactor testing; 3 - in-reactor testing at representative nuclear, thermal-hydraulic and water chemistry conditions. While new fuel compositions might also be considered, it is expected that the focus will be on new cladding materials, mainly because the cladding is the main barrier to radioactive release and because it constitutes the fuel rod overall structural support. The cladding reaction with steam at high temperatures as can occur in beyond design basis LOCA conditions also constitutes a weakness for the cladding materials currently used in the industry. The report is intended to outline the types of in-reactor testing that are currently required when the nuclear industry proceeds in the development of new fuel types or materials, and that are likely to be needed also in the development of ATF fuel. The existing experience at Halden and at other laboratories constitutes the basis for the above. The presentation focuses on in-reactor testing and in particular on the following items: 4 - in-reactor material property assessments; The tests considered under this item can normally be carried out with small size specimens, as well as with one or more fuel rods suitably designed for the purpose. These tests should be carried out at representative LWR conditions or at somewhat more demanding conditions in terms of radiation, temperature or water chemistry environment. Typical tests are: - stress corrosion cracking susceptibility; - corrosion and irradiation growth; - irradiation creep and stress relaxation; 5 - rod integrity in normal operation; These tests are carried out with fuel rod segment, and follow the same procedure used today for validating modified fuel designs. The tests are normally performed at the upper envelope of the normal operating conditions, or at somewhat more demanding conditions. Typical tests discussed in the presentation are: - integrity of un-fuelled rods (for e.g. first SiC cladding qualification); - integrity of fuel rods, normal operation; - PCI/PCMI margin in power ramps. 6 - rod behaviour in transients; Currently, only reactivity initiated accident (RIA) and design basis (DB) LOCA tests are considered for new fuel developments. However, it is expected that ATF demonstrations will also require studies in beyond-design basis (BDB) conditions. Typical tests addressed in the presentation are: - RIA transients; - Design Basis LOCA; - Beyond Design Basis LOCA (severe accident scenarios)

[en] Experimental facilities in nuclear energy are key to addressing safety issues. The recent loss of some critical infrastructure, from facilities to industry expertise, has therefore become a concern for many countries. In response, the Nuclear Energy Agency (NEA) has launched several efforts to address the matter as outlined in this report. Current safety issues, research needs and research facilities associated with currently operating water-cooled reactors in NEA countries are all addressed. Also included is an assessment of the present needs to maintain experimental databases. The Senior Group of Experts on Nuclear Safety Research, which produced this update of the 2007 report on the same issue, noted the success of previous reviews in helping maintain critical infrastructure and make a number of recommendations to preserve key research facilities and capabilities

[en] Highlights: • Three different thorium containing fuel types are test irradiated. • Fuel temperatures, cladding elongation and rod pressures are monitored online. • Addition of 7% thorium to the uranium matrix lowers fuel temperatures. • Measurements also indicate lowered temperatures of thorium–plutonium MOX fuel. - Abstract: Thorium based fuels are being tested in the Halden Research Reactor in Norway with the aim of producing the data necessary for licensing of these fuels in today’s light water reactors. The fuel types currently under irradiation are thorium oxide fuel with plutonium as the fissile component, and uranium fuel with thorium as an additive for enhancement of thermo-mechanical and neutronic fuel properties. Fuel temperatures, rod pressures and dimensional changes are monitored on-line for quantification of thermo-mechanical behavior and fission gas release. Preliminary irradiation results show benefits in terms of lower fuel temperatures, mainly caused by improved thermal conductivity of the thorium fuels. In parallel with the irradiation, a manufacturing procedure for thorium–plutonium mixed oxide fuel is developed with the aim to manufacture industrially relevant high-quality fuel pellets for the next phase of the irradiation campaign

[en] Highlights: • Irradiation testing of uranium oxide fuel enhanced with Cr, SiO₂/TiO₂ and ThO₂. • Irradiation testing of cladding coated with (Fe,Cr,Al) and (Cr,Al) alloys. • All tested materials perform as expected, or better. • Lowered fuel temperatures observed for fuel enhanced with Cr microcell structure. • Unexpectedly low temperatures observed for SiO₂/TiO₂ and 40% ThO₂ enhanced fuel. - Abstract: Enhanced uranium oxide fuel types are being tested in the Halden Research Reactor in Norway with the aim is to assess the effect that these enhancements have on fuel performance. Fuel temperatures, rod pressures and dimensional changes are being monitored online and an extensive post-irradiation examination programme is planned. Preliminary data show that fuel centerline temperatures can be lowered by addition of ThO₂ to the fuel matrix, or by incorporating Cr or SiO₂-TiO₂ as a network structure within the fuel. In parallel, two types of cladding coatings are tested in order to investigate their in-core properties. No abnormal behaviour has been noted during the first 100 days of irradiation.