[en] Fuel Safety Research Specialists' Meeting, which was organized by Japan Atomic Energy Research Institute, was held on March 4-5, 2002 at JAERI in Tokai Establishment. Purposes of the Meeting are to exchange information and views on LWR fuel safety topics among the specialist participants from domestic and foreign organizations, and to discuss the recent and future fuel research activities in JAERI. In the Meeting, presentations were given and discussions were made on general report of fuel safety research activities, fuel behaviors in normal operation and accident conditions, FP release behaviors in severe accident conditions, and JAERI's ''Advanced LWR Fuel Performance and Safety Research Program''. A poster exhibition was also carried out. The Meeting significantly contributed to planning future program and cooperation in fuel research. This proceeding integrates all the pictures and papers presented in the Meeting. The 10 of the presented papers are indexed individually. (J.P.N.)

[en] The Phebus-FP programme is a wide international effort to investigate LWR severe accident phenomena, through a series of in-pile integral experiments, dealing with fuel degradation, hydrogen production, fission product release and subsequent transport/deposition in the Reactor Coolant System down to the containment building. Three tests (FPT0-2) simulating a low-pressure cold-leg break have so far been successfully performed, which provide experimental data of high interest to characterise and improve LWR safety in the event of a severe accident. The data analysis of FPT0 and FPT1 is now terminated (and available in public domain literature for FPT0), while that of FPT2 is still under progress. The present paper gives a brief overview of the lessons learnt so far regarding fission product release, emphasising latest experimental results on the FPT2 test. (author)

[en] The Phebus FP (fission Product) project is an international reactor safety project led by IPSN Cadarache and with European Commission as partner along with other EU national institutes as well as 5 other non-EU countries. The project simulates severe reactor accidents using a 20-rod bundle of medium burn-up fuel rods, and monitors the degradation of the bundles as well as the release of fission products and structural materials along the simulated primary circuit and into the containment. ITU (Institute for Transuranium Elements) Karlsruhe was involved in the macroscopic and microscopic examination of the degraded FPT1 bundle and the molten pool of corium, and the examination of deposits from various filters positioned along the primary circuit mainly the hotter (upstream) points. An examination of the vertical line deposits directly above the overheating bundle were also carried out. These were examined as part of a supplementary shared cost action project (Phebus Revaporisation Project) examining the potential for volatilisation of deposits at later stages of a reactor accident. Selected results will be given to illustrate the main features of the reactor accident and comparison with the initial, base-line test (FPT0-using only trace-irradiated fuel) indicates the degradation mechanisms. That is, how fuel rods deform, and liquid phases (corium) collects to form a pool, as well as how the fission products are released and deposited along the primary circuits and that these distributions may change in the course of the accident. (author)

[en] The research of Zr1%Nb fuel rod claddings of VVER type reactor behaviour and characteristics in loading conditions simulating accidents with loss of the coolant and active zone quenching experimental data are presented. The experimental data on the deformation and oxidation behaviour of Zr-1%Nb fuel rod claddings in LOCA simulating conditions are illustrated. As for the thermal shock experiments under loading conditions simulating accidents with loss of the coolant (temperature, environment, deforming, limitation of the claddings axial deforming, quenching rate, irradiation) experimental data are presented. The experimental data allow to estimate type and numerical value of the embrittlement criteria parameters (ECR) taking into consideration Zr1%Nb fuel rod claddings resistance during quenching and further fuel rod claddings assemblies removing and transportation. It is shown, that the mechanical characteristics of the oxidized unirradiated and irradiated claddings material (Zr1%Nb) after thermal shock (residual ductility) are sufficient to withstand quenching and further removing and transportation. (author)

[en] UO₂ fuel rod performance analysis code, INFRA (INtegrated Fuel Rod Analysis) has been being developed since 1997. Fuel performance models such as the prediction of rim microstructure formation, thermal conductivity degradation in the rim microstructure region, radial power and burnup distribution inside the pellet, mechanistic fission gas release model, fission gas bubble swelling, cladding corrosion and creep-out models, and finite element analysis model for pellet cladding mechanical interaction were developed to predict UO₂ fuel behavior which occurs or becomes more important at high burnup. INFRA code was verified by comparison with the in-pile fuel test data for both the specific models and the integral parameters such as fuel temperature and fission gas release. It showed that INFRA code could predict the UO₂ fuel behavior quite well up to the burnup higher than 60 MWD/kgU-rod avg.. (author)

[en] The Expansion-Due-to-Compression (EDC) test has been developed, by Studsvik Nuclear, for the study of irradiated cladding failure, under hoop strain rates as high as 1 to 10 s^-1. During this test, which is proposed for the mechanical simulation of a reactivity initiated accident or in-pile test, a piece of cladding tube is circumferentially loaded in tension due to the expansion of a polymer pellet, axially compressed inside the tube. A finite element simulation of the EDC-test has been performed. Considering the material properties at room temperature, both zirconium alloy cladding and polymer pellet were considered to be elastic-plastic. Three different values of cladding material yield stress were considered as well as both hardening and non-hardening behavior. Thus the effect of irradiation on cladding material properties is implicitly taken into account. The distributions of important field quantities, with respect to the damage of the cladding, as well as the evolution of their maximum values, during loading, are studied. It is shown that, before cladding yielding as well as after substantial plastic deformation, the radial displacement, on the external surface, and the total energy per unit volume, when appropriately normalized, vary along the cladding axis according to specific distributions, which do not depend on the level of loading. The calculated profile of the cladding, after deformation, is in agreement with experimental measurements. (J.P.N.)

[en] Formerly, nuclear safety models were designed with wide margins. This enabled to cope with the uncertainties both in the database and in the knowledge of the phenomena prevailing in safety transients. Further, conservative scenarios were taken as a basis and standard setting for regulatory deliberations. The margins provided by this conservative approach, however, risk of being gradually eroded by more aggressive demands on plant and core operation posed by the industry. This will require not only that the database for accident analyses is extended to broader conditions and materials, but also that the uncertainties associated with such database and its use for specific applications is adequately quantified. Considering fuel safety aspects, licensing guidelines for nuclear design are primarily directed toward ensuring that various fuel design limits or acceptance criteria are not exceeded during normal operation or during anticipated operational transients. As deigns and operational requirement evolve, these guidelines need continuous update through experiments devised to improve or extend the database. This might involve small-scale, out-of-pile examinations or experiments, normally intended to provide materials property data. However, this is largely insufficient for most safety (and reliability) applications, which do require representative reactor conditions. To this end, experiments must be conducted in test reactors. For fuel safety tests, single or multiple test arrangements can be considered depending on the type of investigation. In general, integral in-pile experiments run with representative materials and at representative conditions are essential for accident analyses. But equally important is the analytical effort that should always accompany the experimental work, thanks to separate effects tests. The results allow the final code assessment in terms of reactor applicability and simulation completeness. According to this context, IPSN (Institute de Protection et de Surete Nucleaire) is preparing new experimental programmes in the CABRI and PHEBUS facilities. This paper provides an overview of the IPSN future programmes in the PHEBUS facility which consist in a two aspects. The High Burnup Fuel LOCA (loss of coolant accident) experimental programme. One of the important aspect of this programme is in-Pile experiments involving bundle geometries. The Phebus-2000 programme. Its aim is to provide new information about the impact of high burn-up and MOX fuels and of reflooding of a degraded core on core additional degradation, potential threat to containment and Source Term. (J.P.N.)

[en] The HCE4 (Hot-Cell Experiment 4) experiment was performed to study fission-product releases (FPR) from CANDU fuel at 1650degC. Three sets of tests were conducted to assess the effects of fuel element length, atmosphere, and heating rate on fission-product release. The fuel samples were from the Darlington reactor (burnup of ∼8.8MWd/kgU, maximum linear power of 42 kW/m) and an irradiation in the NRU reactor (burnup of 6.3MWd/kgU, maximum linear power of 64kW/m). The first set of tests was performed to assess the effect of fuel element length on fission-product release from clad samples of CANDU fuel at 1650degC in a 76mol% steam and 0.5mol%H₂ (balance Ar) environment. For most fission products released under the test conditions (Xe, Kr, Cs and I), no significant differences were found between the releases from clad fuel samples of different lengths (20 and 100 mm). The fractional Te-129m release seems to have been higher from one of the shorter fuel samples, but the uncertainty in this result is large due to poor counting statistics. The difference in the Te-129m releases may arise from its affinity for unoxidized Zircaloy and local differences in the extent of clad oxidation. These results indicate that it may not be necessary to account specifically for any effect of fuel element length on releases of volatile fission products from the fuel in CANDU safety analysis. The possible exceptions to this conclusion are those species that are known to have a specific chemical interaction with the cladding that may be affected by steam ingress into the fuel element, e.g., Te and Sb. The second set of tests was performed to assess the effect of environment (inert, steam and air) on fission-product release from clad samples of CANDU fuel at 1650degC. The release of the volatile fission products (Kr, Xe, Cs and I) from the clad fuel samples were low (<10%) in the Ar/2%H₂ environment, and high (75-100%) in the steam and air environments. The Te-129m release in the inert atmosphere test was negligible. The Te-129m releases in the oxidizing atmospheres varied significantly (30-80%) and did not appear to depend on the nature of the oxidant (steam or air). Some Ru release (∼15%) was observed beginning about 3600 s after the onset of rapid Cs release in the air environment. No Ru release was observed in the inert or steam atmosphere tests. No statistically significant releases of Ba, Eu, La, Nb, Nd, Pr, Y and Zr were observed in these tests. The third set of tests was performed to assess the effect of heating rate on fission-product release from full-pellet-cross-section fuel fragment samples in an inert (Ar/2%H₂) environment. Good results were obtained on releases of volatile fission products (Cs, I) in these tests, but the noble gas (Kr, Xe) release data are not as reliable because of experimental difficulties. There appeared to be little effect of heating rates between 0.2 and 6degC/s on FPR from fuel samples heated to ∼1650degC in an inert (Ar/2%H₂) atmosphere. Heating rate effects are thought to be more significant for fission products located on grain boundaries and the fuel used in the HCE4 experiment had a relatively small grain-boundary inventory. The data from the HCE4 fission-product release experiment appear to be suitable for use in the validation of the SOURCE fission-product release code. (J.P.N.)

[en] In 1998 NOK (Nordest Schweizeische Kraftwerke) modified Beznau Nuclear Power plants operating strategy for a cycle concept with an annual reload batch of 20 Fuel Assemblies (1/6 of the core). This approach has a significant benefit for back-end cost but is faced with two current licensed limits, fuel burn-up level and enrichment. A surveillance programme has been proposed on a series of lead test fuel Assemblies, UO2 and MOX, having an enrichment of up to 4.55 w/o 2³⁵U and 4.75 w/o Pu_fiss respectively. The fuel enrichment limit had to be raised on basis of acceptable calculations for the matrix of stored fuel in the pools having sufficient conservatism, both in the assumptions made and the methodologies applied. An overview of the surveillance programme estimated to run up to 2007, and the Burn-up credit approach used to increase the enrichment limits in Beznau pool facilities are presented in this paper. (author)

[en] A large quantity of experimental data on fission gas release is now available in the public domain. It covers a wide variety of fuel types and burnups of up to more than 70 GWd/tU. This data, together with gas release measurements from British Energy's AGRs, has been used to build a comprehensive validation database for the fuel performance code ENIGMA. Validation of ENIGMA version 5.11 against this database has identified a requirement for model development to improve predictions at high burnup. A modified gas release model has been produced and tested. (author)