PWR Accident Research Articles

Behaviour of fission products under severe PWR accident conditions. The VERCORS experimental programme—Part 3: Release of low-volatile fission products and actinides

The VERCORS analytical programme consisted of a series of tests carried out on irradiated PWR fuel samples. The tests – funded jointly by EDF and IRSN – were carried out by the Commissariat à l’Energie Atomique (CEA) at their Grenoble site. They were performed in a hot cell belonging to the Active Materials Analysis Laboratory (LAMA). The general outline of the programme was set out in a first article (of a series of 3), which described the different levels of fission products (FP) volatility and their characteristics. This led to a classification into five main categories of volatility and/or behaviour: (1) Volatile FP including fission gases, iodine, caesium, antimony, tellurium, cadmium, rubidium and silver; (2) Semi-volatile FP, a category made up of molybdenum, rhodium, barium, palladium and technetium; (3) Low-volatile FP comprising ruthenium, cerium, strontium, yttrium, europium, niobium and lanthanum with generally low but significant release; (4) Non-volatile FP including zirconium, neodymium and praseodymium; and lastly (5) Actinides which group together uranium, plutonium, neptunium, americium and curium. The specific behaviour of fission gases and volatile FP is dealt with in the second article, which also includes the specific characteristics of volatile FP regarding transport. The main variables (i.e. temperature, which is the main variable at least until loss of sample geometry, oxidising–reducing conditions, burn-up, interactions with the cladding and/or the structural components, the nature of the fuel, and finally the state of the fuel) affecting the kinetics and/or the released fraction of these same FP could also be identified. This final article represents the Third Part of the series. It concerns the release of actinides and less volatile FP, in keeping with the classification by categories previously identified, which are as follows: (1) semi-volatile FP, comprising of Mo, Ba, Rh, Pd, Tc, (2) low-volatile FP, comprising of Sr, Y, Nb, Ru, La, Ce, Eu, (3) non-volatile FP, comprising of Zr, Nd and Pr and lastly (4) actinides. The main parameters favouring their release are also highlighted.

Behaviour of fission products under severe PWR accident conditions: The VERCORS experimental programme—Part 2: Release and transport of fission gases and volatile fission products

The VERCORS programme is an analytical experimental programme focusing on the release of fission products (FP) and actinides from an irradiated fuel rod during a severe accident in a PWR. Following on from the HEVA programme, VERCORS was conducted in a similar manner in a high activity cell at the CEA Grenoble Centre over a 14-year period between 1989 and 2002. A total of 17 tests were carried out, in 3 experimental phases, on a facility that continued to evolve over time. The instrumentation progressively become more sophisticated, thus making it possible to extend the knowledge database to include the transport of FP (HT loop), or on the contrary focus on the release objective by associating the study of actinides (RT loop). The general description of the programme has already been presented in a previously published article, which is the first in a series of three articles on the same subject. The specific behaviour of fission gases (Xe and Kr) and volatile FP (I, Cs, Te, Sb, Ag, Rb and Cd) in terms of global release, release rate and transport is the subject of this second article. The third and the final part of this series of articles focusing on the VERSCORS programme will study the release of actinides and less volatile FP, respecting the classification previously identified, i.e. (1) semi-volatile FP composed of Mo, Ba, Rh, Pd, Tc, (2) very low-volatile FP composed of Sr, Y, Nb, Ru, La, Ce, Eu, and finally (3) non-volatile FP composed of Zr, Nd and Pr.

Behaviour of fission products under severe PWR accident conditions VERCORS experimental programme—Part 1: General description of the programme

VERCORS is an analytical experimental programme focusing on the release of fission products (FP) and actinides from an irradiated fuel rod, under conditions representative of those encountered during a severe PWR accident. The 17 tests – financed jointly by EDF and IRSN – were conducted by the CEA on its Grenoble site in a specific high-activity cell at the Laboratory for Active Materials (LAMA) over a 14-year period (1989–2002), in accordance with three test phases. A first series of six tests (VERCORS 1–VERCORS 6) was conducted between 1989 and 1994 on UO 2 fuel close to the relocation. Next, two different test series – VERCORS HT (three tests) and RT (eight tests) – were alternately performed between 1996 and 2002 at higher temperature up to the fuel sample collapse. These tests focused on UO 2 and MOX fuels with a variety of initial configurations (intact or debris beds). This programme made it possible to precisely quantify fission product releases in all the situations explored, as well as to identify similar behavioural patterns between some of these fission products, thus making it possible to classify them schematically into four groups with decreasing volatility: (1) volatile FP including fission gases, iodine, caesium, antimony, tellurium, cadmium, rubidium and silver with very high releases (practically total release) at temperatures of around 2350 °C; (2) semi-volatile FP, a category composed of molybdenum, rhodium, barium, palladium and technetium with releases of 50–100%, but very sensitive to oxygen potential and with marked redeposits nearby the emission point; (3) FP that are low volatile, such as ruthenium, cerium, strontium, yttrium, europium, niobium and lanthanum, with significant releases of around 3–10% on average, but capable (for some elements and under particular conditions) of reaching 20–40%; (4) non-volatile FP composed of zirconium, neodymium and praseodymium, for which no release can be measured by gamma spectrometry for the envelope conditions of the VERCORS test grids. Actinides each have their own type of behaviour. They can nevertheless be subdivided into two categories, the first including U and Np, with releases of up to 10% and behaviour similar to that of the low volatile FP, and the second (Pu) with very low releases, typically less than 1%.

Experimental study and modelling of liquid ejection through orifices by sparging gas

Corium coolability after a severe PWR accident involving core meltdown and RPV failure is one of the main issues in nuclear safety. The case considered here is a situation in which the corium is assumed to spread over a concrete floor and is flooded by water. In this framework, many research efforts are performed to study the physical phenomena, which may enhance the heat transfer between the corium and the water pool. Among these phenomena, melt entrainment above the corium crust by the sparging gas, released by the concrete ablation appears to be a potentially efficient cooling mechanism. The main target of the PERCOLA experimental program is to provide qualitative and quantitative information on this entrainment phenomenon. It consists in analytical tests performed with simulant materials of increasing viscosity to simulate the enrichment of the corium in SiO 2 during the MCCI. After a short reminder about the assessment of the efficiency of the ejection mechanism, the second part of this paper is devoted to a general description of the test program and to the presentation of the main results. In the last part, attention focuses on modelling the liquid entrainment phenomenon and on comparison between the experimental data and the calculation results of two different entrainment models.

Modeling oxidic molten core-concrete interaction in WECHSL

The WECHSL code aims at modeling the physical and chemical phenomena governing the molten core-concrete interaction in a severe reactor accident, when the molten core has penetrated the pressure vessel. Attention has recently been drawn to the influence of the melt properties, among others of the viscosity of oxidic melts, on the heat transfer, and on the melt behavior during MCCI. In the first step, the viscosity model in WECHSL was changed in order to better fit the experimental viscosities. To take into account the variation of the viscosity with temperature, an iterative scheme has been introduced in WECHSL to determine the gradient from the pool bulk temperature to the interface temperature in the melt boundary layer. In addition, the influence of porosity of the melt on the thermal conductivity was modeled. Finally, a method was implemented to correct the heat transfer relations. In this paper the WECHSL code including the above modifications is applied to a PWR accident scenario.

Effect of rainwater on the remobilization and dissolution of 134Cs and 85Sr contained in aerosols similar to those discharged in a nuclear accident and deposited in an urban environment

An accident in a PWR could lead to discharge of radioactive aerosols into the atmosphere. The present study consisted of establishing, under laboratory conditions, how the caesium and strontium contained in such aerosols are remobilized by rainwater after dry deposition in an urban environment. Aerosols representative of those discharged in a PWR accident were generated and deposited on five building materials used in France. A rainfall simulator was used to provide varying intensities of rain. The percentages remobilized, both as particles and dissolved in rainwater, were then measured. Remobilization is dependent on the nature of the material, the radioelement and the intensity of the rain. It varies from 5 to 60% for caesium and from 5 to 30% for strontium. Remobilization increases as rain intensity decreases. Caesium is mainly remobilized in dissolved form while strontium is remobilized in the form of particles.

Investigation of PWR accident situations at BETHSY facility

For several years an extensive programme of separate effect tests related to PWR safety has been conducted in France and in particular at the Nuclear Center of Grenoble. Recently the BETHSY integral test facility - three identical loops, full height and pressure - has been constructed with the main objectives of contributing to the verification of the CATHARE calculation code of accidents and to the validation of the physical bases of Emergency Operating Procedures. So far several tests have been (November 1988) carried out, among which natural circulation under various conditions (single phase, two-phase, symmetric or asymmetric conditions, different core power, variable steam generator liquid level), 2 inches cold leg break, steam bubble formation and collapse in the upper head. Summarized results of some of these tests are introduced and compared to a first CATHARE calculation as far as the 2″ cold leg break is concerned.

Oxidation kinetics and related phenomena of zircaloy-4 fuel cladding exposed to high temperature steam and hydrogen-steam mixtures under PWR accident conditions

With respect to the behavior of Nuclear Pressurized Water Reactor fuel cladding during accidents the oxidation kinetics of Zircaloy-4 tubing in steam and hydrogen-steam mixtures and the related changes in the mechanical properties have been investigated. Short tube sections were exposed to steam in surplus between 600 and 1600°C under isothermal and temperature transient conditions and to steam of limited supply and hydrogen-steam mixtures between 800 and 1300°C under isothermal conditions. Tube capsules were creep-rupture tested in steam under isothermal/isobaric (600–1300°C, 7–150 bar) and various temperature/pressure transient conditions. The oxidation kinetics of Zircaloy-4 is described in case of short-term reaction by simple rate laws. The long-term behavior is proved to be highly influenced by the breakaway transition and finally by total consumption of the tube wall. On the basis of the isothermal behavior the short-term temperature transient case can be understood and also modelled. Deviations from predictable behavior are the consequences of Zr- and ZrO 2-phase transformations. The protective character of preexistent scales is lost above 1100°C for the same reason. The limits of steam supply rate dependent oxidation and its interrelation with the competing hydrogen take-up are described. The consequences of oxidation on the mechanical properties are increased strength and decreased ductility, whereas the response to creep deformation is an overall increase in oxidation due to oxide cracking.