Fission release from fuel

A volatile iodine species, neither elemental nor organic, which has been found in steam/air atmospheres, has been identified as hypoiodous acid (HOI) (Cartan et al., 1968). In water-cooled power reactors, any fission products released from fuel will pass into hot alkaline water and thence to a steam-air mixture. These conditions are thought to favour the formation of HOI (Keller et al., 1970), but the evidence is indirect. For example, tests for elemental iodine or iodine with an oxidation state higher than that of HOI gave negative results. 

Accidents 4 and 5 both involve release of fission products from fuel failures induced by off-design transients. It is of interest that the applicant... 

As to accident 9, here, of course, is the crux of the matter so far as population hazard is concerned. One is left to make up one s own mind as to what parameters to assume—as is the case with what may be thought to be corresponding though different circumstances for gas-cooled reactors (see section II,B). At the present state of the art, this is perhaps inevitable. So far as fission product release from fuel is concerned, we clearly have to consider fuel at temperatures up to the melting point of uranium dioxide, 2800 C some representative values are given in Section III. 

The models and material property data for predicting fission metal release from fuel particles and fuel elements are described in Ref. 4. The transport of fission metals through the kernel, coatings, fuel rod matrix, and fuel element graphite is modeled as a transient diffusion process in the TRAFIC code (Section 4.2.5,2.2.1.2). The sorption isotherms which are used in the calculation of the rate of evaporation of volatile metals from graphite surfaces account for an increase in graphite sorptivity with increasing neutron fluence. 

The thermal power is 450 MW. For "° Ag inventory calculation, plutonium fission fraction and thermal flux are assumed to be 60% and 4x lO cm s , respectively. Fractional release from fuel... 

Highly heat-conducting fuel rods - for reducing the maximum operating fuel temperature to T < 1,000°C, which provides a small (less than 7%) fission gas release from fuel and its low pressure on claddings, as well as a relatively small amount of thermal energy in fuel. 

Fission product containment 1 Fuel clad/sheath 2 Heat transport system pressure boundary 3 Containment envelope P Multiple bamers to release with m-servace inspection mcludmg activi release from fuel, pressure tube leak-bef(xe-break and on-lme detection of leak rate from contamm t... 

The fission gas bubbles present at the grain boundaries can be liberated by mechanical impact. Ruhmann et al. (1987) reported that by impact testing of typical LWR fuel pellets (irradiated to a burnup of 33 MWd/kg U) at ambient temperature with a specific load of about 1000 N m/g, 6 to 7% of the total Kr inventory of the pellet was released. Since these investigations were directed to the study of fission gas release from fuel caused by accidental mechanical impact, the analytical technique used did not aim at a quantitative liberation of the gases present at the grain boundaries. For this reason, the reported values are assumed to represent only a small fraction of the total grain boundary inventory. 

Cesium and iodine atoms which are released from fuel specimens into a high-temperature steam-hydrogen environment are thermodynamically unstable and will be rapidly converted into species that are stable under these conditions. Since the chemical form of iodine in particular will considerably influence its transport and retention behavior within the reactor pressure vessel and the primary system, it is important to know the kinetics of these conversion reactions. A kinetics assessment of the most essential reactions (Cronenberg and Osetek, 1988) has shown that for extremely low concentrations of iodine and cesium in steam (e. g. mole ratio I H2O < 10" ), the predominant form of iodine is HI and that of cesium is CsOH. This is due to the fact that because the concentrations of iodine and cesium are so dilute, the elements are much more likely to collide and react with H2O and H2 than with each other. Low concentrations of iodine and cesium increase the time for thermochemical equilibrium to be established for their reaction products. For mixtures which are so dilute in fission products, the reaction times may approach tens of seconds or longer, so that for high effluent rates the environmental conditions may change (e. g. by transport into the next control volume showing other conditions) before thermochemical equilibrium has been achieved. Under such conditions, certain limitations caused by reaction kinetics may exist. 

Hobbins, R. R., Petti, D. A., Hagrman, D. L. Fission product release from fuel under severe accident conditions. Nucl. Technology 101, 270-280 (1993)... 

The MARS safety concept provides for retaining of radionuclides in the fuel both in normal operation and accidents, so that radiation exposure of the personnel and population around the NPP falls within the limits prescribed by the regulations. Barriers to the release of fission products from fuel are provided primarily by the coatings applied to each fuel particle. The required retention of radioactivity is ensured by the long-term operability of the fUel elements, which maintain their performance at temperatures up to 1250 C with the fast neutron (E>0.18 MeV) fluence to the coatings of up to 2.2 10 cm, and by quality assurance during fuel fabrication. 

The N Reactor used a confinement system based on the concept of releasing the Initial burst of steam resulting from a postulated reactor-coolant-pipe break. When the confinement pressure subsided, the steam vents were closed and ventilation valves opened. The ventilated steam was filtered through charcoal and high-efficlency filters to prevent any release of fission products from fuel failure (WHC 1989a). Figure 3-14 Illustrates the confinement system process. 

Considerable reactor operating experience at Hanford combined with an analysis of the limited nund>er of non-standard operations, which have occurred suggests that accidents which could result in significant fission product release from fuel elements in the reactor would most likely result... 

It is possible, on the basis of.the data discussed above on fuel behaviour, fission product release from fuel, containment leak rates, and fission product cleanup, to assess the lodlne-131 release to atmosphere for a range of possible accidents. The basic fission product release assessments are made assuming that all protective systems work as intended. 

The purpose of VICTORIA is to enable the prediction of the magnitude, chemistry, physical properties, and timing of fission product release from the RCS of a nuclear reactor undergoing a severe accident. VICTORIA does not treat thermal-hydraulics, but requires such information as input. The heart of the code is in its mechanistic treatment of fission product release from fuel, chemistry, aerosol physics, transport, and decay heating. The coupled treatment of these phenomena make VICTORIA unique in its predictive capabilities. The ultimate goai is to identify, wherever possible, the essential physics and to develop simplified models that can be incorporated into system-level codes that can treat complete accident sequences. 

Table 4.1. Aerosol and Fission Product Vapors Released from Fuel in ST-1... 

Off-Gas Treatment. Before the advent of the shear, the gases released from the spent fuel were mixed with the entire dissolver off-gas flow. Newer shear designs contain the fission gases and provide the opportunity for more efficient treatment. The gaseous fission products krypton and xenon are chemically inert and are released into the off-gas system as soon as the fuel cladding is breached. Efficient recovery of these isotopes requires capture at the point of release, before dilution with large quantities of air. Two processes have been developed, a cryogenic distillation and a Freon absorption. 

Relocation of the decay heat source as fission products are released from the fuel and transported to the containment. 

Parker, G. W. etal., Out-of-Pile Studies of Fission Product Release from Overheated Reactor Fuels at ORNL, 1955-1965, ORNL-3 981 (1967)... 

Concern about fission-product release from coated reactor fuel particles and fission-product sorption by fallout particles has provided stimulus to understand diffusion. In a fallout program mathematics of diffusion with simple boundary conditions have been used as a basis for (1) an experimental method of determining diffusion coefficients of volatile solutes and (2) a calculational method for estimating diffusion profiles with time dependent sources and. time dependent diffusion coefficients. The latter method has been used to estimate the distribution of fission products in fallout. In a fission-product release program, a numerical model which calculates diffusion profiles in multi-coated spherical particles has been programmed, and a parametric study based on coating and kernel properties has provided an understanding of fission product release. 

An important parameter in the evaluation of the safety of a reactor system is the release of fission products from the fuel. The fuel in the high-temperature gas-cooled reactor (HTGR) consists of spherical particles (U, ThC2) that are coated with a material presenting a diffusion... 

To evaluate fission product release in a reactor, it is necessary to supply the appropriate particle geometry, diffusion coefficients, and distribution coefficients. This is a formidable task. To approach this problem, postirradiation fission product release has been studied as a function of temperature. The results of these studies are complex and require considerable interpretation. The SLIDER code without a source term has proved to be of considerable value in this interpretation. Parametric studies have been made of the integrated release of fission products, initially wholly in the fueled region, as a function of the diffusion coefficients and the distribution coefficients. These studies have led to observations of critical features in describing integrated fission product releases. From experimental values associated with these critical features, it is possible to evaluate at least partially diffusion coefficients and distribution coefficients. These experimental values may then be put back into SLIDER with appropriate birth and decay rates to evaluate inreactor particle fission product releases. Figure 11 is a representation of SLIDER simulation of a simplified postirradiation fission product release experiment. Calculations have been made with the following pertinent input data ... 

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