Fission product release

Rodgers, S. J., Udaveak, R. J. and Mausteller, J. W. In International Symposium of Fission Product Release and Transport under Accident Conditions, Oak Ridge National Laboratory, TN, USA, 1965, pp. 1204 1215. 

Code of Federal Regulations (lOCFR) part 100 provides reactor siting criteria. It specifies that the fission product release calculated for major hypothetical accidents shall produce a w hole... 

Computer Codes for Fission Product Release and In-Plant Transport... 

An accident sequence source term requires calculating temperatures, pressures, and fluid flow rates in the reactor coolant system and the containment to determine the chemical environment to which fission products are exposed to determine the rates of fission product release and deposition and to assess the performance of the containment. All of these features are addressed in the... 

Of these phenomena, the first three in particular, involve thermal hydraulics beginning with the pre-accident conditions. Items 4 through 7 address the meltdown of the core and its influence on (1) hydrogen production, which affects containment loads, (2) fuel temperatures, which affect in-vessel fission product releases, (3) thermal-... 

Computer sensitivity studies show that hole size strongly affects the fraction of fission products released from the containment. The failure location determines mitigation due to release into another building in which condensation and particulate removal occur. The quantity released depends on the time of containment fails relative to reactor vessel failure. If containment integrity is maintained for several hours after core melt, then natural and engineered mechanisms (e.g., deposition, condensation, and filtration) can significantly reduce the quantity and radioactivity of the aerosols released to the atmosphere. 

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

Spheres Diffusion-Controlled Fission Product Release and Absorption... 

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. 

Fission Product Release from Coated Particles... 

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 ... 

Postactivation annealing fission product release data can be analyzed on the basis of a model that assumes the release to be a diffusion-con-trolled process from a spherical particle. The release is limited by diffusion either in the kernel or in the coating. This model is discussed by Norman and Winchell (9) in another chapter in this volume. 

Postactivation annealing fission product release measurements have shown that the excellent retention of iodine and noble fission gases by pyrolytic carbon-coated fuel particles is unaffected by irradiation up to 24% FIMA. These data are interpreted as indicating that the pyrolytic carbon coatings remain undamaged by these high levels of irradiation. 

Maximum Release. The analytical model described above assumes that the liquid phase is completely stagnant. While this may be true in an ideal laboratory experiment where a small sample can be kept isothermal at a specified temperature, in large scale systems where non-isothermal conditions exist, both natural convection and molecular diffusion will contribute to mass transfer. This combined effect, which is often very difficult to assess quantitatively, will result in an increase in fission-product release rate. Therefore, in making reactor safety analyses, it is desirable to be able to estimate the maximum release under all possible conditions. 

Equilibrium Vaporization. The cesium release results presented in this chapter may also be used to demonstrate our earlier conclusion that equilbirium vaporization represents the upper limit for the fractional fission-product release as a function of sodium vaporization. Figure 6 shows three cesium release curves. Curve A was calculated from the Rayleigh Equation in conjunction with the partial molar excess free energy of mixing of infinitely dilute cesium—sodium solutions reported... 

In PWRs, the fuel is U02, enriched typically to 3.3% 235U while for BWRs, the fuel is U02, enriched to 2.6%. (Natural uranium is 0.72% 235U). The fuel elements are clad in Zircaloy, a zirconium alloy that includes tin, iron, chromium, and nickel that prevents fission product release and protects them against corrosion by the coolant. The control rod material in BWRs is B4C, while PWRs have Ag-In-Cd or Hf control materials. 

Osborne, M.F., Collins, T.L., Lorenz, R.A. Strain, R.V. (1986) Fission product release and fuel behaviour in tests of LWR fuel under accident conditions. In Source Term Evaluation for Accident Conditions, IAEA, Vienna, pp. 89-104. 

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. 

Public interest in radioactive aerosols began in the mid-1950s, when world-wide fallout of fission products from bomb tests was first observed. The H-bomb test at Bikini Atoll in 1954 had tragic consequences for the Japanese fisherman, and the inhabitants of the Ronge-lap Atoll, who were in the path of the fallout. In 1957, radio-iodine and other fission products, released in the accident to the Windscale reactor, were tracked over much of Europe, and these events were repeated on a much larger scale after the Chernobyl accident. 

Reactor accidents and fission product release (after Vinjamuri et al., 1982)... 

The second step is to dissolve the metal oxide fuel using strong nitric acid. The object is to bring all the fission products, uranium, and transuranics, into solution to feed the extraction process. Some of the fission products exceed solubility limits and the fine solids formed must be removed before extraction. Provisions to recover nitrogen oxides and collect gaseous fission products released during this step must be in place. The stainless steel and zircaloy fuel jackets from the fuel assemblies do not dissolve and are separated from the solution, washed, checked for radioactivity, and packaged for disposal as low-level radioactive waste. 

The sorption of the fission products Cs and Sr by the graphitic materials, from which the core and the fuel elements of High-Temperature Gas-Cooled Reactors (HTGRs) are made, is important for the prediction of fission product release in the case of an accident. Hilpert et al. [564, 565] determined, therefore, Cs and Sr partial pressures over such graphitic materials with different Cs and Sr concentrations. The vaporization enthalpies obtained showed a strong chemisorption of Cs and Sr by these materials. The vaporization enthalpy of Sr exceeds that of the pure Sr metal by about 210kJmol at 1500 K, if fuel element matrix graphite with a Sr concentration of 4.0 mmol kg is considered [564]. This value for Cs amounts to about 230 kJmol" at 1250 K for a similar concentration of 4.2 mmol kg[565]. In addition, sorption isotherms were evaluated. 

Safety is clearly a major consideration and research reactors are designed to fail-safe to prevent fission product release. Reactors operate under a triple containment philosophy. The first container is the cladding of the fuel itself, the second is the swimming pool which is made from heavy, 1.5 m thick, concrete lined with stainless steel. Finally the whole reactor is housed inside a reinforced building that is kept at a slightly sub-ambient pressure and is accessed by an air-lock. 

The importance of corrosion product mass transfer was realized first in the early operation of NRU. Here the solubility of the oxide formed on the aluminum fuel sheathing led to the production of a colloidal alumina floe in the heavy water. The mechanism for its formation, means to control it, and the role it played in transporting uranium and fission products released from failed fuel were studied (55, 56). 

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