Fuel and fission product behaviour

Validation of Predictive Methods for Fuel and Fission Product Behaviour in GCRs,... 

The experience base for GCR fuel behaviour under accident conditions was reviewed at an IAEA Specialists Meeting in 1990 [Ref. 2], and a CRP on Validation of Predictive Methods for Fuel and Fission Product Behaviour in GCRs was initiated in 1993. Countries participating in this CRP include China, France, Japan, Poland, Germany, the USA and the Russian Federation. Within this CRP, participants are documenting the status of the experimental data base and predictive methods, cooperating in methods verification and validation and will identify and document the additional needs for methods development and experimental validation data. 

Validation of predictive methods for fuel and fission product behaviour in gas cooled... 

Fuel Performance and Fission Product Behaviour in Gas Cooled Reactors, IAEA-TECDOC-978, International Atomic Energy Agency, Vienna, Austria (1997). 

XV-7] INTERNATIONAL ATOMIC ENERGY AGENCY, Fuel performance and fission product behaviour in gas cooled reactors, IAEA-TECDOC-978, Vienna (November 1997), 246-252. 

It should be noted that as fuel is exposed in the reactor, U-235 densities are burnt down and both plutonium and fission products are produced. This isotopic change can lead directly to some variation in the behaviour of both void and power coefficients. There can also be an important indirect effect when the reactivity is high (e.g. with fresh fuel) and fixed absorbing rods are inserted to compensate, a large thermal reactor tends to behave as a number of small, admittedly linked, reactors with a different balance of capture, leakage and production of neutrons. This indirectly affects the various reactivity coefficients. 

Lewis, B. J. Fundamental aspects of defective nuclear fuel behaviour and fission product release. J. Nucl. Materials 160, 201-217 (1988)... 

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. 

Strasser A, Sunderland D (1992) A review of recent LWR fuel failures. In IAEA Tech Committee Meeting on Fuel Failure in Normal Operation of Water Reactors Experience, Mechanisms and Management, Dimitrovgrad, Russian Federation, 26-29 May 1992 IAEA (1992) Proc IAEA Topical Meeting on Behaviour of Core Materials and Fission Product Release in Accident Conditions in LWRs, Cadarache, France, 16-20 March 1992 Rau P (1993) Application of experiences to the structure of fuel assemblies for WER. In Proc Meeting NUSIM 93, Landshut, 20-22 January 1993... 

Since the water movement will be very slow compared with the rate at which the wastes dissolve, we are concerned first and foremost with equilibrium solubility. Also, if only to relate behaviour on the geological time scale to that on the laboratory time scale, we will need to know about the mechanisms and kinetics of dissolution and leaching. The waste forms envisaged at present are glass blocks containing separated fission products and residual actinides fused into the glass and, alternatively, the uranium dioxide matrix of the used fuel containing unseparated fission products and plutonium. In the... 

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. 

When more them one solute is involved in the consideration of the process design, the situation becomes much more complex since the extraction behaviours of the different solutes will usually be interdependent. In the case of irradiated thermal reactor fuels the solvent extraction process will be dealing with uranium containing up to ca. 4% of fission products and other actinides. These will have only a minor effect on uranium distribution so that a single-solute model may be adequate for process design. However, in some cases nitric acid extraction may compete with U02 extraction and a two-solute model may be needed. In the case of breeder reactor fuels the uranium may contain perhaps 20% of plutonium or thorium. Neptunium or protactinium levels in such fuels may also not be negligible and, under these circumstances, the single-solute... 

Molybdenum provides an example of a fission product which would appear in the dissolver solution in oxidation state (VI). However, the half lives of the molybdenum isotopes produced are sufficiently short for decay processes largely to eliminate this element before fuel reprocessing is undertaken. The extraction of Mo by TBP from aqueous nitrate solutions of pH ().5-3.0 has been investigated. It was found that at Mg concentrations above 10 M, slow aqueous phase polymerization reactions gave rise to time dependent extraction behaviour. Thus Dmo between TBP and a solution containing 1.0 M KNOj and 1.5xlO M Mg at pH 3 rose to 0.1 over a period of 12 hours. At Mg concentrations below 10 M, equilibrium was rapidly attained and in this case was at a maximum of 0.08 between pH 0.5 and pH 1.0, falling off steeply at... 

In-depth analytical studies and irradiation experiments to examine the degradation of fuel pellet thermal conductivity under increased bum-up, swelling of the fission product gas pores generated around Pu-rich spots, and fuel rod deformation behaviour... 

The analyses were conducted for a singe rod, which is assumed to have the highest bum-up in the RMWR core. The models or material properties applied to the analysis, such as fuel thermal conductivity, fission product (FP) gas diffusion and release and creep rate are derived or extrapolated from those in the usual analysis of LWR fuel rods. For the first analysis, particular focus was on the thermal behaviour, such as FP gas release and internal pressure increase induced by the fuel temperature rise. 

The basic principle adopted In the safety design and analysis of the prototype SGHWR has been to understand euid to be able to predict the behaviour of the reactor under all conditions. The vented containment system depends upon the detailed analysis of the course of each accident and the fate of fission products released from the fuel for a demonstration of Its satisfactory performance. 

Less is known about the fate of the fission and neutron capture products that could result in the precipitation of unique alteration phases depending on the availability of these species in the fuel matrix. Burns et al. (1997) theorized that many of the U(VI) alteration phases may be capable of incorporating the long-lived radiotoxic isotopes, including 237Np, 99Tc, and 239Pu. In this chapter, we will discuss the evidence for Np incorporation into U(VI) phases and the behaviour of Pu in corroded spent nuclear fuel (SNF). 

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