Release and detection of fission products from defective fuel pins

To test the long term behaviour of UO₂-pins with artificial cladding failures a special FR2 inpile steam loop was built. The individual activity concentrations of gaseous and volatile fission products were measured with the aid of a Ge(Li)-system and in addition the integral activity concentration was recorded by a precipitator. Typical results and time behaviours are given. With the aid of a prototype-DND-monitor for the SNR the delayed neutrons of the fission products are measured. Considerable deviations hitherto unknown were observed between the measured values and the values calculated according to the recoil model. The ratio of the both values, called k factor, was in the range of 0.7–533; in the case of fresh fuel the k factor decreases to one-tenth of the initial value after one or two weeks; the k factor of fuel with a burn-up of 46 000 MWd/t was only 1/30 compared to unirradiated fuel and about 1.In the meantime, also in the case of sodium cooling, k factors > 1 were observed, and the GfK defect pin program will focus attention on this effect. The release values of the gaseous fission products were about the same as the values for the neutron precursors, Br and J.     

Reduction of fission gas release at extended burnups in light water reactors by decreasing fuel temperature

The effects of fuel temperature on fission gas release in light water reactor UO₂ fuel at extended burnups of up to 56 effective full power months (EFPMs) are evaluated using a simple fission gas release mechanistic model. The model is first described and then model validation comparisons are made against experimental fission gas release date. The study shows that by decreasing the maximum operating fuel temperature to below 1200°C, it is possible to reduce the amount of released fission gas at 56 EFPMs to less than that at the current design burnup of 36 EFPMs.     

Silicon carbide encapsulated fuel pellets for light water reactors

In this paper a new fuel element concept is proposed. In addition to the cladding tube the fuel is encapsulated in SiC capsules. The basis of the concept is a bonding technique that allows to join SiC without loosing the excellent features of the SiC base material like high mechanical strength and good oxidation resistance. The results of different tests and calculations performed in the on-going R&D program with regard to the mechanical, thermal, chemical and neutron-physical behaviour have shown that this fuel element concept fulfils the safety requirements like negative fuel temperature coefficient and negative moderator void coefficient. Furthermore the maximum fuel temperatures can be reduced. Due to a new developed joining technique, which is one of the major topics of the R&D program the density of the ceramic encapsulation remains unchanged as well during irradiation tests as under oxidation and thermal shock loading conditions.     

Plutonium isotopic composition of high burnup spent fuel discharged from light water reactors

The isotopic composition and amount of plutonium (Pu) in spent fuel from a high burnup boiling water reactor (HB-BWR) and a high burnup pressurized water reactor (HB-PWR), each with an average discharge burnup of 70 GWd/t, were estimated, in order to evaluate fast breeder reactor (FBR) fuel composition in the transition period from LWRs to FBRs.     

Recoil release of fission products from nuclear fuel

An analytical approximation is developed for calculating recoil release from nuclear fuel into gas filled interspaces. This expression is evaluated for a number of interspace geometries and shown to be generally accurate to within about 10% by comparison with numerical calculations. The results are applied to situations of physical interest and it is demonstrated that recoil can be important when modelling fission product release from low temperature CAGR pin failures. Furthermore, recoil can contribute significantly in experiments on low temperature fission product release, particularly where oxidation enhancement of this release is measured by exposing the fuel to CO₂. The calculations presented here are one way of allowing for this, other methods are suggested.     

Advanced fuel elements for passive pressure tube light water reactors

This paper focuses on the development of advanced fuel elements for innovative pressure tube light water reactors. Considerations and constraints that affect the design process and various possible options are discussed. The two most promising fuel designs, which can survive a loss-of-coolant accident without primary coolant replenishment, while having sufficient margins to fuel design limits, are proposed, described and evaluated. It is demonstrated that this key objective can be achieved, provided that reliable SiC cladding or coating, which can withstand operating and accident conditions without failure, can be manufactured. Recent advances in ceramic coating technologies and experimental tests of coated specimens indicate that the attainment of this goal is feasible.     

Indirect neutron radiography experiment on dummy nuclear fuel rods for pressurized water reactors at CMRR

Nuclear energy is a vital source of clean energy that will continue to play an essential role in global energy production for future generations. Nuclear fuel rods are core components of nuclear power plants, and their safe utilization is paramount. Due to its inherent high radioactivity, indirect neutron radiography(INR) is currently the only viable technology for irradiated nuclear fuel rods in the field of energy production. This study explores the experimental technique of indirect neutron computed tomography(INCT) for radioactive samples. This project includes the development of indium and dysprosium conversion screens of different thicknesses and conducts resolution tests to assess their performance. Moreover, pressurized water reactor(PWR) dummy nuclear fuel rods have been fabricated by self-developing substitute materials for cores and outsourcing of mechanical processing. Experimental research on the INR is performed using the developed dummy nuclear fuel rods. The sparse reconstruction technique is used to reconstruct the INR results of 120 pairs of dummy nuclear fuel rods at different angles, achieving a resolution of 0.8 mm for defect detection using INCT.     

Comparisons of escape mechanisms of gaseous fission products from polycrystalline fuel

Translated from Atomnaya Énergiya, Vol. 69, No. 4, pp. 250–251, October, 1990.     

Dynamics of the escape of gaseous fission products from melted nuclear fuel

Institute of Fast Nuclear Power Reactors, Russian Academy of Sciences. Translated from Atomnaya énergiya, Vol. 79, No. 2, pp. 138–145, August, 1995.     

Modeling the diffusion yield of radioactive fission products from uranium dioxide fuel

The numerical aspects of a systematic solution of the problem of diffusion and yield of radioactive fission products from a homogeneous sphere, simulating a uranium dioxide fuel kernel, with realistic boundary conditions are discussed. The numerical scheme is based on a one-group method of calculating the production and radioactive mutual transmutations of fission products in combination with the standard expansion of the radial dependence of the concentration in terms of the eigenfunctions of the Laplace operator. It is demonstrated on illustrative examples that the approach is highly economical.     

Fuel oxidation and thermal conductivity model for operating defective fuel rods

A model has been developed to describe the fuel oxidation behaviour, and its influence on the fuel thermal conductivity, in operating defective nuclear fuel rods. The fuel-oxidation model is derived from adsorption theory and considers the influence of the high-pressure environment that results from coolant entry into the fuel-to-clad gap. This model is in agreement with the fuel-oxidation kinetics observed in high-temperature annealing experiments conducted at 1473-1623 K in steam over a range of pressure from 0.001 to 0.1 MPa. Using a Freundlich adsorption isotherm, the current model is also consistent with recent experiments conducted at a higher pressure of 7 MPa. The model also considers radiolytic effects as a consequence of fission fragment bombardment in the fuel-to-clad gap. This treatment suggests that radiolysis-assisted oxidation is insignificant in operating defective rods (as compared to thermal effects), as supported by limited in-reactor data. The effects of diffusion of the interstitial oxygen ions in the solid in the operating rod is further discussed.     

Fundamental aspects of defective nuclear fuel behaviour and fission product release

An in-reactor research program with individual, purposely defected, nuclear fuel elements has provided a fundamental understanding of the physical processes of fission product release from defective fuel. On the basis of these experiments, an analytical model has been developed to describe the release of radioactive iodine and noble gas from defective fuel into the primary coolant. An analytic treatment has also been used to model the low-temperature release of fission products from small particles of uranium-bearing compounds (uranium contamination) deposited on in-core surfaces. As a result of this study, a methodology is established whereby release from surface uranium contamination can be distinguished from that resulting from fuel pin failure. Application of this work to power reactor operation is discussed.     

Effect of fission products yield library on the calculation of the nuclide composition of the fuel in sodium-cooled fast reactors

The calculation of the composition of irradiated fuel for different degrees of burnup is a basic problem in the analysis of nuclear-radiological safety of objects holding spent fuel assemblies. The yield of fission products is one of the important initial indicators in burnup calculations. Methods for compiling libraries of fission products yield on the basis of the ENDF/B up-to-date evaluated nuclear data files are described. The nuclide composition of uranium oxide and uranium-plutonium-zirconium metal fuel in sodium-cooled fast reactors is analyzed by means of high-precision calculations performed with different fission product yields libraries using different computer codes MONTEBURNS–MCNP5–ORIGEN2 and the results are presented.     

Conversion of high enriched uranium in thorium-232-based oxide fuel for light and heavy water reactors: MOX-T fuel

This paper discusses the possibility of using military high enriched uranium and plutonium in thorium oxide fuel for light and heavy water reactors (LWRs and HWRs). It is shown that such a fuel has several important advantages: (i) ²³⁹Pu and other long-living actinides are generated in quantities which are at least 100 times less than in conventional fuel; (ii) neutron emission is lower by a factor of more than 100; (iii) ²³³U is generated and burnt (the conversion factor for LWRs is 0.64–0.68 and for HWRs about 0.88); (iv) thorium is utilized and the total available amount of nuclear fuel is increased. The problem of non-proliferation of fissile material is also discussed and it is shown that the supervision of such fuel does not differ essentially from the supervision of low enriched uranium fuel with plutonium generation.