LWR Fuel Research Articles

Corrosion Tests of Lwr Fuels – Nuclide Release

ABSTRACTTwo high burnup BWR fuels [64 and 71 (MWd)/kgU], one of which contained 2% Gd, and two low burnup PWR fuels [30 and 45 (MWd)/kgU], are tested by dripping groundwater on the fuel under oxidizing and hydrologically unsaturated conditions for 2.4 and 8.2 yr, respectively, at 90°C. The 99Tc, 129I, 137Cs, 97Mo, and 90Sr releases are presented to show the effects of reaction time and of gadolinium (Gd) on nuclide release. For the PWR fuels, the five nuclides have similar fractional release rates at 8.2 yr. For the BWR fuels, the presence of 2% Gd reduced the 99Tc cumulative release fraction after 2.4 yr by about an order of magnitude from that of a fuel with a similar burnup but without added Gd.

A comparison of different neutron spectroscopy systems at the reactor facility VENUS

The VENUS facility is a zero-power research reactor mainly devoted to studies on LWR fuels. Localised high-neutron rates were found around the reactor, with a neutron/gamma dose equivalent rate ratio as high as three. Therefore, a study of the neutron dosimetry around the reactor was started some years ago. During this study, several methods of neutron spectroscopy were employed and a study of individual and ambient dosemeters was performed. A first spectrometric measurement was done with the IPSN multisphere spectrometer in three positions around the reactor. Secondly, the ROSPEC spectrometer from the Fraunhofer Institut was used. The spectra were also measured with the bubble interactive neutron spectrometer. These measurements were compared with a numerical simulation of the neutron field made with the code TRIPOLI-3. Dosimetric measurements were made with three types of personal neutron dosemeters: an albedo type, a track etch detector and a bubble detector.

Analysis of high burnup fuel behavior in Halden reactor by FEMAXI–V code

The author developed a code FEMAXI–V to analyze the behaviors of high burnup LWR fuels. FEMAXI–V succeeded the basic structure of code FEMAXI–IV, and incorporated such new models and functions as fuel thermal conductivity degradation with burnup, alliance with burnup analysis code which gives radial power profile and fast neutron flux, etc. In the present analysis, coolant conditions, detailed power histories and specifications of the fuel rods DH and DK of IFA-519.9 irradiated in Halden reactor were input, and calculated rod internal pressures were compared with experimental data for the range of ∼25–93 MWd kg −1 UO 2, and factors affecting pellet temperature were discussed. Also some sensitivity studies were conducted with respect to the effect of swelling rate and grain growth. As a result, it is found that the prediction is sensitive to the models of thermal conductivity and swelling rate of fuel, and FEMAXI–V analytical system proved to give a reasonable prediction even in the high burnup region.

High burn-up rim structure: evidences that xenon-depletion, pore formation and grain subdivision start at different local burn-ups

An experimental LWR fuel with high 235U enrichment (8.6%) and large UO 2 grain size (15–20 μm) was analysed by optical microscopy, SEM and EPMA. Although the high burn-up reached (69.8 GWd/tM), the porosity growth in the rim zone was found equivalent to a standard LWR fuel with barely 40 GWd/tM burn-up. Also, the grain subdivision associated with the rim structure was absent or scarcely visible around the pores of the outermost periphery, being evidenced that the three typical features of this structure, i.e. Xe depletion, pore formation and grain subdivision, did not appear simultaneously, but sequentially in this order as the local burn-up increased. According to the results, the use of higher 235U enrichments in LWR fuels may help to shift all these steps to higher local burn-ups, whereas the grain size of the UO 2 matrix may play only a secondary role in the initiation of these processes.

Fission Gas Induced Cladding Deformation of LWR Fuel Rods under Reactivity Initiated Accident Conditions

Behavior of irradiated fuel rods under power burst conditions by accidental reactivity insertion in light water reactors (LWRs) has been studied in the Nuclear Safety Research Reactor (NSRR). In the experiments, cladding hoop deformation, which reached up to about 10%, was much larger than that of the fresh rods. The current LWR fuel behavior analysis codes, which only take account of the thermal expansion of the fuel pellets for the deformation calculation, under-predicted the plastic deformation of the cladding to be less than about 1%. Fission gas release during the pulse irradiation tests reached as high as 22% in the NSRR irradiated fuel tests. In order to describe these test results, a model of grain boundary fission gases to cause the cladding deformation has been developed and installed in a fuel behavior simulation code, FRAP-T6. In the model, the over-pressurized gases by the pulse irradiation cause grain boundary separation and stress the cladding during the tests. The model assumes that the gases remain in the fuel during the early part of pulse irradiation and are released to the open volume in the rod after the cladding deformation. The model, in combination with a fuel thermal expansion model, GAPCON, which was validated through fresh fuel tests, reproduces the NSRR test results reasonably well.

Study of fission-product segregation in used CANDU fuel by X-ray photoelectron spectroscopy (XPS) II

A thorough investigation of the grain-boundary chemistry of used CANDU fuel from one intact element has been conducted by X-ray photoelectron spectroscopy (XPS). Selected findings from more extensive XPS measurements on other used CANDU fuels exposed to storage conditions are included for comparison. Cesium, rubidium, tellurium and barium have been commonly observed, often reaching high degrees of surface enrichment, although their relative abundances can vary widely with a complex dependence on the fuel irradiation history. Lower concentrations of cadmium, molybdenum, strontium and iodine have also been occasionally detected. Except for iodine, chemical-shift data are indicative of oxidized species, possibly uranates. Segregation at monolayer-level coverages has been demonstrated by sequential XPS analysis and argon-ion sputtering. Calculations based on an idealized thin-film model are consistent with the depth profiles. The interpretation of these results is discussed in the context of previous studies, especially on LWR fuels.

Development of LWR Fuel Performance Analysis Codes

Abstract This paper is to cover brief history of the development of LWR fuel performance analysis codes in Japan. This includes the FEMAXI-III code which calculates local stress and strain of the cladding at power ramp due to the pellet cladding mechanical interaction (PCMI) using axisymmetric finite element method. From this code some improved codes were developed, such as FEMAXI IV, IRON and EIMUS. The FEMAXI-IV is developed by J AERI to add capacity of transient analysis and che IRON code was developed by Shikoku Electric Power Co. to analyze fuel reliability at load following operation and the EIMUS code was developed by CRIEPI for high burnup analysis. The major objective and contents of adapted submodels in these codes are presented with some calculational results. The future development and utilization of the codes, especially for performance analyses at high burnup, will be briefly discussed.

Gap and Grain-Boundary Inventories of Cs, Tc, and Sr in Spent LWR Fuel

ABSTRACTSoluble radionuclides concentrated within the gap and grain-boundary regions of spent fuel could dissolve relatively rapidly were the waste container to fail and the fuel to be contacted by water in a geologic repository. To provide an estimate of the quantities of radionuclides that may be rapidly released, fractional inventories of Cs, Tc, and Sr concentrated within the fuel/cladding gap region have been measured for U.S. LWR spent fuels with fission gas release (FGR) values ranging from 0.25% to 18%. Separate measurements of the grain-boundary inventories of Cs, Tc, and Sr have been made for the same fuels. The Cs gap inventories were generally found to be about one fourth of the FGR values. The Cs grain-boundary inventories were generally less than 1% of the total Cs inventories and were not strongly correlated with FGR. Both the gap and grain-boundary inventories of Tc and Sr were near the detection limits of the methods used, less than 0.2% of the total inventories of these elements. However, some of the Tc may reside at the grain boundaries in the form of relatively insoluble metallic particles and not be detected by these experiments. Data obtained by comparing the dissolution behavior of fuel fragments with that of fuel grains were used to estimate the dissolution rate of Cs from the grain boundaries of one of the fuels. Surface-area normalized dissolution rates determined for fuel fragments in these same tests exceeded those determined for grains. A likely explanation is that the estimated fragment surface area did not take into account the “effective” grain-boundary surfaces.

使用済軽水炉燃料溶解時の揮発性核種の分析方法と挙動

The amounts of induced volatile nuclides (I, Kr, Xe and gaseous tritium) and the isotopic compositions of Kr and Xe were examined during nitric acid dissolution of 10 pieces of spent LWR fuels of various burnups. Each component was analyzed radiometrically, except of Xe, which was determined through gas chromatography with Kr. The released amounts of I, Kr, Xe and gaseous tritium increased with the burnup of fuel. The large amount of I volatilized during dissolution of fuel and the small remained in the liquid. The ratios of 85Kr to all of krypton nuclides were about 5% over the burnups. The amount of Xe showed to be around 0.4w/o of the initial amount of U and gaseous tritium around 0.5% in all of induced tritium.

Migration of fission product barium in UO 2 fuel under transient conditions

Gamma scanning and electron probe microanalysis results are reported for LWR fuels transient-tested in the first Risø Fission Gas Project. The fuels, which had been prepared from powders supplied by Risø National Laboratory (R fuel) and the UKAEA (U fuel), had been transient-tested to powers in the range 30.1 to 44.4 kW m −1 at a burn-up between 3.6 and 4.5 at%. It was found that considerable radial redistribution of Ba had occurred during the transient tests. Evaporation-condensation of BaO had led to the formation of BaSiO 3in pores and cracks at intermediate radial positions. As a result of thermal diffusion up the temperature gradient, Ba had also accumulated in the centre of the fuel. Although radial redistribution of barium occurred in both the R and U fuels, only the R fuel released Ba to the rod free volume. In rods that had been transient-tested to powers greater than 41 kW m −1, accumulations of 140Ba were found at the pellet interfaces. The release of Ba is ascribed to the evaporation of BaO in the central region of the fuel at temperatures above 2000° C. Indications were found that the U fuel operated at a lower temperature than the R fuel. Therefore, it is concluded that the U fuel never reached 2000° C, even at the highest terminal power of 44.4 kW m −1.

Latest safety related LWR fuel behaviour studies in Japan, overviewed.

Safety-related LWR fuel studies in Japan during the past several years are overviewed. Fuel Specialists Group (NEN-ANSEN), JAERI and NUPEC have been the main members in this field. Participation in International Projects at Halden, Studsvik, Battelle etc. has been useful and fruitful in obtaining the wide range of data for LWR fuel behaviour in normal and transient operating conditions. So-called “reliability tests” were carried out at NUPEC supported by the Government to verify the integrity of then-current design LWR fuels. Demonstration of the higher-duty fuel program is also overviewed. Burn-up Extension Program has been going on also as an activities of NUPEC. For fuel behaviour under accident conditions, JAERI made important contributions through the experiments in NSRR. Fuel behaviour during severe-core damage has been considered based on the data obtained from the International Corporation Programs.

UO2. Matrix Dissolution Rates and Grain Boundary Inventories of Cs, Sr, and Tc in Spent Lwr Fuel

ABSTRACTExperimental methods have been developed for measuring the grain-boundary inventories of radionuclides and for determining whether the UO2 matrix of spent light-water reactor fuel dissolves congruently. Both methods depend upon first separating the fuel into individual grains. With the grain boundaries thus exposed, the associated inventories of radionuclides can be completely dissolved and measured. To determine whether the UO2 matrix of spent fuel dissolves congruently, the fuel grains were placed in a flow-through column and water was pumped through the column at a rate sufficient to maintain the concentration of U in the column effluent far below saturation. Once the grain-boundary material has dissolved, the forward dissolution rate of the UO2 matrix can be measured and, by measuring the concentrations of other radionuclides in the column effluent, the degree of congruency of the dissolution process can be determined. Data obtained to date indicate that the grain-boundary inventories of Cs, Tc, and Sr are approximately equal to gap inventories and that the fractional dissolution rate of Cs from the U02 matrix is approximately equal to that of U, i.e., the Cs and U dissolved nearly congruently. In addition, the data appear to show a gradient in the concentrations of Cs and Sr across the individual UO2 grains.

Development of Impact Protection for Irradiated LWR Fuel Transport Flasks

AbstractThe development and refining of impact limited systems is described for BNFL's ‘Excellox’ irradiated fuel flasks. The process began with the adaptation of the aluminium cage type lid shock absorber developed for the original Excellox 1 flask to the longer and heavier Excellox 2 and 3 designs and finally to the 100 ton Excellox 4 flask. The importance of protecting the components and contents of the flask against high g base impacts is now realised. The development of compact balsa-filled lid and base shock absorbers which would cushion the flask and its contents against the IAEA 9 m impact test is described, and their refinement to protect flask orifices and seals against penetrative impact and fire. This development process entailed extensive testing of quarter-scale models at the CEGB Structural Test Centre and the tests, described in this paper, have provided a wealth of practical data which is already proving very useful in designing further flasks.

Latest Safety Related LWR Fuel Behaviour Studies in Japan, Overviewed

Latest Safety Related LWR Fuel Behaviour Studies in Japan, Overviewed

Achievement of world's best performance in Japanese LWR, through the successful development on fuels and materials

Research and development on fuel and cladding materials in the past 40 years, which led to the excellent performance of today's Japanese LWR are overviewed. Most of the programs for the development of LWR fuels and Zircaloy cladding tube have been conducted by the author, starting from the metal uranium fuel rod for the research reactor JRR-3, then going on to the domestic production of fuel for the power-demonstration reactor JPDR. LWR fuel fabrication has been successfully performed on these bases together with the research and development work on Zircaloy tube fabrication technology among metal industries. Fundamental works and co-operative studies among specialists from every related field have been effectively combined with the development works among industry, which contributed to the achievement of the world's top level fuel and thus to the excellent performance of 36 LWRs in Japan, today.

LWR Fuel Safety Research with Particular Emphasis on RIA/LOCA and Other Conditions

Fuel safety research at Japan Atomic Energy Research Institute (JAERI) is reviewed on the major subjects including studies on fuel behavior under postulated Reactivity Initiated Accident (RIA), postulated Loss of Coolant Accident (LOCA) and normal operating conditions. Nuclear Safety Research Reactor (NSRR) at JAERI has been utilized extensively for the studies of fuel behavior under RIA conditions. For the studies of fuel rod and cladding behavior under LOCA conditions, outpile experiments were conducted. The work on this subject has been concluded. Pellet Cladding Interaction (PCI) has been major subject on fuel integrity study during normal operating conditions. Irradiation experiments at Halden Boiling Water Reactor (HBWR) as well as code development are described.

Fuel Fragmentation and Mechanical Energy Conversion Ratio at Rapid Deposition of High Energy in LWR Fuels

Fuel Fragmentation and Mechanical Energy Conversion Ratio at Rapid Deposition of High Energy in LWR Fuels

Fuel fragmentation and mechanical energy conversion ratio at rapid deposition of high energy in LWR fuels.

The fuel fragmentation is one of the important subjects in the field of molten fuel-coolant interaction (MFCI) since it is one of basic processes of the MFCI, and it has not yet been made clear enough. Accordingly, U02 fuel fragmentation was studied in a postulated reactivity initiated accident (RIA) condition by the Nuclear Safety Research Reactor (NSRR). The distribution of the size of fuel fragments was obtained through the experiments and the mechanism of fuel fragmentation was studied. Also, the relation between the conversion ratio of the mechanical energy to the thermal and the degree of fuel fragmentation was obtained experimentally. It was revealed that the distribution of fuel fragments was well described in the form of logarithmic Rosin-Rammler's distribution law. The fuel fragmentation was found to be explained by the Weber-type hydraulic instability model and the internal pressurization model. It was also shown that the mechanical energy conversion ratio was inversely proportional to the volu...

Source Term for the Activity Release from a Repository for Spent LWR Fuel

AbstractIn order to provide basic data for the calculation of the dose burden from a repository for spent LWR fuel in a salt dome, a formula is given for determining activity release into the surrounding salt brine under accident conditions. The main characteristic of this source term is a time-independent, instantaneous release of part of the activity and mobilization of another part at a continuous and temperature dependant rate, using nuclide-specific leach rates. Changes in leach rates resulting from reduction of the waste surface during mobilization and precipitation are taken into account. The model also takes into account axial and radial activity distribution in the fuel rods.

The determination of 232U and 236Pu in solutions of irradiated reactor fuel by alpha-spectrometry

Anion-exchange from HCl and HBr media was applied to the separation of uranium and plutonium from solutions of irradiated LWR fuels. The minor isotopes 232U and 236Pu were then determined by alpha spectrometry relative to 238U and 239+240Pu respectively. Successful analyses were carried out with the requisite high decontamination factors. Samples of standard PWR uranium and MOx fuel and BWR uranium fuel were analysed. The results were compared with values calculated using the program KORIGEN. Only in some cases was good agreement found between the experimental and calculated values. The possible causes are discussed and it is concluded that a large variable factor arises from decay processes occurring in the fuel before irradiation.