Einfluß schneller Neutronen auf das Transportverhalten fester Spaltprodukte in pyrokohlenstoffbeschichteten Kernbrennstoffteilchen

The transport behaviour of metallic fission products in pyrocarbon-coated fuel particles is investigated by means of postirradiation annealing at 1400°C as a function of the fast neutron fluence. The release of barium and strontium are both increased by fast neutrons, and that even, if the kernels contain alumina additives, which improve the fission product retention by forming aluminates (e.g. BaAl $_{2}$O$_{4}$). The release of these two fission products is controlled by the effective diffusion coefficient in the kernel, which is in the case of Sr 90 in the range of 5,7·10$^{-15}$cm$^{2}$s$^{-1}$ (nonirradiated) and 5,3·10$^{-13}$ cm$^{2}$s$^{-1}$ (fluence 2· 10$^{22}$ cm$^{-2}$ at E> 0,1 MeV). The most significant increase is observed in the low fluence region. Ruthenium and ceriüm are released from the kernels by recoil only, not by diffusion. The caesium release is controlled by diffusion in the outer coating layer and depends on the microstructure of the pyrocarbon: The diffusion coefficient is decreasedby fast neutrons, if the amount of fiber component in the pyrocarbon is high. At low amounts of fiber component the diffusion coefficient is increased by fast neutrons

Bestrahlungsverhalten von beschichteten Brennstoffteilchen mit spaltproduktbindenden Kernadditiven

The four irradiation experiments FRJ2-P17, FRJ2-P18, FRJ2-P19, and FRJ2-P20 for testing the efficiency of fission product-retaining kernel additives in coated fuel particles are described. The evaluation of the obtained experimental data led to the following results:- zirconia and alumina kernel additives are not suitable for an effective fission product retention in oxide fuel kernels,- alumina-silica kernel additives reduce the in-pile release of Sr 90 and Ba 140 from BISO-coated particles at temperatures of about 1200°C by two orders of magnitude, and the Cs release from kernels by one order of magnitude,- effective transport coefficients including all parameters which contribute to kernel release are given for (Th,U)O$_{2}$ mixed oxide kernels and low enriched UO$_{2}$ kernels containing 5 wt. % alumina-silica additives: log D$_{K}$/cm$^{2}$s$^{-1}$ = - $\frac{36028}{T}$ + 6,261 (Sr 90), log D$_{K}$/cm$^{2}$s$^{-1}$ = - $\frac{29646}{T}$ + 5,826(Cs134/137), -alumina-silica kernel additives are ineffective for retaining Ag 110 m in coated particles. However, also an intact SiC-interlayer was found not to be effective at temperatures above 1200 C.- the penetration of the buffer layer by fission product containing eutectic additive melt during irradiation can be avoided by using additives which consist of alumina and mullite without an excess of silica,- annealing of LASER-failed irradiated particles and the irradiation test FRJ2-P20 indicate that the efficiency of alumina-silica kernel additives is not altered if the coating becomes defect

Bestrahlungsverhalten von beschichteten Brennstoffteilchen für das Abbrand/Brut Partikelsystem

The cross evaluation of 30 irradiation experiments which were carried out in the last ten years in order to test fuel particles for the separate use of high enriched uranium in fissile particles and thorium in fertile particles, led to the following results: - An oxide-based fissile/fertile particle system (UO$_{2}$ fissile kernel/ThO$_{2}$ fertile kernel) can be used as well as the American carbide/oxide particle system (UC$_{2}$ fissile kernel/ThO$_{2}$ fertile kernel) under the operation conditions of a high temperature reactor with spherical fuel elements. - The swelling of fissile kernels as a consequence of fission gas pores is much more pronounced in UO$_{2}$ than in UC$_{2}$ fissile kernels but the buffer layer topes with the swelling without any problems. - Ceramic kernel additives (e. g. Al$_{2}$O$_{3}$) as well as carbon additives proved not to be suitable because they deteriorate the mechanical properties of the fissile kernels. - Kernel migration in a temperature gradient ("amoeba effect") is observed during irradiation of UO$_{2}$ fissile particles but this does not cause any coating failure. The amoeba effect is suppressed completely by 10 % UC$_{2}$ additives to the UO$_{2}$ kernel. - The silicon carbide interlayer is absolutely necessary for an efficient retention of the solid fission products and has also proved sucessfully for fertile particles. Inner corrosion of the SIC layer by traces of chlorine can be avoided by a suitable coating process. The SIC corrosion caused by metallic fission products (especially Pd) starts at high temperatures which, however, decrease wich increasinq fast neutron fluentes. - A measurable fraction of-defective particle coatings was not observed before exceeding the target values of burnup and fast neutron fluence. The failure fraction is described quantitatively and correctly by a model which takes account of the increasing internal pressure due to burnup and the decreasing ultimate tensile strength and the SiC layer due to the influence of fast neutrons. - The irradiation-induced dimensional changes of the graphite matrix are Independent of the fuel volume loading and have no influence on the irradiation behaviour of the embedded fissile and fertile particles

Die chemischen Grundlagen des Hydrolyseverfahrens zur Herstellung spärischer Kernbrennstoffteilchen

The preparation of UO$_{2}$ microspheres by the Hydrolysis Process is based on the solidification of droplets of a concentrated uranyl " nitrate solution in hot oil. The uranyl nitrate is complexed by urea and partly hydrolized by adding solid hexamethylentetramine. The temperature depending solidification reaction of the metastable solution is a continuous polymerisation of [(U0$_{2}$ (OH)$_{2}$)$_{n}$ UO$_{2}$]$^{2+}$ ions forming kolloidal UO$_{2}$(OH)$_{2}$. The crystallite size of the spherical gel particles is increased by washing with hot water. Heat treatment of the washed and dried particles in reducing atmosphere leds to UO$_{2}$ with excellent sintering behaviour. The process was also successfully applied to the preparation of ThO$_{2}$ and (U,Th)O$_{2}$ microspheres

Untersuchungen zur Getterung metallischer Spaltprodukte im Primärkühlkreislauf eines Hochtemperaturreaktors

Supplementary to the recommended method to retain fission products by ceramic coatings (e.g. pyrocarbon, silicon carbide) in small spherical fuel particles, a new concept of gettering silver 110m and cesium 134/137 in the primary circuit of a High Temperature Reactor (HTR) is presented. It is based upon the known fact that the vapor pressure of metallic fission products in solid or liquid solutions is lower compared with that of the pure fission products. Although metallic additives were found not to influence the silver release from oxide fuel kernels, the effective diffusion coefficient of silver 110m in graphite matrix is reduced by about two orders of magnitude by small amounts of the metallic Cu, Ge, Sn or Au additions. However, these reduced silver diffusion coefficients are not sufficiently low in order to retain silver 110m in the fuel-free zone of spherical HTR fuel elements. On the other hand, metallic additives were found to be very efficient in gettering silver 110m from the gaseous phase: During a contact time of only 0,15 seconds at 950°C more than 80 %,at 850°C even more than 99 % of the Ag 110m could be absorbed from the streaming gas by using a metal-containing graphite filter. The best results were obtained by using tin or gold additives. By optimizing the filter geometry further increase of the efficiency should be possible

Transportverhalten von Plutonium und Americium in niedrig angereicherten beschichteten Brennstoffteilchen bei hohenBestrahlungstemperaturen

Low enriched coated fuel particles with different kernel composition (oxide, carbide, oxicarbide) were irradiated at high temperatures in the range of 1100-1600°C reaching burnup values of 11-12 % FIMA. By mechanical separation of kernel and coating of single irradiated particles followed by chemical separation and alphaspectrometric determination of plutonium and americium, the internal release of both transuranium elements was measured. In coated particles with U0$_{2}$ kernels the amount of Pu and Am in the coating was the saure as the amount of uranium before irradiation : any internal release of Pu and Am from UO$_{2}$ kernels could not be observed. From UC$_{2}$ and UCO kernels both transuranium elements were released, the fractional release of americium was always higher than that of plutonium. Effective diffusion coefficients of the transuranium elements describing their release behaviour from UC$_{2}$ kernels were found to be in the range of (2,0$\pm$0,8)-10$^{-14}$ cm$^{2}$s$^{-1}$ (plutonium) and (2,4$\pm$0,8)-10$^{-13}$ cm$^{2}$ (americium) at an average irradiation temperature of about 1350°C. Using UO$_{2}$ kernels these coefficients are surely below 2,5 $\cdot$ 10$^{-18}$ cm$^{2}$s$^{-1}$. Plutonium and americium diffused through pyrocarbon coatings nearly equally with average diffusion coefficients of 1,6$\cdot$10$^{-11}$ cm$^{2}$s$^{-1}$ (plutonium) and 2,3$\cdot$10$^{-11}$ cm$^{2}$s$^{-1}$ (americium). These data agree within their standard deviation with earlier published diffusion coefficients of plutonium in pyrocarbon coatings

Untersuchungen zur Herstellung kugelförmiger Brennstoffteilchen nach einem Sol-Gel-Verfahren

Als Kernbrennstoff für gasgekühlte Hochtemperaturreaktoren werden Carbide oder Oxide des Spaltstoffs (U$^{235}$ , U$^{233}$ , Pu$^{239}$ ) und des Brutstoffs (Th$^{232}$, U$^{238}$) in Form von beschichteten Teilchen ("coated particles") eingesetzt. Diese Teilchen bestehen aus einem kugelförmigen Brennstoffkern mit einem Durchmesser von etwa 0,5 mm und einer Beschichtung aus mehreren Schichten Pyrokohlenstoff und häufig einer Siliciumcarbid-Schicht . Die Beschichtung ist nahezu gasdicht und soll die Freisetzung der im Reaktor gebildeten Spaltprodukte verhindern . In der Praxis liegt die relative Freisetzungrate für Spaltgase in der Größenordnung 1o$^{ -5}$ , für einige feste Spaltprodukte wie Cs, Sr, Ba jedoch auch höher. Zur Verbesserung der Spaltproduktrückhaltung im Brennstoff, insbesondere im Hinblick auf die Entwicklung von Hochtemperaturreaktoren mit einer Heliumturbine im Primärkreislauf, sind umfangreiche Untersuchungenund Bestrahlungstests an Teilchen mit variierten Schicht- und Kernparametern erforderlich . Ferner besteht Interesse an der Entwicklung von Hochtemperaturbrennstoffen mit erhöhter Schwermetalldichte, was durch Einsatz von beschichteten Teilchen mit größeren Kernen (Durchmesser bis 1 mm) erreicht werden kann . Für diese Untersuchungen sollte daher auf der Grundlage bereits bekannter Verfahren ein möglichst vielseitiges Verfahren zur Herstellung dererforderlichen Brennstoffkerne erarbeitet werden

Untersuchungen zur Verringerung der Freisetzung fester Spaltprodukte aus beschichteten Brennstoffteilchen durch Zusätze hochschmelzender Oxide zum Brennstoffkern

For increasing the retention of solid fission products in coated fuel particles, investigations are made on oxide fuel kernels containing special refractory oxides, e.g. Al$_{2}$O$_{3}$. It is the aim of this work to test the possibility of retaining solid fission products in the kernel by formation of stable ternary oxides. Using the Hydrolysis Process, U0$_{2}$ and (U,Th) O$_{2}$ kernels are prepared which contain the solid fission products Sr, Ba, Ce in concentrations corresponding to a burnup of about 10 % fima. After an out-of-pile heat treatment of these kernels of the fission product release is measured. The Sr release at 1250 °C is found tobe smaller by a factor 20 to 100 if the kernels contain Al$_{2}$O$_{3}$ in amounts of 1 to 5 weight percent. The solid fission products are enriched in the inclusions of Al$_{2}$O$_{3}$ which is not soluble in the UO$_{2}$ matrix of the kernel