FORMAL REPORT
GERHTR-159
UNITED STATES-GERMAN HIGH TEMPERATURE REACTOR RESEARCH EXCHANGE PROGRAM
Original report number ______Title Improvement in Retention of Solid Fission Products in HTGR Fuel Particles by Ceramic Kernel Additives
Authorial R. Forthmann, E. Groos and H. Grobmeier Originating Installation Kemforschtmgsanlage Juelich, West Germany. Date of original report issuance August 1975______Reporting period covered ______
In the original English
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page
1. INTRODUCTION 2
2. FUNDAMENTAL STUDIES 3
3. IRRADIATION EXPERIMENT FRJ2-P17 5
3.1 Results of the Fission Product 8 Release Measurements Improvement in Retention of Solid 3.2 Electron Microprobe Investigations Fission Products in HTGR Fuel Particles 8 by Ceramic Kernel Additives. 4. IRRADIATION EXPERIMENT FRJ2-P18 16
4.1 Release of Solid Fission Products 19
4.2 Electron Microprobe Studies 24 by
R. Forthmann, E. Groos, H. GrObmeier 5. SUMMARY AND CONCLUSIONS 27
6. ACKNOWLEDGEMENT 28
7. REFERENCES 29 2 -
X. INTRODUCTION Kernforschungs- anlage JUTich JOL - 1226 August 1975 Considerations of the core design of advanced High-Temperature Gas-cooled GmbH IRW Reactors (HTGRs) led to increased demands concerning solid fission product retention in the fuel elements. This would be desirable not only for HTGR power plants with a helium-turbine in the primary circuit (HHT project), but also for the application of HTGRs as a source of nuclear process heat.
Improved solid fission product retention of pyrocarbon-coated fuel particles can be obtained by two different methods:
Improvement in R etention of Solid (a) Deposition of carbide interlayers (e.g. silicon carbide or zirconium carbide) as an additional diffusion barrier in the pyrocarbon coating, Fission Products in HTGR Fuel Particles by Ceramic K ernel Additives (b) Improvement of the kernel retention by chemical reaction of the fission products with suitable kernel additives forming stable fission product by compounds. R. Forthmann E. Groos H. Griibmeier
ABSTRACT
Increased requirements concerning the retention of long-lived solid fission products in fuel elements for use in advanced High Tempera Silicon carbide interlayer Ceramic kernel additives ture Gas-cooled Reactors led to the development of coated particles with improved fission product retention of the kernel, which re present an alternative to silicon carbide-coated fuel particles. Two irradiation experiments have shown that the release of strontium, Fig. 1: METHODS OF IMPROVED FISSION PRODUCT RETENTION IN COATED FUEL barium, and caesium from pyrocarbon-coated particles can be reduced by orders of magnitude if the oxide kernel contains alumina-silica PARTICLES AT HIGH TEMPERATURES additives. It was detected by electron microprobe analysis that the improved retention of the mentioned fission products in the fuel kernel is caused by formation of the stable aluminosilicates The principle of the two methods is shown in figure 1. At the left side of SrAlgSijOg, BaAlgSi^Og and CsAlSi^Og in the additional alumina- the figure the kernel releases solid fission products (marked by arrows) silica phase of the kernel. which easily penetrate the porous and dense inner pyrocarbon layer (PyC = pyrocarbon), until they are retained by the silicon carbide layer (SiC = silicon carbide). This method is well known and has been investigated at
1 many laboratories. At the right side of the figure the coated particle contains
ceramic kernel additives (e.g. AljOg + SiC>2), the solid fission products are does not form solid solutions with U02 and is found in the ceramic inclusions retained in this second ceramic phase of the kernel by solid state chemical together with zirconium and uranium forming BaZrOg and probably BaUG^. Most reactions. In this case an additional silicon carbide layer is not necessary. of the caesium is present as metal vapor, only small amounts form the rather volatile compounds CsJ and CSgMoO^. Since Sr and Ba form zirconates in kernels without additives, an excess of Zr02 in the fuel kernel cannot be expected 2. FUNDAMENTAL STUDIES to be very effective for retaining these fission products. The BaO vapor pressure of BaAlo0. was found to be about one order of magnitude lower than Out-of-pile investigations on coated particles containing artificial solid that of BaZrO^ . Therefore an improvement of the Ba retention and possibly fission products, which had been introduced during the kernel fabrication of the Sr retention by A1203 additives can be expected. The best out-of-pile process, gave preliminary information on the fission product retention results, however, were obtained by using a combination of A1203 and Si02 efficiency of ceramic kernel additives. Table I shows some of these kernel additives. These additives are insoluble in U02 as well as in Th02 additives and their possible fission product compounds. and appear as a second phase in the fuel kernel. In this second ceramic phase not only Sr and Ba can be retained by formation of very stable
THERMAL NEUTRON aluminosilicates of the feldspar type, but also Os by forming stable alumino CROSS SECTION MELTING POINT ADDITIVE FISSION PRODUCT COMPOUND silicates as CsAlSiO., or CsAlSio0c. OF THE METAL ( °C ) 4 2b (barns) An important question is the compatibility of these additives with the pyro carbon coating of the fuel particles. During the ijrradiation oxide fuel kernels Zr02 0.188 SrZrOg 2800 produce an increasing carbon monoxide pressure with increasing heavy BaZrOg metal burn-up which may reach values up to 100 bars in the coated particle. SrNb_0c If the equilibrium carbon monoxide pressures of the additives Zr02, A^Q^, 2 o Nb.Os 1.15 BaNb_0c 1455 Si02 in contact with carbon are plotted as a function of temperature CsNb2,6°7 1416 (Fig. 2), it can be seen that these CO pressures are below the observed CO pressures in irradiated coated particles. The same is valid for the fission A12°3 0.235 SrAl204 1790 5) product compounds as far as thermodynamic data are available. It can be BaAl204 1815 concluded that the additives and their fission product compounds will be thermodynamically stable in coated particles at HTGR operating conditions. A12°3 0.235 SrA12Si2°8 - 1 These thermodynamic considerations were confirmed out-of-pile,using coated + sio2 0.16 BaAl2Si208 S. >1700 particles with kernel additives and artificial solid fission products by CsAlSi-0c 2 b heat treatmant in the temperature range of 1000 - 1800°C. 2 CsAlSi04
Table I: KERNEL ADDITIVES AND FISSION PRODUCT COMPOUNDS
The chemical state of solid fission products in irradiated U0o kernels is 3) ^ known from electron microprobe analysis : Strontium is oxydized to SrO and forms partially solid solutions in the U02 phase, the other part is concentrated in zirconia containing ceramic inclusions forming SrZrO^. Barium - 5 - 6
1200 K
Irradiated particles
1000 1100 1200 1300 1400 Temperature (®CI
Fig. 2: CARBON MONOXIDE PARTIAL PRESSURES
3. IRRADIATION EXPERIMENT FRJ2-P17
In the FRJ2-17 experimentspyrocarbon-coated fuel particles with fission product retaining kernel additives were irradiated for the first time. The e \ kernels were prepared by using the H-process» the starting solution of £tJ02(1 ^N03^2 g U/l) was mixed with calculated amounts of
ZrO(NOg>2, A1(N03>3 and SiO^ powder (grain size < 50 ^um) respectively. After sintering,the additives were present in the form of oxides in the UO^ kernel matrix. The kernels were coated with two pyrocarbon layers (BIS0- coating): a porous layer (density = 1.0 g cm by pyrolysis of acetylene and a dense layer (density = 2.0 g cm ) by pyrolysis of propene. The coated particles were embedded in a graphite matrix (Fig. 3) and the fuel compacts were irradiated at temperatures ranging from 1100 to 1200°C. The main data of three particle varieties of this experiment are given in table II. Fig. 3: MICROGRAPH OF A FUEL COMPACT SECTION (FRJ2-P17) 3.1 Results_of_the_Fission__Product_Release_Measurements
KERNEL 1 2 3 Fractional release values are available only for fission products remaining in material U°2 U02/Zr02 U02/Al,03/Si02 the graphite matrix of the compacts. The values of Zr, Sr and Ba are composition summarized in table III for three different kernel compositions. The fractional (wt. %) 95 release of Zr is comparable with the fraction of free uranium on the U02 100 89,7 95,1 _7 , particle surface (below 5.10 ) and is apparently caused by fission of this Zr02 - 10,3 - free uranium. As expected, zirconia additives do not retain Sr 90, but A12°3 - - 2,2 alumina-silica additives reduce the Sr 90 release by two orders of magnitude, sio2 - - 2,7 and the Ba 140 release by three orders of magnitude. This retention efficiency enrichment is comparable with the retention properties of silicon carbide coated particles. (% 235y) 15 15 15 diameter (^um) 684 682 671 The fission products Cs 134, Cs 137 and Ag 110m did not remain in the graphite matrix of the fuel compacts. About 70 - 85 % of the caesium was released from
COATING the compacts and was found on the inner graphite spine, on the outer graphite tube, and on the steel wall of the capsule. It must be assumed that this material PyC PyC PyC released caesium contaminated the fuel compacts with lower caesium release. total thickness (^um) 180 199 178 As compacts with different kernel varieties had been irradiated in one capsule, it was not possible to obtain definite fractional release data for
IRRADIATION caesium. From measurements of the activity of Ag 110m very high fractional release values resulted (between 4.4 . 10 and 9.2 . 10 ), which were time 114 114 (full power days) 114 independent of the presence of alumina or alumina-silica kernel additives. temperature Only the two particle varieties with zirconia kernel additives showed a 1200 jh 60 1180 + 60 1140 + 50 “1 (°C) lower silver release value (1.4 . 10 ), but this cannot be designated as burn-up a marked retention effect. The silver was totally released from the fuel (% FIMA) 8,3 7,1 8,2 compact matrix and mainly collected in the outer graphite tube (more than 90 %), where the surface temperature remained below 1000°C. The same distribution 7) Table II: IRRADIATION EXPERIMENT FRJ2-P17: FUEL AND IRRADIATION DATA behaviour of silver was observed by other laboratories.
After irradiation,the fuel compacts were disintegrated chemically for 3.2 Electron_Microgrobe_Inyestigations measuring the fission product distribution between coated particles, graphite matrix, and the surrounding graphite and steel components of the capsule. Ceramographic sections of coated particles with alumina-silica kernel 95 90 140 The fission products Zr, Sr and Ba remained completely in the graphite additives (Fig. 4) show a heterogeneous structure of the kernel. The matrix of the fuel compacts while and ^^mAg were released from AljOg/SiOj phase (grey) is insoluble in the 00^ kernel matrix and was the compacts and distributed to all components of the capsule. precipitated in the grain bounderies of the U02 grains (bright) . After irradiation this structure remained nearly unchanged (Fig. 5): There is the grey alumina-silica phase in the grain bounderies of the uranium oxide. 10 -
PARTICLE DATA U02(95,1) U02(89,7) Kernel material U02(100) AI203!2,2) Zr02(10,3) (wt %] Si02 (2,7) 235 u f 4,3'10'7 3.6-lO-7 4,5-lO"7 (Particle surface) Fig. 4 : MICROGRAPH OF A COATED PARTICLE WITH ALUMINA-SILICA KERNEL ADDITIVES BEFORE IRRADIATION FRACTIONAL RELEASE
95Zr 4,9-lO"7 < 10"7 1,4-10-6
9°Sr 8.0-10"3 4.9 tO-3 5,0-lO-5
140 Ba 8,6 10-3 5,3-10"4 1,4-10-6
Table III: FRACTIONAL RELEASE OF SOLID FISSION PRODUCTS (FRJ2-P17)
Fig. 5: MICROGRAPH OF A SECTION OF AN ALUMINA-SILICA CONTAINING KERNEL AFTER IRRADIATION (BURN-UP: 8,2 % FIMA) - 12 -
additionally there are some fission gas pores and bright spots of metallic fission product inclusions. By means of an electron microprobe analyser the distribution of uranium, aluminium, silicon and the fission products Sr, Zr, Mo, Tc, Ru, Rh, Pd, Cs, Ba, Ce was investigated. As shown in Pig. 6, the fission products Cs, Sr and Ba are localized at the same positions where aluminium and silicon
..-T—
Fig. 6: DISTRIBUTION OF KERNEL ADDITIVES AND FISSION PRODUCTS IN THE KERNEL MATERIAL (FRJ2-P17)
are present: This indicates a strong chemical interaction between these fission products and the additives. Zirconium and tare earth elements were not found in the alumina-silica phase but uniformly distributed in the kernel matrix. Metallic fission products such as molybdenum (Fig.6) have been precipitated Fig. 7: DISTRIBUTION OF A1 AND Si IN THE CERAMIC KERNEL ADDITIVES independently of the alumina-silica additives. In addition to molybdenum, (BEFORE IRRADIATION) these separate metallic inclusions also contained technetium, ruthenium, rhodium and in some cases palladium.
The alumina-silica kernel additives are not homogeneous with regard to the atomic ratio of silicon to aluminium. By line scanning of the additional ceramic 13 14 -
inclusions two different phases were identified (Fig. 7): one phase with a constant silicon to aluminium atomic ratio of Si : A1 = 1 : 39 and a second © » phase on the surface of the inclusions with higher silicon and lower aluminium 1
Different atomic ratios of the sum of fission product atoms to the sum of Cs Al Si206 aluminium and silicon atoms, resulted from point analyses at different positions of the alumina-silica kernel inclusions after irradiation. The atomic Compounds Sr A*2Si2°8 ^ Cs + Sr + Ba . . . ratio ---Al"T"si— was useci as an approximation of the fission product BaA^Si2 Qg concentration in the alumina-silica additives and was plotted as a function
of the Si/Al atomic ratio (Fig. 8). The maximum of this plot at 1 ^Si/Al$ 2 Fig. 9: RESULTS OF THE POINT ANALYSES OF ALUMINA-SILICA KERNEL ADDITIVES indicates that preferentially fission product compounds with theseSi/Al ratios were formed. Furthermore, there were always at least two silicon atoms per fission product atom, even at very low Si/Al atomic ratios. These experimental 3 •3 results agree very well with the assumption that the three compounds shown
in Fig. 9 are formed by chemical reaction of the fission products in the An improved Cs retention can be expected from the fact that Cs-alumino- alumina-silica additives. silicates form in the ceramic kernel additives. As it was not possible to measure the in-pile fractional release of caesium (see 3.1.), single irradiated particles were annealed in closed graphite capsules at 1200°C.
After the heat treatment the particle was removed from the capsule, followed o by the measurement of the Cs and Cs activity by means of a Ge(Li) detector. The results were reproducible if different irradiated particles of the same kernel composition were annealed. The fractional release 137 values of Cs obtained by these experiments were plotted as a function 137 of the annealing time (Fig. 10). The fractional Cs release of pyrocarbon coated particles without kernel additives reached the in-pile measured value of 1.5 . 10 ^ after an annealing time of about 20 days ( - 500 hours). The observed 137Cs release of particles with 5 wt. % alumina-silica kernel -7 additives was independent of the annealing time and remained below 2 . 10 This value is comparable with the uranium contamination of the particle surface (see table III), i.e. no caesium diffusion from the kernel to the particle surface had occurred.
Si/Al atomic ratio
Fig. 8: CONCENTRATION OF THE SOLID FISSION PRODUCTS Sr, Ba, Cs■IN THE ALUMINA- SILICA PHASE OF IRRADIATED FUEL KERNELS - 15 -
4. IRRADIATION EXPERIMENT FRJ2-P18
In the second experiment on coated particles with fission product retaining kernel additives each capsule contained only one particle variety. The coated particles were irradiated in graphite trays with three concentric grooves (Fig. 11). Each capsule consisted of 20 trays stuck on a central graphite spine. Between 7 and 15 trays per capsule were fueled with loose coated particles according to the desired fissile material loading. The other Cs Release trays remained empty. One particle variety contained a 500 ThO^ kernel at 1200 °C
in-pile ^ _ UO2 kernel without additives
UO2 kernel with 5 wt.% AI2D3-Si02
Fig. 11: GRAPHITE TRAYS WITH COATED PARTICLES AFTER IRRADIATION (FRJ2-P18)
235 Fig. 10: CS RELEASE FROM IRRADIATED COATED PARTICLES AT 1200 C with alumina-silica additives and sufficient U for reaching 5 % burn-up (FRJ2-P17) (’’breed particle”), the kernels of the other three varieties were small and contained high-enriched uranium (’’feed particles”). The main data of the fuel particles and the irradiation are summarized in table IV. 17 - 18 -
KERNEL © © © O material (U,Th)02 + uo2 + U°2 CM A1203/Si02 A12°3/Si02 IA composition Q. O (wt. %) U°2 17.5 70.6 100 100 UO2 kernel Th02 77.5 - - - without additives A12°3 2.3 20.0 - - 31,6% FIFA/28,4%FIMA 1150 °C Si02 2.7 9.4 - -
enrichment 89.9 89.9 93.1 (% 235 u) 93.1
diameter 525 315 206 205 (^um)
COATING BISO BISO TRISO BISO UO2 kernel with AlaOs/SiC^additives material Pyc PyC PyC/SiC PyC 34,1% FIFA/30,7% FIMA
total thickness 183 203 186 180 1250 °C (^um) i'rh,U)02kemel with IRRADIATION M20^S\02 additive 32,7% FI FA / 5,4% FIMA time 58.5 58.5 58.5 58.5 (full power days) 1220 °C temperature 1220 1250 1190 1150 (°C) top bottom burn-up 5.4 30.7 29.8 28.4 (% FIMA) 1 cm
Table IV: IRRADIATION EXPERIMENT FRJ2-P18: FUEL AND IRRADIATION DATA Fig. 12: GAMMA ACTIVITY OF THE EMPTY GRAPHITE TRAYS (FRJ2-P18) 4.1 Release of Solid_Fission Products
The evaluation of the solid fission product release measurements is not yet finished, therefore only some preliminary results can be presented.
After irradiation the capsules were opened in the hot cells and all particles were removed from the graphite trays. The total gamma activity of the empty trays was measured in the same sequence of the trays as they had been irradiated in the capsule. The results of the three capsules containing pyrocarbon-coated particles are shown in Fig. 12: the gamma activity of the released fission products is plotted as a function of the position of the tray in the capsule. The total gamma activity of the released fission products was lower by at least two orders of magnitude , if the kernels contained alumina-silica additives, although the irradiation temperature of the particles with kernel additives was about 100°C higher than the temperature of the particles without kernel additives. The same result was obtained from beta autoradiographs of the empty trays: taking into account the sensitivity of the used film the blackness after the same time of exposure was about two orders of magnitude lower, if the kernels contained alumina-silica additives. In the case of the silicon carbide-coated particles (TRISO, variety (f)) an unexpected behaviour was observed. Contrary to the pyrocarbon-coated feed particles (BISO) without kernel additives (variety ®) from which a high and very uniform fission product release was observed, the silicon carbide- coated particles (TRISO) showed fission product release only from some single particles (Fig. 13). It should be mentioned that in this case each fueled tray contained about 300 particles. The coatings of the releasing particles
CORROSION 21 - - 22 -
were found to be intact. Even by scanning electron microscopy no microcracks
were observed on the particle surface. The ceramographic section of the As a result of this estimation it can be stated that the irradiation time
TRISO particles, however, showed considerable silicon carbide corrosion of of 58.5 full power days was sufficient for a complete penetration of the some particles (Fig. 14). Contrary to the experience of the Oak Ridge outer pyrocarbon layer (40 ^um) of the TRISO particle by caesium, if the National Laboratory ^ silicon carbide corrosion occurred in this case in SiC layer had become permeable: the break-through time would be about particles with oxide kernel. This is probably the reason for the observed 22 days. On the other hand the BISO particles without kernel additives increase in solid fission product release for some of these particles. (variety ©) did not reach equilibrium release conditions: at 1150°C the
break-trough time of caesium in the 87 ^um thick dense pyrocarbon layer In this context the first results of gamma spectrometric measurements on would be about 250 days. This may be the reason for the relatively small the graphite trays are of interest. One single graphite tray was taken from caesium release of this particle variety. As the estimated caesium break the center of each capsule, the results are shown in table V. The zirconium through times of the varieties © and © with alumina-silica kernel release is nearly independent of kernel additives and irradiation temperature additives are comparable with the irradiation time, it can be concluded that and is probably caused by the direct fission of the uranium contamination the observed low caesium release of these particles is caused by improved on the particle surface. The caesium fraction in the trays of BISO particles kernel retention. with kernel additives is about 1G“6 and without kernel additives one order
of magnitude higher, although these particles were irradiated at a lower The behaviour of barium-140 (table V) was quite different: the highest temperature. The maximum caesium fraction was found in the trays of the release value was observed for BISO particles without kernel additives TRISO particles: In this case part of the SiC layers evidently became perme (variety ©), the release from the TRISO variety ® was lower and again able by SiC corrosion, and the irradiation time was sufficient for complete the BISO particles with alumina-silica kernel additives had the lowest penetration of the thin outer pyrocarbon layer by caesium. release values. According to the rather high diffusion coefficient of barium in propene-derived pyrocarbon the estimated break-through times are
The time for reaching complete penetration of the outer pyrocarbon layer very short: they are below 6 hours at 1200°C.Consequently, equilibrium with caesium ("break-through time”) tCg was estimated from the diffusion release conditions were reached for all particle varieties, therefore the coefficient of Cs in pyrocarbon DCg and the layer thickness 1 using the barium release is controlled by the kernel release, and, in the case of equation the TRISO variety, by the fraction of permeable silicon carbide layers. The total penetration of the intact SiC layers by barium was not reached during Cs (1) Cs irradiation (break-through time t^ (SiC) y 110 days). Because of the relatively fast diffusion of barium in the pyrocarbon coating the efficiency of alumina-silica kernel additives for retaining barium The diffusion coefficients of caesium in propene-derived pyrocarbon were 90 calculated from equation (2) given by MORGAN and coworkers 9'): becomes evident. Similar results can be expected for strontium (the Sr release data are not yet available), as were found in the experiment DCg = 168 • exp (- 4.74 . 104/T) (2) FRJ2-P17 (see 3.1.).
These diffusion coefficients agree rather well with values given by BURNETTE and recently by UKAEA Harwell . The diffusion coefficient t of Ba in propene-derived pyrocarbon is about D- = lO-9 cm2s~1 at o 9) 83 1200 +_ 50 C , the diffusion coefficient of Ba in SiC can be estimated from Sr-data 12} to be < 10~13 cm2s"1. 23
4.2 Electron_Microgrobe-Studies iQ (£3 3-
Ba rH T—t Tp The ceramographic sections of irradiated (U,Th)02 kernels with alumina-
1 W CO CT> CO to silica additives (variety (l)) looked very similar to the irradiated UO^ CM 3- 3- H kernels with alumina-silica additives of the experiment FRJ2-P17 (see
Fraction r- ■ ! t i o o o o Cs rH rH I—t and the burn-up is below 10 % FIMA. 137 to co co d- CO tH LO -H Quite a different irradiation behaviour was observed for particles with alumina-silica diluted high enriched U02 kernels (feed particles), which r- r-" r-- t-" lli! O O O O reached bum-up values of about 30 % FIMA (variety ©). The kernels contain Fission Product Fission »H rH rH Zr 29.4 wt. % additives corresponding to about 60 vol. % alumina-silica phase. 95 CM O Zj- CT) to to CO Before irradiation,the coated particles were heated for one hour at 1800°C. During this heat treatment the kernels became liquid resulting in a marked C layer. bO e o -p CO w w CO fd •H M as M After irradiation the kernel structure was completely changed (Fig. 15 B). cq m oq o The gap disappeared and was filled with a new phase. In the center of the CM CS kernel the distribution of the initial two phases became very fine and there o o > CO CO were additionally small fission gas pores and metallic fission product •f- inclusions. The first results of electron microprobe studies on the irradiated •H 1 1 XI o oJ < CM CM kernels (Fig. 16) can be summarized as follows: H rH < >> material and in the ceramic phase between kernel and porous layer. •H © © © © u > Silica forms inclusions together with alumina in the kernel and in the carbon phase between kernel and coating. 25 26 - Fig. 16: DISTRIBUTION OF URANIUM, CARBON, KERNEL ADDITIVES AND SOLID FISSION PRODUCTS AFTER IRRADIATION (FRJ2-P18, VARIETY ®, 30,7 % FIMA) Fig. 15: FEED PARTICLE WITH ALUMINA-SILICA DILUTED UC>2 KERNEL (FRJ2-P18) A) BEFORE IRRADIATION B) AFTER IRRADIATION (30.7 % FIMA) - 27 - - The fission products Cs, Sr, Ba are localized at the positions of the - efficiency of alumina-silica kernel additives after failure of the alumina-silica inclusions not only in the kernel but also in the carbon pyrocarbon coating, phase between kernel and coating. - behaviour of kernel additives during the reprocessing of irradiated fuel. - The fission products Zr and Nd are distributed homogeneously in the UC^ phase and not concentrated in the alumina-silica inclusions. 6. ACKNOWLEDGEMENT - The metallic fission products Mo, Tc, Ru form mainly small separate inclusions, partially these fission products migrated to the kernel It is not possible to list all persons who have contributed with their work surface and were deposited at the border between kernel and the carbon to the two irradiation experiments. We wish to thank particularly G. BlaB phase. It still has to be investigated whether in this region carbides (kernel preparation, coordination of the preirradiation work, annealing of the metallic fission products have been formed. studies and fission product release measurements), A.K. Gupta (coating), K. Tauber (characterization), H. Hoven(ceramography), W. Jungen (chemical analysis), A. Schirbach (compact manufacture), B, Hurttlen (irradiation), 5. SMART AND CONCLUSIONS F. Schmidt (rig manufacture), J. Rau (coordination of the hot cell work), B. Thiele (hot ceramography and beta autoradiography), Ch. Bauer (compact By first irradiation experiments on pyrocarbon-coated particles with alumina- desintegration), R. SchrSder (gamma spectrometric measurements), W. Kuhnlein silica kernel additives the improved retention of the fission products caesium, (evaluation of gamma spectrometric measurements), K.D. Rohe and W. Tirtey strontium and barium in the additional ceramic phase could be demonstrated. (electron microprobe investigations). Electron microprobe analysis of the fission product distribution in the irradiated kernels provided the first information on the chemical composition of the fission product compounds in the alumina-silica phase. It was shown that this new method of improved fission product retention in the kernel can also be applied, if higher burn-up values are desired. The results of these experiments have shown that this new type of HTGR fuel may become an important alternative to SiC-coated particles for use in advanced HTGRs. The advantages of pyrocarbon-coated particles with fission product retaining kernel additives compared with SiC-coated particles are the lower fabrication costs, the simple coating design, a sure retention mechanism and a good irradiation behaviour. Further investigations will contribute to the following aims: - calculation of the desired amount of alumina and silica in the kernel as a function of the bum-up, - test of feed particles and breed particles with alumina-silica kernel additives up to the maximum values of burn-up and fast neutron fluence. - 29 - 7. REFEREMCES j X. K. Ehlers, K. R61iig9 R. Rotterdam: V Reaktortagung des Deutschen Atomforums, Berlin 1974, Tagungsberioht 1 p. 416-419 2. R. Forthmann, E. Gyarmati, H. Nickel, K. Hilpert: Proc. IAEA-Symposium on Principles and Standards of Reactor Safety, Julich, Feb. 1973, IAEA-SM-169/40 p. 583-598 3. R. Forthmann, H. Griibmeier, H. Kleykamp, A. Naoumidis: Proc. lAEA-Symposium on Thermodynamics of Nuclear Materials, Vienna, Okt. 1974, IAEA-SB-190/35 p. 147-162 4. K. Hilpert, R. Forthmann, H. Nickel: J. Nucl. Mater. 52 (1974) 89-94 5. I. Barin, 0. Khacke: Thermochemical properties of inorganic substances, Springer-Verlag Berlin Heidelberg New York, Verlag Stahleisen mbH Dusseldorf (1973) 6. R. Forthmann: Die chemischen Grundlagen des Hydrolyseverfahrens zur Herstellung sphSrischer Kembrennstoffteilchen , Jul-950-RW (Mai 1973) 7. H. Nabielek, P.E. Brown: Reaktortagung des Deutschen Atomforums, Nurnberg 1975, Tagungsberioht p. 370-373 - 8. T.D. Gulden, J.L. Scott, C. Moreau: Present Thorium-Cycle Concepts and Performance Limitations, Proc. ANS Topical Meeting on Gascooled Reactors, Gatlinhurg/Tenn./USA (May 1974), C0NF-740501, p. 176-200 9. M.T. Morgan, H.J. de Nordwall, R.L. Towns: Release of Fission Products from Pyrocarbon-Coated HTGR Fuel Particles During Postirradiation Anneals, 0RNL-TN-4539 (Dec. 1974) 10. R.D. Burnette, W.E. Bell, N.L. Baldwin: Fission Product Retention Characteristics of HTGR Fuel, Paper No. 16, Proc. BNES Conf. on Nuclear Fuel Performance, London (Oct. 1973) 11. R. Flowers, R. Faircloth, F.E. Brown: Private Communication 12. B. Chinaglia, T. Corte, G. Constanzo, G. Mosoa, P. Novario, E. Voice: The Diffusion of Strontium in Pyrolytic Silicon Carbide, D.P. Report 805 (July 1972)