Hajime Sekino

Fission product release from ZrC-coated fuel particles during post-irradiation heating at 1800 and 2000°C

Abstract The ZrC coating layer is a candidate to replace the SiC coating layer of the Triso-coated fuel particles for high-temperature gas-cooled reactors. Post-irradiation heating tests of the ZrC-Triso coated UO 2 particles were performed at 1800°C for 3000 h and at 2000°C for 100 h to study the release behavior of fission products. The fission gas release monitoring and the X-ray microradiography revealed that no through-coating failure occurred during the heating tests. The high cesium retention of the ZrC-Triso coated fuel particles was confirmed up to 1800°C. The diffusion coefficient for cesium in the ZrC layer was more than two orders smaller than that in the SiC layer at 1800°C. The diffusion coefficient for ruthenium in the ZrC layer was almost the same as that for cesium in the SiC layer.

Irradiation experiment on ZrC-coated fuel particles for high-temperature gas-cooled reactors

The ZrC coating layer is a candidate to replace the SiC coating layer of the Triso-coated fuel particle. To compare the irradiation performance of the ZrC Triso-coated fuel particles with that of the normal Triso-coated fuel particles at high temperatures, a capsule irradiation experiment was performed, where both types of the coated fuel particles were irradiated under identical conditions. The burnup was 4.5% FIMA and the irradiation temperature was 1400 to 1650°C. The postirradiation measurement of the through-coating failure fractions of both types of coated fuel particles revealed better irradiation performance of the ZrC Triso-coated fuel particles. The optical microscopy and electron probe microanalysis on the polished cross section of the ZrC Triso-coated fuel particles revealed no interaction of palladium with the ZrC coating layer nor accumulation of palladium at the inner surface of the ZrC coating layer, whereas severe corrosion of the SiC coating layer was observed in the normal Triso-coated fuel particles. Although no corrosion of the ZrC coating layer was observed, additional evaluations need to be made of this layer’s ability to satisfactorily retain the fission product palladium.

Fission product release from ZrC-coated fuel particles during postirradiation heating at 1600°C

Abstract Release behavior of fission products from ZrC-coated UO 2 particles was studied by a postirradiation heating test at 1600°C (1873 K) for 4500 h and subsequent postheating examinations. The fission gas release monitoring and the postheating examinations revealed that no pressure vessel failure occurred in the test. Ceramographic observations showed no palladium attack and thermal degradation of ZrC. Fission products of 137 Cs 134 Cs, 106 Ru, 144 Ce, 154 Eu and 155 Eu were released from the coated particles through the coating layers during the postirradiation heating. Diffusion coefficients of 137 Cs and 106 Ru in the ZrC coating layer were evaluated from the release curves based on a diffusion model. 137 Cs retentiveness of the ZrC coating layer was much better than that of the SiC coating layer.

Release behavior of metallic fission products from HTGR fuel particles at 1600 to 1900°C

Abstract Release behavior of metallic fission products from the Triso-coated UO 2 particles was studied by postirradiation heating tests in the temperature range 1600 to 1900°C (1873 to 2173 K) and subsequent postheating examinations. The fission gas release monitoring and the postheating examinations revealed that no pressure vessel failure occurred in the tests. Ceramographic observations showed no palladium attack and thermal decomposition of SiC, 137 Cs, 134 Cs, 110m Ag, 154 Eu and 155 Eu were released from the coated particles through the coating layers during postirradiation heating. The diffusion coefficient of 137 Cs in the SiC layer was evaluated from the release curves based on a simple diffusion model assuming a one-layer coated particle. Fractional release measurements suggested that the diffusion coefficient of 110m Ag in SiC be larger than that of 137 Cs.

Deterioration of ZrC-coated fuel particle caused by failure of pyrolytic carbon layer

Abstract The ZrC coating layer is a candidate to replace the SiC coating layer of the Triso-coated fuel particles for high-temperature gas-cooled reactors. To understand the behavior of the ZrC-Triso-coated fuel particles at 1800 to 2000°C, a ceramographic examination and an electron probe microanalysis were performed on the ZrC-Triso-coated fuel particles after the post-irradiation heating tests and a thermodynamic analysis of the ZrCUO system was carried out. Based on the results of the examination and analyses, a mechanism of the deterioration of the ZrC-Triso-coated fuel particles was proposed. The deterioration of the ZrC-Triso-coated fuel particles observed at 1800 to 2000°C was caused by the failure of the inner pyrolytic carbon layer.

Retention of fission product caesium in ZrC-coated fuel particles for high-temperature gas-cooled reactors

Abstract The ZrC coating layer is a candidate to replace the SiC coating layer of the Triso-coated fuel particle for high-temperature gas-cooled reactors. To understand mechanisms of the good retention capabilities for fission product caesium of the ZrC Triso-coated fuel particles, the particles after post-irradiation heating tests were examined individually with X-ray microradiography and the caesium inventories of the fuel kernel and coating layers of each particle were measured with gamma-ray spectrometry. The fractional content of 137Cs in the fuel kernel was found to be different from particle to particle though 137Cs was not released from the particles practically. The particles, which showed relatively good retention of 137Cs in the fuel kernels, had radially broken inner pyrolytic carbon layers and deformed fuel kernels. The ZrC layer developed the caesium retention capabilities of the fuel kernel through interaction of ZrC with the fuel kernel.

A model to predict the ultimate failure of coated fuel particles during core heatup events

In this paper a model to predict the ultimate failure of TRISO-coated fuel particles under hypothetical core heatup events is proposed. Features of the model include the ability to treat the statistical variation of the number of coated fuel particles and to make a thermodynamic estimation of the stoichiometry of irradiated UO{sub 2} kernels and the equilibrium CO pressures. The model predictions agree well with the results of postirradiation heating tests. The thermal creep of pyrolytic carbon, however, must be taken into account to further improve the accuracy of the prediction.

Release of metal fission products from UO2 kernel of coated fuel particle

Abstract Release behaviors of Cs 137 , Cs 134 , Sb 125 , Ce 144 , Ru 106 and Ag 110m during irradiation from the high-density UO 2 kernel of Triso-coated fuel particle were studied. The intact coated fuel particles were carefully cracked after irradiation; the amounts of fission products remaining in the UO 2 and their partition between the UO 2 and the coating were measured hv gamma-ray spectrometry. The results are discussed in view of the high-temperature chemistry of the fission products within the coated fuel particles.

MAGNETIZATION OF CERAMIC Y-BA-CU-O AND BI-SR-CA-CU-O AFTER NEUTRON IRRADIATION

Magnetization of ceramic Y-Ba-Cu-O and Bi-Sr-Ca-Cu-O superconductors was studied after neutron irradiation in the fluence from 2.4×1021/m2 to 1.8×1022/m2 at 60°C. The area of hysteresis loops was enhanced at low neutron fluence, followed by saturation and then a decrease at high fluence. In Bi-Sr-Ca-Cu-O, the degree of enhancement was smaller than in Y-Ba-Cu-O and the enhancement reached saturation at lower neutron fluence.

Integrity Confirmation Tests and Post-irradiation Test Plan of the HTTR First-Loading Fuel

Since the first-loading fuel of the High Temperature Engineering Test Reactor (HTTR) is the first mass-production High Temperature Gas-cooled Reactor (HTGR) fuel in Japan, their quality should be carefully inspected. For the quality control related to the fabrication process, Japan Atomic Energy Research Institute (JAERI) carried out the tests to certify the fuel integrity during operation. The tests comprise (1) as-fabricated SiC failure fraction measurement, (2) high-temperature heatup test of irradiated fuel and (3) accelerated irradiation test. For (1), the SiC failure fraction was measured independently in JAERI in addition to the measurement in the fabrication process. The measured failure fractions agreed within 95% confidence limit. In order to confirm the integrity of the SiC layer with respect to the 1,600°C criterion, the high-temperature heatup test of irradiated fuel compact was carried out. The result showed that no failed particle was present in the fuel compact after heating. The diffusion coefficient of metallic fission products in SiC layer was also examined in a series of post-irradiation heating tests. The measured diffusion coefficient of 137Cs showed a good holding ability as those obtained for research and development fuel specimen. The measured fission gas release rate in accelerated irradiation test showed no additional failure up to 60 GWd/t which was about two times higher than 33 GWd/t of the maximum burnup in the HTTR core. Through the tests, integrity of as-fabricated first-loading fuel of the HTTR was finally confirmed. The future post-irradiation test plan, which will be carried out to confirm the fuel irradiation performance and to obtain the data on its irradiation characteristics in the core, is also described.