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The inert fluid perfluorocarbon (PFC) has been used since about 1960 in liquid respiration and artificial blood for mammals. PFC has been used to successfully resuscitate tardigrades that had been dried and exposed to a high barometric pressure of 6,000 atmospheres. Next, scientists attempted to experimentally preserve organs that had been removed from animals, dried, and immersed in PFC. Since 1998 preservation and resuscitation experiments have been conducted with mammalian hearts using 2,015 rats and 70 pigs. Among those experiments, the maximum time after desiccation until successful resuscitation was 26 days for a rat heart and 37 days for a pig heart. However, these results could not be reproduced. Finally, in 2005, this laboratory demonstrated that a rat heart removed under 2 atmospheres pressure and a CO2 partial pressure of 400 hPa, followed by desiccation for 24 h, could be revived and heterotypically transplanted. Moreover, these results were reproducible. The preservation time can be extended to 72 h if, after immersing isolated rat hearts in PFC, they are dried by air exposure under a CO2 partial pressure of 100 hPa. The present report documents the resuscitation of this heart after 72 h of preservation followed by heterotrophic transplantation.

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...

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.

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 t...

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.

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.

TiNi and TiPd shape memory alloys (SMA) reveals good workability and shape memory properties and these SMAs seem to be one of hopeful functional materials in a severe irradiation field. The effects of irradiation induced residual strain on the deformation processes of TiNi and TiPd SMAs after neutron irradiation with fluences ( E > 1 MeV) up to 3.9 × 10 24 m −2 at Japan Materials Testing Reactor (JMTR) were investigated by the remote controlled X-ray diffraction measurement. Residual strains of TiNi SMAs took place over damage of 0.1 dpa and the strains were not completely removed by post-irradiation annealing at 473, 523 and 573 K. On the other hand, residual strains of TiPd SMAs scarcely occurred after irradiation, and furthermore, the TiPd SMA seems to be an irradiation-resistant material. This may be explained by a difference between the irradiation response of a parent phase and that of a martensitic phase to neutron irradiation. Deformation processes of SMAs are associated with stress fields generated by irradiation in a parent phase with B2 type ordered structure or in a martensitic phase with 2H and 9R type close-packed structures.

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 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.

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.

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.