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Journal Articles

High-temperature interaction between zirconium and UO$$_2$$

Shirasu, Noriko; Suzuki, Akihiro*; Nagae, Yuji; Kurata, Masaki

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05

High temperature interaction tests between UO$$_{2}$$ and Zr were performed at around 2173 K, to make clear the UO$$_{2}$$/ $$alpha$$-Zr(O) interaction and the mechanism of degradation, for developing the improved models for advanced severe accident analysis codes. A Zr plate was inserted in a UO$$_{2}$$ crucible, and heat treated at 2173 K in stream of Ar. After the heat-treatment, the samples were subjected to surface microanalysis. The middle region of Zr sample shows streak-like structures which are extended towered the top. It is confirmed that the streak-like structures were mainly consist of U from the EDX results, and the structures revealed that the U-rich phase was liquid during the heat-treatment. It seems that the U-rich liquid grew selectively toward the area where the oxygen concentration was low.

Journal Articles

Fuel behavior analysis for accident tolerant fuel with sic cladding using adapted FEMAXI-7 code

Shirasu, Noriko; Saito, Hiroaki; Yamashita, Shinichiro; Nagase, Fumihisa

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 8 Pages, 2017/09

Silicon carbide (SiC) is an attractive candidate of accident tolerant fuel (ATF) cladding material because of its high chemical stability, high radiation resistance and low neutron absorption. FEMAXI-ATF has been developed to analysis SiC cladding fuel behaviors. The thermal, mechanical and irradiation property models were implemented to FEMAXI-7, which is a fuel behavior analysis code being developed in JAEA. Fuel rod behavior analysis was performed under typical boiling water reactor (BWR) operating conditions with a model based on a 9$$times$$9 BWR fuel (Step III Type B), in which the cladding material was replaced from Zircaloy to SiC. The SiC cladding shows large swelling by irradiation. It increases the gap size and decreases cladding thermal conductivity. The mechanism of relaxation of stress is also different from the Zircaloy cladding. The experimental data for SiC materials are still insufficient to construct the models, especially for evaluating fracture behavior.

Journal Articles

The Applicability of SiC-SiC fuel cladding to conventional PWR power plant

Furumoto, Kenichiro*; Watanabe, Seiichi*; Yamamoto, Teruhisa*; Teshima, Hideyuki*; Yamashita, Shinichiro; Saito, Hiroaki; Shirasu, Noriko

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

Since 2015, Mitsubishi Nuclear Fuel (MNF) has joined in a Japanese R&D project of ATF founded by the Ministry of Economy, Trade and Industry (METI) as a subcontractor to Japan Atomic Energy Agency (JAEA) which is the prime contractor to METI. In this program, MNF plans to evaluate an influence of Silicon Carbide (SiC) composite cladding upon fuel rod behavior in current pressurized water reactors (PWR). This paper reports the evaluation result of the applicability of fuel rod with SiC composite cladding for a conventional PWR. For the applicability evaluations of SiC composite to conventional PWR, both of analytical evaluations and out-of-pile tests for SiC composite were conducted. Analytical evaluations were performed by Mitsubishi's own fuel rod design code and the fuel rod behavior evaluation code developed by JAEA. These codes were modified to evaluate the behavior of the fuel rod with SiC composite cladding. As out-of-pile tests, thermal diffusivity measurement and autoclave corrosion test for SiC composite samples were performed. Test apparatus were developed for evaluation of performance of SiC composite under the condition simulated design basis accident (DBA).

Journal Articles

Corium stratification test using intermediate products of degraded core materials in severe accident of BWR

Tokushima, Kazuyuki; Shirasu, Noriko; Hoshino, Kuniyoshi*; Ohara, Hiroshi*; Kurata, Masaki

Proceedings of Annual Topical Meeting on LWR Fuels with Enhanced Safety and Performance (TopFuel 2016) (USB Flash Drive), p.1055 - 1063, 2016/09

At the fuel assembly degradation stage in severe accidents, chemical features of the intermediate products are expected to be changed depending upon the accident progressions. These differences are originated from the differences in oxygen potential and temperature, and are highly important for evaluating the relocation and stratification progress of the fuel debris. Two types of sim-test with the different oxygen potentials were performed to investigate these tendencies. The chemical features of the intermediate materials used in the tests were determined from the observations for the control blade and channel box degradation in our previous study. The present results indicate that the U concentration in the metallic layer is largely varied depending upon the oxygen potential of the atmosphere. Also, when the B$$_{4}$$C-Fe alloy, as of a typical intermediate product, coexists with UO$$_{2}$$ and Zr, the apparent red-ox reaction rate between UO$$_{2}$$ and Zr are mitigated.

Journal Articles

Fundamental experiments on phase stabilities of Fe-B-C ternary systems

Sudo, Ayako; Nishi, Tsuyoshi; Shirasu, Noriko; Takano, Masahide; Kurata, Masaki

Journal of Nuclear Science and Technology, 52(10), p.1308 - 1312, 2015/10

 Times Cited Count:4 Percentile:61.74(Nuclear Science & Technology)

For understanding the control blade degradation mechanism of BWR, the thermodynamic database for the fuel assembly materials is a useful tool. Although iron, boron, and carbon ternary system is a dominant phase diagram, phase relation data is not sufficient for the region in which the boron and carbon compositions are richer than the eutectic composition. The phase relations of three samples were analyzed by X-ray diffraction, scanning electron microscope and energy dispersed X-ray spectrometry. The results indicate that Fe$$_{3}$$(B,C) phase only exists in the intermediate region at 1273 K and that the solidus temperature widely maintains at about 1400 K for all three samples, which are different from the calculated data using previous thermodynamic database. The difference might be originated from the over-estimations of the interaction parameter between boron and carbon in Fe$$_{3}$$(B,C).

Journal Articles

Investigation of single-cycle separation process based on forward and backward extractions of actinides and fission products

Sasaki, Yuji; Tsubata, Yasuhiro; Shirasu, Noriko; Morita, Keisuke; Suzuki, Tomoya

Nippon Genshiryoku Gakkai Wabun Rombunshi, 14(3), p.202 - 212, 2015/09

We have been developing the new partitioning method of high-level radioactive waste by single-cycle extraction process. This process is composed of extraction of actinides (An) and fission products (FP, e.g., Pd, Ru, Mo and Tc), and mutual separation by back-extraction. The extractant employed in this process is required to extract soft, hard acid metals and oxonium anions simultaneously. The NTAamide (hexaoctyl-nitrilotriacetamide) is one of the candidate extractants. After extraction of An and FP, the mutual separation by back-extraction should be set up. Pd and Ru extracted by NTAamide can be back-extracted by complexing agents such as thiourea, systeine, diethylenetriamine, and trisaminoethylamine, and the back-extraction of Mo can be performed by methylimino-diethylacetamide (MIDEA), NTAamide(C2) and iminodimethylphosphoric acid, and Re can be done by aqueous phase with high pH.

Journal Articles

Concept for the single cycle process based on mutual separation by reverse extraction of actinides and fission products

Sasaki, Yuji; Tsubata, Yasuhiro; Shirasu, Noriko; Morita, Keisuke; Suzuki, Tomoya

Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.1653 - 1656, 2015/09

The concept for the new partitioning method of high-level radioactive waste (HLW) by single-cycle extraction process has been investigated. This process is based on extraction of actinides (An) and fission products (FP), and mutual separation by reverse extraction. Solo extractant and several stripping reagents will be utilized in this process. The extractant employed in this process is required to extract soft (platinum metals), hard acid metals(An), and oxonium anions (Mo, Tc) simultaneously. NTAamide is one of the candidate extractants. After extraction of An and FP by NTAamide(C8), the mutual separation among these metals by reverse extraction will be followed using the suitable water-miscible reagents. The extraction of An and FP, and the masking effect by some water-miscible reagents has been studied.

Journal Articles

Thermodynamic evaluation on chemical reaction between degraded nuclear fuel and B$$_{4}$$C control rod in severe accident of LWR

Shirasu, Noriko; Kurata, Masaki; Ogawa, Toru*

Proceedings of 2014 Water Reactor Fuel Performance Meeting/ Top Fuel / LWR Fuel Performance Meeting (WRFPM 2014) (USB Flash Drive), 6 Pages, 2014/09

In the accident of Fukushima-Daiichi Nuclear Power Plant, degraded fuels containing Zircaloy probably reacted with B$$_{4}$$C control blades containing stainless steel cladding or blade sheath. Since light elements like B and C are able to react easily with various elements and form various chemical species, several concerns are pointed out, such as variation in volatility and heat generation by oxidation of B and C. The chemical states of degraded fuel were evaluated on the assumption of thermodynamic equilibrium under various conditions of oxygen potential and temperature. The chemical behavior of B affects significantly the variation in oxygen potential with progressing severe accident, and many kinds of volatile compounds are formed by oxidation. The behavior of B causes the changes of volatility of FPs, such as Sr, Cs and Mo.

Journal Articles

Characterization of solidified melt among materials of UO$$_{2}$$ fuel and B$$_{4}$$C control blade

Takano, Masahide; Nishi, Tsuyoshi; Shirasu, Noriko

Journal of Nuclear Science and Technology, 51(7-8), p.859 - 875, 2014/07

 Times Cited Count:13 Percentile:17.44(Nuclear Science & Technology)

To predict phase relationships in the solidified core melt of Fukushima Daiichi Nuclear Power Plants, the solidified melt samples among core materials were prepared by arc melting. Phases and compositions in the samples were determined by X-ray diffraction, microscopy and elemental analysis. The only oxide phase formed is (U,Zr)O$$_{2}$$. The stable metallic phases are Fe-Cr-Ni alloy and Fe$$_{2}$$Zr-type (Fe,Cr,Ni)$$_{2}$$(Zr,U) intermetallic. The borides, ZrB$$_{2}$$ and (Fe,Cr,Ni)$$_{2}$$B, are solidified in the metallic part. Annealing at 1773 K under an oxidizing atmosphere resulted in the oxidation of uranium and zirconium in the alloy and ZrB$$_{2}$$, instead the (Fe,Cr,Ni)$$_{2}$$B and Fe-Cr-Ni alloy became dominant. The metallic zirconium content in the melt is found to be a key factor that determines the phase relationships. As a basic mechanical property, the microhardness of each phase was measured. The borides showed notably higher hardness than any other oxide and metallic phases.

JAEA Reports

"Development of mutual separation technology of minor actinides by the novel hydrophilic and lipophilic diamide compounds" summary of the researches for three years (Contract research)

Sasaki, Yuji; Tsubata, Yasuhiro; Kitatsuji, Yoshihiro; Sugo, Yumi; Shirasu, Noriko; Ikeda, Yasuhisa*; Kawasaki, Takeshi*; Suzuki, Tomoya*; Mimura, Hitoshi*; Usuda, Shigekazu*; et al.

JAEA-Research 2014-008, 220 Pages, 2014/06


The researches on Development of mutual separation technology of minor actinides by the novel hydrophilic and lipophilic diamide compounds, entrusted to Japan Atomic Energy Agency by the Ministry of Education, Culture, Sports, Science and Technology of Japan, from 2010 to 2012 are summarized. This project was composed of three themes, those are (1) Development of total recovery of MA+Ln: basic researches for new extractant, DOODA, (2) Development of mutual separation of Am/Cm/Ln: basic researches of Ln-complex, solvent extraction, and extraction chromatography, and (3) Evaluation of separation technique: process simulation. For topic (1), we summarized the information on characteristic of DOODA extractant. For topic (2), we summarized the information on structures of Ln-complexes, solvent extraction and chromatography. For topic (3), we summarized the information on conditions of mixer-settler and evaluation of each fraction separated.

Journal Articles

Utilization of rock-like oxide fuel in the phase-out scenario

Nishihara, Kenji; Akie, Hiroshi; Shirasu, Noriko; Iwamura, Takamichi*

Journal of Nuclear Science and Technology, 51(2), p.150 - 165, 2014/02

 Times Cited Count:2 Percentile:77.15(Nuclear Science & Technology)

Utilization of rock-like oxide (ROX) fuel in light water reactors for plutonium (Pu) burning was studied by material balance analysis for a case of Japanese phase-out scenario under investigation after the Fukushima accident. For the analysis, the nuclear material balance analysis (NMB) code was developed with features of accurate burn-up calculation, flexible combination of reactors and fuels and an ability to estimate waste and repository. Three scenario-groups of once-through, Pu burning in mixed oxide (MOX) fuel and in ROX fuel were analyzed. By construction of two full MOX- or ROX- reactors, Pu amount is reduced to about a half and isotopic vector of Pu is deteriorated as nuclear weapon especially in terms of spontaneous fission neutron. Effects by ROX are more significant than MOX in both amount and vector. Repository footprint and potential radio-toxicity is not reduced by MOX and ROX because heat and toxicity of MOX and ROX spent fuel is considerably high.

Journal Articles

Thermodynamic evaluation on effect of sea water to degraded nuclear fuel in severe accident of LWR

Kurata, Masaki; Shirasu, Noriko; Ogawa, Toru*

Nippon Genshiryoku Gakkai Wabun Rombunshi, 12(4), p.286 - 294, 2013/12

In the severe accident of Fukushima-Daiichi Nuclear Power Plant, large amount of sea water was introduced into reactor pressure vessels. Not only sodium chloride but also several minor elements contained in sea water possibly reacted with degraded fuel debris or molten corium. Varous concerns are pointed out, such as volatilization of FP, characterization of fuel debris, formation of corrosive gases, and etc. Thermodynamic evaluation can give useful information on the general tendency of these sea-water effects. Volatility of Cs, Sr, and Te is potentially increased due to the change in the chemical species. Corrosive gases, such as HCl, H$$_{2}$$S, and etc. are possibly generated from sea-water heated at high temperature. These phenomena are predicted to be varied with change in oxygen potential.

Journal Articles

Variation in the surface morphology of polycrystalline UO$$_{2}$$ powder induced by helium precipitation

Serizawa, Hiroyuki; Matsunaga, Junji*; Shirasu, Noriko; Nakajima, Kunihisa; Kashibe, Shinji*; Kaji, Yoshiyuki

Journal of Asian Ceramic Societies, 1(3), p.289 - 295, 2013/09

This report addresses the precipitation of helium in polycrystalline UO$$_{2}$$, which deforms the morphology of the grains and their surfaces Helium was injected into pulverized UO$$_{2}$$ particles at 1473 K by hot isostatic pressing (HIP). The specific surface area measured by volumetric gas adsorption instrument implied that small pores should exist on the as-helium-treated sample surface. Field-emission scanning electron microscopy observations showed that numerous shallow basins (approximately 500 nm in radius) with hexagonal fringe were formed on the surface. The basin resembles a ruptured blister whose lid has been forced open. SEM observations showed a uniform polygonal-shaped section of the gas bubble on the fracture surface; this implies that precipitated helium forms a negative crystal in the grain.

Journal Articles

Mutual separation of Am/Cm/Ln by the use of novel-triamide, NTAamide and water-soluble diglycolamide

Sasaki, Yuji; Tsubata, Yasuhiro; Kitatsuji, Yoshihiro; Sugo, Yumi; Shirasu, Noriko; Morita, Yasuji

Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.1079 - 1082, 2013/09

Mutual separation of Am, Cm and lanthanides (Ln) is important to develop the partitioning process of high-level radioactive liquid waste, because the application to their different disposal methods are advantageous. Namely, Am is studied for transmutation due to the reduction of long half-life radionuclides, Cm should be kept in interim storage in order to reduce the calorific value, and Ln should be present in the vitrified radioactive waste toward the geological disposal. However, this mutual separation method is difficult to establish because they have very similar chemical behavior, same oxidation state (III) and similar ionic radii. The development of their mutual separation is termed as the challenging study. In order to obtain the satisfactory results, the property of extractant requires the differentiation of actinide (An) from Ln, high preferability to different ionic radii between Am and Cm, and high extractability to hard acids. Therefore, the extractant have to include both N atom, whose soft donor has high selectivity between An and Ln, and O atoms for the strong extractability to An. The new extractant, NTAamide (N,N,N',N',N'',N''-hexaoctyl-nitrirotriacetamide) is a triamide having N donor at the center of backbone, then NTAamide has hybrid performance of complexation to metals by soft N and three hard amidic O atoms. It is clear that NTAamide can extract trivalent An at diluted HNO$$_{3}$$ with small D(Ln), the separation of An from Ln can be carried out at that condition. The SF of Am/Cm by NTAamide is approximate 1.8, which is not so high to separate each other. The combination of NTAamide of extractant and TEDGA (N,N,N',N'-tetraethyl-diglycolamide) as a masking agent in the aqueous phase shows very high SF(Am/Cm) of maximal 6.5. It is obvious that NTAamide is a promising extractant to achieve the mutual separation among Am/Cm/Ln.

Journal Articles

Multiplier effect on separation of Am and Cm with hydrophilic and lipophilic diamides

Sasaki, Yuji; Tsubata, Yasuhiro; Kitatsuji, Yoshihiro; Sugo, Yumi; Shirasu, Noriko; Morita, Yasuji

Procedia Chemistry, 7, p.380 - 386, 2012/00

 Times Cited Count:8 Percentile:5.36

Following the nuclear properties, the different disposal methods for Am, Cm and lanthanides in HLW have been investigating, e.g., Am; transmutation, Cm; interim storage and Ln; geological disposal. The mutual separation is an important task. However, these trivalent Ln and An are difficult to separate due to their very similar chemical behavior, same oxidation state and similar ionic radii. We try to use both hydrophilic and lipophilic diamides in an extraction system simultaneously in order to attain the effective mutual separation. In this work, lipophilic DOODA or DGA are used as the extractant and some hydrophilic diamides are employed as the masking agents. The results of mutual separation of Am/Cm/Ln are discussed in this work.

Journal Articles

The Solubility and diffusion coefficient of helium in uranium dioxide

Nakajima, Kunihisa; Serizawa, Hiroyuki; Shirasu, Noriko; Haga, Yoshinori; Arai, Yasuo

Journal of Nuclear Materials, 419(1-3), p.272 - 280, 2011/12

 Times Cited Count:16 Percentile:17.43(Materials Science, Multidisciplinary)

The solubility and diffusion coefficient of helium in the single-crystal UO$$_{2}$$ samples were determined by a Knudsen-effusion mass-spectrometric method. The measured helium solubilities were found to lie within the scatter of the available data, but to be much lower than those for the polycrystalline samples. The diffusion analysis was conducted based on a hypothetical equivalent sphere model and the simple Fick's law. The helium diffusion coefficient was determined by using the pre-exponential factor and activation energy as the fitting parameters for the measured and calculated fractional releases of helium. The optimized diffusion coefficients were in good agreement with those obtained by a nuclear reaction method reported in the past. It was also found that the pre-exponential factors of the determined diffusion coefficients were much lower than those analyzed in terms of a simple interstitial diffusion mechanism.

Journal Articles

Fundamental research on behavior of helium in MA-bearing oxide fuel

Arai, Yasuo; Serizawa, Hiroyuki; Nakajima, Kunihisa; Takano, Masahide; Sato, Isamu; Katsuyama, Kozo; Akie, Hiroshi; Suzuki, Motoe; Shirasu, Noriko; Haga, Yoshinori; et al.

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 8 Pages, 2011/12

High amount of He is generated in MA-bearing fuel during irradiation and storage periods compared with that in U or U-Pu fuel. Laboratory scale experiments, post irradiation examinations and modeling study were carried out in order to understand the He behavior in MA-bearing oxide fuel. Diffusion characteristics of He in single-crystal UO$$_{2}$$ were investigated by the Knudsen effusion mass spectrometry. Effects of the He accumulation on lattice and bulk expansion of oxide pellets were examined by use of alpha-decay of $$^{244}$$Cm. Post irradiation examinations of 0.5%Am-MOX fuel irradiated at a fast test reactor JOYO were carried out, concentrating on the He behavior in the fuel pellets. A model describing the He behavior in MA-MOX fuel was constructed based on the principle processes, such as generation, diffusion, equilibrium and release to outer gaseous phase. By use of the model as a subroutine of a conventional fuel behavior analysis code, the He behavior in MA-MOX fuel for fast reactors was simulated.

Journal Articles

Separation of Am, Cm and Ln by combination of two novel-diamide compounds at $$n$$-dodecane and HNO$$_{3}$$ system

Sasaki, Yuji; Kitatsuji, Yoshihiro; Tsubata, Yasuhiro; Sugo, Yumi; Shirasu, Noriko; Morita, Yasuji

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12

Mutual separation of Am, Cm and lanthanides (Ln) is important to develop the partitioning process of high-level radioactive waste, although this method is difficult to establish due to their very similar chemical behavior, same oxidation state (III) and similar ionic radii. Relative high separation values for these metals can be seen when the multidentate ligands, e.g., diethylene-triamine-N,N,N',N'',N''-pentaacetic acid (DTPA), and softN-donor extractants are employed. We examine to use the combination of two-multidentate ligands having N atoms in their structures in behalf of the high separation of Am, Cm and Ln. Various diamide-type ligands can be synthesized from the initial materials of carboxylic acid and amine, and the aimed materials can be tailored to have high solubility in either HNO$$_{3}$$ orn-dodecane, in case short or long alkyl chains are attached in amidic Natoms. Taking the simultaneous use of these materials into consideration, hydrophilic and lipophilic diamide compounds with different complexing ability for trivalent Ln and actinide (An) are used into both of the aqueous and the organic phases in order to enlarge their separation factors. This technique is applicable in the salt-free system and diamide can be gasificated by combustion, which reduces the secondary-radioactive waste. Two diamide compounds, DGA (diglycolamide) and DOODA (dioxaoctanediamide) as the representative multidentate ligands, have the different features for Ln-complexation, namely DGA has higher D values for middle-Ln than those of light-Ln, and DOODA shows the opposite trend to DGA. TDdDGA (N,N,N',N'-tetradodecyl-diglycolamide) and DOODA(C2) (N,N,N',N'-tetraethyl-dioxaoctanediamide) have high solubility in n-dodecane and HNO$$_{3}$$, respectively, and the condition of 0.1 M TDdDGA in n-dodecane and 0.5 MDOODA(C2) in 3 M HNO$$_{3}$$ gives the separation factors of La/Gd: 1128 and Am/Cm:3.26.

Journal Articles

Fuel performance evaluation of rock-like oxide fuels

Shirasu, Noriko; Kuramoto, Kenichi; Nakano, Yoshihiro; Yamashita, Toshiyuki; Kimura, Yasuhiko; Nihei, Yasuo; Ogawa, Toru

Journal of Nuclear Materials, 376(1), p.88 - 97, 2008/05

 Times Cited Count:12 Percentile:31.78(Materials Science, Multidisciplinary)

The concept of the rock-like oxide (ROX) fuel has been developed for the annihilation of excess plutonium in light water reactors. Irradiation tests and post-irradiation examinations were carried out on candidate ROX fuels. The ternary fuel of YSZ-spinel-corundum system, the single-phase fuels of YSZ, the particle dispersed fuels of YSZ in spinel or corundum matrix, and the blended fuels of YSZ and spinel or corundum matrix were fabricated and submitted to irradiation tests. The fuels containing spinel showed chemical instabilities with the vaporization of MgO component, which caused fuel restructuring. The swelling behavior was improved with the particle-dispersed fuels. However, the particle-dispersed fuels showed higher fractional gas release (FGR) than blended type fuels. The FGR of YSZ single-phase fuels were comparable to what would be expected for UO$$_{2}$$ fuels. The YSZ single-phase fuel showed the best irradiation performance among the ROX fuels investigated.

Journal Articles

Study of rock-like oxide fuels under irradiation

Shirasu, Noriko; Kuramoto, Kenichi; Nakano, Yoshihiro; Yamashita, Toshiyuki; Ogawa, Toru

Journal of Power and Energy Systems (Internet), 2(1), p.145 - 152, 2008/00

To evaluate the irradiation behavior of the rock-like oxide fuel, irradiation experiments were carried out. Three fuels were prepared; a single phase fuel of yttria-stabilized zirconia containing UO$$_{2}$$ (U-YSZ) and two types of particle-dispersed fuels of U-YSZ particles in spinel or corundum matrix. These fuels were irradiated in JRR-3 for about 280 days. The burnups were about 11% FIMA. The fission gas release rate (FGR) was determined by puncture test and gas analysis. Corundum-based fuel showed extremely high FGR (88%). The temperature of U-YSZ single-phase fuel pellets was highest among the fuels, because of its low thermal conductivity. Nevertheless the U-YSZ single-phase fuel showed very low FGR (5%). Microstructure analyses of irradiated fuel pellets were carried out by ceramography and EPMA. The restructuring of fuel pellet was not observed in the spinel-based fuel irradiated below 1400 K.

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