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JAEA Reports

Validation of fuel behavior analysis code FEMAXI-8 using fast reactor MOX fuel irradiation tests

Ikusawa, Yoshihisa; Nagayama, Masahiro*

JAEA-Data/Code 2023-006, 24 Pages, 2023/07


Core fuels with stainless steel cladding and high plutonium content mixed oxide (MOX) fuel in a water-cooled environment, such as supercritical water-cooled reactors (SCWR) and reduced-moderation water reactors (RMWR), have been studied. In order to contribute to the research and development of such a core fuel concept, the fuel performance code "FEMAXI-8" was verified based on the results of post irradiation examinations of MOX fuel irradiated in the experimental fast reactor "JOYO". FEMAXI-8 is the latest version of the behavior analysis code developed by JAEA to analyze the behavior of light water reactor fuels under normal operation and transient conditions. This latest code has been improved and developed to allow the selection of stainless steel cladding property models to analyze improved fuels such as accident tolerant fuels. The purpose of this report is to confirm the prediction accuracy of FEMAXI-8 for the irradiation behavior of the new type of core fuel that is currently being developed. As a result of the verification, it was confirmed that FEMAXI-8 has sufficient analysis accuracy for the irradiation behavior of sodium-cooled fast reactor MOX fuel with stainless steel cladding, which exceeds the plutonium content and irradiation conditions of light water reactors. In the future, the analysis accuracy of FEMAXI-8 could be improved by adopting the O/M ratio dependence of MOX fuel thermal conductivity and the irradiation behavior evaluation model at high temperature.

Journal Articles

Materials science and fuel technologies of uranium and plutonium mixed oxide

Kato, Masato; Machida, Masahiko; Hirooka, Shun; Nakamichi, Shinya; Ikusawa, Yoshihisa; Nakamura, Hiroki; Kobayashi, Keita; Ozawa, Takayuki; Maeda, Koji; Sasaki, Shinji; et al.

Materials Science and Fuel Technologies of Uranium and Plutonium mixed Oxide, 171 Pages, 2022/10

Innovative and advanced nuclear reactors using plutonium fuel has been developed in each country. In order to develop a new nuclear fuel, irradiation tests are indispensable, and it is necessary to demonstrate the performance and safety of nuclear fuels. If we can develop a technology that accurately simulates irradiation behavior as a technology that complements the irradiation test, the cost, time, and labor involved in nuclear fuel research and development will be greatly reduced. And safety and reliability can be significantly improved through simulation of nuclear fuel irradiation behavior. In order to evaluate the performance of nuclear fuel, it is necessary to know the physical and chemical properties of the fuel at high temperatures. And it is indispensable to develop a behavior model that describes various phenomena that occur during irradiation. In previous research and development, empirical methods with fitting parameters have been used in many parts of model development. However, empirical techniques can give very different results in areas where there is no data. Therefore, the purpose of this study is to construct a scientific descriptive model that can extrapolate the basic characteristics of fuel to the composition and temperature, and to develop an irradiation behavior analysis code to which the model is applied.

JAEA Reports

Manufacture of substitutive assemblies for MONJU reactor decommissioning

Sakakibara, Hiroshi; Aoki, Nobuhiro; Muto, Masahiro; Otabe, Jun; Takahashi, Kenji*; Fujita, Naoyuki*; Hiyama, Kazuhiko*; Suzuki, Hirokazu*; Kamogawa, Toshiyuki*; Yokosuka, Toru*; et al.

JAEA-Technology 2020-020, 73 Pages, 2021/03


The decommissioning is currently in progress at the prototype fast breeder reactor Monju. Fuel assemblies will be taken out of its core for the first step of the great task. Fuel assemblies stand on their own spike plugged into a socket on the core support plate and support with adjacent assemblies through their housing pads each other, resulting in steady core structure. For this reason, some substitutive assemblies are necessary for the purpose of discharging the fuel assemblies of the core. Monju side commissioned, therefore, Plutonium Fuel Development Center to manufacture the substitutive assemblies and the Center accepted it. This report gives descriptions of design, manufacture, and shipment in regard to the substitutive assemblies.

Journal Articles

Evaluation of breach characteristics of fast reactor fuel pins during steady state irradiation

Oka, Hiroshi*; Kaito, Takeji; Ikusawa, Yoshihisa; Otsuka, Satoshi

Nuclear Engineering and Design, 370, p.110894_1 - 110894_8, 2020/12

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

The objective of this study is to evaluate the reliability of a cumulative damage fraction (CDF) analysis for the prediction of fuel pin breach in fast rector using experimentally obtained fuel pin breach data for the first time. Six breached fuel pins were obtained from steady state irradiation in the EBR-II. Post irradiation examinations revealed that FP gas pressure was the main cause of creep damage in cladding, and that the stress contribution from FCMI was negligible. CDFs evaluated for these pins using in-reactor creep rupture equation, taking into account the irradiation history of cladding temperature and hoop stress due to FP gas pressure, were in the range of 0.7 to 1.4 at the occurrence of breach. This shows clearly that fuel pin breach occurs when the CDF approaches 1.0. The results indicate that CDF analysis would be a reliable method for the prediction of fuel pin breach when appropriate material strength and environmental effects are adopted.

Journal Articles

Oxygen potential and self-irradiation effects on fuel temperature in Am-MOX

Ikusawa, Yoshihisa; Hirooka, Shun; Uno, Masayoshi*

2018 GIF Symposium Proceedings (Internet), p.321 - 327, 2020/05

Research and development of Minor actinides (MAs) bearing MOX fuel for fast reactor has been proceeding from the viewpoint of reducing radioactive waste. In order to develop, MA bearing MOX, it is indispensable to clarify the influence of MA addition on irradiation behavior. The addition of Americium (Am) to MOX affects vapor pressure and thermal conductivity, which are important properties from the perspective of evaluating fuel temperature. This is because vapor pressure affects fuel restructuring, and thermal conductivity affects fuel temperature distribution. Focusing on these physical properties, this study evaluates the influence of Am on fuel temperature using irradiation behavior analysis code to contribute to the development of MA-bearing MOX fuel. An increase in Am content decreases the thermal conductivity and increases the oxygen potential of oxide fuel. Because vapor pressure increases with increasing Am content, pore migration is accelerated, and the central void diameter increases with increasing Am content. As a result, after formation of the central void, the influence of Am content on the fuel center temperature is mild. Alpha particles generated by radioactive decay of transuranium elements cause lattice defects in the oxide fuel pellets. It is well known that this phenomenon, which is called self-irradiation, affects thermal conductivity. Since americium is the typical alpha radioactive nucleus, to evaluate fuel temperature of Am-MOX is necessary to take account of the influence of self-irradiation damage on thermal conductivity. Self-irradiation decreases thermal conductivity, and as the Am content increases, the rate of decrease in thermal conductivity is accelerated. Because it recovers with temperature rise, the decrease in thermal conductivity due to self-irradiation damage has very little effect on fuel center temperature. These results suggest that Am-MOX fuel could be irradiated under the same conditions as conventional MOX fuel.

Journal Articles

The Effects of plutonium content and self-irradiation on thermal conductivity of mixed oxide fuel

Ikusawa, Yoshihisa; Morimoto, Kyoichi; Kato, Masato; Saito, Kosuke; Uno, Masayoshi*

Nuclear Technology, 205(3), p.474 - 485, 2019/03

 Times Cited Count:2 Percentile:22.31(Nuclear Science & Technology)

This study evaluated the effects of plutonium content and self-irradiation on the thermal conductivity of mixed-oxide (MOX) fuel. Samples of UO$$_{2}$$ fuel and various MOX fuels were tested. The MOX fuels had a range of plutonium contents, and some samples were stored for 20 years. The thermal conductivity of these samples was determined from thermal diffusivity measurements taken via laser flash analysis. Although the thermal conductivity decreased with increasing plutonium content, this effect was slight. The effect of self-irradiation was investigated using the stored samples. The reduction in thermal conductivity caused by self-irradiation depended on the plutonium content, its isotopic composition, and storage time. The reduction in thermal conductivity over 20 years' storage can be predicted from the change of lattice parameter. In addition, the decrease in thermal conductivity caused by self-irradiation was recovered with heat treatment, and recovered almost completely at temperatures over 1200 K. From these evaluation results, we formulated an equation for thermal conductivity that is based on the classical phonon-transport model. This equation can predict the thermal conductivity of MOX fuel thermal conductivity by accounting for the influences of plutonium content and self-irradiation.

Journal Articles

Oxide-metal ratio dependence of central void formation of mixed oxide fuel irradiated in fast reactors

Ikusawa, Yoshihisa; Maeda, Koji; Kato, Masato; Uno, Masayoshi*

Nuclear Technology, 199(1), p.83 - 95, 2017/07

 Times Cited Count:3 Percentile:29.5(Nuclear Science & Technology)

Based on thermal computation results obtained using an irradiation behavior analysis code, we have evaluated the effect of O/M ratio on fuel restructuring from the results of PIEs for the B14 irradiation test fuel, which was a mixed oxide fuel and was irradiated in the experimental reactor Joyo. The thermal computation results showed that fuel restructuring in the stoichiometric oxide fuel was accelerated, though the fuel temperature in the stoichiometric oxide fuel was evaluated as lower than that of the hypo-stoichiometric one. We explained this behavior as follows: first, the fuel temperature decreased due to the high thermal conductivity at stoichiometry; second, the pore migration velocity increased due to the increase in vapor pressure caused by the high vapor pressure of UO$$_{3}$$, which was derived from the high oxygen potential at stoichiometry. In addition, our results indicated that the central void diameter strongly depended on not only fuel temperature, but also vapor pressure.

Journal Articles

Design study of a 750 MWe Japan sodium-cooled fast reactor with metal fuel

Ohgama, Kazuya; Ota, Hirokazu*; Ikusawa, Yoshihisa; Oki, Shigeo; Ogata, Takanari*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04

Journal Articles

Thermal expansion measurement of (U,Pu)O$$_{2-x}$$ in oxygen partial pressure-controlled atmosphere

Kato, Masato; Ikusawa, Yoshihisa; Sunaoshi, Takeo*; Nelson, A. T.*; McClellan, K. J.*

Journal of Nuclear Materials, 469, p.223 - 227, 2016/02

 Times Cited Count:9 Percentile:69.2(Materials Science, Multidisciplinary)

Thermal expansion of (U$$_{0.7}$$Pu$$_{0.3}$$)O$$_{2-x}$$ (x = 0, 0.01, 0.02, 0.03) and (U$$_{0.52}$$Pu$$_{0.48}$$)O$$_{2.00}$$ was measured with a dilatometer in an oxygen partial pressure-controlled atmosphere. The oxygen partial pressure was controlled to hold a constant oxygen-to-metal ratio in the (U,Pu)O$$_{2-x}$$ during the measurement. Thermal expansion slightly increased with the decrease in oxygen-to-metal ratio. The relationship was derived to describe thermal expansion.

Journal Articles

Fuel restructuring behavior analysis of MA-bearing MOX fuels irradiated in a fast reactor

Ozawa, Takayuki; Ikusawa, Yoshihisa; Kato, Masato

Transactions of the American Nuclear Society, 113(1), p.622 - 624, 2015/10

A recycle system for minor actinides (MAs), in which MAs are recycled by reprocessing and irradiating them in a fast reactor, is studied to reduce the degree of hazard and the amount of high-level radioactive wastes. MAs would be used as mixed oxide (MOX) fuels with plutonium and uranium in fast reactors. Since MA content of MA-bearing MOX (MA-MOX) to be used in fast reactors is assumed to reach $$sim$$5 wt%HM, the effects on not only fuel properties but also fuel behaviors have to be estimated to use MA-MOX as fast reactor fuels. As the MOX fuels to be used will be irradiated at a comparably high linear power and the fuel center temperature would be assumed to be over 2,273 K during irradiation in the fast reactors, fuel restructuring would take place due to void migration towards the fuel center under the radial temperature gradient, and a central void would be formed. Since the fuel center temperature would be decreased by the effect of formation of the central void, the fuel restructuring is one of the most important behaviors for fast reactor fuels. In this study, the effect of MA content on fuel restructuring behavior was estimated from the results of irradiation experiments such as B11 and B14 performed in Joyo to study the irradiation behaviors of MA-MOX and the calculation results using a fuel restructuring model which can take into account MA-MOX dependence on vapor pressure.

Journal Articles

Development of science-based fuel technologies for Japan's Sodium-Cooled Fast Reactors

Kato, Masato; Hirooka, Shun; Ikusawa, Yoshihisa; Takeuchi, Kentaro; Akashi, Masatoshi; Maeda, Koji; Watanabe, Masashi; Komeno, Akira; Morimoto, Kyoichi

Proceedings of 19th Pacific Basin Nuclear Conference (PBNC 2014) (USB Flash Drive), 12 Pages, 2014/08

Uranium and plutonium mixed oxide (MOX) fuel has been developed for Japan sodium-cooled fast reactors. Science based fuel technologies have been developed to analyse behaviours of MOX pellets in the sintering process and irradiation conditions. The technologies can provide appropriate sintering conditions, irradiation behaviour analysis results and so on using mechanistic models which are derived based on theoretical equations to represent various properties.

Journal Articles

Development and verification of the thermal behavior analysis code for MA containing MOX fuels

Ikusawa, Yoshihisa; Ozawa, Takayuki; Hirooka, Shun; Maeda, Koji; Kato, Masato; Maeda, Seiichiro

Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 6 Pages, 2014/07

In order to develop MA contained MOX (MA-MOX) fuel design method, the analysis models to predict irradiation behavior of MA-MOX fuel have to be developed and the accuracy of irradiation behavior analysis code should be evaluated with the result of post-irradiation examinations (PIEs) for MA-MOX fuels. In this study, we developed the computer module "TRANSIT" to compute thermal properties of MA-MOX fuel. TRANSIT can give thermal conductivity, melting temperature and vapor pressures of MA-MOX. By using this module, we improved the thermal behavior analysis code "DIRAD" and developed DIRAD-TRANSIT code system to compute the irradiation behavior of MA-MOX fuel. This system was verified with the results of PIEs for the conventional MOX fuels and the MA-MOX fuels irradiated in the experimental fast reactor "JOYO". As the result of the verification, it can be mentioned that the DIRAD-TRANSIT system would precisely predict the fuel thermal behavior, i.e. fuel temperature and fuel restructuring, for oxide fuels containing several percent minor actinides.

Journal Articles

Short-term irradiation behavior of low-density americium-doped uranium-plutonium mixed oxide fuels irradiated in a fast reactor

Maeda, Koji; Katsuyama, Kozo; Ikusawa, Yoshihisa; Maeda, Seiichiro

Journal of Nuclear Materials, 416(1-2), p.158 - 165, 2011/09

 Times Cited Count:15 Percentile:74.32(Materials Science, Multidisciplinary)

In order to evaluate the thermal behavior of low-density uranium and plutonium mixed oxide fuels containing several percent of americium (Am-MOX), fuel irradiation test (B14) was conducted using the experimental fast reactor. Pellet-cladding gap width and O/M ratio of oxide fuels were specified as experimental parameters. Four fuel pins were irradiated step-by-step in consideration of fuel restructuring during 48 hours as pre-conditioning before full power reactor operation. The irradiation history, i.e. linear power, was simulated the conventional FBR oxide fuel pins. And the linear power was rapidly increased up to 47 kW/cm for 10 minutes to simulate the transient condition. After the irradiation, ceramography samples were taken from the axial position of each fuel pins where the fuel centerline temperature reached the maximum during irradiation. The result was investigated relative to those of other irradiated fuels.

Journal Articles

MOX fuel performance and database development for MOX fuel use in LWRs

Ozawa, Takayuki; Ikusawa, Yoshihisa

Proceedings of 2010 LWR Fuel Performance Meeting/TopFuel/WRFPM (CD-ROM), p.72 - 81, 2010/09

For the effective utilization of the energy resources, preparations are underway to recycle plutonium separated by reprocessing the spent fuels from nuclear power plants into nuclear fuels in Light Water Reactors (LWRs). In this nuclear fuel cycle, plutonium is reused as uranium-plutonium mixed dioxide (MOX). In Japan, a total of 772 MOX fuel assemblies were used in FUGEN without any failure until the end of its operation in March, 2003, the most MOX fuel usage by a thermal reactor in the world. Several post-irradiation examinations necessary to evaluate the MOX fuel performance were carried out for the MOX fuel assembly irradiated in FUGEN, and consequently we could obtain the usable data to evaluate the irradiation behavior of MOX fuels. Furthermore, several MOX fuel assemblies, which were equipped in-pile instruments, used in the irradiation tests, i.e. the regular operation irradiation tests, the ramp tests, and the load-follow tests, in Norway's "Halden" reactor (HBWR). We developed a MOX fuel database to make the most of our experiences with FUGEN and HBWR in helping improve the reliability of future MOX fuel use in LWRs.

Journal Articles

Burn-up effect on MOX fuel thermal conductivity

Ikusawa, Yoshihisa; Morimoto, Kyoichi; Ozawa, Takayuki; Kato, Masato

Proceedings of Plutonium Futures; The Science 2010 (CD-ROM), p.341 - 342, 2010/09

Thermal conductivity of oxide fuel is important for fuel design and performance analyses. Uranium dioxide and uranium-plutonium mixed oxide (MOX) are used as fuels in light water reactors (LWRs), and the thermal conductivities of these oxide fuels have been measured in various laboratories. In a review of oxide fuel properties, it was reported that the thermal conductivity of oxide fuel would decrease with burn-up increase. In this study, burn-up effect on MOX fuel thermal conductivity was discussed.

JAEA Reports

Fission gas release behavior of MOX fuels under simulated daily-load-follow operation condition; IFA-554/555 test evaluation with FASTGRASS code

Ikusawa, Yoshihisa; Ozawa, Takayuki

JAEA-Technology 2007-070, 27 Pages, 2008/03


IFA-554/555 load-follow tests were performed in HALDEN reactor to study the MOX fuel behavior under the daily-load-follow operation condition in the framework of ATR-MOX fuel development in JAEA. In this report, FP gas release behavior of MOX fuel rod was evaluated under the daily-load-follow operation condition with the examination data from IFA-554/555 by using the computation code FASTGRASS. As the computation results of FASTGRASS code, which could compute the FP gas release behavior under the transient condition, it could be concluded that FP gas would release due to the relaxation of fuel pellet inner stress, which were caused by the cyclic power change during the daily-load-follow operation. In addition, since the amount of released FP gas would decrease during the steady operation after the daily-load-follow, it could be mentioned that the total of FP gas release at the end of life with the daily-load-follow would not be so much different from that without the daily-load-follow.

JAEA Reports

Verification of annular fuel design code "CEPTAR"; Verification with the irradiation data of JOYO Mk-II driver fuel

Ikusawa, Yoshihisa; Ozawa, Takayuki

JAEA-Technology 2007-013, 38 Pages, 2007/03


The following generation MONJU core fuel is considered using the high density solid pellet. Although the fuel design code "CEPTAR" was developed for annular fuel pellet, CEPTAR code was not verified with the data of high density solid pellet. In this study, CEPTAR code was verified with irradiation data of JOYO Mk-II driver fuel that used high density solid pellet. To estimate irradiation behavior of JOYO Mk-II driver fuel, the following new equations were added to CEPTAR code; The swelling equation and irradiation creep equation of PNC316. The pellet swelling equation evaluated with the PIE data of JOYO Mk-II driver fuel. As a result of verification by using the irradiation data of JOYO Mk-II driver fuel, the calculated values with CEPTAR code were in agreement with the observed values from the result of PIEs up to pellet peak burn-up $$sim$$ 76,000MW d/t.

JAEA Reports

Development of MOX fuel database

Ikusawa, Yoshihisa; Ozawa, Takayuki

JAEA-Technology 2007-010, 44 Pages, 2007/03


We developed MOX Fuel Database, which included valuable data from several irradiation tests in FUGEN and Halden reactor, for help of LWR MOX use. This database includes the data of fabrication and irradiation, and the results of post-irradiation examinations for seven fuel assemblies, i.e. P06, P2R, E03, E06, E07, E08 and E09, irradiated in FUGEN. The highest pellet peak burn-up reached $$sim$$48 GWd/t in MOX fuels, of which the maximum plutonium content was $$sim$$6 wt%, irradiated in E09 fuel assembly without any failure. Also the data from the instrumented MOX fuels irradiated in HBWR to study the irradiation behavior of BWR MOX fuels under the steady state condition (IFA-514/565 and IFA-529), under the load-follow operation condition (IFA-554/555) and under the transit condition (IFA-591) are included in this database. The highest assembly burn-up reached $$sim$$56 GWd/t in IFA-565 steady state irradiation test, and the maximum linear power of MOX fuel rods was 58.3-68.4 kW/m without any failure in IFA-591 ramp test. In addition, valuable instrument data, i.e. cladding elongation, fuel stack elongation, fuel center temperature and rod inner pressure were obtained from IFA-554/555 load-follow test.

Journal Articles

Fuel-cladding chemical interaction in MOX fuel rods irradiated to high burnup in an advanced thermal reactor

Tanaka, Kosuke; Maeda, Koji; Sasaki, Shinji; Ikusawa, Yoshihisa; Abe, Tomoyuki

Journal of Nuclear Materials, 357(1-3), p.58 - 68, 2006/10

 Times Cited Count:11 Percentile:60.46(Materials Science, Multidisciplinary)

The performance of MOX fuel irradiated in the advanced thermal reactor, FUGEN, to the burnup of 47.5 GWd/t, was investigated by using a telescotpe, optical microscope, SEM and EPMA. Observations focused on elucidating the corrosion behavior of the cladding inner surface. A reaction layer was observed at burnups higher than about 35 GWd/t. The relationship between the thickness of the reaction layer and burnup was similar to that reported in the literature for conventional UO$$_{2}$$ fuel and other MOX fuels. The existence of a plutonium spot near the outer surface of the fuel pellet had no significant effect on the thickness of the reaction layer. A bonding layer was observed on the cladding inner surface. Its morphology and elemental distributions were not so different from those in BWR UO$$_{2}$$ fuel pins irradiated to high burnup, in which the fission gas release rate is high. In addition, the dependences of bonding layer formation on the burnup and linear heat rating were similar to results of UO2 fuel pins. It was, thus, suggested that the bonding layer formation mechanism was similar in both UO2 and MOX fuel pins.

JAEA Reports

Post Irradiation Examination for The FUGEN High Burn-up MOX Fuel Assembly (III) Final Report

Ikusawa,Yoshihisa; Kikuchi, Keiichi; Ozawa, Takayuki; Nakazawa, Hiroaki; Isozaki,Takao*; Nagayama, Masahiro*

JNC TN8410 2005-012, 113 Pages, 2005/08


The E09 fuel assembly was irradiated in the FUGEN from February 1990 to January 1997. The fuel assembly was the highest burn-up assembly in FUGEN and the pellet peak burn-up reached about 48 GWd/t. The E09 fuel assembly was transported to Japan Atomic Energy Research Institute (JAERI) Tokai in 2001. Post Irradiation Examinations (PIE) were started in July 2001, and all PIE items were completed by March 2005. The irradiation behavior of E09 MOX fuel was evaluated from the result of PIE. The major results are as follows; The integrity of E09 fuel assembly and fuel rods was confirmed. The corrosion behavior of ATR MOX fuel cladding was similar to that of LWR-UO2 fuel cladding. The central void was observed in outer ring samples irradiated with the maximum linear power over 45kW/m. A porous fine structure, similar to the rim structure seen in LWR-UO$$_{2}$$ pellet, was observed in the circumferential region of MOX pellet and around the plutonium-rich spots. The MOX fuel properties irradiated up to ~48 GWd/t, which are pellet swelling, thermal conductivity, pellet melting temperature and diffusivity of fission gas, were similar to LWR-UO$$_{2}$$ fuel properties. These results will be used for CANDU-OPTION program, which is one of Russian surplus weapon plutonium disposition programs with AECL in Canada, and available for LWR plutonium recycle program in Japan.

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