Transient response of LWR fuels (RIA)

Udagawa, Yutaka; Fuketa, Toyoshi*

Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08

https://doi.org/10.1016/B978-0-12-803581-8.11682-0

Transient response of LWR fuels (RIA)

Fuketa, Toyoshi

Comprehensive Nuclear Materials, 2, p.579 - 593, 2012/02

https://doi.org/10.1016/B978-0-08-056033-5.00044-6

Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi

Journal of Nuclear Materials, 414(2), p.303 - 308, 2011/07

https://doi.org/10.1016/j.jnucmat.2011.04.020

In order to estimate the behavior of high burnup mixed-oxide (MOX) fuel, it is important to evaluate fuel temperature accurately. The thermal conductivity formula of MOX fuel pellet which is needed to evaluate the fuel temperature was proposed. By using Klemens's theory and reported thermal conductivities of unirradiated (U, Pu)O and irradiated UO pellets, the thermal conductivity formula which contains the effects of burnup and plutonium (Pu) addition was obtained. Temperature of high burnup MOX fuel was evaluated based on the above-mentioned formula and the thermal conductivity integral method, and was compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it is considered that the proposed thermal conductivity formula of MOX pellets is adequate.

Behavior of high burnup LWR fuels during design-basis accidents; Key observations and an outline of the coming program

Fuketa, Toyoshi; Nagase, Fumihisa; Sugiyama, Tomoyuki; Amaya, Masaki

Proceedings of 2010 LWR Fuel Performance Meeting/TopFuel/WRFPM (CD-ROM), p.244 - 253, 2010/09

In order to evaluate adequacy of present safety criteria and safety margins and to provide a database for future regulation on higher burnup UO and MOX fuels, new cladding and pellets, an extensive program has been performed in the Japan Atomic Energy Agency (JAEA). The research program "Advanced LWR Fuel Performance and Safety" (ALPS) is comprised primarily of tests simulating a reactivity-initiated accident (RIA) and a loss-of-coolant accident (LOCA) on high burnup fuels shipped from European nuclear power plants. The paper describes key observations in the first phase (FY2002 to FY2007) and an outline of the second phase (FY2008 to FY2013) of the program.

Evaluation of initial temperature effect on transient fuel behavior under simulated reactivity-initiated accident conditions

Sugiyama, Tomoyuki; Udagawa, Yutaka; Fuketa, Toyoshi

Journal of Nuclear Science and Technology, 47(5), p.439 - 448, 2010/05

https://doi.org/10.3327/jnst.47.439

Relationship between changes in the crystal lattice strain and thermal conductivity of high burnup UO pellets

Amaya, Masaki; Nakamura, Jinichi; Fuketa, Toyoshi; Kosaka, Yuji*

Journal of Nuclear Materials, 396(1), p.32 - 42, 2010/01

https://doi.org/10.1016/j.jnucmat.2009.10.049

Two kinds of disk-shaped UO samples were irradiated in a test reactor up to about 60 and 130 GWd/t, respectively. The microstructures of the samples were investigated by means of optical microscopy, SEM/EPMA and micro-X-ray diffractometry. Thermal diffusivities of the irradiated samples were also measured and their thermal conductivities were evaluated. The thermal conductivity measurement results suggest that the amount of accumulated irradiation-induced defects depends on the irradiation condition of each sample. From the comparison of the changes in the lattice parameter and strain energy density before and after the thermal diffusivity measurements, it is likely that the thermal conductivity recovery in the temperature region from 1200 to 1500 K is related to the migration of dislocation.

Stress intensity factor at the tip of cladding incipient crack in RIA-simulating experiments for high burnup PWR fuels

Udagawa, Yutaka; Suzuki, Motoe; Sugiyama, Tomoyuki; Fuketa, Toyoshi

Journal of Nuclear Science and Technology, 46(10), p.1012 - 1021, 2009/10

https://doi.org/10.3327/jnst.46.1012

Numerical analysis and simulation of behavior of high burnup PWR fuel pulse-irradiated in reactivity-initiated accident conditions

Suzuki, Motoe; Sugiyama, Tomoyuki; Udagawa, Yutaka; Nagase, Fumihisa; Fuketa, Toyoshi

Proceedings of OECD/NEA Workshop on Nuclear Fuel Behaviour during Reactivity Initiated Accidents (CD-ROM), 11 Pages, 2009/09

The fuel behavior during the fast transients in the two cases of NSRR experiments using high burnup PWR fuel rods are analyzed by using the RANNS code. In one case, cladding failure occurred whereas in the other case the rod survived but gave rise to departure from nucleate boiling. The analytical results are compared with the metallographic observations of failed part of the cladding to discuss the failure-determining condition in terms of incipient crack depth, temperature and stress at the depth. Based on these evaluations, sensitivity study with respect to the effect of initial temperature on the stress/strain of cladding is conducted. In addition, simulations are performed in commercial reactor conditions. The results were compared with each other and failure capability of cladding is discussed.

Microstructure and mechanical property changes in fuel cladding during RIA-type temperature transients

Nagase, Fumihisa; Sugiyama, Tomoyuki; Fuketa, Toyoshi

Proceedings of OECD/NEA Workshop on Nuclear Fuel Behaviour during Reactivity Initiated Accidents (CD-ROM), 8 Pages, 2009/09

Laboratory-scale experiments are performed to examine microstructure and mechanical property changes of the fuel cladding during temperature transients and estimate the fuel behavior for a wide temperature range under various RIA conditions. Ductility of the cladding generally increases with the temperature. The ductility increase of the hydrided cladding is obvious between 300 and 473 K, while it is small between 473 and 573 K. Ductility increase due to recrystallization is not expected during the RIA-type quick transient. On the other hand, phase transitions above 1100 K occur so quickly and affect the mechanical property of the cladding. The influence of the hydride rim at the cladding periphery would not be seen above 870 K since hydrides in the rim are in solid solution for a very short time.

Comparative analysis on behavior of high burnup PWR fuels pulse-irradiated in reactivity-initiated accident conditions

Suzuki, Motoe; Sugiyama, Tomoyuki; Udagawa, Yutaka; Nagase, Fumihisa; Fuketa, Toyoshi

Proceedings of Top Fuel 2009 (DVD-ROM), p.473 - 479, 2009/09

Two cases of the Reactivity-Initiated-Accident-simulating pulse irradiation experiments conducted in the NSRR using a high burnup PWR fuel are analyzed by using the RANNS code. One case was performed in a stagnant coolant water of room temperature and at atmospheric pressure, and the other in the stagnant coolant of high temperature at 7 MPa pressure. The two cases resulted in failure. Behavior in a fast transient of the fuels is investigated by comparing the two experiments. Specifically, the temperature of rod, pellet-clad mechanical interaction, and stress/strain of cladding etc. are calculated and the results are discussed in comparison with the experimental observations. The metallography of cladding shows that the cladding failed by macroscopic shear gliding from a tip of a crack. The calculated plastic strains of claddings at failure reasonably coincide with the observed local strains if the latter strains are averaged over the total circumference.