Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi*
Nuclear Technology, 208(3), p.484 - 493, 2022/03
An Ag-In-Cd control rod alloy was heated in argon or oxygen at 1073-1673 K for 60-3600 s and the release behavior of the elements was examined. Complete liquefaction of the alloy occurred between 1123 and 1173 K, and elemental release was quite limited below the liquefaction temperature. In argon, almost all of the Cd content was released within 3600 s at 1173 K and within 60 s at 1573 K, while the released fractions of Ag and In were 3% and 8%, respectively. In oxygen, the release of Cd, which was quite small at temperatures up to 1573 K, drastically increased to 30-50% at 1673 K for short periods. Releases of Ag and In were also small in oxygen under the examined conditions. Comparison with the experimental data suggests that conventional empirical release models may underestimate the Cd release at lower temperatures just after control rod failure in severe accidents.
Nagase, Fumihisa; Uetsuka, Hiroshi
Journal of Nuclear Science and Technology, 49(1), p.96 - 102, 2012/01
Thermal properties of molten and mixed core materials are required to be known for effective analysis of core damage in severe accidents at nuclear power plants. The specific heat capacity, thermal expansion, thermal diffusivity, and melting temperature were measured or estimated on the core debris samples of the Three Mile Island Unit 2 (TMI-2) reactor and simulated debris (SIMDEBRIS), which had chemical composition and porosity similar to the TMI-2 debris. The thermal diffusivity of the TMI-2 debris, which is mainly composed of (U,Zr)O, was as low as 10% to 25% of UO at room temperature but was comparable above 1500 K. The melting temperature of SIMDEBRIS was about 2840 K, which is equivalent to the liquidus temperature of (U,Zr)O with the same ZrO/UO ratio. Accordingly, other core materials less than 10% in weight were observed to have no influence on the melting temperature.
Ichimura, Toshio; Uetsuka, Hiroshi
Denki Hyoron, 92(2), p.68 - 89, 2007/02
Japan Atomic Energy Agency (JAEA) was established as the result of the integration of Japan Atomic Energy Research Institute (JAERI) and Japan Nuclear Cycle Development Institute (JNC) on Oct. 1, 2005. This paper introduces the current status of research and development at JAEA.
Ichimura, Toshio; Uetsuka, Hiroshi
Denki Hyoron, 91(2), p.62 - 80, 2006/02
no abstracts in English
Suzuki, Motoe; Uetsuka, Hiroshi; Saito, Hiroaki*
Nuclear Engineering and Design, 229(1), p.1 - 14, 2004/04
Mechanical load on cladding induced by fuel swelling in a high burn-up BWR type rod has been analyzed by a fuel performance code FEMAXI-6. The code has been developed for the analysis of LWR fuel rod behaviors in normal operation and transient conditions using FEM. During a power ramp for the high burn-up rod, instantaneous pellet swelling can significantly exceed the level that is predicted by a "steady-rate" swelling model, causing a large circumferential strain in cladding. This phenomenon has been simulated by a new swelling model to take into account the fission gas bubble growth, and as a result it has been found that the new model can give reasonable predictions on cladding diameter expansion in comparison with post-irradiation data. In addition, a pellet-clad bonding model which has been incorporated in the code to assume firm mechanical coupling between pellet outer surface and cladding inner surface has predicted the generation of bi-axial stress state in the cladding during ramp.
Nakamura, Takehiko; Sasajima, Hideo; Yamashita, Toshiyuki; Uetsuka, Hiroshi
Journal of Nuclear Materials, 319, p.95 - 101, 2003/06
Pulse irradiation tests under simulated RIA conditions were performed with three types of ROX fuels. Single phase YSZ, homogeneous mixture of YSZ/spinel and YSZ particle dispersed in spinel type ROX fuels were pulse irradiated in the Nuclear Safety Research Reactor (NSRR). Mode and threshold of the fuel failure including its consequences were investigated under the RIA conditions. The fuel failure occurred in a burst type mode in all the three types of ROX fuel tests with considerable fuel melting. Even though the mode was quite different from those of UO fuel, failure threshold enthalpies of the ROX fuels were close to that of UO fuel at about 10 GJ m. The consequence of the failure of the ROX fuels was different from the one of UO fuel, because molten fuel dispersal occurred at lower enthalpies in the ROX fuel tests. Change of the fuel structure and material interaction in the transient heating conditions were examined through optical and secondary electron microscopy, and electron probe micro analysis.
Nakamura, Takehiko; Nakamura, Jinichi; Sasajima, Hideo; Uetsuka, Hiroshi
Journal of Nuclear Science and Technology, 40(5), p.325 - 333, 2003/05
In order to examine high burnup fuel performance and to confirm its integrity under unstable power oscillation conditions arising during an ATWS in BWRs, two tests of irradiated fuels under simulated power oscillation conditions were conducted in the NSRR. Irradiated fuels at burnups of 25 and 56GWd/tU were subjected to four to seven power oscillations, which peaked at 50 to 95kW/m with intervals of 2s. The power oscillation was simulated by quick withdrawal and insertion of six regulating rods of the NSRR with a computerized control. Deformation of the fuel cladding of the test rods was comparable to those observed in shorter transient tests, which simulated RIAs, at the same fuel enthalpy level up to 368J/g. The fuel deformation was mainly caused by PCMI and was roughly proportional to the fuel enthalpy. Enhanced cladding deformation due to ratcheting by the cyclic load was not observed. Fission gas release, on the other hand, was considerably smaller than in the RIA tests, suggesting different release mechanisms in the two types of transients.
Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi
Journal of Nuclear Science and Technology, 40(4), p.213 - 219, 2003/04
Isothermal oxidation tests in flowing steam were performed on low-Sn Zircaloy-4 cladding tubes over the wide temperature range from 773 to 1573 K in order to obtain oxidation kinetics applicable to various loss-of-coolant accident conditions of LWRs. The oxidation generally obeys a parabolic rate law for the examined time range up to 3600s at temperatures from 1273 K to 1573K, and for a limited time range up to 900s from 773 to 1253 K. A cubic rate law is preferable for evaluating the longer-term oxidation at 1253 K and below. The parabolic rate law constant and the cubic rate law constant for measured weight gain were evaluated at every examined temperature, and Arrhenius-type equations were determined in order to describe the temperature dependence of the rate constants. It was indicated that the change of the oxidation kinetics from the cubic to the parabolic rate and the discontinuities in the temperature dependence of the rate constants are caused by the monoclinic/tetragonal phase transformation of ZrO.
Hidaka, Akihide; Kudo, Tamotsu; Nakamura, Takehiko; Kanazawa, Toru; Kiuchi, Toshio; Uetsuka, Hiroshi
JAERI-Tech 2003-009, 30 Pages, 2003/03
The VEGA (Verification Experiments of radionuclides Gas/Aerosol release) program is being performed at JAERI to clarify mechanisms of radionuclides release from irradiated fuel during severe accidents and to improve source term predictability. The fifth VEGA-5 test was conducted in January 2002 to confirm the reproducibility of decrease in cesium release under elevated pressure that was observed in the VEGA-2 test and to investigate the release behavior of short-life radionuclides. The PWR fuel of 47GWd/tU after 8.2 years of cooling was re-irradiated at Nuclear Safety Research Reactor (NSRR) for 8 hours before the heat-up test. After that, the two pellets of 10.9g without cladding were heated up to about 2,900K at 1.0MPa under the inert He condition. The experiment reconfirmed the decrease in cesium release under elevated pressure. The release data on short-life radionuclides such as Ru-103 and Ba-140 that has never been observed in the previous VEGA tests without re-irradiation was obtained using the gamma ray measurement.
Suzuki, Motoe; Uetsuka, Hiroshi
IAEA-TECDOC-CD-1345 (CD-ROM), p.217 - 238, 2003/03
A fuel performance code FEMAXI-6 has been developed for the analysis of LWR fuel rod behaviors. The code uses FEM analysis, and has incorporated thermal and mechanical models of phenomena anticipated in high burn-up fuel rods. In the present study, PCMI induced by swelling in a high burn-up BWR type fuel rod has been analyzed. During a power ramp for the high burn-up fuel, instantaneous pellet swelling been simulated by a new swelling model which has been installed in the code to take into account the FP gas bubble growth, and the new model can give satisfactory predictions on cladding diametral expansion. In addition, a pellet-clad bonding model in the code, which assumes firm mechanical coupling between pellet outer surface and cladding inner surface, predicts an increased tensile stress in the axial direction of cladding during the power ramp, indicating the generation of bi-axial stress state in the cladding.
Sasajima, Hideo; Sugiyama, Tomoyuki; Nakamura, Takehiko; Fuketa, Toyoshi; Uetsuka, Hiroshi
Proceedings of 7th International Topical Meeting on Research Reactor Fuel Management (ENS RRFM2003), p.109 - 113, 2003/03
Uranium-zirconium hydride (U-ZrHx) fuel has been widely utilized in the world as TRIGA reactor fuel. In order to obtain the fuel performance data under accident conditions and to enhance accountability of the safety assessment of the reactors using the fuel, irradiation tests under power burst type accident conditions have been conducted in the NSRR. Five pulse irradiation tests have been performed at peak fuel enthalpies ranging from 187 J/g to 483 J/g. Cladding surface temperature increased rapidly at the pulse and DNB occurred in peak fuel enthalpy over 187 J/g in the tests. The DNB occurred at lower fuel enthalpy in the U-ZrH1.6 fuel than in the UO fuel rods. The rod internal pressure rose up to as high as 1MPa in the transient heating tests, suggesting considerable release of the hydrogen decomposed from the fuel. The peak pressure was lower than equilibrium hydrogen pressure of 1.5MPa at the peak temperature, suggesting the transient effect. Considerable PCMI was observed in the tests, through cladding elongation up to 3.3 mm synchronized to the pellet stack deformation.
Kusagaya, Kazuyuki*; Sugiyama, Tomoyuki; Nakamura, Takehiko; Uetsuka, Hiroshi
JAERI-Tech 2002-105, 24 Pages, 2003/01
High-temperature and high-pressure influence on the destructive force at the fuel rod failure in reactivity-initiated-accident (RIA) simulating experiment using the NSRR (Nuclear Safety Research Reactor) is estimated, for the purpose of mechanical designing of a new experimental capsule for simulating the temperature and pressure condition of typical commercial BWR. When knowledge on pressure impulse and water hammer, which are the cause of the destructive force, and steam property dependence on temperature and pressure are taken into account, one can qualitatively estimate that the destructive force in the BWR operation condition is smaller than that in the room temperature and atmospheric pressure condition. The water column velocity, which determines the impact by water hammer, is further investigated quantitatively by modeling the experimental system and the water hammer phenomenon. As a result, the maximum velocity of water column in the BWR operation condition is calculated to be only about 10% of that in the room temperature and atmospheric pressure condition.
Nakamura, Takehiko; Kusagaya, Kazuyuki*; Sasajima, Hideo; Yamashita, Toshiyuki; Uetsuka, Hiroshi
Journal of Nuclear Science and Technology, 40(1), p.30 - 38, 2003/01
Pulse irradiation tests of three types of ROX fuel, i.e. YSZ single phase, finely mixed two phase composite of YSZ and spinel, and the other composite of larger YSZ particles dispersed in spinel matrix, were conducted in the NSRR to investigate their behavior under RIA conditions. Owing to their lower melting temperatures than that of UO fuel, melting of ROX fuel occurred while the cladding was still solid and intact in the accident conditions. Therefore, consequence of the ROX fuel failure was quite different from that of UO fuel. When the ROX fuels failed, a considerable amount of the molten fuel was released out to the surrounding coolant water. In spite of the release, no significant mechanical energy generation due to fuel/coolant thermal interaction was observed in the tested enthalpy range below 12 GJ/m. In terms of the failure threshold, on the other hand, the ROX fuels failed at fuel volumetric enthalpies above 10 GJ/m, which was comparable to that of UO fuel. The results highlighted controlling parameters on the fuel behavior under the RIA conditions.
JAERI-Review 2002-027, 147 Pages, 2002/11
The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the NSRR, the JMTR, the JRR-3 and the Reactor Fuel Examination Facility of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The research conducted in the year 2001 produced many important data and information. They are, for example, the fuel behavior data under BWR power oscillation conditions in the NSRR, the data on failure-bearing capability of hydrided cladding under LOCA conditions and the FP release data at very high temperature in steam which simulate the reactor core condition during severe accidents.This report summarizes the outline of research activities and major outcomes of the research executed in 2001 in the Fuel Safety Research Laboratory.
Nagase, Fumihisa; Uetsuka, Hiroshi
JAERI-Research 2002-023, 23 Pages, 2002/11
To obtain basic data to evaluate fuel rod integrity during abnormal transient and accident of LWRs, high burnup PWR fuel claddings were heated for 0 to 600s at temperatures of 673 through 1173K, and the mechanical property changes were examined by using ring tensile test at room temperature. As a result of the test, it was shown that strength and ductility of the cladding are changed depending on heating temperature and time. The mechanical property changes by temperature transients are considered to be correspondent mainly to recovery of irradiation defect, recovery and recrystallization of the Zircaloy, phase transformations, and associated change of the hydride distribution and morphology. Comparison with unirradiated claddings suggested that irradiation effects are not completely annealed out by the short-term annealing at high temepratures. Radial change of hydrogen concentration was measured for the high burnup PWR fuel cladding and very high hydrogen concentration of about 2400wtppm was detected at the cladding periphery.
Nagase, Fumihisa; Tanimoto, Masataka*; Uetsuka, Hiroshi
IAEA-TECDOC-1320, p.270 - 278, 2002/11
With a view to obtaining basic data for evaluating high burnup fuel behavior under LOCA conditions, a systematic research program is being conducted at JAERI. High-temperature oxidation tests with non-irradiated cladding have been performed to investigate separate effects of pre-oxidation and pre-hydriding on the oxidation kinetics. "Integral thermal shock tests" have been conducted simulating a LOCA condition to examine the influence of pre-hydriding on failure-bearing capability of oxidized cladding upon quenching. Test results showed almost no influence of absorbed hydrogen on the threshold value for oxidation amount under no axial restraint condition. On the other hand, it was shown that the threshold value is reduced by absorbed hydrogen for the restraint condition.
Nakamura, Jinichi; Nakamura, Takehiko; Sasajima, Hideo; Suzuki, Motoe; Uetsuka, Hiroshi
HPR-359, Vol.2, p.34_1 - 34_16, 2002/09
In BWR, power oscillations can occur due to the void fraction fluctuation. To investigate the fuel behavior during power oscillation of BWRs, two types of irradiated fuel rods were tested under simulated power oscillation conditions in the Nuclear Safety Research Reactor(NSRR). One is high burnup BWR fuel (56GWd/t) test, with 4 power oscillation cycles, to clarify the behavior of high burnup fuel. The second one is high enriched fuel(20%,25GWd/t) test, with 7 power cycles, to perform the test under high power conditions. The fuel behavior data, such as cladding elongation, fuel stack elongation, cladding temperature, etc. were obtained in these tests. The DNB did not occur in these tests. The PCI was observed through cladding elongation and fuel stack elongation during the power oscillations, but the residual strain of cladding was very small. Fuel behavior under simulated power oscillations is discussed based on in-pile data and PIE data and is compared with FEMAXI-6 and FRAP-T6 calculation.
Hidaka, Akihide; Kudo, Tamotsu; Nakamura, Takehiko; Uetsuka, Hiroshi
Journal of Nuclear Science and Technology, 39(7), p.759 - 770, 2002/07
no abstracts in English
Nakamura, Takehiko; Kusagaya, Kazuyuki*; Fuketa, Toyoshi; Uetsuka, Hiroshi
Nuclear Technology, 138(3), p.246 - 259, 2002/06
no abstracts in English
Nagase, Fumihisa; Uetsuka, Hiroshi
NUREG/CP-0176, p.335 - 342, 2002/05
no abstracts in English