Release behaviors of elements from an Ag-In-Cd control rod alloy at temperatures up to 1673 K

Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi*

Nuclear Technology, 208(3), p.484 - 493, 2022/03

https://doi.org/10.1080/00295450.2021.1905472

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.

Current status of research and development at Japan Atomic Energy Agency

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.

Analysis of mechanical load on cladding induced by fuel swelling during power ramp in high burn-up rod by fuel performance code FEMAXI-6

Suzuki, Motoe; Uetsuka, Hiroshi; Saito, Hiroaki*

Nuclear Engineering and Design, 229(1), p.1 - 14, 2004/04

https://doi.org/10.1016/j.nucengdes.2003.11.017

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.

Irradiated fuel behavior under power oscillation conditions

Nakamura, Takehiko; Nakamura, Jinichi; Sasajima, Hideo; Uetsuka, Hiroshi

Journal of Nuclear Science and Technology, 40(5), p.325 - 333, 2003/05

https://doi.org/10.3327/jnst.40.325

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.

Radionuclides release from re-irradiated fuel under high temperature and pressure conditions; -ray measurements of VEGA-5 test

Hidaka, Akihide; Kudo, Tamotsu; Nakamura, Takehiko; Kanazawa, Toru; Kiuchi, Toshio; Uetsuka, Hiroshi

JAERI-Tech 2003-009, 30 Pages, 2003/03

JAERI-Tech-2003-009.pdf:1.73MB

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.

Behavior of uranium-zirconium hydride fuel under reactivity initiated accident conditions

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.

Behavior of YSZ based rock-like oxide fuels under simulated RIA conditions

Nakamura, Takehiko; Kusagaya, Kazuyuki*; Sasajima, Hideo; Yamashita, Toshiyuki; Uetsuka, Hiroshi

Journal of Nuclear Science and Technology, 40(1), p.30 - 38, 2003/01

https://doi.org/10.3327/jnst.40.30

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.

Mechanical properties changes of high burnup PWR fuel cladding by temperature transient

Nagase, Fumihisa; Uetsuka, Hiroshi

JAERI-Research 2002-023, 23 Pages, 2002/11

JAERI-Research-2002-023.pdf:1.94MB

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.

NSRR tests under simulated power oscillation conditions of BWRs

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.

High burnup BWR fuel behavior under simulated reactivity initiated accident conditions

Nakamura, Takehiko; Kusagaya, Kazuyuki*; Fuketa, Toyoshi; Uetsuka, Hiroshi

Nuclear Technology, 138(3), p.246 - 259, 2002/06

https://doi.org/10.13182/NT02-A3292

no abstracts in English