Nippon Nuclear Fuel Development

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Sakamoto K.,Nippon Nuclear Fuel Development | Une K.,Nippon Nuclear Fuel Development | Aomi M.,Global Nuclear Fuel Japan | Hashizume K.,Kyushu University
Progress in Nuclear Energy | Year: 2012

To understand the basic oxidation kinetics of alloying elements which is considered to be strongly related with the corrosion and hydrogen pickup, the depth profiles of chemical states of alloying elements (Cr and Fe) were measured in the oxide layer of Zr-0.5Sn-1.0Cr-0.5Fe alloys. The depth profiles were obtained by combinations of a surface-sensitive XANES and an extremely low energy Ar ion sputtering. The XANES measurements revealed that the chemical states of alloying elements (Fe and Cr) varied with the depth in the oxide layer. Especially in the oxide layer formed in steam, a decrease of the fractions of oxidation states was significant rather than that in LiOH solution. In the oxide layer formed in steam, the oxidation rate of chromium was faster than iron by a factor of approximately 2. © 2011 Elsevier Ltd. All rights reserved.


Kubo T.,Nippon Nuclear Fuel Development | Kobayashi Y.,M.O.X. Co. | Uchikoshi H.,Nippon Nuclear Fuel Development
Journal of Nuclear Materials | Year: 2012

Delayed hydride cracking (DHC) tests of Zircaloy-2 cladding tubes were performed in the chamber of a scanning electron microscope (SEM) to directly observe the crack propagation and measure the crack velocity in the radial direction of the tubes. Pre-cracks were produced at the outer surfaces of the tubes. Hydrogen contents of the tubes were from 90 ppm to 130 ppm and test temperatures were from 225°C to 300°C. The crack velocity followed the Arrhenius law at temperatures lower than about 270°C with apparent activation energy of about 49 kJ/mol. The upper temperature limit for DHC, above which DHC did not occur, was about 280°C. The threshold stress intensity factor for the initiation of the crack propagation, K IH, was from about 4 MPa m 1/2 to 6 MPa m 1/2, almost independent of temperature. An increase in 0.2% offset yield stress of the material accelerated the crack velocity and slightly decreased K IH. Detailed observations of crack tip movement showed that cracks propagated in an intermittent fashion and the propagation gradually approached the steady state as the crack depth increased. The SEM observations also showed that hydrides were formed at a crack tip and a number of micro-cracks were found in the hydrides. It was presumed from these observations that the repetition of precipitation and fracture of hydrides at the crack tip would be responsible for the DHC propagation. © 2012 Elsevier B.V. All rights reserved.


Turnbull J.A.,United Kingdom | Yagnik S.K.,EPRI | Hirai M.,Nippon Nuclear Fuel Development | Staicu D.M.,Institute for Transuranium Elements
Nuclear Science and Engineering | Year: 2015

To investigate the potential disintegration to powder of high-burnup fuel pellets during a rapid temperature transient, the Nuclear Fuels Industry Research (NFIR) Program commissioned two independent scoping studies. The first investigated the effect of hydrostatic restraint pressure on fission gas release during a series of fast temperature ramps. In the second study laser heating was used to investigate the temperature at which small samples of fuel fragmented. From the observations made in these studies, local burnup and temperature thresholds of 71 MWd/kg HM and 645° C were identified for fuel pulverization during a loss-of- coolant accident (LOCA). It is shown that fine fragment production in integral LOCA tests performed in other independent investigations at Studsvik and Halden was generally well predicted using these thresholds of burnup and temperature. The NFIR investigations also reveal that the degree of pulverization and resulting fragment size are dependent on the temperature ramp rate. Moreover, they confirm that pulverization can be substantially reduced by the imposition of hydrostatic pressure.


Kubo T.,Nippon Nuclear Fuel Development | Kobayashi Y.,M.O.X. Co.
Journal of Nuclear Materials | Year: 2013

Delayed hydride cracking (DHC) of Zircaloy-2 is one possible mechanism for the failure of boiling water reactor fuel rods in ramp tests at high burnup. Analyses were made for hydrogen diffusion around a crack tip to estimate the crack velocity of DHC in zirconium alloys, placing importance on effects of precipitation of δ-hydride. The stress distribution around the crack tip is significantly altered by precipitation of hydride, which was strictly analyzed using a finite element computer code. Then, stress-driven hydrogen diffusion under the altered stress distribution was analyzed by a differential method. Overlapping of external stress and hydride precipitation at a crack tip induces two stress peaks; one at a crack tip and the other at the front end of the hydride precipitate. Since the latter is larger than the former, more hydrogen diffuses to the front end of the hydride precipitate, thereby accelerating hydride growth compared with that in the absence of the hydride. These results indicated that, after hydride was formed in front of the crack tip, it grew almost steadily accompanying the interaction of hydrogen diffusion, hydride growth and the stress alteration by hydride precipitation. Finally, crack velocity was estimated from the calculated hydrogen flux into the crack tip as a function of temperature, stress intensity factor and material strength. There was qualitatively good agreement between calculation results and experimental data. © 2013 Published by Elsevier B.V.


Kubo T.,Nippon Nuclear Fuel Development | Kobayashi Y.,M.O.X. Co. | Uchikoshi H.,Nippon Nuclear Fuel Development
Journal of Nuclear Materials | Year: 2013

The fracture strength of δ-zirconium hydrides embedded in a zirconium matrix was determined at temperatures between 25 °C and 250 °C by ring tensile tests using Zircaloy-2 tubes. Essentially all of the present hydrides in the tubes were re-oriented in the radial direction by a temperature cycling treatment and then tensile stress was applied perpendicular to the hydrides to ensure that brittle fracture would occur at the hydrides. The hydrides failed in a brittle manner below 100 °C where-as the zirconium matrix itself underwent ductile fracture without hydride cracking at temperatures above 200 °C under plane stress condition. Brittle fracture of the hydrides continued to occur at temperatures up to 250 °C under plane strain condition, suggesting that the upper limit temperature for hydride fracture, T upper, was raised by the triaxial stress state under the plane strain condition. The apparent fracture strength of the hydrides, σhydridef, was determined at temperatures below Tupper from the measured fracture strength of the tubes, making a correction for the compressive transformation stress in the hydrides. σhydridef was about 710 MPa at temperatures between 25 °C and 250 °C at both plane stress and plane strain conditions. The temperature dependency was very small in this temperature range. Tupper was almost equivalent to the cross-over temperature between σhydridef and the ultimate tensile strength (UTS), which suggests that, at temperatures above Tupper, the zirconium matrix would undergo ductile fracture before the stress in the hydride is raised above σhydridef, since UTS is smaller than σhydridef. © 2013 Elsevier B.V. All rights reserved.


Ishiyama Y.,Nippon Nuclear Fuel Development | Rogge R.B.,National Research Council Canada | Obata M.,Toshiba Corporation
Journal of Nuclear Materials | Year: 2011

Weld beads on plate specimens made of type 316L stainless steel were neutron-irradiated up to about 2.5 × 1025 n/m2 (E > 1 MeV) at 561 K in the Japan Material Testing Reactor (JMTR). Residual stresses of the specimens were measured by the neutron diffraction method, and the radiation-induced stress relaxation was evaluated. The values of σx residual stress (transverse to the weld bead) and σy residual stress (longitudinal to the weld bead) decreased with increasing neutron dose. The tendency of the stress relaxation was almost the same as previously published data, which were obtained for type 304 stainless steel. From this result, it was considered that there was no steel type dependence on radiation-induced stress relaxation. The neutron irradiation dose dependence of the stress relaxation was examined using an equation derived from the irradiation creep equation. The coefficient of the stress relaxation equation was obtained, and the value was 1.4 (×10-6/MPa/dpa). This value was smaller than that of nickel alloy. © 2010 Elsevier B.V. All rights reserved.


Takagi I.,Kyoto University | Une K.,Nippon Nuclear Fuel Development | Miyamura S.,Kyoto University | Kobayashi T.,Kyoto University
Journal of Nuclear Materials | Year: 2011

In situ diffusion experiments of the hydrogen isotope deuterium in the oxide layer formed on zirconium alloys were carried out to clarify the hydrogen diffusion mechanism in the layer. The experiments were done in deuterium plasma for the temperature range from 523 to 673 K by using a nuclear reaction analysis for D(3He,p)4He. The zirconium alloys used were GNF-Ziron (high iron Zircaloy-2 type alloy) and VB (high iron and chromium alloy), which had been corroded in 673 K H2O- or D2O-steam for 10-15 days. The oxide thickness ranged from1.4 to 1.7 μm of pre-transition condition. The results showed that the steam oxides had a double-layer structure composed of the outside non-protective oxide with faster diffusivity and the inside barrier layer with slower diffusivity. The barrier layer thickness was about 0.8-0.9 μm and unchanged for the two alloys. For the in situ deuterium plasma diffusion experiments, the diffusion coefficient of deuterium in the barrier layer of GNF-Ziron was given as, D (cm2/s) = 4.5 × 10-13exp (-17,000/RT). The diffusion coefficient in the VB oxide at 573 K was approximately half of that in the GNF-Ziron oxide. This factor for the diffusivity was roughly consistent with their hydrogen absorption performance. For the deuterium release experiments in a vacuum subsequent to the in situ deuterium diffusion experiments, the diffusion in the barrier oxide layer was further retarded, suggesting lower diffusivity than for the case of the in situ deuterium plasma atmosphere. © 2011 Elsevier B.V. All rights reserved.


Muta H.,Osaka University | Etoh Y.,Nippon Nuclear Fuel Development | Ohishi Y.,Osaka University | Kurosaki K.,Osaka University | Yamanaka S.,Osaka University
Journal of Nuclear Science and Technology | Year: 2012

Hydrogen diffusion in monoclinic and tetragonal zirconium oxides has been studied by electronic state calculations. In both structures, the optimized hydrogen site lies near the center of a distorted fluorite structure. The activation energy was calculated to be 120-200 kJ/mol, which is similar to experimentally measured values. The effects of compressive stress, alloying elements, and oxygen defects are considered individually. Compressive stress reduces the hydrogen diffusion coefficient by 40%/GPa. Oxygen defects and substituted Fe and Cr are thought to act as trapping sites for hydrogen, which probably reduces hydrogen diffusion in zirconium oxide. © 2012 Atomic Energy Society of Japan.


Fukuya K.,Japan Institute of Nuclear Safety System | Nishioka H.,Japan Institute of Nuclear Safety System | Fujii K.,Japan Institute of Nuclear Safety System | Miura T.,Japan Institute of Nuclear Safety System | Kitsunai Y.,Nippon Nuclear Fuel Development
Journal of Nuclear Materials | Year: 2013

Local strain near grain boundaries under tensile stress was examined using the EBSD technique for cold-worked SUS316 stainless steel irradiated to 73 dpa. Distribution of misorientation in the same areas was analyzed while macroscopically deforming the specimen at an elastic strain level of 0.03% and a plastic strain level of 3%. A clear increase in local strain was detected within 4 μm from the grain boundaries at the 3% plastic strain. It was confirmed that high local strain was produced at the 0.03% elastic strain near the grain boundaries which exhibited higher misorientations at the plastic strain. Detailed analysis in areas within 1 μm from individual grain boundaries also revealed that local strain was increased at some boundaries at the 0.03% elastic strain. © 2012 Elsevier B.V. All rights reserved.


Fukuya K.,Japan Institute of Nuclear Safety System | Nishioka H.,Japan Institute of Nuclear Safety System | Fujii K.,Japan Institute of Nuclear Safety System | Miura T.,Japan Institute of Nuclear Safety System | Torimaru T.,Nippon Nuclear Fuel Development
Journal of Nuclear Materials | Year: 2011

Local deformation behavior in cold worked SUS316 stainless steels irradiated to 73 dpa at 573 K was examined by SEM and EBSD after deformation to plastic strain of ∼2% at 593 K at a strain rate of 7.8 × 10 -8/s in an argon gas environment. Grain boundary separation occurred at random grain boundaries which lay almost normal to the tensile direction and had coarse dislocation channels in grains on one side. Misorientation maps indicated that a local high strain field existed near grain boundaries when coarse channels impinged on grain boundaries forming dislocation pileups. The grain boundary separation occurred more frequently at the grain boundaries where the difference in Schmidt factor was larger between two adjacent grains. These results indicated that the grain boundary separation was triggered by the local high stress field induced by dislocation pileups and high normal component of tensile stress. © 2011 Elsevier B.V. All rights reserved.

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