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Oral presentation

Modelling of Cs chemisorption behaviour under LWR severe accident

Nishioka, Shunichiro; Nakajima, Kunihisa; Suzuki, Eriko; Miradji, F.; Osaka, Masahiko

no journal, , 

In order to contribute to Cs distribution analysis in Fukushima Daiichi Nuclear Power Station reactors by using severe accident analysis code, the influence of chemical factors (temperature, atmosphere etc.) on the Cs chemisorption behaviour onto stainless steel was investigated experimentally. An improved Cs chemisorption model was then established considering the effect of chemical factors obtained by the experiment. The accuracy of the present model is greatly improved by 1 order of magnitude compared with that incorporated in the existing SA codes.

Oral presentation

Ultra-high temperature creep and transient burst strength of ODS steel cladding tube

Yano, Yasuhide; Sekio, Yoshihiro; Kato, Shoichi; Tanno, Takashi; Inoue, Toshihiko; Oka, Hiroshi; Otsuka, Satoshi; Furukawa, Tomohiro; Uwaba, Tomoyuki; Kaito, Takeji; et al.

no journal, , 

ODS steels have been noticed as a prospective candidate for long-life fuel claddings of FR due to their high temperature strength and radiation resistance. It is necessary to acquire data on tensile strength and creep rupture resistance of core materials at ultra-high temperature for use in safety design. The authors reported some data of tensile properties of ODS steel cladding at ultra-high temperatures, however, there are few data on creep rupture strength. In this study, ultra-high temperature creep rupture strength and transient burst properties with wide range of heating rates have been investigated for ODS claddings. Internal creep rupture tests for 9Cr-ODS claddings were carried out at temperatures between 650 and 1000$$^{circ}$$C, and ring creep tests were performed at 1000$$^{circ}$$C. The temperature-transient-to-burst tests were performed on both steel cladding tubes. Creep-rupture curves for 9Cr-ODS claddings were linear shape as a function of long rupture time, although it is well known that those curves for conventional martensitic steels have a sigmoidal shape to long rupture times. The failure temperatures for 9Cr-ODS steels decreased with decreasing the heating rate in a manner to similar to PNC-FMS claddins. However, transient burst strengths for 9Cr-ODS were much higher than those for PNC-FMS at all conditions. Using these data, discussions were carried out on a technique for evaluating rupture life of ODS steel cladding after various load-time-temperature histories.

Oral presentation

Sintering experiments of Dy$$_{0.3}$$Zr$$_{0.7}$$N under a variety of milling conditions

Takaki, Seiya; Harada, Makoto; Takano, Masahide

no journal, , 

It is necessary to control the pellet densities for the purpose of securing margin against swelling for nitride fuel for transmutation of minor actinide (MA). Appropriate polymer particles will be applied as a pore former in order to decrease the density. This study aimed to investigate the influence of various milling parameters on the densities of sintered Dy$$_{0.3}$$Zr$$_{0.7}$$N solid solution as surrogate nitride fuel in order to obtain fundamental knowledge for controlling sintered density with pore former. The sintered pellet densities are studied under various milling time (from 10 to 150 min) and employing either tungsten carbide (WC) or silicon nitride (Si$$_{3}$$N$$_{4}$$) milling cup and balls. The measurement of specific surface area clarifies that finer powder can be obtained with WC. However, the achieved densities of the sintered pellets, as a function of specific surface area of the milled powder, show that the denser pellets can be obtained with Si$$_{3}$$N$$_{4}$$ in spite of the smaller specific surface area. These results suggest that the distortion in the powder particles influences the behaviour of grain growth during the sintering.

Oral presentation

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

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

no journal, , 

Oral presentation

Cesium migration effects on irradiation behavior of fast reactor MOX fuel pins

Tanno, Takashi; Oka, Hiroshi; Ikusawa, Yoshihisa; Uwaba, Tomoyuki; Otsuka, Satoshi; Kaito, Takeji; Maeda, Seiichiro

no journal, , 

Cesium (Cs), which is a volatile fission product (FP), migrates axially from hot core region to cold top and/or bottom ones in fast reactor (FR) fuel pins. The Cs accumulated in cold region such as axial UO$$_{2}$$ blanket fuel can form Cs-U-O compounds having lower density than that of the fuel pellet, causing fuel cladding mechanical interaction (FCMI). The severe FCMI would arouse concern about the integrity of fuel pins. This work aims to understand Cs axial migration behavior and its effect on FR fuel pins. Two MOX fuel pins irradiated in EBR-II were evaluated the Cs migration behavior. Pin diameter measurement and gamma-scanning were carried out, and calculations with the ORIGEN-2 code was also done to estimate FP inventory with burnup in the pins. It was found from the comparison between the calculated pellet swelling by the Cs-U-O compounds and the measured pin diameter increase that the localized pin diameter increases at the MOX fuel-blanket interfaces were due to the FCMI caused by the pellet swelling associated with the formation of the Cs-U-O compounds.

Oral presentation

The Analytical prediction of inventories and physicochemical composition of spallation products produced in Lead-Bismuth Eutectic of Accelerator Driven System

Miyahara, Shinya*; Arita, Yuji*; Ohdaira, Naoya*; Sasa, Toshinobu; Maekawa, Fujio; Matsuda, Hiroki

no journal, , 

Lead-Bismuth Eutectic (LBE) is used as a spallation neutron target and coolant materials of Accelerator Driven System (ADS), and many kinds of elements are produced as spallation products. It is important to evaluate the release and transport behavior of the spallation products in the LBE. The inventories and the physicochemical composition of the spallation products produced in LBE have been investigated for an LBE loop in the ADS Target Test Facility (TEF-T) in J-PARC. The inventories of the spallation products in the LBE were estimated using the PHITS code. The physicochemical composition of the spallation products in the LBE was calculated using the Thermo-Calc code under the conditions of the operation temperatures of LBE from 350$$^{circ}$$C to 500$$^{circ}$$C and the oxygen concentrations in LBE from 10 ppb to 1 ppm. The calculation showed that the 5 elements of Hg, Tl, Au, Os, Tc were soluble in LBE under the all given conditions and any kinds of compound were not formed in LBE. It was suggested that the oxides of Ce, Zr and Y were stable as CeO$$_{2}$$, ZrO$$_{2}$$ and Y$$_{2}$$O$$_{3}$$ in the LBE.

Oral presentation

Thermophysical property measurements of molten stainless steel containing 10mas%B$$_{4}$$C by electromagnetic levitation technique

Fukuyama, Hiroyuki*; Higashi, Hideo*; Yamano, Hidemasa

no journal, , 

In this study, densities, surface tensions, normal spectral emissivities, heat capacities and thermal conductivities of molten SUS316L and SUS316L containing 10mass%B$$_{4}$$C were measured by the electromagnetic levitation technique in a static magnetic field.

Oral presentation

Development of the viscosity measurement apparatus of molten nickel and stainless steel

Kokubo, Hiroki*; Nishi, Tsuyoshi*; Ota, Hiromichi*; Yamano, Hidemasa

no journal, , 

It is necessary to obtain viscosity data of eutectic molten material of boron carbide (B$$_{4}$$C) and stainless steel (SS) for severe accident analyses. To measure the viscosity, in this study, a viscometer apparatus has been developed using Nuckel and stainless steel as a mother alloy.

Oral presentation

Viscosities of stainless steel and boron carbide alloy melts

Nishi, Tsuyoshi*; Ota, Hiromichi*; Kokubo, Hiroki*; Yamano, Hidemasa

no journal, , 

The viscosity measurements of the SS+5mass%B$$_{4}$$C and SS+10mass%B$$_{4}$$C alloy melts were performed in the temperature range from 1713 to 1823 K and from 1773 to 1813 K, respectively. In these results, there was no difference as expected for the viscosities of the SS+5mass%B$$_{4}$$C alloy melts and the SS+10mass%B$$_{4}$$C alloy melts. Moreover, the viscosities of the SS+5 mass%B$$_{4}$$C alloy melts can be fitted by Arrhenius's equation. It was also found that the viscosities of both SS+B$$_{4}$$C alloy melts were higher than that of the SS melts. It can be presumed that the viscosity of the SS+B$$_{4}$$C alloy melts increased with B$$_{4}$$C components.

Oral presentation

EPMA analysis of multi-component interaction products between boron carbide and stainless steel under a postulated core disruptive accident in SFR

Nakamura, Kinya*; Ota, Hirokazu*; Takai, Toshihide; Yamano, Hidemasa

no journal, , 

To investigate in-depth redistribution behavior of the constituent elements of B$$_{4}$$C and SS, electron probe microanalysis was applied to the multi-component interaction products held at 1500$$^{circ}$$C for 8 min approximately in argon gas atmosphere in this study. The results indicated that relatively low-density boron was preferentially diffused the upper surface of the molten SS horizontally to form acicular and high melting point compound (Cr,Fe)$$_{2}$$B along with the eutectic microstructure composed of (Fe,Cr,Ni) solid solution and (Cr,Fe)$$_{2}$$B. According to the macroscopic quantitative analysis, average chemical concentration of boron is almost uniform in the horizontal direction regardless of the distance from the B$$_{4}$$C pellet but decreased toward the bottom in the molten SS monotonously in the vertical direction.

Oral presentation

Enthalpy measurement and evaluation of heat capacity on (U, Pu)O$$_{2}$$

Morimoto, Kyoichi; Ogasawara, Masahiro*

no journal, , 

The heat capacity of MOX fuel is one of important thermophysical properties for the evaluation of its thermal conductivity and for the evaluations of transient event and severe accident of nuclear reactor. In this study, MOX sample with Pu-content of nearly 50% (50%Pu-MOX) was measured because it was predicted that the influence of Pu addition appeared to be significant. The raw material powder was cold-pressed and was sintered to make pellet. The oxygen to metal ratio of this pellet was adjusted to 2.00. The enthalpy of this sample was measured with drop calorimeter in in high temperature range. The rhenium container was used to prevent the reaction between the sample and the sample container. It was found that the enthalpy increased at a nearly constant rate up to about 1900 K with increasing temperature and that its rate rose at about 1900 K or more. It means that the heat capacity is approximately constant up to 1900K and starts to rise at about 1900K.

Oral presentation

Corrosion properties of PCV equivalent materal for simulated severe accident

Nakano, Hiroko; Takeuchi, Tomoaki; Otsuka, Kaoru; Hirota, Noriaki; Tsuchiya, Kunihiko

no journal, , 

no abstracts in English

Oral presentation

Kinetic behavior of eutectic melting reaction between stainless steel and boron carbide

Kikuchi, Shin; Yamano, Hidemasa

no journal, , 

In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), eutectic reaction between boron carbide (B$$_{4}$$C) and stainless steel (SS) may probably occur. Elucidation on the behavior of cited eutectic reaction is very important in terms of evaluation of core disruptive accidents in SFRs. For the first step to clarify the kinetic feature of B$$_{4}$$C-SS eutectic reaction, the preliminary thermogavimetry-differential thermal analysis (TG-DTA) measurements using individual reagent were performed to obtain the fundamental information and to confirm the applicability of sample crucibles. It was found that alumina crucible was applicable in terms of eutectic behavior. Based on the DTA curves at different heating rates, the kinetic parameters were roughly estimated by using Kissinger method.

Oral presentation

Effect of dissolved oxygen and hydrogen on mechanical property of AISI 316 stainless steels in a simulated PWR water conditions

Takeuchi, Tomoaki; Nakano, Hiroko; Otsuka, Kaoru; Hirota, Noriaki; Tsuchiya, Kunihiko

no journal, , 

no abstracts in English

Oral presentation

Corrosion behavior of Ni-based alloy in molten FLiNaK salt as a fundamental research on molten salt reactors

Ogasawara, Koji*; Sekiguchi, Yuma*; Terai, Takayuki*; Kawamura, Hiroshi*; Tsuchiya, Kunihiko; Watanabe, Takashi*

no journal, , 

no abstracts in English

Oral presentation

High temperature chemical reaction of strontium vapor species with Stainless steel

Afiqa, B. M.; Nakajima, Kunihisa; Miwa, Shuhei; Osaka, Masahiko; Oishi, Yuji*; Muta, Hiroaki*; Kurosaki, Ken*

no journal, , 

The release of strontium (Sr) from a fuel would be enhanced by the formation of volatile SrCl$$_{2}$$ in case of seawater injection into the core of Fukushima Dai-ichi Nuclear Power Plant. We investigated the chemisorption behavior of releasing SrCl$$_{2}$$ vapor onto type 304 stainless steel. It was observed that stable Sr-Si-O compounds were formed on the stainless steel surface.

Oral presentation

Lattice and bulk expansion of $$^{244}$$Cm-doped nitride induced by self-irradiation damage at room temperature

Takano, Masahide; Takaki, Seiya

no journal, , 

To understand behavior of the nitride fuel for minor actinide transmutation, lattice and bulk expansion of the nitride fuel pellet induced by $$alpha$$ self-irradiation at room temperature was investigated. Lattice parameter and dimensions of the (Pu,Cm,Zr)N nitride fuel pellet were repeatedly measured at room temperature as a function of storage period, and their relationship was considered. The lattice expansion followed the model equation well, and saturated to 0.49%, which is greater than that for CmN. The higher density of metallic elements in (Pu,Cm,Zr)N can be the main cause of the greater expansion. On the other hand, both the pellet diameter and height increased with the expansion curve similar to the lattice expansion, and saturated approximately to 0.5%. From this similarity we have found that the main cause of the bulk expansion is the lattice expansion due to the accumulation of Frenkel defects.

Oral presentation

Materials challenges for future nuclear fuel cycles; R&D on nitride fuel cycle for MA transmutation

Takano, Masahide

no journal, , 

We have studied nitride fuels for the accelerator driven system (ADS) as a potential candidate for MA transmutation fuels in these two decades. We have demonstrated the synthesis of MA nitrides from the oxides and the sintering of fuel pellets. Material properties such as thermal conductivity and thermal expansion have been also measured to develop the database. Based on these experience and fundamental knowledge, now we strive toward the R&D in engineering scale. As for the fuel fabrication, application of sol-gel process and economical use of $$^{15}$$N-enriched nitrogen gas are main technical challenges. Development of the fuel performance code optimized for the nitride fuel is also important. In addition to the existing data, we are studying thermomechanical properties, compatibility with cladding materials, and behavior of accumulated helium to develop models describing high temperature phenomena in the fuel pins.

Oral presentation

Thermophysical properties of stainless steel containing 10mass%-B$$_{4}$$C in the solid state

Takai, Toshihide; Furukawa, Tomohiro; Yamano, Hidemasa

no journal, , 

Oral presentation

Numerical simulation of distribution of melt component in reactor

Sato, Takumi; Hirata, Naoya*; Oikawa, Katsunari*; Nagae, Yuji; Kurata, Masaki

no journal, , 

Macroscopic segregation of molten core and melt components occurs with slow cooling rate in the accident of Fukushima Daiichi Nuclear Power Plants. In this study, solidification and microscopic segregation are simulated with the Scheil model and thermal properties calculated by Thermo-calc in order to investigate an influence of cooling conditions on macroscopic segregation. A macroscopic segregation behaviour has been calculated for UO$$_{2}$$-ZrO$$_{2}$$-FeO system, which are major oxides of molten core materials. According to calculated results, UO$$_{2}$$ and ZrO$$_{2}$$ was concentrated in initial solidification area. On the other hand, FeO were strongly concentrated in later solidification area. In addition, macroscopic segregation tends to be suppressed in the conditions of fast solidification rate and slow velocity of solidification interface.

25 (Records 1-20 displayed on this page)