Teruo Kikuchi

Irradiation behavior of HTGR coated particle fuel at abnormally high temperature

Abstract Irradiation behavior of high temperature gas-cooled reactor (HTGR) coated particles under temperature transient conditions was investigated in accordance with a design-base accident scenario for HTTR, a 30 MWth HTGR under construction at JAERI. One of the scenarios predicts that the fuel temperature of the block-type fuel elements rises to abnormally high temperature by blocking a coolant channel with some foreign substance. For simulating this scenario the fuel compacts incorporating the coated particles were irradiated at normal temperature in three capsules, followed by temperature transient up to a maximum of approximately 2000°C. The post-irradiation examinations, including surface inspection, metrology, ceramography and a measurement of coated particle failure were applied to the fuel compacts to investigate the thermal-transient effect on the fuel integrity. Integrity of the fuel compact was also assessed by an estimation of tangential stress introduced into the compact by the temperature transient.

Capability of energy selective neutron irradiation test facility (ESNIT) for fusion reactor materials testing and the status of ESNIT program

Abstract The concept of a high energy neutron irradiation facility named Energy Selective Neutron Irradiation Test Facility (ESNIT) was conceived at Japan Atomic Energy Research Institute (JAERI) as a facility for materials research in generic nuclear applications for which major potential research incentives were originated from nuclear fusion application. Technological survey and feasibility studies of ESNIT have continued since 1988 and two international workshops have been held in 1989 and 1991. In this paper, the capability of ESNIT for fusion reactor materials testing and the status of technological survey activities including deuteron accelerator system, target system and experimental system are summarized.

Characteristics of an energy selective neutron irradiation test facility (ESNIT) for material irradiation studies

Abstract The concept of an energy selective neutron irradiation test facility (ESNIT) was generated at the Japan Atomic Energy Research Institute, and a program for the preliminary technical survey was initiated in 1988. The potential benefit of such a facility is that it makes possible that neutron-irradiation experiments on materials can be carried out with well-defined energy spectra up to reasonably high flux and fluence levels and with high manipulability. A significant fraction of its applications will be focused on fundamental aspects of radiation damage in fusion reactor materials.

In-pile and post-irradiation creep of type 304 stainless steel under different neutron spectra

In addition to post-irradiation creep tests, in-pile creep tests were performed using newly developed technology with in situ measurement under different neutron spectra. The in-pile creep properties of type 304 stainless steel at 550°C appear to depend on neutron spectrum, but a spectral effect on post-irradiation creep properties is not clearly seen. The rupture time of in-pile creep under a high thermal neutron flux condition is the shortest. The order of the rupture time following the high thermal flux condition is post-irradiation creep, in-pile creep with a thermal neutron shield condition and finally creep of unirradiated material, all in increasing order. It is suggested that the acceleration of creep deformation and fracture observed in irradiation creep tests may be related to enhancement of thermal creep in terms of FMD increased under a high thermal neutron flux in addition to increased helium embrittlement.

Distribution of fission products in irradiated graphite materials of HTGR fuel assemblies: Third and fourth OGL-1 fuels

Abstract Axial, circumferential and radial distributions of fission products in the graphite sleeve, inner-tube and block of irradiated high temperature gas-cooled reactor (HTGR) fuel assemblies were measured by gamma spectrometry, lathe sectioning and beta counting with ion-exchange separation. Some distinctive peaks of 144Ce and 125Sb in their axial profiles, together with the very high activity level of fission products are ascribed to the failure of coated fuel particles. The effective retention capability of the graphite sleeve was observed for 90Sr, 106Ru, 125Sb, 144Ce and 155Eu; whereas not for 134Cs and 137Cs. Silver-110m was detected in graphite materials of the fourth OGL-1 fuel assembly with an increased burnup of 1.96% fissions per initial metal atom (FIMA). Effective in-pile diffusion coefficients of 90Sr, 125Sb and 144Ce in the graphite sleeves have been estimated using the Fickian diffusion theory.

Distribution of fission and activation products in the graphite sleeves of HTGR fuel rods: first and second OGL-1 fuels

Abstract In order to develop the fuel of a high temperature gas-cooled reactor, axial and radial profiles of fission and activation products in the graphite sleeves of the first and second fuel assemblies irradiated in the in-pile gas loop OGL-1 were measured by means of lathe sectioning, gamma spectrometry, and ion-exchange separation. Several sharp axial profile peaks of 90Sr, 134Cs, and 137Cs were observed for the second fuel sleeves. The peaks are assigned to the failed coated fuel particles; the rest of the profiles to the contamination uranium contained in the fuel compact matrix. Fission products with lower diffusivity such as 90Sr, 106Ru, and 144Ce were completely retained in the second fuel sleeve, whereas similar effectiveness in retention was not observed for 134Cs and 137Cs. Diffusion coefficients of cesium and strontium have been roughly estimated through the comparison between the measured and the simply calculated profiles in the sleeve.

Present status of ESNIT (energy selective neutron irradiation test facility) program

Abstract The present status of technical studies of a high energy neutron irradiation facility, ESNIT (energy selective neutron irradiation test facility), is summarized. Technological survey and feasibility studies of ESNIT have continued since 1988. The results of technical studies of the accelerator, the target and the experimental systems in ESNIT program were reviewed by an International Advisory Committee in February 1993. Recommendations for future R&D on ESNIT program are also summarized in this paper.

Uranium contamination in coating and in matrix material of unirradiated coated particle fuel.

A study was made of uranium contamination in (a) the coating layers of TRISO particles (a-1) before compacting and (a-2) separated from once-compacted fuel heat-treated at 1,400 or 1,800°C, and (b) in the matrix material of the same compacts. The contamination in the pyrocarbon layers of the coating was determined, after mechanically separating the coating layers, by a procedure of neutron activation, burn-off and 133Xe trapping. For the silicon carbide coating layer, the fragments of coating left from the above procedure were fused into alkaline melt, and the 133Xe released at each heating step was trapped. For the matrix material, the fuel compacts were deconsolidated electrolytically or mechanically, followed by activation analysis. The results of the foregoing measurements proved the uranium contamination in pyrocarbon and silicon carbide coating to be at most of the order of 10−4 in reference to uranium content in kernel, while the corresponding value for particles sampled from fuel compacts heat-tre...

Studies on Glassification of Purex-type Waste Solution

Purex 1 WW型模擬廃液に正リン酸を加えて, 1100℃に加熱してP2O5含量が65重量%のリン酸ガラスを造り, これを用いて, 蒸留水, 水道水および海水による137CsとPとの浸出を調べた。その結果, (1) 137Csの浸出は蒸留水によるよりも水道水や海水によるほうが多かったが, Pの浸出は水の種類にあまり影響されない, (2) 浸出量は温度とともに急激に増大する, (3) 浸出測定結果は, ガラス成分と水との反応における活性化エネルギーが広い範囲に分布する場合の解析を適用することにより, 活性化エネルギー初期スペクトルによって整理できる, などが明らかにされた。