核技术在药型罩密度检测中的应用

介绍了用核技术检测石油射孔弹的药型罩密度的基本结构以及工作原理，给出了使用该技术检测的药型罩装弹后靶试结果与检测结果的对比。实验证明靶试结构与检测结果相符，一致性较好。     

Nondestructive measurement of the coating thickness for simulated TRISO-coated fuel particles by using phase contrast X-ray radiography

The TRISO-coated fuel particle for a HTGR (high temperature gas-cooled reactor) is composed of a nuclear fuel kernel and outer coating layers. The coating layers consist of a buffer PyC (pyrolytic carbon) layer, an inner PyC (I-PyC) layer, a SiC layer, and an outer PyC (O-PyC) layer. X-ray radiography is one of the nondestructive alternatives to measure a coating thickness without generating a radioactive waste. Phase contrast X-ray radiography technology is more powerful for acquiring a radiograph with clear boundaries, when compared with a conventional X-ray radiography. The contrast can be enhanced for weakly absorbing materials in a phase contrast X-ray radiograph by detecting an intensity variation due to the variation of a phase of the X-rays in the boundary between two objects. Phase contrast X-ray radiograph was acquired from a simulated TRISO-coated fuel particle with a micro-focus X-ray imaging system. The coating thickness was nondestructively measured from the phase contrast X-ray image for the fuel particle.     

Measurement of implanted helium particle transport by a flowing liquid lithium film

Due to its low atomic number, low sputtering yield, high sputtered ion fraction and excellent thermal properties, liquid lithium has been proposed as a potential candidate for advanced plasma-facing components (PFC). Using a liquid material opens the possibility of a continuously flowing, self-regenerating plasma-facing surface with a small residence time. This would allow such component to handle very high heat loads that are expected. There are, however, multiple unanswered questions regarding how such a liquid PFC would interact with the plasma in the reactor. The issue of particle control is critical, and it can be a factor to determine the feasibility of these advanced concepts. Hydrogen and helium are important in this regard: hydrogen transport by a flowing PFC impacts the reactor fuel recycling regime and tritium inventory; helium transport can help quantify ash removal by the flowing PFC. The flowing liquid-metal retention experiment (FLIRE) was built at the University of Illinois to answer some of the questions regarding particle transport by flowing liquid films exposed to plasmas. Experimental results regarding helium transport by a flowing lithium film after irradiation with an energetic He ion beam are presented in this work. Retained fraction values up to 2% were measured for the experimental conditions, and the retention was found to increase linearly with implanted ion energy. A pure diffusion model was used to describe the helium transport by the Li film, and it was found that such model predicts a diffusion coefficient of (2.8 ± 0.6) × 10⁻¹¹ m²/s, based on the experimental retention measurements. Preliminary evidence of long-term trapping of helium will also be presented.     

Measurement of time varying thickness of liquid film flowing with high speed gas flow by a constant electric current method (CECM)

A constant-electric-current method (CECM) developed by the present author is a kind of conductance method. The characteristics of the CECM are (1) a constant-current power source is used for supplying the electric power and (2) two kinds of electrodes are installed. One is used for supplying electric power and the other is for detecting the information of hold-up or film thickness. The main merits of the CECM are (1) the output from the sensor electrode is independent of the location of gas phase, for example radial location in a tube cross-section, (2) the sensitivity of detecting the change in the hold-up is higher in the case of the thinner film thickness, and (3) the interaction among the electrodes is negligible. The basic idea, calibration and examples of the application of the CECM will be discussed in the present paper.     

Measurement of carbon distribution in nuclear fuel pin cladding specimens by means of a secondary ion mass spectrometer

Cladding carburization during irradiation of advanced mixed uranium plutonium carbide fast breeder reactor fuel is possibly a life limiting fuel pin factor. The quantitative assessment of such clad carbon embrittlement is difficult to perform by electron microprobe analysis because of sample surface contamination, and due to the very low energy of the carbon K_α X-ray transition.The work presented here describes a method developed at the Swiss Federal Institute for Reactor Research (EIR) to use shielded secondary ion mass spectrometry (SIMS) as an accurate tool to determine radial distribution profiles of carbon in radioactive stainless steel fuel pin cladding. Compared with nuclear microprobe analysis (NMA) [1], which is also an accurate method for carbon analysis, the SIMS method distinguishes itself by its versatility for simultaneous determination of additional impurities.     

On the thermal conductivity of UO2 nuclear fuel at a high burn-up of around 100 MWd/kgHM

A study of the thermal conductivity of a commercial PWR fuel with an average pellet burn-up of 102 MWd/kgHM is described. The thermal conductivity data reported were derived from the thermal diffusivity measured by the laser flash method. The factors determining the fuel thermal conductivity at high burn-up were elucidated by investigating the recovery that occurred during thermal annealing. It was found that the thermal conductivity in the outer region of the fuel was much higher than it would have been if the high burn-up structure were not present. The increase in thermal conductivity is a consequence of the removal of fission products and radiation defects from the fuel lattice during recrystallisation of the fuel grains (an integral part of the formation process of the high burn-up structure). The gas porosity in the high burn-up structure lowers the increase in thermal conductivity caused by recrystallisation.     

放射性核素45Ca2+示踪技术测定核Ca2+摄取和释放功能

介绍细胞核＾４５Ｃａ＾２＋摄取、释放功能测定,＾４５Ｃａ＾２＋向内核膜囊泡转运测定等技术,从而为研究核Ｃａ＾２＋摄取和释放系统在生理及病理条件下的作用提供有力的技术手段。     

Study of the vaporization of Ba from UO2 nuclear fuel particles by high-temperature mass spectrometry

In connection with improving the retention of solid fission products in gas-cooled high-temperature reactor fuels, the vaporization of Ba from UO₂ model nuclear fuel particles with and without a pyrocarbon coating was studied by high-temperature mass spectrometry using a Knudsen cell. The UO₂ kernels of the particles were doped with BaO. In addition, some of them contained Al₂O₃. Whereas BaO mainly evaporated from the surface of the kernels as BaO, only Ba could be observed over the coated particles. Moreover, the BaO vapor pressure over kernels with and without the addition of Al₂O₃ was determined. From this it was determined that the BaO vapor pressure could be diminished by approximately two orders of magnitude by the admixture of Al₂O₃. Finally it was proved that the diminution of the BaO vapor pressure was caused by the formation of the compound BaAl₂O₄.     

Study of the retention of Ba in UO2 nuclear fuel particles by ZrO2 as getter

A new concept to retain solid fission products in the coated particles of gas-cooled high-temperature reactors is the addition of refractory oxides to the fuel kernels as a getter. The suitability of this concept for the retention of BaO by ZrO₂ was determined. Therefore, the evaporation of BaZrO₃ was studied by high temperature mass spectrometry with a Knudsen cell made of Mo and the loss of Ba from model kernels heated for 100 h at different temperatures in a furnace was determined. From the results obtained with each of the two methods which were carefully discussed by theoretical considerations it follows independently that Ba can be retained by ZrO₂ but that Al₂O₃ is even more efficient for the retention of Ba than ZrO₂. Moreover, it could be shown that BaO is the main species in the gaseous phase over BaZrO₃.     

Post-facta analysis of the TMI accident (II): Analysis of fuel rod behavior and core damage estimation by use of TOODEE2-J

By use of the TOODEE2-J computer program, an analysis was carried out of the fuel rod behavior, and core damage was estimated for the TMI-2 reactor during the first three hours of the accident on March 28, 1979. The boundary conditions (e.g. core mixture level, steam flow rate and core inlet flow) are based on a thermal-hydraulic analysis by the RELAP4/MOD6/U4/J2 computer program. The calculated results suggest that bursting of almost all rods except peripheral low-powered rods occurred, and that a large part of the zircaloy cladding exceeded the eutectic temperature to form a liquid phase of Zr---U---O. A total of 43.5% of the zircaloy in the fueled part of the core had been converted to zirconium-dioxide by three hours into the accidient, and major damage to the fuel rods had also occured by then.     

Finite element analysis of local overheating within plutonium enriched UO2 fuel rods caused by PuO2 islands

Within natural UO₂ fuel elements enriched with plutonium, this last material should form PuO₂ solid solutions inside the UO₂ pellets, in a wide range of concentrations. If the solutions are obtained by mechanical mixing of the oxides, PuO₂ islands are formed in the UO₂ matrix. These islands may be the source of several problems in the fuel behaviour, the most important being the overheating of the matrix in the neighbourhood of the particles. It is caused by the large fission cross section of plutonium compared with that of uranium.A detailed study of the thermal effects produced by PuO₂ particles in the UO₂ matrix and the cladding is then important for the specification of their permissible size. A portion of the fuel rods with spherical particles in the most significant places was studied. In order to obtain the dimensionless overheating of the fuel and cladding produced by the presence of those particles, the spacial distribution of temperature was calculated, solving the stationary and linear bidimensional equation of heat conducting using a finite element code. Several geometrical variables and material properties have been taken as dimensionless parameters. A satisfactory convergence of the numerical results to an asymptotic limit with a wellknown exact solution, has been obtained.