[en] The health effects resulting from severe accidents of 1,000MWe KSNP(Korea Standard Nuclear Plant) PWR and typical 600MWe CANDU(CANada Deuterium Uranium) plants were estimated and compared. The population distribution of the site extending to 80km for both site were considered. The release fraction for various Source Term Categories(STC) and core inventories were used in the estimation of the health effects risks by using the MACCS2(MELCOR Accident Consequence Code System2) code. Individuals are assumed to evacuate beyond 16km from the site. The health effects considered in this comparative study are early and cancer fatality risk, and the results are presented as CCDF(Complementary Cumulative Distribution Function) curves considering the occurrence probability of each STC's. According to the results, the early and cancer fatality risks of PHWR plants are lower than those of PWR plants. This is attributed the fact that the amount of radioactive materials that released to the atmosphere resulting from the postulated severe accidents of PHWR plants are smaller than that of PWR plants. And, the dominating initiating event of STC that shows maximum early and cancer fatality risk is SGTR(Steam Generator Tube Rupture) for both plants. Therefore, the appropriated actions must be taken to reduce the occurrence probability and the amounts of radioactive materials released to the environment in order to protect the public for both PWR and PHWR plants

[en] In this study, we investigate the dispersion behavior of debris and debris cloud generated by high-velocity impacts using the smoothed particle hydrodynamics (SPH) technique. The projectile and target plate were made of aluminum, and we confirm the validity of the SPH technique by comparing the measured major and minor axis lengths of the debris cloud in the reference with the predicted values obtained through the SPH analysis. We perform high-velocity impact and fracture analysis based on the verified SPH technique within the velocity ranges of 1.5~4 km/s, and we evaluate the dispersion behavior of debris induced by the impact in terms of its kinetic energy. The maximum dispersion radius of the debris on the witness plates located behind the target plate was increased with increasing impact velocity. We derive an empirical equation that is capable of predicting the dispersion radius, and we found that 95% of the total kinetic energy of the debris was concentrated within 50% of the maximum dispersion radius

[en] Wound evidence derived from various physical threats is closely related to the direct elements of the threat. To understand the causality of threat elements and wound evidence, a finite element analysis was conducted for material behavior and damage according to the threat elements. Then, the aspect of wound evidence was analyzed based on the results. In addition, the relativeness between threat elements and wound evidence was evaluated quantitatively through a statistical comparison via the cross correlation analysis. Moreover, the shape similarity between the wound and the sectional shapes of the threat was calculated. As a result, it was confirmed that the correlation coefficient between wound evidence and the sectional shape of the threat generating it was 0.9692, indicating a very good agreement. The validity of the evaluation technique was verified by using the correlation coefficient, which was derived from the compared threats. Concludingly, the shape similarity with the threat that caused the wound evidence was higher than the other threats compared. The correlation between the threat elements and wound evidence together with the evaluation methodology of shape similarity proposed in this study could be used as a physical reasoning method in future forensic investigations.

[en] A corrosion fatigue crack propagation test for 430 stainless steel and its heat affected zone was conducted in pH buffer solutions, and the results were compared with model predictions. The bare corrosion effect on fatigue crack propagation, particularly in corrosive environments was evaluated by means of a modified Forman equation. As shown in the results, the average corrosion rate determined from the ratio of corrosion induced crack length to the entire crack length under a cycle load was 0.11 and 0.37 for the base metal and heat affected zone, respectively, with a load ratio of 0.5, frequency of 0.5, and a pH 10.0 environment. The modeling and experimental processes demonstrate a step towards a methodology enabling the corrosion effects on fatigue crack propagation behavior to be determined.

[en] This paper revisits the Fukushima accident to draw lessons in the aspect of nuclear safety considering the fact that the Fukushima accident resulted in core damage for three nuclear power plants simultaneously and that there is a high possibility of a failure of the integrity of reactor vessel and primary containment vessel. A brief review on the accident progression at Fukushima nuclear power plants is discussed to highlight the nature and characteristic of the event. As the severe accident management measures at the Fukushima Daiich nuclear power plants seem to be not fully effective, limitations of current severe accident management strategy are discussed to identify the areas for the potential improvements including core cooling strategy, containment venting, hydrogen control, depressurization of primary system, and proper indication of event progression. The gap between the Fukushima accident event progression and current understanding of severe accident phenomenology including the core damage, reactor vessel failure, containment failure, and hydrogen explosion are discussed. Adequacy of current safety goals are also discussed in view of the socio-economic impact of the Fukushima accident. As a conclusion, it is suggested that an investigation on a coherent integrated safety principle for the severe accident and development of innovative mitigation features is necessary for robust and resilient nuclear power system.

[en] It is dangerous to work inside reactor vessel because of high radioactivity. So robot system is necessary to work inside reactor vessel instead of workers. And robot also has advantage of mobility over conventional equipment which has limitation of the range of accessibility. In this paper, we describe design consideration and criteria of robot system for reactor vessel, and component of developed system. Reliability, usability and convenience of robot system were considered to design and fabrication

[en] Structures are often subjected to various types of loading such as static, dynamic, or impact loading. Therefore, experimental and numerical methods have been employed to find adequate material properties according to the conditions. The Split Hopkinson pressure bar (SHPB) test has frequently been used to test engineering materials, particularly those used under high strain rates. In this study, the compressive deformation behaviors of aluminum alloy under impact conditions have been investigated by means of the SHPB test. The experimental results were then compared with those of finite element analyses. It was shown that reasonably good agreement with the true stress strain curves was obtained at strain rates ranging from 1000s'-1' to 2000s'-1'. When the strain rate increased by 30%, the peak stress in particular increased by 17%, and the strain also increased by 20%

[en] A two-dimensional array of planar Langmuir probes on a 200 mm diameter silicon wafer was used in an inductively coupled plasma reactor to follow the spatial and temporal variation of ion flux impinging on the wafer in the presence of an instability. Small amplitude low frequency (∼2 Hz) oscillations superimposed on a steady state ion flux distribution were observed in SF₆ plasmas. The magnitude and phase of the oscillations depend on position on the wafer and analysis of these variations reveals that these low frequency oscillations correspond to waves that move across the wafer

[en] Physicians and nurses stand with their back towards the C-arm fluoroscope when using the computer, taking things out of closets and preparing drugs for injection or instruments for intervention. This study was conducted to investigate the relationship between the type of lead apron and radiation exposure to the backs of physicians and nurses while using C-arm fluoroscopy. We compared radiation exposure to the back in the three groups: no lead apron (group C), front coverage type (group F) and wrap-around type (group W). The other wrap-around type apron was put on the bed instead of on a patient. We ran C-arm fluoroscopy 40 times for each measurement. We collected the air kerma (AK), exposure time (ET) and effective dose (ED) of the bedside table, upper part and lower part of apron. We measured these variables 30 times for each location. In group F, ED of the upper part was the highest (p < 0.001). ED of the lower part in group C and F was higher than that in group W (p = 0.012). The radiation exposure with a front coverage type apron is higher than that of the wrap-around type and even no apron at the neck or thyroid. For reducing radiation exposure to the back of physician or nurse, the wrap-around type apron is recommended. This type of apron can reduce radiation to the back when the physician turns away from the patient or C-arm fluoroscopy. (authors)

[en] The purpose of this paper is to confirm the event timings and the magnitude of fission product aerosol release from the Fukushima accident. Over a few hundreds of technical papers have been published on the environmental impact of Fukushima Daiichi accident since the accident occurred on March 11, 2011. However, most of the research used reverse or inverse method based on the monitoring of activities in the remote places and only few papers attempted to estimate the release of fission products from individual reactor core or from individual spent fuel pool. Severe accident analysis code can be used to estimate the radioactive release from which reactor core and from which radionuclide the peaks in monitoring points can be generated. The basic material used for this study are the initial core inventory obtained from the report JAEA-Data/Code 2012–018 and the given accident scenarios provided by Japanese Government or Tokyo Electric Power Company (TEPCO) in official reports. In this research a forward method using severe accident progression code is used as it might be useful for justifying the results of reverse or inverse method or vice versa. The release timing and amounts to the environment are estimated for volatile radioactive fission products such as noble gases, cesium, iodine, and tellurium up to 184 hours (about 7.7 days) after earthquake occurs. The in-plant fission product behaviors and release characteristics to environment are estimated using the severe accident progression analysis code, MELCOR, for Fukushima Daiichi accident. These results are compared with other research results which are summarized in UNSCEAR 2013 Report and other technical papers. Also it may provide the physically based arguments for justifying or suspecting the rationale for the scenarios provided in open literature. The estimated results by MELCOR code simulation of this study indicate that the release amount of volatile fission products to environment from Units 1, 2, and 3 cores is well within the range estimated by the reverse or inverse method, which are summarized in UNSCEAR 2013 report. But this does not necessarily mean that these two approaches are consistent