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Oh S.,Nuclear Engineering and Technology Institute | Jang C.,KAIST | Kim J.H.,Korea Atomic Energy Research Institute | Jeong Y.H.,Korea Atomic Energy Research Institute
Materials Science and Engineering A | Year: 2010

The effects of Nb on hydride Zr alloys were investigated. Various Zr alloys with different Nb content up to 2.0% were prepared in a sheet shape and charged with hydrogen up to 850 ppm. It was found that the fraction of recrystallized grains was reduced with increasing Nb content during the heat treatment. Intergrain and intragrain hydrides were tangled in unalloyed Zr, which has fully recrystallized grains. On the other hand, Nb containing Zr alloys that have partially cold-worked grains had only intergrain hydrides precipitated along the rolling direction. In tensile tests, elongation was decreased significantly with increasing hydrogen content for unalloyed Zr. However, the reduction of elongation was less significant for Nb-containing alloys. The softening of the Zr matrix was caused by hydrogen, which was arrested in cold-worked grain. And β-Nb precipitates, which have high solubility of hydrogen, retarded the precipitation of hydrides, thereby enhancing the resistance of hydride embrittlement. Fracture surface of the tensile tested specimen was observed by SEM. From the SEM observation, secondary cracks caused by hydrides were found on the fracture surface of the hydrogen-charged specimens. Cleavage facets were observed on the fracture surface of unalloyed Zr; however, a mixture of dimple and cleavage facets was observed on the surface of the Nb-containing alloys, which indicates that Nb increased the resistance to hydride embrittlement of Zr alloys. © 2009 Elsevier B.V. All rights reserved.

Kim J.,Korea Atomic Energy Research Institute | Kim J.,Paul Scherrer Institute | Park J.,Korea Atomic Energy Research Institute | Jang S.C.,Korea Atomic Energy Research Institute | Shin Y.C.,Nuclear Engineering and Technology Institute
Human Factors and Ergonomics In Manufacturing | Year: 2011

As computer-based design features are adopted in main control rooms of nuclear power plants, a human reliability analysis (HRA) method dealing with the effects of these design features on human behavior is needed. This article provides experimental results of human diagnostic performance characteristics in a computer-based, full-scope, dynamic simulator to inform some insights on developing an HRA method for a computer-based advanced control room. In comparison to the performance time for diagnostic actions, it showed more or less faster performance in the computer-based control room with a computer-based emergency operating procedure (EOP) for an event scenario with an apparent diagnostic symptom than in the conventional control room with a paper-based EOP, but it is also revealed that the diagnosis time is highly dependent on the situational characteristics of simulated events. Regarding the aspect of human error occurrence, a decision maker showed the potential for leading to a wrong conclusion regarding the plant state when he makes a situational assessment or makes a decision based on abnormal information by himself without communicating or consulting with other operators. Finally, regarding the aspects of error recovery, it showed that the error recovery potential becomes much higher for the advanced control room than for the conventional control room due to the information sharing and access capability of the advanced control room between and for all the crew members. It is expected that an HRA method for an advanced control room environment should adequately reflect these characteristics of human behavior in a computer-based control room. © 2010 Wiley Periodicals, Inc.

Park J.,Korea Atomic Energy Research Institute | Jung W.,Korea Atomic Energy Research Institute | Jonghyun K.O.,Nuclear Engineering and Technology Institute
Nuclear Engineering and Technology | Year: 2010

According to wide-spread experience in many industries, a procedure is one of the most effective countermeasures to reduce the possibility of human related problems. Unfortunately, a systematic framework to evaluate the complexity of procedural tasks seems to be very scant. For this reason, the TACOM measure, which can quantify the complexity of procedural tasks, has been developed. In this study, the appropriateness of the TACOM measure is investigated by comparing TACOM scores regarding the procedural tasks of high speed train drivers with the associated workload scores measured by the NASA-TLX technique. As a result, it is observed that there is a meaningful correlation between the TACOM scores and the associated NASA-TLX scores. Therefore, it is expected that the TACOM measure can properly quantify the complexity of procedural tasks.

Jang C.,KAIST | Cho P.-Y.,KAIST | Kim M.,KAIST | Oh S.-J.,Nuclear Engineering and Technology Institute | Yang J.-S.,KEPCO E&C
Materials and Design | Year: 2010

The effects of weld microstructure and residual stress distribution on the fatigue crack growth rate of stainless steel narrow gap welds were investigated. Stainless steel pipes were joined by the automated narrow gap welding process typical to nuclear piping systems. The weld fusion zone showed cellular-dendritic structures with ferrite islands in an austenitic matrix. Residual stress analysis showed large tensile stress in the inner-weld region and compressive stress in the middle of the weld. Tensile properties and the fatigue crack growth rate were measured along and across the weld thickness direction. Tensile tests showed higher strength in the weld fusion zone and the heat affected zone compared to the base metal. Within the weld fusion zone, strength was greater in the inner weld than outer weld region. Fatigue crack growth rates were several times greater in the inner weld than the outer weld region. The spatial variation of the mechanical properties is discussed in view of weld microstructure, especially dendrite orientation, and in view of the residual stress variation within the weld fusion zone. It is thought that the higher crack growth rate in the inner-weld region could be related to the large tensile residual stress despite the tortuous fatigue crack growth path. © 2009 Elsevier Ltd. All rights reserved.

Ko D.-Y.,Nuclear Engineering and Technology Institute
Nuclear Engineering and Technology | Year: 2011

The construction technology for reactor vessel internals (RVI) modularization is one of the most important factors to be considered in reducing the construction period of nuclear power plants. For RVI modularization, gaps between the reactor vessel (RV) core-stabilizing lug and the core support barrel (CSB) snubber lug must be measured using a remote method from outside the RV. In order to measure RVI gaps remotely at nuclear power plant construction sites, certain core technologies must be developed and verified. These include a remote measurement system to measure the gaps between the RV core-stabilizing lug and the CSB snubber lug, an RVI mockup to perform the gap measurement tests, and a new procedure and schedule for RVI installation. A remote measurement system was developed previously, and a gap measurement test was completed successfully using the RVI mockup. We also developed a new procedure and schedule for RVI installation. This paper presents the new and improved installation procedure and schedule for RVI modularization. These are expected to become core technologies that will allow us to shorten the construction period by a minimum of two months compared to the existing installation procedure and schedule.

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