Nuclear Engineering and Technology Institute

Yuseong gu, South Korea

Nuclear Engineering and Technology Institute

Yuseong gu, South Korea
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Cho D.-K.,Korea Atomic Energy Research Institute | Sun G.-M.,Korea Atomic Energy Research Institute | Choi J.,Korea Atomic Energy Research Institute | Hwang D.-H.,Nuclear Engineering and Technology Institute | And 3 more authors.
Journal of Nuclear Science and Technology | Year: 2011

The sensitivity of parameters related with reactor physics on the source terms of decommissioning wastes from a CANDU reactor was investigated in order to find a viable, simplified burned core model of a Monte Carlo simulation for decommissioning waste characterization. First, a sensitivity study was performed for the level of nuclide consideration in an irradiated fuel and implicit geometry modeling, the effects of side structural components of the core, and structural supporters for reactive devices. The overall effects for computation memory, calculation time, and accuracy were then investigated with a full-core model. From the results, it was revealed that the level of nuclide consideration and geometry homoge-nization are not important factors when the ratio of macroscopic neutron absorption cross section (MNAC) relative to a total value exceeded 0.95. The most important factor affecting the neutron flux of the pressure tube was shown to be the structural supporters for reactivity devices, showing an 10% difference. Finally, it was concluded that a bundle-average homogeneous model considering a MNAC of 0.95, which is the simplest model in this study, could be a viable approximate model, with about 25% lower computation memory, 40% faster simulation time, and reasonable engineering accuracy compared with a model with an explicit geometry employing an MNAC of 0.99. ©Atomic Energy Society of Japan.

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.

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.

Cho D.-K.,Korea Atomic Energy Research Institute | Sun G.-M.,Korea Atomic Energy Research Institute | Choi J.,Korea Atomic Energy Research Institute | Yang H.-Y.,Nuclear Engineering and Technology Institute | Hwang T.-W.,Nuclear Engineering and Technology Institute
Journal of Nuclear Science and Technology | Year: 2011

The method for the establishment of an equilibrium core model proposed in the previous paper and the source term calculation method proposed in this paper for the characterization of decommissioning waste were verified by comparing the nuclide inventory estimated by MCNP/ORIGEN2 simulations with the measured nuclide inventory according to a chemical assay in an irradiated pressure tube discharged from Wolsong Unit 1 in 1994. At first, the time-average pseudoequilibrium full-core model of Wolsong Unit 1 was developed on the basis of the previously proposed modeling method for the activation of in-core and ex-core structural components. Then, the application level of the neutron flux and cross section in the radionuclide buildup calculation were compromised. Fourteen major actinides and fission products were considered to represent the irradiated fuel condition, and a geometry simplification was also introduced in the burned full-core model for MCNP simulation. The assumption of a constant neutron flux and capture cross section as a function of the irradiation time was applied in the radionuclide buildup calculation in ORIGEN2. As a result, the values estimated from the analysis system agreed with the measured data within a difference range of 30%. Therefore, it was found that the MCNP/ORIGEN system and source term characterization method proposed can be viable to estimate the source terms of the decommissioning waste from a CANDU reactor. © 2011 Atomic Energy Society of Japan, 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.

Lim S.-G.,Nuclear Engineering and Technology Institute | Lee S.-H.,Nuclear Engineering and Technology Institute | Kim H.-G.,Nuclear Engineering and Technology Institute
Nuclear Engineering and Design | Year: 2010

A passive flow controller or a fluidic device (FD) is used for a safety injection system (SIS) for efficient use of nuclear reactor emergency cooling water since it can control the injection flow rate in a passive and optimal way. The performance of the FD is represented by pressure loss coefficient (K-factor) which is further affected by the configuration of the components such as a control port direction and a nozzle angle. The flow control mechanism that is varied according to the water level inside a vortex chamber determines the duration of the safety injection. This paper deals with a computational fluid dynamics (CFD) analysis for simulating the flow characteristics of the FD using the ANSYS CFX 11.0. The CFD analysis is benchmarked against existing experimental data to obtain applicability to the prediction of the FD performance in terms of K-factor. The CFD calculation is implemented with Shear Stress Transport (SST) model for a swirling flow and a strong streamline curvature in the vortex chamber of the FD, considering a numerical efficiency. Based on the benchmark results, parametric analyses are performed for an optimal design of the FD by varying the control port direction and the nozzle angle. Consequently, the FD performance is enhanced according to the angle of the control port nozzle. © 2010 Elsevier B.V. All rights reserved.

Kim H.-S.,Nuclear Engineering and Technology Institute | Park J.-K.,Nuclear Engineering and Technology Institute
Proceedings of the International Conference on Radioactive Waste Management and Environmental Remediation, ICEM | Year: 2011

The programs for estimating the decommissioning cost have been developed for many different purposes and applications. The estimation of decommissioning cost is required a large amount of data such as unit cost factors, plant area and its inventory, waste treatment, etc. These make it difficult to use manual calculation or typical spreadsheet software such as Microsoft Excel. The cost estimation for eventual decommissioning of nuclear power plants is a prerequisite for safe, timely and cost-effective decommissioning. To estimate the decommissioning cost more accurately and systematically, KHNP, Korea Hydro and Nuclear Power Co. Ltd, developed a decommissioning cost estimating computer program called "DeCAT-Pro", which is Decommission-ing Cost Assessment Tool - Professional. (Hereinafter called "DeCAT") This program allows users to easily assess the decommissioning cost with various decommissioning options. Also, this program provides detailed reporting for decommissioning funding requirements as well as providing detail project schedules, cash-flow, staffing plan and levels, and waste volumes by waste classifications and types. KHNP is planning to implement functions for estimating the plant inventory using 3-D technology and for classifying the conditions of radwaste disposal and transportation automatically. Copyright © 2011 by ASME.

Ko D.-Y.,Nuclear Engineering and Technology Institute | Lee J.-G.,Nuclear Engineering and Technology Institute
Nuclear Engineering and Design | Year: 2010

This paper describes the development of a measuring system to measure gaps between the reactor vessel (RV) and the core support barrel (CSB) remotely with the aim of reactor vessel internals (RVI)-modularization. A remote measurement system was developed for use at actual construction sites of nuclear power plants using a measurement sensor, a threaded connection jig, and a zero-point adjustment device. With these, a reduced-scale model system was validated. With the remote measurement system, experiments and analyses were performed using mockups for both the RV and the CSB to verify the applicability of the proposed system in a construction project. From the data acquired by the remote measurement system, shims were separately made and adjusted. After installing the shims on RV core-stabilizing lugs, the gaps satisfied requirements within the permissible range of 0.381-0.508 mm. We evaluated the reliability and applicability of the remote measurement method and concluded that the remote measurement system enables RVI-modularization with a significantly reduced construction period. © 2010 Elsevier B.V. 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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