• Nuclear Engineering and Technology
• /
• 제39권4호
• /
• pp.313-326
• /
• 2007

In this paper, an algorithm to reconstruct two orthogonal images into a three-dimensional image is developed in order to measure the bubble size and volume in a two-phase boiling flow. The central-active contour model originally proposed by P. $Szczypi\'{n}ski$ and P. Strumillo is modified to reduce the dependence on the initial reference point and to increase the contour stability. The modified model is then applied to the algorithm to extract the object boundary. This improved central contour model could be applied to obscure objects using a variable threshold value. The extracted boundaries from each image are merged into a three-dimensional image through the developed algorithm. It is shown that the object reconstructed using the developed algorithm is very similar or identical to the real object. Various values such as volume and surface area are calculated for the reconstructed images and the developed algorithm is qualitatively verified using real images from rubber clay experiments and quantitatively verified by simulation using imaginary images. Finally, the developed algorithm is applied to measure the size and volume of vapor bubbles condensing in a subcooled boiling flow.

• Nuclear Engineering and Technology
• /
• 제42권6호
• /
• pp.704-711
• /
• 2010

Modal testing requires measuring the vibration of many points, for which an accelerometer, a gab sensor and laser vibrometer are generally used. Conventional modal testing requires mounting of these sensors to all measurement points in order to acquire the signals. However, this can be disadvantageous because it requires considerable measurement time and effort when there are many measurement points. In this paper, we propose a method for modal testing using a camera image. A camera can measure the vibration of many points at the same time. However, this task requires that the measurement points be classified frame by frame. While it is possible to classify the measurement points one by one, this also requires much time. Therefore, we try to classify multiple points using pattern recognition. The feasibility of the proposed method is verified by a beam experiment. The experimental results demonstrate that we can obtain good results.

• Nuclear Engineering and Technology
• /
• 제45권4호
• /
• pp.477-488
• /
• 2013

The PLC (Programmable Logic Controller) has been widely used to implement real-time controllers in nuclear RPSs (Reactor Protection Systems). Increasing complexity and maintenance cost, however, are now demanding more powerful and cost-effective implementation such as FPGA (Field-Programmable Gate Array). Abandoning all experience and knowledge accumulated over the decades and starting an all-new development approach is too risky for such safety-critical systems. This paper proposes an RPS software development process with a platform change from PLC to FPGA, while retaining all outputs from the established development. This paper transforms FBD designs of the PLC-based software development into a behaviorally-equivalent Verilog program, which is a starting point of a typical FPGA-based hardware development. We expect that the proposed software development process can bridge the gap between two software developing approaches with different platforms, such as PLC and FPGA. This paper also demonstrates its effectiveness using an example of a prototype version of a real-world RPS in Korea.

• Nuclear Engineering and Technology
• /
• 제53권9호
• /
• pp.2803-2815
• /
• 2021

This paper addresses the uncertainty quantification and sensitivity analysis of a depleted light-water fuel assembly of the Turkey Point-3 benchmark. The uncertainty of the fuel assembly decay heat and isotopic densities is quantified with respect to three different groups of diverse parameters: nuclear data, assembly design, and reactor core operation. The uncertainty propagation is conducted using a two-step analysis code system comprising the lattice code STREAM, nodal code RAST-K, and spent nuclear fuel module SNF through the random sampling of microscopic cross-sections, fuel rod sizes, number densities, reactor core total power, and temperature distributions. Overall, the statistical analysis of the calculated samples demonstrates that the decay heat uncertainty decreases with the cooling time. The nuclear data and assembly design parameters are proven to be the largest contributors to the decay heat uncertainty, whereas the reactor core power and inlet coolant temperature have a minor effect. The majority of the decay heat uncertainties are delivered by a small number of isotopes such as ²⁴¹Am, ¹³⁷Ba, ²⁴⁴Cm, ²³⁸Pu, and ⁹⁰Y.

• Nuclear Engineering and Technology
• /
• 제53권10호
• /
• pp.3438-3448
• /
• 2021

Dynamic seals are one of the critical components of Absorber Rod Drive Mechanism of Fast Breeder Reactors, requiring separate experimental development. Their significance can't be overemphasized considering that the availability and re-usability of Control Rod Drive Mechanisms of Fast Breeder Test Reactor is governed by the failure rate of dynamic seals (bellows). For prototype and subsequent Fast Breeder Reactors in India, choice of the dynamic seal is changed to an in-house designed & developed labyrinth type V-ring seal. The present work is related to the numerical investigations carried out to gain insights into the sealing mechanism and the thermal behaviour of these seals. The results indicate that the seal geometry is very important for obtaining optimum performance. By changing the geometry of the present seal, performance enhancement by more than 50% in important indices is obtained. Further, the results point out that caution shall be exercised when the seal material & its operating temperature are changed. Also, the numerical model developed in this study will be useful for developing more robust dynamic seals in future.

• Nuclear Engineering and Technology
• /
• 제53권10호
• /
• pp.3182-3195
• /
• 2021

The implementation and validation of multi-dimensional (multi-D) features in thermal-hydraulic system codes aims to extend the application of these codes towards multi-scale simulations. The main goal is the simulation of large-scale three-dimensional effects inside large volumes such as piping or vessel. This novel approach becomes especially relevant during the simulation of accidents with strongly asymmetric flow conditions entailing density gradients. Under such conditions, coolant mixing is a key phenomenon on the eventual variation of the coolant temperature and/or boron concentration at the core inlet and on the extent of a local re-criticality based on the reactivity feedback effects. This approach presents several advantages compared to CFD calculations, mainly concerning the model size and computational efforts. However, the range of applicability and accuracy of the newly implemented physical models at this point is still limited and needs to be further extended. This paper aims at contributing to the validation of the multi-D features of the system code ATHLET based on the simulation of the Tests 1.1 and 2.1, conducted at the test facility ROCOM. Overall, the multi-D features of ATHLET predict reasonably well the evolution from both experiments, despite an observed overprediction of coolant mixing at the vessel during both experiments.

• Nuclear Engineering and Technology
• /
• 제53권11호
• /
• pp.3465-3473
• /
• 2021

For a long time, research into integrated deterministic-probabilistic safety assessment has been continuously conducted to point out and overcome the limitations of classical ET (event tree)/FT (fault tree) based PSA (probabilistic safety assessment). The current paper also attempts to assert the reason why a technical transformation from classical PSA is necessary with a re-interpretation of the categories of risk. In this study, residual risk was classified into interpolating- and extrapolating-censored categories, which represent risks that are difficult to identify through an interpolation or extrapolation of representative scenarios due to potential nonlinearity between hardware and human behaviors intertwined in time and space. The authors hypothesize that such risk can be dealt with only if the classical ETs/FTs are freely relocated, entailing large-scale computation associated with physical models. The functional elements that are favorable to find residual risk were inferred from previous studies. The authors then introduce their under-development enabling techniques, namely DICE (Dynamic Integrated Consequence Evaluation) and DeBATE (Deep learning-Based Accident Trend Estimation). This work can be considered as a preliminary initiative to find the bridging points between deterministic and probabilistic assessments on the pillars of big data technology.

• 센서학회지
• /
• 제30권5호
• /
• pp.326-330
• /
• 2021

Reactions between gold (Au) surface plasmon resonance (SPR) chips and bovine serum albumin (BSA) dissolved in solutions of different pH were investigated. The charge on the BSA depends on the pH of the solution in which it is dissolved. Thus, dissolving BSA in different pH solutions resulted in different charges of BSA. Among the BSA dissolved in solutions with pH 4.01, 7.4, and 10.01, the SPR response was the highest for BSA dissolved in the solution of pH 4.01. To eliminate the response variation owing to the difference in the refractive indices of the solutions, phosphate buffered saline (PBS) was injected into the system after the reaction of BSA with the Au SPR chip had happened. In this case too, the BSA dissolved in the solution with pH 4.01 exhibited the highest response. This may be attributed to the non-uniform distribution of ionic patches on the BSA, which can induce electrostatic attraction to the surface even though BSA has a positive charge at pH 4.01, and the absolute values of the net charge of BSA at pH 4.01 and 7.4 were very close.

• Nuclear Engineering and Technology
• /
• 제53권6호
• /
• pp.2056-2065
• /
• 2021

The transition toward clean energy is an issue of great importance with growing debate in climate change mitigation. The complex nature of nuclear energy-CO₂ emissions nexus makes it difficult to predict whether or not nuclear acts as a clean energy source. Hence, we examined the relationship between nuclear energy consumption and CO₂ emissions in the context of the IPAT and Environmental Kuznets Curve (EKC) framework. Dynamic Auto-regressive Distributive Lag (DARDL), a newly modified econometric tool, is employed for estimation of long- and short-run dynamics by using yearly data spanning from 1971 to 2018. The empirical findings of the study revealed an instantaneous increase in nuclear energy reduces environmental pollution, which highlights that more nuclear energy power in the Indian energy system would be beneficial for climate change mitigation. The results further demonstrate that the overarching effect of population density in the IPAT equation stimulates carbon emissions. Finally, nuclear energy and population density contribute to form the EKC curve. To achieving a cleaner environment, results point out governmental policies toward the transition of nuclear energy that favours environmental sustainability.

• Nuclear Engineering and Technology
• /
• 제53권2호
• /
• pp.647-656
• /
• 2021

Knowing a material's true stress-strain curve is essential for performing a nonlinear finite element analysis to solve an elastoplastic problem. This study presents a simple methodology to determine the true stress-strain curve of type 304 and 316 austenitic stainless steels in the full range of strain from a typical tensile test. Before necking, the true stress and strain values are directly converted from engineering stress and strain data, respectively. After necking, a true stress-strain equation is determined by iteratively conducting finite element analysis using three pieces of information at the necking and the fracture points. The Hockett-Sherby equation is proposed as an optimal stress-strain model in a non-uniform deformation region. The application to the stainless steel under different temperatures and loading conditions verifies that the strain hardening behavior of the material is adequately described by the determined equation, and the estimated engineering stress-strain curves are in good agreement with those of experiments. The presented method is intrinsically simple to use and reduces iterations because it does not require much experimental effort and adopts the approach of determining the stress-strain equation instead of correcting the individual stress at each strain point.