• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.681-688
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• 2022

Dose monitoring in CT patients requires accurate dose estimation but most of the CT dose calculation tools are based on Caucasian computational phantoms. We established a library of organ dose conversion coefficients for Korean adults by using four Korean adult male and two female voxel phantoms combined with Monte Carlo simulation techniques. We calculated organ dose conversion coefficients for head, chest, abdomen and pelvis, and chest-abdomen-pelvis scans, and compared the results with the existing data calculated from Caucasian phantoms. We derived representative organ doses for Korean adults using Korean CT dose surveys combined with the dose conversion coefficients. The organ dose conversion coefficients from the Korean adult phantoms were slightly greater than those of the ICRP reference phantoms: up to 13% for the brain doses in head scans and up to 10% for the dose to the small intestine wall in abdominal scans. We derived Korean representative doses to major organs in head, chest, and AP scans using mean CTDI_vol values extracted from the Korean nationwide surveys conducted in 2008 and 2017. The Korean-specific organ dose conversion coefficients should be useful to readily estimate organ absorbed doses for Korean adult male and female patients undergoing CT scans.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.689-700
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• 2022

In the present study, iodine-131 S values (r_T ← thyroid) were calculated for 30 target organs and tissues using the most recently developed Korean reference computational phantoms. The calculated S values were then compared with those of the International Commission on Radiological Protection (ICRP) reference computational phantoms to investigate the dosimetric impact of the Korean S values against those of the ICRP reference phantoms. The results showed significant differences in the S values due to the different anatomical/morphological characteristics between the Korean and ICRP reference phantoms. Most target organs/tissues showed that the S values of the Korean reference phantoms are lower than those of the ICRP reference phantoms, by up to about 4 times (male spleen and female thymus). Exceptionally, three target organs/tissues (gonads, thyroid, and extrathoracic region) showed that the S values of the Korean reference phantoms are greater, by 1.5-3.7 times. We expect that the S values calculated in the present study will be beneficially used to estimate organ/tissue doses of Korean patients under radioiodine therapy.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.896-908
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• 2022

As a part of probabilistic risk (or safety) assessment (PRA or PSA) of nuclear power plants (NPPs), the primary role of human reliability analysis (HRA) is to provide credible estimations of the human error probabilities (HEPs) of safety-critical tasks. In this regard, it is vital to provide credible HEPs based on firm technical underpinnings including (but not limited to): (1) how to collect HRA data from available sources of information, and (2) how to inform HRA practitioners with the collected HRA data. Because of these necessities, the U.S. Nuclear Regulatory Commission and the Korea Atomic Energy Research Institute independently developed two dedicated HRA data collection systems, SACADA (Scenario Authoring, Characterization, And Debriefing Application) and HuREX (Human Reliability data EXtraction), respectively. These systems provide unique frameworks that can be used to secure HRA data from full-scope training simulators of NPPs (i.e., simulator data). In order to investigate the applicability of these two systems, two papers have been prepared with distinct purposes. The first paper, entitled "SACADA and HuREX: Part 1. The Use of SACADA and HuREX Systems to Collect Human Reliability Data", deals with technical issues pertaining to the collection of HRA data. This second paper explains how the two systems are able to inform HRA practitioners. To this end, the process of estimating HEPs is demonstrated based on feed-and-bleed operations using HRA data from the two systems.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.926-939
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• 2022

The area of this study will cover the location-wise seismic response variation of an electrical cabinet in nuclear power point (NPP) based on classical reliability analysis. The location-based seismic ground motion (GM) selection is carried out with the help of probabilistic seismic hazard analysis using PSHRisktool, where the variation of reliability analysis can be understood from the relation between the reliability index and intensity measure. Two different approaches such as the first-order second moment method (FOSM) and Monte Carlo Simulation (MCS) are helped to evaluate and compare the reliability assessment of the cabinet. The cabinet is modeled with material uncertainty utilizing Steel01 as the material model and the fiber section modeling approach is considered to characterize the section's nonlinear reaction behavior. To verify the modal frequency, this study compares the FEM result with recorded data using Least-Squares Complex Exponential (LSCE) method from the impact hammer test. In spite of a few investigations, the main novelty of this study is to introduce the reader to check and compare the seismic reliability assessment variation in different seismic locations and for different earthquake levels. Alongside, the betterment can be found by comparing the result between two considered reliability estimation methods.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.72-83
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• 2022

This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt. The coolability limit due to external reactor vessel cooling is associated with the natural circulation flow rate around the lower head of the reactor vessel. For an elaborate prediction of the natural circulation flow rate using a thermal-hydraulic system code, MARS-KS1.5, a three-dimensional computational fluid dynamics (CFD) simulation is conducted to estimate the flow rate and pressure distribution of a liquid-state coolant at the brink of significant void generation. The CFD calculation results are used to determine the loss coefficient at major flow junctions, where substantial pressure losses are expected, in the nodalization scheme of the MARS-KS code such that the single-phase flow rate is the same as that predicted via CFD simulations. Subsequently, the MARS-KS analysis is performed for the two-phase natural circulation regime, and the transient behavior of the main thermal-hydraulic variables is investigated.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2625-2634
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• 2022

Discretization errors are extremely challenging conundrums of discrete ordinates calculations for radiation transport problems with void regions. In previous work, we have presented a multi-collision source method (MCS) to overcome discretization errors, but the efficiency needs to be improved. This paper proposes a goal-oriented algorithm for the MCS method to adaptively determine the partitioning of the geometry and dynamically change the angular quadrature in remaining iterations. The importance factor based on the adjoint transport calculation obtains the response function to get a problem-dependent, goal-oriented spatial decomposition. The difference in the scalar fluxes from one high-order quadrature set to a lower one provides the error estimation as a driving force behind the dynamic quadrature. The goal-oriented algorithm allows optimizing by using ray-tracing technology or high-order quadrature sets in the first few iterations and arranging the integration order of the remaining iterations from high to low. The algorithm has been implemented in the 3D transport code ARES and was tested on the Kobayashi benchmarks. The numerical results show a reduction in computation time on these problems for the same desired level of accuracy as compared to the standard ARES code, and it has clear advantages over the traditional MCS method in solving radiation transport problems with reflective boundary conditions.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3299-3309
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• 2022

Nuclear energy has the potential to play an influential role in energy transition efforts than is now anticipated by many countries. Realizing sustainable human development and reducing global climate crises will become more difficult without significantly increasing nuclear power. This paper aims to probe the role of nuclear energy, external debt, and financial globalization in sustaining human development and environmental conditions simultaneously in BRICS (Brazil, Russia, India, China, and South Africa) countries. This study applied a battery of second-generation estimation approaches over the period from 1990 to 2019. These methods are useful and robust to cross-countries dependencies, slope heterogeneity, parameters endogeneity, and serial correlation that are ignored in conventional approaches to generate more comprehensive and reliable estimates. The empirical findings indicate that nuclear energy and financial globalization contribute to human development, whereas external debt inhibits it. Similarly, financial globalization accelerates ecological deterioration, but nuclear energy and external debt promote environmental sustainability. Moreover, the study reveals bidirectional feedback causalities between human development, carbon emissions and nuclear energy consumption. The study offers useful policy guidance on accomplishing sustainable and inclusive development in BRICS countries.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.1085-1097
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• 2022

A printed circuit steam generator (PCSG) is being considered as the component for pressurized water reactor (PWR) type small modular reactor (SMR) that can further reduce the physical size of the system. Since a steam generator in many PWR-type SMR generates superheated steam, it is expected that dryout front oscillation can potentially cause thermal fatigue failure due to cyclic thermal stresses induced by the transition in boiling regimes between convective evaporation and film boiling. To investigate the fatigue issue of a PCSG, a reference PCSG is designed in this study first using an in-house PCSG design tool. For the stress analysis, a finite element method analysis model is developed to obtain the temperature and stress fields of the designed PCSG. Fatigue estimation is performed based on ASME Boiler and pressure vessel code to identify the major parameters influencing the fatigue life time originating from the dryout front oscillation. As a result of this study, the limit on the temperature difference between the hot side and cold side fluids is obtained. Moreover, it is found that the heat transfer coefficient of convective evaporation and film boiling regimes play an essential role in the fatigue life cycle as well as the temperature difference.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1331-1344
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• 2021

This study is to evaluate the seismic capacity of the fire-damaged cabinet facility in a nuclear power plant (NPP). A prototype of an electrical cabinet is modeled using OpenSees for the numerical simulation. To capture the nonlinear behavior of the cabinet, the constitutive law of the material model under the fire environment is considered. The experimental record from the impact hammer test is extracted trough the frequency-domain decomposition (FDD) method, which is used to verify the effectiveness of the numerical model through modal assurance criteria (MAC). Assuming different temperatures, the nonlinear time history analysis is conducted using a set of fifty earthquakes and the seismic outputs are investigated by the fragility analysis. To get a threshold of intensity measure, the Monte Carlo Simulation (MCS) is adopted for uncertainty reduction purposes. Finally, a capacity estimation model has been proposed through the investigation, which will be helpful for the engineer or NPP operator to evaluate the fire-damaged cabinet strength under seismic excitation. This capacity model is presented in terms of the High Confidence of Low Probability of Failure (HCLPF) point. The results are validated by the proper judgment and can be used to analyze the influences of fire on the electrical cabinet.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2969-2974
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• 2020

Looking at the recent studies, nuclear energy and carbon dioxide (CO₂) emissions nexus shows inconclusive result. To further explain nuclear energy-pollution nexuses this study is an attempt to analyze the impact of nuclear energy on pollution reduction for BRICS countries covering data for the period from 1993 to 2017. This study conducts advanced panel techniques such as Continuously-Updated Fully-Modified (CUP-FM) and Continuously-Updated Bias-Corrected (CUP-BC) for long run estimation. Our results support the notion that nuclear energy reduce CO₂ emissions. Also, renewable energy corrects environmental pollution in BRICS countries. The magnitude of the coefficient of nuclear energy is less as compared to renewable energy, implying that nuclear is less effective in reducing environmental pollution. The findings offer significant policy understandings and suggestions not only for BRICS economies but for developing countries as well in designing suitable nuclear energy-growth-carbon policies.