• Journal of Preventive Medicine and Public Health
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• v.49 no.1
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• pp.53-60
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• 2016

Objectives: The purpose of this study is to examine the magnitude of and the factors associated with the downward mobility of first-episode psychiatric patients. Methods: This study used the claims data from the Korean Health Insurance Review and Assessment Service. The study population included 19 293 first-episode psychiatric inpatients diagnosed with alcohol use disorder (International Classification of Diseases, 10th revision [ICD-10] code F10), schizophrenia and related disorders (ICD-10 codes F20-F29), and mood disorders (ICD-10 codes F30-F33) in the first half of 2005. This study included only National Health Insurance beneficiaries in 2005. The dependent variable was the occurrence of downward mobility, which was defined as a health insurance status change from National Health Insurance to Medical Aid. Logistic regression analysis was used to assess factors associated with downward drift of first-episode psychiatric patients. Results: About 10% of the study population who were National Health Insurance beneficiaries in 2005 became Medical Aid recipients in 2007. The logistic regression analysis showed that age, gender, primary diagnosis, type of hospital at first admission, regular use of outpatient clinic, and long-term hospitalization are significant predictors in determining downward drift in newly diagnosed psychiatric patients. Conclusions: This research showed that the downward mobility of psychiatric patients is affected by long-term hospitalization and medical care utilization. The findings suggest that early intensive intervention might reduce long-term hospitalization and the downward mobility of psychiatric patients.

• Computers and Concrete
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• v.31 no.3
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• pp.253-265
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• 2023

In this research, the gene expression programming (GEP) technique was employed to provide a new model for predicting the maximum loading capacity of concrete-encased steel (CES) columns. This model was developed based on 96 CES column specimens available in the literature. The six main parameters used in the model were the compressive strength of concrete (f_c), yield stress of structural steel (f_ys), yield stress of steel rebar (f_yr), and cross-sectional areas of concrete, structural steel, and steel rebar (A_c, A_s and A_r respectively). The performance of the prediction model for the ultimate load-carrying capacity was investigated using different statistical indicators such as root mean square error (RMSE), correlation coefficient (R), mean absolute error (MAE), and relative square error (RSE), the corresponding values of which for the proposed model were 620.28, 0.99, 411.8, and 0.01, respectively. Here, the predictions of the model and those of available codes including ACI ITG, AS 3600, CSA-A23, EN 1994, JGJ 138, and NZS 3101 were compared for further model assessment. The obtained results showed that the proposed model had the highest correlation with the experimental data and the lowest error. In addition, to see if the developed model matched engineering realities and corresponded to the previously developed models, a parametric study and sensitivity analysis were carried out. The sensitivity analysis results indicated that the concrete cross-sectional area (A_c) has the greatest effect on the model, while parameter (f_yr) has a negligible effect.

• Journal of radiological science and technology
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• v.47 no.2
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• pp.107-116
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• 2024

In this study, we analyzed the use of general radiography imaging and effective dose in inpatients. Our aim is to help reduce national medical radiation exposure doses and develop rational health-care financial policies. The effective dose for each general radiography was calculated using the ALARA-GR program for 53 types (total: 260 codes) general radiography codes selected from 'National Health Insurance Care Benefit Cost'. The usage of general radiography was analyzed in the 2018 inpatient patient data of the Health Insurance Review and Assessment Service, and the effective dose for each general radiography was analyzed. 89.00% of inpatients undergo general radiography imaging at least once, with an average of 12.63 scans per person and an effective dose of 1.00 mSv. Those who received support from Medical Aid showed a higher value compared to those who were insured by National Health Insurance, with 17.39 cases and 1.43 mSv (p<.001). Chest had the highest usage rate at 23.12% for general radiography imaging, while L-spine had the highest effective dose at 24.53%. It is estimated that 420 inpatients patients undergo 121 to 820 general radiography imaging procedures per year, and 233 inpatients are estimated to have an annual effective dose of >20.00~58.25 mSv. Rational use of health-care finances and the practice of medical radiation safety management are essential for the well-being of individuals, the enhancement of quality of life, and the improvement of health-care quality.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2367-2375
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• 2022

In this work, we propose a new mechanistic model for the treatment of helium behaviour at the grain boundaries in oxide nuclear fuel. The model provides a rate-theory description of helium inter-granular behaviour, considering diffusion towards grain edges, trapping in lenticular bubbles, and thermal resolution. It is paired with a rate-theory description of helium intra-granular behaviour that includes diffusion towards grain boundaries, trapping in spherical bubbles, and thermal re-solution. The proposed model has been implemented in the meso-scale software designed for coupling with fuel performance codes SCIANTIX. It is validated against thermal desorption experiments performed on doped UO₂ samples annealed at different temperatures. The overall agreement of the new model with the experimental data is improved, both in terms of integral helium release and of the helium release rate. By considering the contribution of helium at the grain boundaries in the new model, it is possible to represent the kinetics of helium release rate at high temperature. Given the uncertainties involved in the initial conditions for the inter-granular part of the model and the uncertainties associated to some model parameters for which limited lower-length scale information is available, such as the helium diffusivity at the grain boundaries, the results are complemented by a dedicated uncertainty analysis. This assessment demonstrates that the initial conditions, chosen in a reasonable range, have limited impact on the results, and confirms that it is possible to achieve satisfying results using sound values for the uncertain physical parameters.

• Structural Engineering and Mechanics
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• v.31 no.5
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• pp.507-526
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• 2009

Based on numerical and experimental methods, a systematic structural evaluation of a steel natatorium in service was carried out in detail in this paper. Planning of inspection tasks was proposed firstly according to some national codes in China in order to obtain the economic and reliable results. The field visual inspections and static computation were conducted in turn under in-service environmental conditions. Further a three-dimensional finite element model was developed according to its factual geometry properties obtained from the field inspection. An analytical modal analysis was performed to provide the analytical modal properties. The field vibration tests on the natatorium were conducted and then two different system identification methods were used to obtain the dynamic characteristics of the natatorium. A good correlation was achieved in results obtained from the two system identification methods and the finite element method (FEM). The numerical and experimental results demonstrated that the main structure of the natatorium in its present status is safe and it still satisfies the demand of the national codes in China. But the roof system such as purlines and skeletons must be removed and rebuilt completely. Moreover the system identification results showed that field vibration test is sufficient to identify the reliable dynamic properties of the natatorium. The constructive suggestion on structural evaluation of the natatorium is that periodic assessment work must be maintained to ensure the natatorium's safety in the future.

• Journal of Radiation Protection and Research
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• v.48 no.2
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• pp.90-99
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• 2023

Background: Quantitative risk assessments should be accompanied by uncertainty analyses of the risk models employed in the calculations. In this study, we aim to develop a computational code named SUMRAY for use in cancer risk projections from radiation exposure taking into account uncertainties. We also aim to make SUMRAY publicly available as a resource for further improvement of risk projection. Materials and Methods: SUMRAY has two versions of code written in R and Python. The risk models used in SUMRAY for all-solid-cancer mortality and incidence were those published in the Life Span Study of a cohort of the atomic bomb survivors in Hiroshima and Nagasaki. The confidence intervals associated with the evaluated risks were derived by propagating the statistical uncertainties in the risk model parameter estimates by the Monte Carlo method. Results and Discussion: SUMRAY was used to calculate the lifetime or time-integrated attributable risks of cancer under an exposure scenario (baseline rates, dose[s], age[s] at exposure, age at the end of follow-up, sex) specified by the user. The results were compared with those calculated using another well-known web-based tool, Radiation Risk Assessment Tool (RadRAT; National Institutes of Health), and showed a reasonable agreement within the estimated confidential interval. Compared with RadRAT, SUMRAY can be used for a wide range of applications, as it allows the risk projection with arbitrarily specified risk models and/or population reference data. Conclusion: The reliabilities of SUMRAY with the present risk-model parameters and their variance-covariance matrices were verified by comparing them with those of the other codes. The SUMRAY code is distributed to the public as an open-source code under the Massachusetts Institute of Technology license.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1083-1095
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• 2016

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V&V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodiumshowed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

• International Journal of High-Rise Buildings
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• v.3 no.1
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• pp.65-71
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• 2014

Baku, the capital of Azerbaijan, has seen a boom in construction in recent years. The old Baku city has been rapidly transforming into a new hub of high-rise buildings and lively cultural centers hosting the Euro Vision Song Contest in 2012 and European Games in 2015. A major population shift to Baku from its suburbs and the countryside has resulted in the doubling of Baku's population in the 4 years between 2009 and 2013. As of January 2013, Baku's population reached four million people, 43% of the citizens in Azerbaijan according to The State Statistical Committee of Azerbaijan. With this trend, the city needs more high-rise buildings to accommodate rapidly increasing demands for more housing and business space. Until the Azerbaijan Seismic Building Code was published in 2010 and became effective, many different seismic criteria, in terms of building codes and seismic intensities, were used for all new high-rise projects in Baku. Some designers used the SNIP (Russian) code with seismic level 9 or level 8 with 1 point penalty. Others used the Turkish code with Seismic Zone 1, UBC 97 with Zone 2 through 4, or IBC with Sa = 0.75 g through 1.0 g. The seismic intensity is now clarified with the Azerbaijan Seismic Building Code. However, the Azerbaijan Seismic Building Code is appropriate for low-rise buildings applications but may be inappropriate for high-rise project applications. This is because the code-defined response spectrum yields unrealistically conservative seismic forces for high-rise buildings with long periods, as compared to those determined by other internationally accepted building codes. This paper provides observations and recommendations for code-based seismic load assessment of high-rise buildings in the Baku area.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.807-816
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• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.10
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• pp.1251-1259
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• 2013

In this study, high temperature design and creep-fatigue damage evaluation of a decay heat exchanger (DHX) in the decay heat removal systems of a sodium-cooled fast reactor (SFR) have been performed. Detail design and 3D finite element analysis have been conducted for the DHXs to be installed in active and passive decay heat removal systems in Korean Generation IV SFR, and the DHX installed in the STELLA-1(Sodium integral effect test loop for safety simulation and assessment) at KAERI (Korea Atomic Energy Research Institute). Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two Mod.9Cr-1Mo steel heat exchangers according to the elevated temperature design codes of ASME Section III Subsection NH and RCC-MR code. Code comparisons were made based on the creep-fatigue damage evaluation and issues on conservatisms of the design codes were discussed.