서울시건강가정지원센터 서울가족학교 아동기 부모교실의 효과성: 2020년 비대면 운영사례 분석

이재림 ( Lee¸ Jaerim ),노신애 ( Noh¸ Shinae ),박지수 ( Park¸ Jisu ) 한국가족관계학회 2021 한국가족관계학회지 Vol.26 No.3

연구목적: 공공영역의 가족생활교육 프로그램에 대한 국내의 효과성 연구가 매우 제한적인 상황에서, 서울시 전역의 건강가정지원센터에서 실시하고 있는 서울가족학교 아동기 부모교실의 효과성을 평가하고, 참여자 특성 및 교육 특성별 효과성의 차이를 살펴보았다. 연구방법: 2020년 6-10월에 서울시 14개 자치구 건강가정지원센터에서 실시한 총 17건의 비대면 아동기 부모교실에 참여했으며, 사전조사와 사후조사에 모두 응답한 213명의 자료를 paired t-test와 mixed ANOVA를 사용하여 분석하였다. 연구결과: 프로그램 참여자의 양육지식, 양육스트레스, 양육효능감이 프로그램 이전과 비교해 이후에 유의미하게 긍정적으로 변화하였다. 이 중 양육지식의 효과크기가 가장 컸다. 우울로 분류된 참여자의 변화폭이 그렇지 않은 참여자에 비해 컸으며, 컴퓨터나 태블릿 PC를 사용하여 교육에 참여한 집단의 양육지식이 스마트폰으로 참여한 집단보다 큰 폭으로 향상되었다. 결론: 코로나19로 건강가정지원센터의 교육사업이 비대면으로 전환된 2020년 아동기 부모교실의 효과성을 분석했다는 의의가 있으나, 후속연구에서는 무작위 대조군 연구와 장기적 효과성 추적이 필요하다. Objectives: We examined the effectiveness of the Childhood Parenting Education program, which is part of the Seoul Family School curricula offered at 25 local Healthy Family Support Centers throughout Seoul. We also examined whether the effectiveness differed by participant and delivery characteristics. This study is unique because of the focus on online delivery of family life education in public agencies during COVID-19. Method: Our data were collected from 213 parents who attended one of the 17 online parenting classes offered by 14 local centers between June and October 2020, and who also participated in both pre- and post-testing. We used paired t-tests and mixed ANOVA. Results: The participants’ parenting knowledge, efficacy, and stress significantly improved after attending the program. The effect size was the largest for parenting knowledge. Depressed participants experienced greater changes in all of the effectiveness indicators compared to non-depressed parents. Participants who used computers and tablet PCs to attend the classes reported gaining more parenting knowledge compared to those who used smart phones. Conclusion: This study is a critical step given that research on the effectiveness of public family life education in Korea is extremely limited. We suggest that future effectiveness research should be conducted using rigorous methods such as randomized controlled trials and longitudinal evaluation.

Has the Korean Kinship Become Bilateral? Gender Differences of Kinship Perceptions and Interactions

Jaerim Lee,Hee-Keum Cho 가정과삶의질학회 2011 한국가정관리학회 학술발표대회 자료집 Vol.2011 No.5

청년의 세대관계 유형화: 세대 간 결속의 하위차원을 중심으로

이재림 ( Jaerim Lee ),박제인 ( Jane Park ),김혜지 ( Hyeji Kim ),오상민 ( Sangmin Oh ),권소영 ( Soyoung Kwon ) 대한가정학회 2020 Human Ecology Research(HER) Vol.58 No.1

The literature on parent-child relationships in young adulthood lacks a systemic approach that considers the multifaceted nature of intergenerational relationships. In this paper, we identify the latent profiles of young Korean adults’ relationships with their parents based on the six dimensions of intergenerational solidarity (structural, associational, affectual, consensual, functional, and normative solidarity) as indicators. We considered solidarity for the mother and the father separately for structural, associational, affectual, and consensual dimensions. In terms of functional and normative solidarity, we measured both upstream and downstream intergenerational support. The sample included 1,015 young adults who were 19-34 years old, never married, not in secondary school, and had both parents living. The latent profile analysis revealed four profiles: (a) independent but intimate (22.7%), (b) coresident, outwardly intimate (32.4%), (c) detached from father (6.7%), and (d) coresident, intimate (38.2%). The factors that predicted each profile included the young adults’ education, income, subjective socioeconomic status, and experiencing the features of emerging adulthood along with the parents’ marital status, father’s employment, and overparenting. The “coresident, intimate” group reported significantly higher levels of life satisfaction and fewer depressive symptoms compared to the other three groups. The lowest levels of psychological adjustment were found in the “coresident, outwardly intimate” group and “detached from father” group.

Perceived Helicopter Parenting and Korean Emerging Adults’ Psychological Adjustment: The Mediational Role of Parent–Child Affection and Pressure from Parental Career Expectations

Lee, Jaerim,Kang, Sieun Springer US 2018 Journal of child and family studies Vol.27 No.11

<P>Examining helicopter parenting in a Confucian culture that values parental authority and involvement can extend previous helicopter parenting research that has mostly focused on a U.S. college student context. In this study, we aim to help clarify the mixed results in the literature regarding the psychological outcomes of helicopter parenting in emerging adulthood by examining the processes underlying the relationship between helicopter parenting and psychological adjustment in the Korean context. Using a diverse sample of Korean emerging adults, we investigated the relationships among perceived helicopter parenting, parent–child affection, pressure from parental career expectations, and psychological adjustment (depressive symptoms and life satisfaction). The data came from 562 Korean emerging adults (269 women and 293 men) aged 19–34 years who were either full-time students or unemployed and unmarried and had at least one living parent. The structural equation modeling used in this study revealed that higher levels of helicopter parenting were directly associated with greater depressive symptoms. Higher levels of helicopter parenting also had an indirect relationship with greater depressive symptoms through higher levels of pressure from parental career expectations. However, higher levels of helicopter parenting were indirectly linked to better psychological adjustment (fewer depressive symptoms, greater satisfaction with life) through higher levels of parent–child affection. Our results indicate that in the East Asian context, helicopter parenting could be related to both negative and positive psychological outcomes depending on the mediating factors.</P>

Validation of UNIST Monte Carlo code MCS for criticality safety analysis of PWR spent fuel pool and storage cask

Jang, Jaerim,Kim, Wonkyeong,Jeong, Sanggeol,Jeong, Eun,Park, Jinsu,Lemaire, Matthieu,Lee, Hyunsuk,Jo, Yongmin,Zhang, Peng,Lee, Deokjung Elsevier 2018 Annals of nuclear energy Vol.114 No.-

<P><B>Abstract</B></P> <P>This paper presents the validation of the continuous-energy Monte Carlo neutron-transport code MCS with the ENDF/B-VII.0 neutron cross-section library for the criticality safety analysis of PWR spent fuel pools and storage casks. The MCS code is developed by the COmputational Reactor physics and Experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for the analysis of Pressurized Water Reactors (PWRs) with high fidelity and high performance. The validation is conducted with 279 selected critical benchmarks from the International Criticality Safety Benchmark Evaluation Project (ICSBEP). The 279 validation cases are representative of PWR spent fuel pools and storage casks with <SUP>235</SUP>U enrichment ranging from 2.35 wt% to 4.74 wt%, pin pitch ranging from 1.075 cm to 2.540 cm, moderator to fuel ratio (H/U) ranging from 0.4683 and 11.5398, Energy of the Average Lethargy causing Fission (EALF) ranging from 0.0109 eV to 1.0600 eV, without soluble boron and with soluble boron concentration ranging from 0.015 g/L to 5.030 g/L. The calculation of the effective neutron multiplication factor by MCS is validated by the comparison between experiment and calculation for the selected critical benchmarks. The Upper Safety Limit (USL) of the MCS code is established in accordance to the NUREG/CR-6698 guideline recommended by the NRC (US National Regulatory Commission). The full validation process and determination of USL based on the selected critical benchmarks was also repeated with the MCNP6.1 and SERPENT2.1.27 codes in order to compare the performances of MCS with other reactor analysis codes. This paper demonstrates the capability of the MCS code for the criticality safety analysis of PWR spent fuel pools and storage casks.</P> <P><B>Highlights</B></P> <P> <UL> <LI> Validation of UNIST Monte Carlo code MCS with ENDF/B.VII.0 nuclear data library. </LI> <LI> Criticality safety analysis of PWR spent fuel pool and storage cask. </LI> <LI> Upper safety limits derived with single-sided tolerance band, limit and non-parametric methods. </LI> <LI> Code/code comparison against MCNP6.1 and SERPENT2.1.27 Monte Carlo codes. </LI> </UL> </P>

이륜 밸런싱 로봇에 대한 비선형 모델 기반 외란보상 기법

유재림(Jaerim Yu),김용국(Yongkuk Kim),권상주(SangJoo Kwon) 제어로봇시스템학회 2016 제어·로봇·시스템학회 논문지 Vol.22 No.10

A two-wheeled balancing mobile robot (TWBMR) has the characteristics of both nonlinear and underactuated system. In this paper, the disturbances acting on a TWBMR are classified into body disturbance and wheel disturbance. Additionally, we describe a nonlinear disturbance observer, which is suitable as a single input multi-output (SIMO) system for the longitudinal motion of TWBMR. Finally, we propose a reasonable disturbance compensation technique that combines the indirect reference input of equilibrium point and the direct torque compensation input. Simulations and experimental results show that the proposed disturbance compensation method is an effective way to achieve robust postural stability, specifically on inclined terrains.

Uncertainty quantification in decay heat calculation of spent nuclear fuel by STREAM/RAST-K

Jang, Jaerim,Kong, Chidong,Ebiwonjumi, Bamidele,Cherezov, Alexey,Jo, Yunki,Lee, Deokjung Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.9

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.

Validation of spent nuclear fuel decay heat calculation by a two-step method

Jang, Jaerim,Ebiwonjumi, Bamidele,Kim, Wonkyeong,Park, Jinsu,Choe, Jiwon,Lee, Deokjung Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.1

In this paper, we validate the decay heat calculation capability via a two-step method to analyze spent nuclear fuel (SNF) discharged from pressurized water reactors (PWRs). The calculation method is implemented with a lattice code STREAM and a nodal diffusion code RAST-K. One of the features of this method is the direct consideration of three-dimensional (3D) core simulation conditions with the advantage of a short simulation time. Other features include the prediction of the isotope inventory by Lagrange non-linear interpolation and the use of power history correction factors. The validation is performed with 58 decay heat measurements of 48 fuel assemblies (FAs) discharged from five PWRs operated in Sweden and the United States. These realistic benchmarks cover the discharge burnup range up to 51 GWd/MTU, 23.2 years of cooling time, and spanning an initial uranium enrichment range of 2.100-4.005 wt percent. The SNF analysis capability of STREAM is also employed in the code-to-code comparison. Compared to the measurements, the validation results of the FA calculation with RAST-K are within ±4%, and the pin-wise results are within ±4.3%. This paper successfully demonstrates that the developed decay heat calculation method can perform SNF back-end cycle analyses.

Room temperature cyclization of arylpropiolic acid anhydride: Synthesis of naphtho[2,3-<i>c</i>]furan-1,3-dione derivatives

Park, Jaerim,Kim, Ji Dang,Raja, Gabriel Charles Edwin,Choi, Hyun Chul,Lee, Sunwoo Informa UK (TaylorFrancis) 2017 Synthetic communications Vol.47 No.21

<P>Cyclic anhydrides such as naphtho[2,3-c]furan-1,3-dione derivatives were synthesized from the reaction of arylpropiolic acids and 2-chloro-4,6-dimethoxy-1,3,5-triazine in the presence of N-methylmorpholine at room temperature. This mild condition provided the naphtho[2,3-c]furan-1,3-dione derivatives in good yields. Spectroscopic analysis suggested that the formation of arylpropiolate is the rate-determining step.</P>

Verification and validation of isotope inventory prediction for back-end cycle management using two-step method

Jang, Jaerim,Ebiwonjumi, Bamidele,Kim, Wonkyeong,Cherezov, Alexey,Park, Jinsu,Lee, Deokjung Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.7

This paper presents the verification and validation (V&V) of a calculation module for isotope inventory prediction to control the back-end cycle of spent nuclear fuel (SNF). The calculation method presented herein was implemented in a two-step code system of a lattice code STREAM and a nodal diffusion code RAST-K. STREAM generates a cross section and provides the number density information using branch/history depletion branch calculations, whereas RAST-K supplies the power history and three history indices (boron concentration, moderator temperature, and fuel temperature). As its primary feature, this method can directly consider three-dimensional core simulation conditions using history indices of the operating conditions. Therefore, this method reduces the computation time by avoiding a recalculation of the fuel depletion. The module for isotope inventory calculates the number densities using the Lagrange interpolation method and power history correction factors, which are applied to correct the effects of the decay and fission products generated at different power levels. To assess the reliability of the developed code system for back-end cycle analysis, validation study was performed with 58 measured samples of pressurized water reactor (PWR) SNF, and code-to-code comparison was conducted with STREAM-SNF, HELIOS-1.6 and SCALE 5.1. The V&V results presented that the developed code system can provide reasonable results with comparable confidence intervals. As a result, this paper successfully demonstrates that the isotope inventory prediction code system can be used for spent nuclear fuel analysis.