• 한국동물학회지
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• 제37권3호
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• pp.318-323
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• 1994

The association of replication origins/termini with nuclear matrix during S phase was investigated by DNase digestion of halo structures in synchronized mouse LPI-1 cells. The binding of parental DNA to nuclear matrix was constant throughout S phase. When nuclear matrix was isolated from the cells pulse-labeled with 3H-thvmidine at various stases of S phase, total 3H-labels associated with nuclear matrix were specifically higher at So, Sa and Ss stages than other stases of S phase, suggesting that the newly synthesized DNAs at those stages are not excluded out of nuclear matrix. Similar patterns were obsenred from the pulse-chase experiments, in which cells were pulse-labeled at each stage of S phase and further incubated for 1 hr. These results suggest that the replication origins and termini are fixed at the nuclear matrix, and that the nuclear matrix binding fractions of DNA at 3C-pause may contain a large population of replication origins and termination sites.

• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2274-2284
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• 2020

The Nuclear Fuel Cycle Code "ANICCA" has been developed by SCK•CEN to answer particular questions about the Belgian nuclear fleet. However, the wide range of capabilities of the code make it also useful for international or regional studies that include advanced technologies and strategies of cycle. This paper shows the main features of the code and the facilities that can be simulated. Additionally, a comparison between several codes and ANICCA has also been made to verify the performance of the code by means of a simulation proposed in the last NEA (OECD) Benchmark Study. Finally, a case study of the Belgian nuclear fuel cycle phase out has been carried out to show the possible impact of the transmutation of the minor actinides on the nuclear waste by the use of an Accelerator Driven System also known as ADS. Results show that ANICCA accomplishes its main purpose of simulating the scenarios giving similar outcomes to other codes. Regarding the case study, results show a reduction of more than 60% of minor actinides in the Belgian nuclear cycle when using an ADS, reducing significantly the radiotoxicity and decay heat of the high-level waste and facilitating its management.

• Nuclear Engineering and Technology
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• 제45권6호
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• pp.821-826
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• 2013

In the present study, the effects of various heat treatments on the microstructure and mechanical properties of dual phase ODS steels were investigated to enhance the high strength at elevated temperature. Dual phase ODS steels have been designed by the control of ferrite and austenite formers, i.e., Cr, W and Ni, C in Fe-based alloys. The ODS steels were fabricated by mechanical alloying and a hot isostatic pressing process. Heat treatments, including hot rolling-tempering and normalizing-tempering with air- and furnace-cooling, were carefully carried out. It was revealed that the grain size and oxide distributions of the ODS steels can be changed by heat treatment, which significantly affected the strengths at elevated temperature. Therefore, the high temperature strength of dual phase ODS steel can be enhanced by a proper heat treatment process with a good combination of ferrite grains, nano-oxide particles, and grain boundary sliding.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2221-2228
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• 2021

One-dimensional oxygen transport relation is indispensable to study the oxygen distribution in the LBE-cooled system with an oxygen control device. In this paper, a numerical research is carried out to study the oxygen transport characteristics in a gas-phase oxygen control device, including the static case and dynamic case. The model of static oxygen control is based on the two-phase VOF model and the results agree well with the theoretical expectation. The model of dynamic oxygen control is simplified and the gas-liquid interface is treated as a free surface boundary with a constant oxygen concentration. The influences of the inlet and interface oxygen concentration, mass flow rate, temperature, and the inlet pipe location on the mass transfer characteristics are discussed. Based on the results, an oxygen mass transport relation considering the temperature dependence and velocity dependence separately is obtained. The relation can be used in a one-dimensional system analysis code to predict the oxygen provided by the oxygen control device, which is an important part of the integral oxygen mass transfer models.

• 한국콘텐츠학회논문지
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• 제19권6호
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• pp.222-235
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• 2019

2011년 발생한 후쿠시마 원전사고 이후 국제사회에서 원자력정책은 침체기에 접어들었으나 우리나라는 박근혜 정부까지 약 60여 년 동안 친원전 정책을 유지하는 경로의존성을 보여주었다. 하지만 2017년 탈원전 정책을 공약으로 내세운 문재인 정부 출범 이후 기존의 친원전 정책은 경로의존성에서 이탈하는 양상을 보이게 되었다. 본 연구는 역사적 제도주의를 중심으로 2008년 이명박 정부 시기부터 2018년 문재인 정부 까지 한국의 원자력정책 변동 과정을 고찰하였다. 연구결과 후쿠시마 원전 사고라는 외부 환경은 중대한 전환점이 되어 정부의 원자력 정책의 변동을 초래하였다. 제도적 맥락 측면에서 원자력정책의 패러다임, 정책결정구조, 법령 등에 영향을 미쳤고 행위자 측면에서도 정치 이념 및 국민 수용성에 영향을 주었다. 2017년 문재인 정부출범 이후 탈원전 정책이 제도적 기반을 확보하고 기존의 친원전 정책에서 탈원전으로 근본적인 정책 변동이 이루어지게 되었다. 이를 통해 문재인 정부의 에너지 정책은 이전 정부들의 원전 중심의 경로의존성에서 벗어나 단절적 균형이 발생하였고 재생에너지 중심으로 정책이 변동되었음을 확인할 수 있었다.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3892-3901
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• 2021

The pressure drop of a moisture separator in a steam generator is the important design parameter to ensure the successful performance of a nuclear power plant. The moisture separators have a wide range of operating conditions based on the arrangement of them. The prediction of the pressure drop in a moisture separator is challenging due to the complexity of the multi-dimensional two-phase vortex flow. In this study, the moisture separator test facility using the air/water two-phase flow was used to predict the pressure drop of a moisture separator in a Korean OPR-1000 reactor. The prototypical steam/water two-phase flow conditions in a steam generator were simulated as air/water two-phase flow conditions by preserving the centrifugal force and vapor quality. A series of experiments were carried out to investigate the effect of hydraulic characteristics such as the quality and liquid mass flux on the two-phase pressure drop. A new prediction model based on the scaling law was suggested and validated experimentally using the full and half scale of separators. The suggested prediction model showed good agreement with the steam/water experimental results, and it can be extended to predict the steam/water two-phase pressure drop for moisture separators.

• Nuclear Engineering and Technology
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• 제44권7호
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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.

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.827-837
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• 2015

Following a severe accident in a nuclear power plant, iodine is a major contributor to the potential health risks for the public. Because the amount of iodine released largely depends on its volatility, iodine's behavior in containment has been extensively studied in international programs such as International Source Term Programme-Experimental Program on Iodine Chemistry under Radiation (EPICUR), Organization for Economic Co-operation and Development (OECD)-Behaviour of Iodine Project, and OECD-Source Term Evaluation and Mitigation. Korea Institute of Nuclear Safety (KINS) has joined these programs and is developing a simplified, stand-alone iodine chemistry model, RAIM (Radio-Active Iodine chemistry Model), based on the IMOD methodology and other previous studies. This model deals with chemical reactions associated with the formation and destruction of iodine species and surface reactions in the containment atmosphere and the sump in a simple manner. RAIM was applied to a simulation of four EPICUR tests and one Radioiodine Test Facility test, which were carried out in aqueous or gaseous phases. After analysis, the results show a trend of underestimation of organic and molecular iodine for the gas-phase experiments, the opposite of that for the aqueous-phase ones, whereas the total amount of volatile iodine species agrees well between the experiment and the analysis result.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.59-68
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• 2020

ECCMIX component was introduced in RELAP5/MOD3 for calculating the interfacial condensation. Compared to other existing components in RELAP5, user experience of ECCMIX component is restricted to developmental assessment applications. To evaluate the capability of the ECCMIX component, ECCS experiment was conducted which included single-phase and two-phase thermal mixing. The experiment was carried out with test sections containing a main pipe (70 mm inner diameter) and a branch pipe (21 mm inner diameter) under the atmospheric pressure. The steam mass flow in the main pipe ranged from 0 to 0.0347 kg/s, and the subcooled water mass flow in the branch pipe ranged from 0.0278 to 0.1389 kg/s. The comparison of the experimental data with the calculation results illuminated that although the ECCMIX component was more difficult to converge than Branch component, it was a more appropriate manner to simulate interfacial condensation under two-phase thermal mixing circumstance, while the two components had no differences under single-phase circumstance.

• Nuclear Engineering and Technology
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• 제44권5호
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• pp.551-560
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• 2012

The Wolsong-1 (W-1) Phase-B pre-simulations were carried out in preparation for tests to be conducted for the restart of the reactor after a major refurbishment project that included replacement of the pressure tube. These pre-simulations for Wolsong-1 Phase-B differ from those in the past that were performed for the Wolsong-1,2,3,4 tests in that these tests use the WIMS/DRAGON/RFSP-IST code suite for verification of the tests and gadolinium instead of the traditional PPV/MULTICELL/RFSP code system and boron as poison in the moderator system. The use of gadolinium is deemed not to have domestically accumulated experience gained from the previous Phase-B tests. Thus, it is appropriate to conduct a study in order to gain a correct understanding and interpretation of potential differences in test results stemming from using gadolinium rather than boron. Although the calibration of the reactivity device will not be noticeably different using boron and gadolinium at a constant moderator temperature, the temperature dependency of the neutronic behavior due to the presence of gadolinium in the moderator system might be pronounced. The results of the pre-simulations using gadolinium revealed that the moderator temperature reactivity coefficients indeed showed significant differences in comparison with those with boron. In order to secure the validity of the analysis results, the newly acquired WIMS/DRAGON/RFSP-IST code suite was verified against the W-2,3,4 Phase-B test results. The results of the new code suite verifications revealed some overall improvements in accuracy; justification of the use of the code can be claimed for the validation of the W-1 Phase-B test results.