• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 추계 학술대회 논문집
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• pp.47-55
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• 2005

가상의 연구로 해체 시설 환경을 설계하는데 필요한 단위 프로그램들의 특성을 검토한 후 결과 자료를 바탕으로 해체 디지털 목업 시스템의 설계가 완료되었다. 단위 프로그램들은 해체 데이터베이스 시스템 모듈 연구로 시설과 제염 및 해체 장비를 3차원으로 모델링하는 모듈, 3차원으로 방사능 오염 분포도를 묘사하는 모듈, 그리고 해체 일정 및 해체 비용을 평가하는 모듈 등으로 구분된다. 독립적으로 운영되는 이들 단위 모듈들을 통합된 시스템으로 만들기 위해 단위 모듈들의 아키텍쳐 설계 연구가 수행되었다. 연구 결과 다양한 모듈들로 구성된 해체 디지털 목업 시스템을 통합된 환경에서 시나리오를 시험 평가할 수 있도록 하기위해 연구로 시설의 제염 및 해체 활동을 시각적으로 보여줄 수 있는 가시화(visualization) 모듈과 해체 일정 및 해체 비용을 평가하고 분석하는 시뮬레이션(simulation) 모듈로 해체 디지털 목업 시스템의 아키텍쳐를 구현하였다.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.937-946
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• 2020

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.

• Journal of Radiation Protection and Research
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• 제41권3호
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• pp.237-244
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• 2016

Background: Korea Atomic Energy Research Institute is developing a fission product transport module for predicting the behavior of radioactive materials in the primary cooling system of a nuclear power plant as a separate module, which will be connected to a severe accident analysis code, Core Meltdown Progression Accident Simulation Software (COMPASS). Materials and Methods: This fission product transport (COMPASS-FP) module consists of a fission product release model, an aerosol generation model, and an aerosol transport model. In the fission product release model there are three submodels based on empirical correlations, and they are used to simulate the fission product gases release from the reactor core. In the aerosol generation model, the mass conservation law and Raoult's law are applied to the mixture of vapors and droplets of the fission products in a specified control volume to find the generation of the aerosol droplet. In the aerosol transport model, empirical correlations available from the open literature are used to simulate the aerosol removal processes owing to the gravitational settling, inertia impaction, diffusiophoresis, and thermophoresis. Results and Discussion: The COMPASS-FP module was validated against Aerosol Behavior Code Validation and Evaluation (ABCOVE-5) test performed by Hanford Engineering Development Laboratory for comparing the prediction and test data. The comparison results assuming a non-spherical aerosol shape for the suspended aerosol mass concentration showed a good agreement with an error range of about ${\pm}6%$. Conclusion: It was found that the COMPASS-FP module produced the reasonable results of the fission product gases release, the aerosol generation, and the gravitational settling in the aerosol removal processes for ABCOVE-5. However, more validation for other aerosol removal models needs to be performed.

• 대한전자공학회:학술대회논문집
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• 대한전자공학회 2003년도 하계종합학술대회 논문집 V
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• pp.2545-2548
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• 2003

Reliability is the very important issue for nuclear fields. In this paper, an analysis method is suggested to evaluate the level of availability improvement by adding the fault diagnosis function in the control system of Reactor Protection System. The Failure Mode Effect Analysis(FMEA), MIL-HDBK-217F, and Makov modelling techniques are used for availability assessment.

• 한국물환경학회지
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• 제25권4호
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• pp.502-506
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• 2009

The removal properties of Polyethylene glycols (PEGs) known as the important group of synthetic polymers of ethylene oxide were examined by the bio-reactor equipped with a mesh filter module. PEG-1000 and PEG-2000 were fairly removed on the basis of TOC, which were 75.1% and 51.6%, respectively. In the case of PEG-20000, the removal efficiency of TOC was less than 15.2% and the favorable acclimation of microbes was not obtained. It was suggested that this system could effectively maintain microbes for the biodegradation of low molecular weight of PEG and TOC removal was significantly influenced by PEG molecular weight.

• 한국지하수토양환경학회지:지하수토양환경
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• 제25권3호
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• pp.23-31
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• 2020

The performance of a downward-irradiation sonoreactor was investigated using calorimetry, KI dosimetry, luminol (Sonochemiluminescence, SCL) method, and aluminium foil erosion method as one of the basic steps for the optimal design of downward-irradiation sonoreactors. The applied frequency was 28 kHz and the input electrical power was 280 - 300 W. The liquid height, from the reactor bottom to the transducer module surface, ranged from 1λ (53.6 mm) to 2λ (107.1 mm). For various liquid heights, the magnitude of calorimetric power and the mass of cavitation-generated I₃^- ion varied significantly. It was found that the additional application of mechanical mixing resulted in higher sonochemical activity, especially in the cavitational active zone, which was induced by violent liquid flow in the reactor. In aluminium foil erosion tests, it was found that less ultrasound energy reached the bottom of the reactor due to the violent liquid flow and no significant sonophysical effect was observed for higher mixing rate conditions (100 and 200 rpm).

• Nuclear Engineering and Technology
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• 제41권3호
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• pp.307-318
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• 2009

KAERI (Korea Atomic Energy Research Institute) has developed the GAMMA+ code for a thermo-fluid and safety analysis of a VHTR (Very High Temperature Gas-Cooled Reactor). A key safety issue of the VHTR design is to demonstrate its inherent safety features for an automatic reactor power trip and power stabilization during an anticipated transient without scram (ATWS) accident such as a loss of forced cooling by a trip of the helium circulator (LOFC) or a reactivity insertion by a control rod withdrawal (CRW). This paper intends to show the ATWS assessment capability of the GAMMA+ code which can simulate the reactor power response by solving the point-kinetic equations with six-group delayed neutrons, by considering the reactivity changes due to the effects of a core temperature variation, xenon transients, and reactivity insertions. The present benchmark calculations are performed by using the safety demonstration experiments of the 10 MW high temperature gas cooled-test module (HTR-10) in China. The calculation results of the power response transients and the solid core temperature behavior are compared with the experimental data of a LOFC ATWS test and two CRW ATWS tests by using a 1mk-control rod and a 5mk-control rod, respectively. The GAMMA+ code predicts the power response transients very well for the LOFC and CRW ATWS tests in HTR-10.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2803-2815
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• 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
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• 제53권9호
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• pp.2816-2829
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• 2021

Decay heat residuals of spent nuclear fuel (SNF), i.e., the differences between calculations and measurements, were obtained previously for various spent fuel assemblies (SFA) using the Polaris module of the SCALE code system. In this paper, we compare decay heat residuals to their uncertainties, focusing on four PWRs and four BWRs. Uncertainties in nuclear data and model inputs are propagated stochastically through calculations using the SCALE/Sampler super-sequence. Total uncertainties could not explain the residuals of two SFAs measured at GE-Morris. The combined z-scores for all SFAs measured at the Clab facility could explain the resulting deviations. Nuclear-data-related uncertainties contribute more in the high burnup SFAs. Design and operational uncertainties tend to contribute more to the total uncertainties. Assembly burnup is a relevant variable as it correlates significantly with the SNF decay heat. Additionally, burnup uncertainty is a major contributor to decay heat uncertainty, and assumptions relating to these uncertainties are crucial. Propagation of nuclear data and design and operational uncertainties shows that the analyzed assemblies respond similarly with high correlation. The calculated decay heats are highly correlated in the PWRs and BWRs, whereas lower correlations were observed between decay heats of SFAs that differ in their burnups.

• Nuclear Engineering and Technology
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• 제54권4호
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• pp.1508-1515
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• 2022

In the present study, the cavity module of the MELCOR code is used for the simulation of molten corium concrete interaction (MCCI) during the late phase of postulated large break loss of coolant (LB-LOCA) accident in the APR1400 reactor design. Using the molten corium composition data from previous MELCOR Simulation of APR1400 under LB-LOCA accident, the ex-vessel phases of the accident sequences with long-term MCCI are recalculated with stand-alone cavity package of the MELCOR code to investigate the impact of water ingression and melt eruption models which were hitherto absent in MELCOR code. Significant changes in the MCCI behaviors in terms of the heat transfer rates, amount of gases released, and maximum cavity ablation depths are observed and reported in this study. Most especially, the incorporation of these models in the new release of MELCOR code has led to the reduction of the maximum ablation depth in radial and axial directions by ~38% and ~32%, respectively. These impacts are substantial enough to change the conclusions earlier reached by researchers who had used the older versions of the MELCOR code for their studies. and it could also impact the estimated cost of the severe accident mitigation system in the APR1400 reactor.