• Proceedings of the KSME Conference
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• 2008.11b
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• pp.3175-3180
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• 2008

A metal-fueled pool-type liquid metal fast reactor (LMFR) provides large margins to sodium boiling and fuel damage under accident conditions. The favorable passive safety results are obtained by both a reactivity feedback mechanism in the core and a passive decay heat removal system. Among the various reactivity feedbacks, the ones by a thermal expansion of a radial dimension of the core and by the control rod drivelines are strongly dependent on the flow conditions in the core and the hot pool, respectively. The effects of multidimensional thermal hydraulic characteristics on these reactivity feedbacks are investigated by the system-wide safety analysis code SSC-K with advanced thermal hydraulics models. Particularly a detailed three dimensional thermal hydraulics reactor core model is integrated into SSC-K for use in a whole system analysis of the passive safety aspects of LMR designs. The model provides fuel and cladding temperatures for every fuel pin in a reactor and coolant temperatures for every coolant sub-channel in the reactor.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.248.1-248
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• 1999

Preliminary analysis of in-reactor thermal performance of three MOX fuel rods, which are going to be irradiated in the Halden reactor beginning in the first Quarter of the year 2000 under the framework of the OECD Halden Reactor Programme, have been conducted by using the computer code COSMOS to ensure their safe operation. Parametric studies have been carried out to investigate the effect of uncertainties on in-reactor behavior by considering the four kinds of uncertainties; thermal conductivity, linear power, manufacturing parameters, and model constants. The analysis shows that, in the case of annular MOX -1 fuel, calculation results for thermal performance vary widely depending on the selection of model constants for fission gas release (FGR). On the contrary, the thermal performance of solid MOX - 3 fuel does not depend on the choice of FGR constants to a large extent as MOX-I, because the fuel temperature is very high in the MOX-3 irrespective of the choice of FGR constants and hence the capacity of grain boundaries to retain gas atoms is not large enough to accommodate the number of gas atoms reaching the grain boundaries. It is planned that when the data on microstructure and thermal conductivity for each type of MOX fuel are available, new analysis will be made using these information. In addition, FGR model constants will be derived from the measured fuel centerline temperature, rod internal pressure and other related data.

• Journal of Applied Reliability
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• v.10 no.1
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• pp.57-64
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• 2010

The HANARO nuclear research reactor had been operated successfully and smoothly up to the year 2008 since its first year of instability in year 1995 just after the completion of construction. But the reliability of the reactor has been degraded from the year 2009 due to new experimental facilities such as Feul Test Loop(FTL) and Cold Neutron Source(CNS) which were installed in the HANARO plant. It turned out that these new facilities contributed unexpected stoppage of the plant. This paper describes causes of stoppage and suggestions to improve the reliability of the plant.

• Journal of the Korea Safety Management & Science
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• v.8 no.3
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• pp.67-75
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• 2006

Scheduling of dismantling old research reactor need to consider time, cost and safety for the worker. The biggest issue when dismantling facility for research reactor is safety for the worker and cost. Large portion of a budget is spending for the labor cost. To save labor cost for the worker, reducing a lead time is inevitable. Several algorithms applied to reduce read time, and safety considered as the most important factor for this project. This research presents three different dismantling scheduling scenarios. Best scenario shows the specific scheduling for worker and machine, so that it could save time and cost.

• Journal of the Korean Society for Nondestructive Testing
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• v.17 no.1
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• pp.31-40
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• 1997

원자로 압력용기 재료의 중성자 조사 취화 문제는 원자력발전소의 안전성 및 수명 관리에 가장 중대 한 영향을 미친다. 재료의 조사 취화를 평가하기 위하여 수행하고 있는 충격 및 인장시험 같은 파괴적 시험 결과는 석출물 크기나 분포, 전위 밀도 등, 재료 자체의 조직학적 특성에 좌우되므로 한정된 시편을 이용한 평가에는 많은 불확실성이 존재하게 된다. 따라서 이와 같은 문제점을 해결하기 위하여 비파괴기술을 이용한 조사 취화 평가에 대한 많은 연구가 진행되고 있다. 현재 원자로 압력용기 재료의 조사 취화에 따른 미세 조직 변화를 분석하기 위하여 응용되고 있는 비파괴기술로는 전기, 자기, 전자기, 초음파 및 경도측정법 등이 있으나 비파괴피험 결과와 미세조직의 변화, 기계적 성질 및 취화 정도 등과의 상관 관계를 정립해야만 기존 파괴적 시험의 대체가 가능하게 된다. 따라서 현재까지 수행되고 있는 여러 비파괴기술을 이용한 조사 취화 평가 연구결과를 비교 분석하여 보다 실현 가능성 있는 비파괴기술을 검토하였다.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.847-858
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• 2013

Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea) in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4-5 $g-U/cm^3$ were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr), additional protective coatings (silicide or nitride), and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

• Proceedings of the Korean Vacuum Society Conference
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• 2015.08a
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• pp.104-104
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• 2015

Due to the demand of the cold neutron flux in the neutron science and beam utilization technology, the cold neutron source (CNS) has been constructed and operating in the nuclear research reactor all over the world. The majority of the heat load removal scheme in the CNS is two-phase thermosiphon using the liquid hydrogen as a moderator. The CNS moderates thermal neutrons through a cryogenic moderator, liquid hydrogen, into cold neutrons with the generation of the nuclear heat load. The liquid hydrogen in a moderator cell is evaporated for the removal of the generated heat load from the neutron moderation and flows upward into a heat exchanger, where the hydrogen gas is liquefied by the cryogenic helium gas supplied from a helium refrigeration system. The liquefied hydrogen flows down to the moderator cell. To keep the required liquid hydrogen stable in the moderator cell, the CNS consists of an in-pool assembly (IPA) connected with the hydrogen system to handle the required hydrogen gas, the vacuum system to create the thermal insulation, and the helium refrigeration system to provide the cooling capacity. If one of systems is running out of order, the operating research reactor shall be tripped because the integrity of the CNS-IPA is not secured under the full power operation of the reactor. To prevent unscheduled reactor shutdown during a long time because the research reactor has been operating with the multi-purposes, the introduction of the standby cooling system (STS) can be a solution. In this presentation, the design considerations are considered how to design the STS satisfied with the following objectives: (a) to keep the moderator cell less than 350 K during the full power operation of the reactor under loss of the vacuum, loss of the cooling power, loss of common electrical power, or loss of instrument air cases; (b) to circulate smoothly helium gas in the STS circulation loop; (c) to re-start-up the reactor within 1 hour after its trip to avoid the Xenon build-up because more than certain concentration of Xenon makes that the reactor cannot start-up again; (d) to minimize the possibility of the hydrogen-oxygen reaction in the hydrogen boundary.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2006.10a
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• pp.33-34
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• 2006

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.28-36
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• 1993

Much effort is being done to standardize the PWR-type nuclear power plant in Korea. This paper presents the development of seismic design criteria for the reactor internals as a part of the standardization program for nuclear power plant. The seismic design loads of the reactor internals are calculated using the reference input motions of reactor vessel taken from Yong-gwang Nuclear Power Plant Units 3 and 4. An overview of analysis related to the basic parameters and methodologies is presented. Also, the response of internal components for the reactor vessel motions is carefully investigated.

• Korean Membrane Journal
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• v.5 no.1
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• pp.68-74
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• 2003

Methane steam reforming (MSR) reaction for hydrogen production was studied in a membrane reactor (MR) using two tubular membranes, one Pd-based and one of porous alumina. A higher methane conversion than the thermodynamic equilibrium for a traditional reactor (TR) was achieved using MRs. The experimental temperature range was 350-500$^{\circ}C$; no sweep-gas was employed during reaction tests to avoid its back-permeation through the membrane and the steam/methane molar feed ratio (m) varied in the range 3.5-5.9. The best results (the difference between the MR conversion and the thermodynamic equilibrium was of about 7%) were achieved with the alumina membrane, working with the highest steam/methane ratio and at 450$^{\circ}C$. Silica membranes prepared at KRICT laboratories were characterized with permeation tests on single gases (N$_2$, H$_2$ and CH$_4$). These membranes are suited for H$_2$ separation at high temperature.