• Nuclear Engineering and Technology
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• 제38권5호
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• pp.437-442
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• 2006

European Material Test Reactors (MTRs) have provided essential support for nuclear power programs over the last 40 years. MTRs are now ageing in Europe and they cannot ensure the securing of experimental capability for the next decades. In this context, a new Material Testing Reactor, named Jules Horowitz Reactor -JHR-, operated as an international user-facility, is under development in Europe. The European MTRs context and the JHR objectives and status will be presented. Emphasis will be put on experiments in the field of nuclear fuels and materials irradiation which are developed in the framework of European and international collaboration.

• Nuclear Engineering and Technology
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• 제49권4호
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• pp.700-709
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• 2017

This work looks at the effect of changes in kinetic parameters on simultaneous reactivity insertions and beam tube flooding in a typical material testing reactor-type research reactor with low enriched high density ($U_3Si_2-Al$) fuel. Using a modified PARET code, various ramp reactivity insertions (from $0.1/0.5 s to $1.3/0.5 s) plus beam tube flooding ($0.5/0.25 s) accidents under uncontrolled conditions were analyzed to find their effects on peak power, net reactivity, and temperature. Then, the effects of changes in kinetic parameters including the Doppler coefficient, prompt neutron lifetime, and delayed neutron fractions on simultaneous reactivity insertion and beam tube flooding accidents were analyzed. Results show that the power peak values are significantly sensitive to the Doppler coefficient of the system in coupled accidents. The material testing reactor-type system under such a coupled accident is not very sensitive to changes in the prompt neutron life time; the core under such a coupled transient is not very sensitive to changes in the effective delayed neutron fraction.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1881-1895
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• 2020

The theoretical aspects behind the reactor depletion capability of the Monte Carlo code MCS developed at the Ulsan National Institute of Science and Technology (UNIST) and practical results of this depletion feature for a Material-Testing Reactor (MTR) with plate-type fuel are described in this paper. A verification of MCS results is first performed against MCNP6 to confirm the suitability of MCS for the criticality and depletion analysis of the MTR. Then, the dependence of the effective neutron multiplication factor to the number of axial and radial depletion cells adopted in the fuel plates is performed with MCS in order to determine the minimum spatial segmentation of the fuel plates. Monte Carlo depletion results with 37,800 depletion cells are provided by MCS within acceptable calculation time and memory usage. The results show that at least 7 axial meshes per fuel plate are required to reach the same precision as the reference calculation whereas no significant differences are observed when modeling 1 or 10 radial meshes per fuel plate. This study demonstrates that MCS can address the need for Monte Carlo codes capable of providing reference solutions to complex reactor depletion problems with refined meshes for fuel management and research reactor applications.

• Nuclear Engineering and Technology
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• 제38권5호
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• pp.417-422
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• 2006

The present paper describes the 45 MW High Flux Reactor (HFR) which is located in Petten, The Netherlands. This paper focuses on selected technical aspects of this reactor and on nuclear fuel irradiation experiments. These fuel experiments are mainly experiments on Light Water Reactor (LWR) and Very/High Temperature Reactor (V/HTR) fuels, but also on Fast Reactor (FR) fuels, transmutation fuels and Material Test Reactor (MTR) fuels.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.953-967
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• 2017

In this paper, a complete station blackout (SBO) or complete loss of electrical power supplies is simulated and analyzed in a downward cooling 5-MW pool-type Material Testing Reactor (MTR). The scenario is traced in the absence of active cooling systems and operators. The code nodalization is successfully benchmarked against experimental data of the reactor's operating parameters. The passive heat removal system includes downward water cooling after pump breakdown by the force of gravity (where the coolant streams down to the unfilled portion of the holdup tank), safety flapper opening, flow reversal from a downward to an upward cooling direction, and then the upward free convection heat removal throughout the flapper safety valve, lower plenum, and fuel assemblies. Both short-term and long-term natural core cooling conditions are simulated and investigated using the RELAP5 code. Short-term analyses focus on the safety flapper valve operation and flow reversal mode. Long-term analyses include simulation of both complete SBO and long-term operation of the free convection mode. Results are promising for pool-type MTRs because this allows operators to investigate RELAP code abilities for MTR thermal-hydraulic simulations without any oscillation; moreover, the Tehran Research Reactor is conservatively safe against the complete SBO and long-term free convection operation.

• Corrosion Science and Technology
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• 제20권4호
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• pp.189-195
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• 2021

A technology for designing and licensing a dedicated radiation shielding facility needs to be developed for safe and efficient operation an R&D center. Technology development is important for smooth operation of such facilities. Causes of damage to internal structures (such as baffle former bolt (BFB) of pressurized water reactor) of a nuclear power reactor should be analyzed along with prevention and countermeasures for similar cases of other plants. It is important to develop technologies that can comprehensively analyze various characteristics of internal structures of long term operated reactors. In high-temperature, high-pressure operating environment of nuclear power plants, cases of BFB cracks caused by irradiated assisted stress corrosion cracks (IASCC) have been reported overseas. The integrity of a reactor's internal structure has emerged as an important issue. Identifying the cause of the defect is requested by the Korean regulatory agency. It is also important to secure a foundation for testing technology to demonstrate the operating environment for medium-level irradiated testing materials. The demonstration testing facility can be used for research on material utilization of the plant, which might have highest fluence on the internal structure of a reactor globally.

• 한국환경과학회지
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• 제27권6호
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• pp.467-475
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• 2018

The lifetime of the electrode is one of the most important factors on the stability of the electrode. Since the lifetime of the DSA (Dimensionally stable anode) electrode is long, an accelerated lifetime test is required to reduce the test time. Beacuse there is no basis or standard method for accelerated lifetime testing, many researchers use different methods. Therefore, there is a need for basis and methods for accelerated lifetime testing that other researchers can follow. We designed a reactor system for accelerated lifetime testing and planned specific methods. Reactor system was circulating batch reactor. Reactor volume and cooling water tank were 12.5 L and 100 L, respectively. Electrode size was $2cm{\times}3cm$ (real electrolysis area, $5cm^2$). In order to maintain the harsh conditions, accelerated lifetime test was carried out in a high current density ($0.6A/cm^2$) and low electrolyte concentration (NaCl, 0.068 mol/L). Maintaining a constant temperature was an important operation parameter for exact accelerated lifetime test. As the accelerated lifetime test progressed, the active component of electrode surface was consumed and desorption occurred. At the point of 5 V rise, corrosion of the surface of the base material(titanium) also started.

• 비파괴검사학회지
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• 제25권5호
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• pp.377-383
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• 2005

석유화학 및 정유설비는 고온이나 고압에서 폭발 위험성을 지닌 유체를 사용하기 때문에 방재기술에 관한 관심이 높다. 이들 설비 중에서도 고압반응기는 특히 고온/고압 하에서 사용되므로 안전성이 요구된다. 본 연구에서는 석유화학 플랜트의 고압반응기 소재로 많이 사용되고 있는 2.25Cr-1Mo강을 대상으로 하였으며, 3가지 온도조건에서 열화시간을 달리하여 총 8 종류의 인공열화 시험편을 준비하였다. 열화는 고압반응기의 사용온도인 $391^{\circ}C$보다 약간 높은 온도에서 둥온 열처리하였다. 미열화재를 포함하여 인공열화재에 대해 비커스경도값과 전기비저항값을 측정하였으며, Larson-Miller parameter와의 상관관계로부터 master curve를 작성하였다. 그리고 현장의 고압반응기에서 비커스경도와 전기비저항을 측정하여 실험실에서 작성한 master curve와 비교하였다. 전기비저항법을 이용한 고압반응기의 현장에서의 열화평가 가능성을 검토하였으며, 현장에서 측정한 전기비저항은 비슷한 열화수준에서의 인공열화재의 전기비저항값과 비슷하였다.

• 한국초전도ㆍ저온공학회논문지
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• 제4권1호
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• pp.30-34
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• 2002

In this Paper, Insulation of current lead in the conduction-cooled DC reactor for the 1.2kV class 3 high-Tc superconducting fault current limiter(SFCL) is studied. Thermal link which conducts heat energy but insulates electrical energy is selected as a insulating device for the current lead in the conduction-cooled Superconducting DC reactor. It consists of oxide free copper(OFC) sheets, Polyimide films, glass fiberglass reinforced Plastics (GFRP) plates and interfacing material such an indium or thermal compound. Through the test of dielectric strength in L$N_2$, polyimide film thickness of 125 ${\mu}{\textrm}{m}$ is selected as a insulating material. Electrical insulation and heat conduction are contrary to each other. Because of low heat conductivity of insulator and contact area between electrical insulator and heat conductor, thermal resistance of conduction-cooled system is increased. For the reducing of thermal resistance and the reliable contact between Polyimide and OFC, thermal compound or indium can be used As thermal compound layer is weak layer in electrical field, indium is finally selected for the reducing of thermal resistance. Thermal link is successfully passed the test. The testing voltage was AC 2.5kVrms and the testing time was 1 hour.

• 비파괴검사학회지
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• 제28권5호
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• pp.436-442
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• 2008

최근 VHTR(very high temperature reactor)에 대한 연구에서 흑연이 노심내 구조재, 반사재, 감속재로서 가장 적절한 재료로 인식되고 있다. 운전온도는 약 $900^{\circ}C$로 원자로 내부에 유입되는 소량의 불순물에도 흑연은 산화되기 쉬우며 산화된 흑연은 공극률이 증가하며 구조재료로서 가져야 할 파괴인성이 낮아진다. 본 연구에서는 산화 전후의 흑연에 대하여 초음파특성을 측정하고, 이를 기반으로 흑연에 대한 초음파탐상검사의 유효성을 타진하였다. 흑연의 초음파특성 측정 결과 초음파속도는 탄소강의 약 1/2, 음파 감쇠는 5배 이상, 신호대잡음비는 약 1/3로 측정되었다. 산화 후, 초음파속도는 as-received 상태에 비하여 미소하게 감소되었으나 초음파감쇠는 200% 이상으로 그 차이가 두드러지게 나타났다. 신호대잡음비에 기반하여 POD (probability of detection)을 산출한 결과, 100 mm 미만의 깊이를 가지는 측면공(SDH; side drilled hole)는 산화 전후에 큰 차이를 나타내지 않으므로 해당 깊이를 가지는 결함에 대해서 초음파탐상검사는 비교적 신뢰성 있는 검사를 수행할 수 있다고 판단된다. 상용 자동초음파탐상 장비에서의 테스트 결과 80 mm이하의 깊이에서는 인적오류가 크지 않을 것으로 예상할 수 있었으며, 위상배열 초음파 기법을 통한 검사를 수행한 결과 역시 양호한 신호대잡음비로 측면공들을 모두 검출할 수 있었다.