• Nuclear Engineering and Technology
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• v.38 no.5
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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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• v.49 no.4
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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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• v.52 no.9
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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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• v.38 no.5
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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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• v.49 no.5
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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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• v.20 no.4
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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.

• Journal of Environmental Science International
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• v.27 no.6
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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.

• Journal of the Korean Society for Nondestructive Testing
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• v.25 no.5
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• pp.377-383
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• 2005

Because explosive fluid is used at high temperature or under high pressure in petrochemistry and refined oil equipment, the interest about safety of equipments is intensive. Specially, the safety of high pressure reactor vessel is required among them. The material selected in this study is 2.25Cr-1Mo steel that is widely used for high pressure reactor vessel material of petrochemical plant. Eight kinds of artificially aged specimens were prepared by differing from aging periods under three different temperatures. The material was iso-thermally heat treated at higher temperatures than $391^{\circ}C$ that is the operating temperature of high pressure reactor vessel. Vickers hardness properties and electrical resistivity properties about artificially aged material as well as un-aged material were measured, and master curves were made out from the correlation with larson-Miller parameter. And electrical resistivity properties as well as Victors hardness properties measured at high pressure reactor vessel of the field were compared with master curves made out in a laboratory. Degradation evaluation possibility in the field of high pressure reactor vessel by using electrical resistivity method was examined. Electrical resistivity property measured in the field is similar with that of artificially aged material in similar aging level.

• Progress in Superconductivity and Cryogenics
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• v.4 no.1
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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.

• Journal of the Korean Society for Nondestructive Testing
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• v.28 no.5
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• pp.436-442
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• 2008

Graphite material has been recognized as a very competitive candidate for reflector, moderator, and structural material for very high temperature reactor (VHTR). Since VHTR is operated up to $900-950^{\circ}C$, small amount of impurity may accelerate the oxidation and degradation of carbon graphite, which results in increased porosity and lowered fracture toughness. In this study, ultrasonic wave propagation properties were investigated for both as-received and degradated material, and the feasibility of ultrasonic testing (UT) was estimated based on the result of ultrasonic property measurements. The ultrasonic properties of carbon graphite were half, more than 5 times, and 1/3 for velocity, attenuation, and signal-to-noise (S/N) ratio respectively. Degradation reduces the ultrasonic velocity slightly by 100 m/s, however the attenuation is about 2 times of as-receive state. The results of probability of detection (POD) estimation based on S/N ratio for side-drilled-hole (SDHs) of which depths were less than 100 mm were merely affected by oxidation and degradation. This result suggests that UT would be reliable method for nondestructive testing of carbon graphite material of which thickness is not over 100 mm. In accordance with the result produced by commercial automated ultrasonic testing (AUT) system, human error of ultrasonic testing is barely expected for the material of which thickness is not over 80 mm.