• Nuclear Engineering and Technology
• /
• v.38 no.5
• /
• pp.437-442
• /
• 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
• /
• v.34 no.4
• /
• pp.344-357
• /
• 2002

The present paper reports spontaneous steam explosions observed in fuel coolant interaction experiments using prototypic reactor materials. Pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$ are used. A high temperature molten material in the form of a jet is poured into a subcooled water pool located in a pressure vessel. An induction skull melting technique is used for the melting of the reactor material. In both tests using pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$, either a quenching or a spontaneous steam explosion was observed. The morphology of debris and pressure profile clearly indicate the differences between the qunching cases and explosion cases. The dynamic pressure. dynamic impulse, water temperature, melt temperature, and static pressure Inside the containment chamber were measured . As the spontaneous steam explosion for the reactor material is firstly observed in the present experiments, the results of present experiments could be a siginificant step forward the understanding the explosion of the reactor material.

• Nuclear Engineering and Technology
• /
• v.55 no.11
• /
• pp.4266-4273
• /
• 2023

In a severe reactor accident, a crust will form on the surface of the molten material during the core melting process. The crust will have a contact melting with the internal components of the reactor. In this paper, the contact melting process of the molten material on the austenitic stainless steel plate bundles is studied. The contact melting model of parabolic molten material on the plate bundles is proposed, and the rule and main effect factors of the contact melting are analyzed. The results show that the melting velocity is proportional to the slope of the paraboloid, the heat flux and the distance between two plates D. The influence of melt gravity and the plate width on melting velocity is negligible. The thickness of the molten liquid film is proportional to the heat flux and plate width, and it is inversely proportional to the gravity. With the increase of D, the liquid film thickness decreases at first and then increases gradually. The liquid film thickness has a minimum against D. When the width of the plate is small, the width of the plate is the main factor affecting the thickness of the liquid film. The parameters are coupled with each other. In a severe reactor accident, the wider internal components of reactor, which can increase the thickness of the melting liquid film and reduce the net input heat flux from the molten material to the components, are the effective measures to delay the melting process.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
• /
• 2001.07a
• /
• pp.499-502
• /
• 2001

In this experiment, an attempt to use the sludge pellets as catalyst for NO removal from simulated gas is experimentally investigated by using cylinder type reactor and packed-bed reactor. An experimental investigation has been conducted for NO concentration of 50[ppm], 100[ppm], 200[ppm] balanced with air, a gas flow rate of 5[1/min]. Ac voltage to discharge the gases was supplied. In the result, NOx removal rate in packed bed reactor is higher than that in cylinder type reactor. it is thought that plasma density in contact point of BaTiO$_3$ is significantly higher than that in cylinder reactor.

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
• /
• v.16 no.12S
• /
• pp.1292-1298
• /
• 2003

In this paper, dc reactor lot the inductive high-Tc superconducting fault current limiter (SFCL) was optimally designed by finite element method(FEM). The Prototype high-Tc do reactor was manufactured and compared to the results of design. This dc reactor consists of 4∼stacked double pancake coils which are wounded with Bi-2223 wire coated with SUS315L. Kapton tape is used for the insulation of turn to turn and layer to layer. Each pancake is connected in series by soldering Finally, optimal design and manufacture method lot the dc reactor is suggested in this paper. Through the comparison of result of optimal design and experimental result of prototype high-Tc superconducting dc reactor, reliance on the design of the high-Tc dc reactor tot the 1.2 kV/80 A SFCL is proved.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
• /
• 2002.08a
• /
• pp.133-137
• /
• 2002

In this paper, the $CF_{4}$ decomposition rate are investigated for a simulated three plasma reactors which are metal particle reactor, spiral wire reactor and reactor with porous dielectric as applied voltage. The $CF_{4}$ decomposition rate by plasma reactor with porous dielectric had a gain of 20~25[%] over that by plasma reactor with spiral wire or metal particle electrode. The $CF_{4}$ decomposition efficiency increases with increasing applied voltage up to the critical voltage for spark formation. The $CF_{4}$ decomposition efficiency of metal particle reactor was about 80[%] at AC 24[kV]. However, decomposition efficiency is more than 90% in case of the reactor with porous dielectric. we think, the reactor with porous dielectric should be much better than other reactors for $CF_{4}$ decomposition.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.12 no.1
• /
• pp.85-92
• /
• 2016

The operating temperature of VHTR components is much higher than that of conventional PWR due to high core outlet temperature of VHTR. Material requirements and technical issues of VHTR reactor components which are mainly dominated by high temperature service condition were discussed. The codification effort for high temperature material and design methodology are explained. The design class for VHTR components are classified as class A or B according to the recent ASME high temperature reactor design code. A separation of thermal boundary and pressure boundary is used for VHTR components as an elevated design solution. Key design characteristics for reactor pressure vessel, control rod, reactor internals, graphite reflector, circulator and intermediate heat exchanger were analysed. Thermo-mechanical analysis of the process heat exchanger, which was manufactured for test, is presented as an analysis example.

• Nuclear Engineering and Technology
• /
• v.47 no.5
• /
• pp.534-545
• /
• 2015

In the very high temperature reactor (VHTR), which is a next generation nuclear reactor system, ceramics are used as a fuel coating material and graphite is used as a core structural material. Even if a depressurization accident occurs and the reactor power goes up instantly, the temperature of the core will change only slowly. This is because the thermal capacity of the core is so high. Therefore, the VHTR system can passively remove the decay heat of the core by natural convection and radiation from the surface of the reactor pressure vessel. The objectives of this study are to investigate the heat transfer characteristics of natural convection of a one-side heated vertical channel with inserted porous materials of high porosity and also to develop the passive cooling system for the VHTR. An experiment was carried out using a one-side heated vertical rectangular channel. To obtain the heat transfer and fluid flow characteristics of the vertical channel with inserted porous material, we have also carried out a numerical analysis using a commercial Computational Fluid Dynamics (CFD) code. This paper describes the thermal performances of the one-side heated vertical rectangular channel with an inserted copper wire of high porosity.

• Nuclear Engineering and Technology
• /
• v.38 no.5
• /
• pp.417-422
• /
• 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.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.10a
• /
• pp.707-712
• /
• 1997

This paper presents the current state of knowledge on molten material relocation into the lower plenum. Consequences of movement of material to the lower head are considered with regardt to the potential for reactor pressure vessel failure from both thermal hydraulic and mechanical standpoints. The models are applied to evaluating various in-vessel retention strategies for the Korean Standard power plant (KSNPP) reactor The results are summarized in terms of thermal response of the reactor vessel from the very relevant severe accident management perspective.