• Nuclear Engineering and Technology
• /
• v.45 no.5
• /
• pp.589-596
• /
• 2013

In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ~2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.

• Transactions on Electrical and Electronic Materials
• /
• v.2 no.3
• /
• pp.7-11
• /
• 2001

Utility boiler operators seeking to gain the greatest economic advantage from their units are faced with three challenges, namely the obligatory light-up fuel costs, the additional expense of supplementary fuel firing should they wish to use a cheaper fuel that may be beyond the original burner manufacturer’s stability and combustion performance assurances and the immediate environmental impact of both. The novel use of plasma arc technology can provide a solution to these challenges. This paper introduces the work being undertaken through a joint collaboration between the EU, Kazahkstan and Russia in order to develop a tried and tested engineering methodology and a mathematical based application and sensitivity analysis approach for the design and optimisation stage of these plasma devices that, as a consequence, their assist in their universal introduction.

• Journal of ILASS-Korea
• /
• v.9 no.3
• /
• pp.29-34
• /
• 2004

To investigate characteristics and micro-explosion of single-droplets of emulsified fuel, water is mixed with diesel oil by using ultrasonic energy fuel feeding system. The fuel characteristics is analysed through H-NMR spectrum and micro-explosion phenomena of the emulsified fuel is also investigated. The life times of droplets of conventional diesel fuel, ultrasonic energy added diesel fuel and emulsified fuel we obtained additionally. According to this study, the micro-explosion phenomena of single-droplets happen in atmospheric pressure condition, a curve form of emulsified fuel's life tim is different from diesel fuel's one and the change of chemical structures is a cause of ultrasonic-energy-added diesel fuel effect.

• Nuclear Engineering and Technology
• /
• v.52 no.1
• /
• pp.123-134
• /
• 2020

The interaction between UN and CdCl₂ in the LiCl-KCl molten eutectic was studied at 773 K. The reaction was controlled by sampling the melt, as well as by analysis of the resulting precipitate. The process was shown to proceed according to several parallel reactions. The summary reaction was determined to have two stages: a fast one and a slow one. The 19-53% UN → UCl₃ conversion was obtained for the molar ratio of CdCl₂/UN = 1.22-14.9. The rest of UN converts into the precipitate of complex composition (UNCl + U₂N₃ + U₄N₇ + UN₂). The increase in the CdCl₂/UN molar ratio from 1.22 to 14.9 resulted in the decrease in duration of the first "fast" stage of the process from 18 h to 1 h.

• Nuclear Engineering and Technology
• /
• v.53 no.6
• /
• pp.1736-1746
• /
• 2021

The results of full-scale numerical experiments of a hybrid thorium-containing fuel cell facility operating in a close-to-critical state due to a controlled source of fusion neutrons are discussed in this work. The facility under study was a complex consisting of two blocks. The first block was based on the concept of a high-temperature gas-cooled thorium reactor core. The second block was an axially symmetrical extended plasma generator of additional neutrons that was placed in the near-axial zone of the facility blanket. The calculated models of the blanket and the plasma generator of D-T neutrons created within the work allowed for research of the neutronic parameters of the facility in stationary and pulse-periodic operation modes. This research will make it possible to construct a safe facility and investigate the properties of thorium fuel, which can be continuously used in the epithermal spectrum of the considered hybrid fusion-fission reactor.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.27 no.3
• /
• pp.275-285
• /
• 2003

A visualization study on the effect of forcing amplitude in tone-excited jet diffusion flames has been conducted. Visualization techniques are employed using optical schemes. which are a light scattering photography. Flame stability curve is attained according to Reynolds number and forcing amplitude at a fuel tube resonant frequency. Flame behavior is globally grouped into two from attached flame to blown-out flame according to forcing amplitude: one sticks the tradition flame behavior which has been observed in general jet diffusion flames and the other shows a variety of flame modes such as the flame of a feeble forcing amplitude where traditionally well-organized vortex motion evolves, a fat flame. an elongated flame. and an in-burning flame. Particular attention is focused on an elongation flame. which is associated with a turnabout phenomenon of vortex motion and on a reversal of the direction of vortex roll-up. It is found that the flame length with forcing amplitude is the direct outcome of the evolution process of the formed inner flow structure. Especially the negative part of the acoustic cycle under the influence of a strong negative pressure gradient causes the shapes of the fuel stem and fuel branch part and even the direction of vortex roll-up to dramatically change.

• Proceedings of the Korean Magnetic Resonance Society Conference
• /
• 2002.08a
• /
• pp.20-20
• /
• 2002

• Nuclear Engineering and Technology
• /
• v.14 no.2
• /
• pp.86-91
• /
• 1982

In 1977, spent fuel storage capacity of Ko-Ri Unit 1 was raised to contain 4-2/3 core, by reducing the center-to-center spacing between fuel assemblies from 53.34cm to 36cm. In this paper the adequacy is discussed in detail by examining the previous design analysis report. According to the analytic method presented by Core Performance Branch, study on credible abnormal moderator density condition is performed by using KENO-IV for the redesigned spent fuel storage facility. Result shows that 36cm for the center-to-center spacing between fuel assemblies is not enough to keep the storage safe at water density of 0.1143g/㎤, which gives the maximum $K_{eff}$ 0.9958$\pm$0.0048, which exceeds the CPB regulation limit 0.98. From sensitivity study regarding to the center-to-center spacing, it should be maintained to space greater than 43cm in order to meet the CPB requirements.s.

• Nuclear Engineering and Technology
• /
• v.54 no.3
• /
• pp.965-974
• /
• 2022

Platinum anodes are widely used for metal oxides reduction in LiCl-Li₂O, however high-cost and low-corrosion resistance hinder their implementation. NiO-Li₂O ceramics is an alternative corrosion resistant anode material. Anode processes on platinum and NiO-Li₂O ceramics were studied in (80 mol.%) LiCl-(20mol.%)KCl and (80 mol.%)LiCl-(20 mol.%)KCl-Li₂O melts by cyclic voltammetry, potentiostatic and galvanostatic electrolysis. Experiments performed in the LiCl-KCl melt without Li₂O illustrate that a Pt anode dissolution causes the Pt²⁺ ions formation at 3.14 V and 550℃ and at 3.04 V and 650℃. A two-stage Pt oxidation was observed in the melts with the Li₂O at 2.40 ÷ 2.43 V, which resulted in the Li₂PtO₃ formation. Oxygen current efficiency of the Pt anode at 2.8 V and 650℃ reached about 96%. The anode process on the NiO-Li₂O electrode in the LiCl-KCl melt without Li₂O proceeds at the potentials more positive than 3.1 V and results in the electrochemical decomposition of ceramic electrode to NiO and O₂. Oxygen current efficiency on NiO-Li₂O is close to 100%. The NiO-Li₂O ceramic anode demonstrated good electrochemical characteristics during the galvanostatic electrolysis at 0.25 A/cm² for 35 h and may be successfully used for pyrochemical treating of spent nuclear fuel.

• Nuclear Engineering and Technology
• /
• v.53 no.7
• /
• pp.2104-2125
• /
• 2021

This paper presents the verification and validation (V&V) of a calculation module for isotope inventory prediction to control the back-end cycle of spent nuclear fuel (SNF). The calculation method presented herein was implemented in a two-step code system of a lattice code STREAM and a nodal diffusion code RAST-K. STREAM generates a cross section and provides the number density information using branch/history depletion branch calculations, whereas RAST-K supplies the power history and three history indices (boron concentration, moderator temperature, and fuel temperature). As its primary feature, this method can directly consider three-dimensional core simulation conditions using history indices of the operating conditions. Therefore, this method reduces the computation time by avoiding a recalculation of the fuel depletion. The module for isotope inventory calculates the number densities using the Lagrange interpolation method and power history correction factors, which are applied to correct the effects of the decay and fission products generated at different power levels. To assess the reliability of the developed code system for back-end cycle analysis, validation study was performed with 58 measured samples of pressurized water reactor (PWR) SNF, and code-to-code comparison was conducted with STREAM-SNF, HELIOS-1.6 and SCALE 5.1. The V&V results presented that the developed code system can provide reasonable results with comparable confidence intervals. As a result, this paper successfully demonstrates that the isotope inventory prediction code system can be used for spent nuclear fuel analysis.