• Nuclear Engineering and Technology
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• 제45권5호
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• pp.589-596
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• 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.

• Nuclear Engineering and Technology
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• 제41권10호
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• pp.1263-1274
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• 2009

Several experimental tests to simulate a reflood phase of a cold-leg LBLOCA of the APR1400 have been performed using the ATLAS facility. This paper describes the related experimental results with respect to the thermal-hydraulic behavior in the core and the system-core interactions during the reflood phase of the cold-leg LBLOCA conditions. The present descriptions will be focused on the LB-CL-09, LB-CL-11, LB-CL-14, and LB-CL-15 tests performed using the ATLAS. The LB-CL-09 is an integral effect test with conservative boundary condition; the LB-CL-11 and -14 are integral effect tests with realistic boundary conditions, and the LB-CL-15 is a separated effect test. The objectives of these tests are to investigate the thermal-hydraulic behavior during an entire reflood phase and to provide reliable experimental data for validating the LBLOCA analysis methodology for the APR1400. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results for the LBLOCA scenario of the APR1400. The ECC water flow rate from the safety injection tanks and the decay heat were simulated from the start of the reflood phase. The simulated core power was controlled to be 1.2 times that of the ANS-73 decay heat curve for LB-CL-09 and 1.02 times that of the ANS-79 decay curve for LB-CL-11, -14, and -15. The simulated ECC water flow rate from the high pressure safety injection pump was 0.32 kg/s. The present experimental data showed that the cladding temperature behavior is closely related to the collapsed water level in the core and the downcomer.

• 대한기계학회논문집B
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• 제31권7호
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• pp.596-603
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• 2007

As increasing interest for soot emission. etc in combustion systems, various studies are being carried out for the reduction and measurement techniques of soot. Especially, laser induced incandescence is the useful measurement technique which has distinguished spatial and temporal resolution for primary particle size, volume fraction and aggregated particle size etc. Time resolved laser induced incandescence is the technique for measuring primary particle size that is decided to solve the signal decay rate which is related to the cooling behavior of heated particle by pulsed laser. The cooling behavior of heated particle is able to represent the heat and mass transfer model which are involved constants of soot property for surround gas temperature on the our previous work. In this study, it is applied to the time-dependence thermodynamic properties for soot temperature instead of constants of soot property for surround gas temperature and compared two different model results.

• 방사성폐기물학회지
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• 제18권spc호
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• pp.1-19
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• 2020

Based on the current high-level radioactive waste management basic plan and the analysis results of spent nuclear fuel characteristics, such as dimensions and decay heat, an improved geological disposal concept for spent nuclear fuel from domestic nuclear power plants was proposed in this study. To this end, disposal container concepts for spent nuclear fuel from two types of reactors, pressurized water reactor (PWR) and Canada deuterium uranium (CANDU), considering the dimensions and interim storage method, were derived. In addition, considering the cooling time of the spent nuclear fuel at the time of disposal, according to the current basic plan-based scenarios, the amount of decay heat capacity for a disposal container was determined. Furthermore, improved disposal concepts for each disposal container were proposed, and analyses were conducted to determine whether the design requirements for the temperature limit were satisfied. Then, the disposal efficiencies of these disposal concepts were compared with those of the existing disposal concepts. The results indicated that the disposal area was reduced by approximately 20%, and the disposal density was increased by more than 20%.

• 대한기계학회논문집
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• 제16권11호
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• pp.2126-2135
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• 1992

본 연구에서는 국내 원자력 발전소중 영광 3,4호기의 설계자료를 토대로 1/6 크기로 축소한 모델실험을 통해서 공기흡입이 발생하는 임계수위를 결정하는 상관식을 개발하였으며 또한 공기흡입구를 reducer type으로 개선함으로써 공기흡입을 방지할 수 있음을 밝혔다.

• Nuclear Engineering and Technology
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• 제40권5호
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• pp.429-438
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• 2008

The objective of the present study is the depth and layout optimizations of a single layer, high level radioactive waste repository in a discontinuous rock mass with special joint set arrangements. A single layer repository model, considering variations in the repository depths, pitches, and tunnel spacings, is used to analyze the thermomechanical interaction behavior. It is assumed that the repository is constructed in saturated granite with joints; the PWR spent fuel in a disposal canister is installed in a deposition drift which is then sealed with compacted bentonite; and the backfill material is filled in the repository tunnel. The decay heat generated by the high level radioactive wastes governs the thermomechanical behavior of the near field rock mass of the repository. The temperature and displacement behavior of the repository is influenced more by the pitch variations than the tunnel spacing and repository depth. However, the stress behavior is influenced more by the repository depth variations than the pitch and tunnel spacing. For the final selection of the tunnel spacing, pitch, and repository depth, other aspects such as the nuclide migration through a groundwater flow path, construction costs, operation costs, and so on should be considered.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2008년 영문 학술대회
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• pp.700-710
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• 2008

A Bubble size distribution model has been developed for the numerical simulation of cryogenic high-speed cavitating flow of the turbo-pumps in the liquid fuel rocket engine. The new model is based on the previous one proposed by the authors, in which the bubble number density was solved as a function of bubble size at each grid point of the calculation domain by means of Eulerian framework with respect to the bubble size coordinate. In the previous model, the growth/decay of bubbles due to pressure difference between bubble and liquid was solved exactly based on Rayleigh-Plesset equation. However, the unsteady heat transfer between liquid and bubble, which controls the evaporation/condensation rate, was approximated by a theoretical solution of unsteady heat conduction under a constant temperature difference. In the present study, the unsteady temperature field in the liquid around a bubble is also solved exactly in order to establish an accurate and efficient numerical simulation code for cavitating flows. The growth/decay of a single bubble and growth of bubbles with nucleation were successfully simulated by the proposed model.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.3102-3113
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• 2023

Following the Fukushima nuclear disaster, the simulation of accidents in the spent fuel pool has become more noticeable. Despite the low amount of decay heat power, the consequences of the accidents in a spent fuel pool (SFP) can be severe due to the high content of long-lived radionuclides and lack of protection by the pressure vessel. In this study, the loss-of-cooling accident (LOFA) for the VVER-1000/V446 spent fuel pool is simulated by employing RELAP5 and MELCOR 1.8.6 as the best estimate and severe accident analysis codes, respectively. For two cases with different total power levels, decay heat of spent fuels is calculated by ORIGEN-II code. For modeling SFP of a VVER-1000, a qualified nodalizations are considered in both codes. During LOFA in SFP, the key sequences such as heating up of the pool water, boiling and reducing the water level, uncovering the spent fuels, increasing the temperature of the spent fuels, starting oxidation process (generating Hydrogen and extra power), the onset of fuel melting, and finally releasing radionuclides are studied for both cases. The obtained results show a reasonable consistency between the RELAP5 and MELCOR codes, especially before starting the oxidation process.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1513-1525
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• 2024

An improved mitigation system for thermally-induced steam generator tube rupture accidents was introduced to prevent direct environmental release of fission products bypassing the containment in the OPR1000. This involves injecting bypassed steam into the containment, cooling, and decontaminating it using a water coolant tank. To evaluate its performance, a severe accident analysis was performed using the MELCOR 2.2 code for OPR1000. Simulation results show that the proposed system sufficiently prevented the release of radioactive nuclides (RNs) into the environment via containment injection. The pool scrubbing system effectively decontaminated the injected RN and consequently reduced the aerosol mass in the containment atmosphere. However, the decay heat of the collected RNs causes re-vaporization. To restrict the re-vaporization, an external water source was considered, where the decontamination performance was significantly improved, and the RNs were effectively isolated. However, due to the continuous evaporation of the feed water caused by decay heat, a substantial amount of steam is released into the containment. Despite the slight pressurization inside the containment by the injected and evaporated steam, the steam decreased the hydrogen mole fraction, thereby reducing the possibility of ignition.

• Korean Chemical Engineering Research
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• 제56권6호
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• pp.878-883
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• 2018

A circulating fluidized bed (CFB) has been used in various chemical industries because of good heat and mass transfer. In addition, the methanol to olefins (MTO) process requiring the CFB reactor has attracted a great deal of interest due to steep increase of oil price. To design a CFB reactor for MTO pilot process, therefore, we has examined the hydrodynamic properties of spherical catalysts with different particle size and developed a correlation equation to predict catalyst holdup in a riser of CFB reactor. The hydrodynamics of micro-spherical catalysts with average particle size of 53, 90 and 140 mm was evaluated in a $0.025m-ID{\times}4m-high$ CFB riser. We also developed a model described by a decay coefficient to predict solid hold-up distribution in the riser. The decay coefficient developed in this study could be expressed as a function of Froude number and dimensionless velocity ratio. This model could predict well the experimental data obtained from this work.