Abstract: In a nuclear reactor, an array of fuel rods containing stacked uranium dioxide pellets clad with zircalloy is the heat source for a thermodynamic cycle of energy conversion from heat to electricity. After fuel is used in a nuclear reactor, the assemblies are stored underwater in a spent nuclear fuel pool at the nuclear power plant while heat generation and radioactive decay rates decrease before it is placed in packages for dry storage or transportation. A computational model of a Boiling Water Reactor spent fuel assembly is modeled using FLUENT, the computational fluid dynamics package. Heat transfer simulations were performed on the two-dimensional 9x9 spent fuel assembly to predict the maximum cladding temperature for different input to the FLUENT model. Uncertainty quantification is used to predict the heat transfer and the maximum temperature profile inside the assembly.
Abstract: Kuosheng nuclear power plant (NPP) is a BWR/6 plant in Taiwan. There is more concern for the safety of Spent Fuel Pools (SFPs) in Taiwan after Fukushima event. In order to estimate the safety of Kuosheng NPP SFP, by using MELCOR2.1 and SNAP, the safety analysis of Kuosheng NPP SFP was performed combined with the mitigation strategy of NEI 06-12 report. There were several steps in this research. First, the Kuosheng NPP SFP models were established by MELCOR2.1/SNAP. Second, the Station Blackout (SBO) analysis of Kuosheng SFP was done by TRACE and MELCOR under the cooling system failure condition. The results showed that the calculations of MELCOR and TRACE were very similar in this case. Second, the mitigation strategy analysis was done with the MELCOR model by following the NEI 06-12 report. The results showed the effectiveness of NEI 06-12 strategy in Kuosheng NPP SFP. Finally, a sensitivity study of SFP quenching was done to check the differences of different water injection time and the phenomena during the quenching. The results showed that if the cladding temperature was over 1600 K, the water injection may have chance to cause the accident more severe with more hydrogen generation. It was because of the oxidation heat and the “Breakaway” effect of the zirconium-water reaction. An animation model built by SNAP was also shown in this study.
Abstract: One of the main characteristics of Heavy Water Moderated Reactors is their high production of plutonium. This article demonstrates the possibility of reduction of plutonium and other actinides in Heavy Water Research Reactor. Among the many ways for reducing plutonium production in a heavy water reactor, in this research, changing the fuel from natural Uranium fuel to Thorium-Uranium mixed fuel was focused. The main fissile nucleus in Thorium-Uranium fuels is U-233 which would be produced after neutron absorption by Th-232, so the Thorium-Uranium fuels have some known advantages compared to the Uranium fuels. Due to this fact, four Thorium-Uranium fuels with different compositions ratios were chosen in our simulations; a) 10% UO2-90% THO2 (enriched= 20%); b) 15% UO2-85% THO2 (enriched= 10%); c) 30% UO2-70% THO2 (enriched= 5%); d) 35% UO2-65% THO2 (enriched= 3.7%). The natural Uranium Oxide (UO2) is considered as the reference fuel, in other words all of the calculated data are compared with the related data from Uranium fuel. Neutronic parameters were calculated and used as the comparison parameters. All calculations were performed by Monte Carol (MCNPX2.6) steady state reaction rate calculation linked to a deterministic depletion calculation (CINDER90). The obtained computational data showed that Thorium-Uranium fuels with four different fissile compositions ratios can satisfy the safety and operating requirements for Heavy Water Research Reactor. Furthermore, Thorium-Uranium fuels have a very good proliferation resistance and consume less fissile material than uranium fuels at the same reactor operation time. Using mixed Thorium-Uranium fuels reduced the long-lived α emitter, high radiotoxic wastes and the radio toxicity level of spent fuel.
Abstract: Heat pipe is considered to be applied as a passive system to remove residual heat that generated from reactor core when incident occur or from spent fuel storage pool. The objectives are to characterized the heat transfer phenomena, performance of heat pipe, and as a model for large heat pipe will be applied as passive cooling system on nuclear spent fuel pool storage. In this experimental wickless heat pipe or two-phase closed thermosyphon (TPCT) is used. Variation of heat flux are 611.24 Watt/m2 - 3291.29 Watt/m2. Variation of filling ratio are 45 - 70%. Variation of initial pressure are -62 to -74 cm Hg. Demineralized water is used as working fluid in the TPCT. The results showed that increasing of heat load leads to an increase of evaporation of the working fluid. The optimum filling ratio obtained for 60% of TPCT evaporator volume, and initial pressure variation gave different TPCT wall temperature characteristic. TPCT showed best performance with 60% filling ratio and can be consider to be applied as passive residual heat removal system or passive cooling system on spent fuel storage pool.
Abstract: Kuosheng nuclear power plant (NPP) is a BWR/6 plant in Taiwan. There is more concern for the safety of NPPs in Taiwan after Japan Fukushima NPP disaster occurred. Hence, in order to estimate the safety of Kuosheng NPP spent fuel pool (SFP), by using TRACE, MELCOR, and SNAP codes, the safety analysis of Kuosheng NPP SFP was performed. There were two main steps in this research. First, the Kuosheng NPP SFP models were established. Second, the transient analysis of Kuosheng SFP was done by TRACE and MELCOR under the cooling system failure condition (Fukushima-like condition). The results showed that the calculations of MELCOR and TRACE were very similar in this case, and the fuel uncover happened roughly at 4th day after the failure of cooling system. The above results indicated that Kuosheng NPP SFP may be unsafe in the case of long-term SBO situation. In addition, future calculations were needed to be done by the other codes like FRAPTRAN for the cladding calculations.
Abstract: TRACE is developed by U.S. NRC for the nuclear
power plants (NPPs) safety analysis. We focus on the establishment
and application of TRACE/FRAPTRAN/SNAP models for Chinshan
NPP (BWR/4) spent fuel pool in this research. The geometry is 12.17
m × 7.87 m × 11.61 m for the spent fuel pool. In this study, there are
three TRACE/SNAP models: one-channel, two-channel, and
multi-channel TRACE/SNAP model. Additionally, the cooling system
failure of the spent fuel pool was simulated and analyzed by using the
above models. According to the analysis results, the peak cladding
temperature response was more accurate in the multi-channel
TRACE/SNAP model. The results depicted that the uncovered of the
fuels occurred at 2.7 day after the cooling system failed. In order to
estimate the detailed fuel rods performance, FRAPTRAN code was
used in this research. According to the results of FRAPTRAN, the
highest cladding temperature located on the node 21 of the fuel rod
(the highest node at node 23) and the cladding burst roughly after 3.7
day.
Abstract: The dry-storage systems of nuclear power plants (NPPs) in Taiwan have become one of the major safety concerns. There are two steps considered in this study. The first step is the verification of the TRACE by using VSC-17 experimental data. The results of TRACE were similar to the VSC-17 data. It indicates that TRACE has the respectable accuracy in the simulation and analysis of the dry-storage systems. The next step is the application of TRACE in the dry-storage system of Kuosheng NPP (BWR/6). Kuosheng NPP is the second BWR NPP of Taiwan Power Company. In order to solve the storage of the spent fuels, Taiwan Power Company developed the new dry-storage system for Kuosheng NPP. In this step, the dry-storage system model of Kuosheng NPP was established by TRACE. Then, the steady state simulation of this model was performed and the results of TRACE were compared with the Kuosheng NPP data. Finally, this model was used to perform the safety analysis of Kuosheng NPP dry-storage system. Besides, FRAPTRAN was used tocalculate the transient performance of fuel rods.
Abstract: This paper focuses on assessing sloshing-induced overflow of the seismically-isolated nuclear tanks based on Fluid-Structure Interaction (FSI) analysis. Typically, fluid motion in the seismically-isolated nuclear tank systems may be rather amplified and even overflowed under earthquake. Sloshing-induced overflow in those structures has to be reliably assessed and predicted since it can often cause critical damages to humans and environments. FSI analysis is herein performed to compute the total cumulative overflowed water volume more accurately, by coupling ANSYS with CFX for structural and fluid analyses, respectively. The approach is illustrated on a nuclear liquid storage tank, Spent Fuel Pool (SFP), forgiven conditions under consideration: different liquid levels, Peak Ground Accelerations (PGAs), and post earthquakes.
Abstract: The presented article deals with the description of a
numerical model of a corridor at a Central Interim Spent Fuel Storage
Facility (hereinafter CISFSF). The model takes into account the
effect of air flows on the temperature of stored waste. The
computational model was implemented in the ANSYS/CFX
programming environment in the form of a CFD task solution, which
was compared with an approximate analytical calculation. The article
includes a categorization of the individual alternatives for the
ventilation of such underground systems. The aim was to evaluate a
ventilation system for a CISFSF with regard to its stability and
capacity to provide sufficient ventilation for the removal of heat
produced by stored casks with spent nuclear fuel.
Abstract: Thermochemcial characteristics of powder fabricated
using oxidation treatment of spent PWR fuel and SIMFUEL were
evaluated for recycling of spent fuel such as DUPIC process.
Especially, the influence of spent fuel burn-ups on the powder
fabrication characteristics was experimentally evaluated, ranging from
27,300 to 65,000 MWd/tU. Densities of powder manufactured from an
oxidation, OREOX and the milling processes at the same process
conditions were compared as a function of the fuel burn-ups
respectively. Also, based on chemical analysis results, homogeneity of
fissile elements in oxidized powder was confirmed.
Abstract: A high performance clarification system has been
discussed for advanced aqueous reprocessing of FBR spent fuel.
Dissolver residue gives the cause of troubles on the plant operation of
reprocessing. In this study, the new clarification system based on the
hybrid of centrifuge and filtration was proposed to get the high
separation ability of the component of whole insoluble sludge. The
clarification tests of simulated solid species were carried out to
evaluate the clarification performance using small-scale test apparatus
of centrifuge and filter unit. The density effect of solid species on the
collection efficiency was mainly evaluated in the centrifugal
clarification test. In the filtration test using ceramic filter with pore
size of 0.2μm, on the other hand, permeability and filtration rate
were evaluated in addition to the filtration efficiency. As results, it was
evaluated that the collection efficiency of solid species on the new
clarification system was estimated as nearly 100%. In conclusion, the
high clarification performance of dissolver liquor can be achieved by
the hybrid of the centrifuge and filtration system.
Abstract: Accident in spent fuel pool (SFP) of Fukushima
Daiichi Unit 4 showed the importance of continuous monitoring of the
key environmental parameters such as water temperature, water level,
and radiation level in the SFP at accident conditions. Because the SFP
water temperature is one of the key parameters indicating SFP
conditions, its behavior at accident conditions shall be understood to
prepare appropriate measures. This study estimated temporal change
in the SFP water temperature at Kori Unit 1 with 587 MWe for 1 hour
after initiation of a loss-of-pool-cooling accident. For the estimation,
ANSYS CFX 13.0 code was used. The estimation showed that the
increasing rate of the water temperature was 3.90C per hour and the
SFP water temperature could reach 1000C in 25.6 hours after the
initiation of loss-of-pool-cooling accident.