• Journal of Radiation Protection and Research
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• v.42 no.2
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• pp.114-129
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• 2017

Background: The purpose of this paper is to confirm the event timings and the magnitude of fission product aerosol release from the Fukushima accident. Over a few hundreds of technical papers have been published on the environmental impact of Fukushima Daiichi accident since the accident occurred on March 11, 2011. However, most of the research used reverse or inverse method based on the monitoring of activities in the remote places and only few papers attempted to estimate the release of fission products from individual reactor core or from individual spent fuel pool. Severe accident analysis code can be used to estimate the radioactive release from which reactor core and from which radionuclide the peaks in monitoring points can be generated. Materials and Methods: The basic material used for this study are the initial core inventory obtained from the report JAEA-Data/Code 2012-018 and the given accident scenarios provided by Japanese Government or Tokyo Electric Power Company (TEPCO) in official reports. In this research a forward method using severe accident progression code is used as it might be useful for justifying the results of reverse or inverse method or vice versa. Results and Discussion: The release timing and amounts to the environment are estimated for volatile radioactive fission products such as noble gases, cesium, iodine, and tellurium up to 184 hours (about 7.7 days) after earthquake occurs. The in-plant fission product behaviors and release characteristics to environment are estimated using the severe accident progression analysis code, MELCOR, for Fukushima Daiichi accident. These results are compared with other research results which are summarized in UNSCEAR 2013 Report and other technical papers. Also it may provide the physically based arguments for justifying or suspecting the rationale for the scenarios provided in open literature. Conclusion: The estimated results by MELCOR code simulation of this study indicate that the release amount of volatile fission products to environment from Units 1, 2, and 3 cores is well within the range estimated by the reverse or inverse method, which are summarized in UNSCEAR 2013 report. But this does not necessarily mean that these two approaches are consistent.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4195-4208
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• 2022

In this study, operator habitability was numerically evaluated in the event of a fire at the main control bench board (MCB) in a reference main control room (MCR). It was investigated if evacuation variables including hot gas layer temperature (HGLT), heat flux (HF), and optical density (OD) at 1.8 m from the MCR floor exceed the reference evacuation criteria provided in NUREG/CR-6850. For a fire model validation, the simulation results of the reference MCR were compared with existing experimental results on the same reference MCR. In the simulation, various input parameters were applied to the MCB panel fire scenario: MCR height, peak heat release rate (HRR) of a panel, number of panels where fire propagation occurs, fire propagation time, door open/close conditions, and mechanical ventilation operation. A specialized-average HRR (SAHRR) concept was newly devised to comprehensively investigate how the various input parameters affect the operator's habitability. Peak values of the evacuation variables normalized by evacuation criteria of NUREG/CR-6850 were well-correlated as the power function of the SAHRR for the various input parameters. In addition, the evacuation time map was newly utilized to investigate how the evacuation time for different SAHRR was affected by changing the various input parameters. In the previous studies, it was found that the OD is the most dominant variable to determine the MCR evacuation time. In this study, however, the evacuation time map showed that the HF is the most dominant factor at the condition of without-mechanical ventilation for the MCR with a partially-open false ceiling, but the OD is the most dominant factor for all the other conditions. Therefore, the method using the SAHRR and the evacuation time map was very useful to effectively and comprehensively evaluate the operator habitability for the various input parameters in the event of MCB fires for the reference MCR.

• 유체기계공업학회:학술대회논문집
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• 2002.12a
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• pp.419-424
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• 2002

The HANARO, a multi-purpose research reactor of 30 MWth open-tank-in-pool type, has been under normal operation since its initial criticality In February, 1995. Many experiments should be safely performed to activate the utilization of the HANARO. A flow simulation facility is being developed for the endurance test of reactivity control units for extended life times and the verification of structural integrity of those experimental facilities prior to loading in the HANARO. This test facility is composed of three major parts; a half-core structure assembly, flow circulation system and support system. The flow circulation system is composed of a circulation pump, a core flow pipe, a core bypass flow pipe and instruments. The system is to be filled with de-mineralized water and the flow should be met the design flow to simulate similar flow characteristics in the core channel of the half-core test facility to the HANARO. This paper, therefore, describes an analytical analysis to study the flow behavior of the system. The computational flow analysis has been performed for the verification of system pressure variation through the three-dimensional analysis program with standard k-$\epsilon$ turbulence model and for the verification of the structural piping integrity through the finite element method. The results of the analysis are satisfied the design requirements and structural piping integrity of flow circulation system.

• Journal of Energy Engineering
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• v.12 no.2
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• pp.74-83
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• 2003

In the advanced nuclear power plants including APR1400, the SDVS (Safety Depressurization and Vent System) is adopted to increase the plant safety using the concept of feed-and-bleed operation. In the case of the TLOFW (Total Loss of Feedwater), the POSRV (Power Operated Safety Relief Value) located at the top of the pressurizer is expected to open due to the pressurization of the reactor coolant system and discharges steam and/or water mixture into the water pool, where the mixture is condensed. During the condensation of the mixture, thermal-hydraulic loads such as pressure and temperature variations are induced to the pool structure. For the pool structure design, such thermal-hydraulic aspects should be considered. Understanding the phenomena of the submerged steam jet condensation in a water pool is helpful for system designers to design proper pool structure, sparger, and supports etc. This paper reviews and evaluates the steam jet condensation in a water pool on the physical phenomena of the steam condensation including condensation regime map, heat transfer coefficient, steam plume, steam jet condensation load, and steam jet induced flow.

• The KSFM Journal of Fluid Machinery
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• v.7 no.1 s.22
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• pp.30-35
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• 2004

The HANARO, a multi-purpose research reactor of 30 MWth open-tank-in-pool type, has been under normal operation since its initial criticality in February, 1995. Many experiments should be safely performed to activate the utilization of the HANARO. HANARO flow simulation facility is being developed for the endurance test of reactivity control units for extended life time and the verification of structural integrity of those experimental equipments prior to loading in the HANARO. This facility is composed of three major parts; a half-core structure assembly, a flow circulation system and a support system. The flow circulation system is composed of a circulation pump, a core flow piping, a core bypass flow piping and instruments. The system is to be filled with de-mineralized water and the flow should be met the design requirements to simulate a similar flow characteristics in the core channel of the half-core structure assembly to the HANARO. This paper, therefore, presents an analytical analysis to study the flow behavior of the system. Computational flow analysis has been performed for the verification of system pressure variation through the three-dimensional analysis program with the standard $k-{\epsilon}$ turbulence model and for the verification of the structural piping integrity through the finite element method. According to the analysis results, it could be said that the design requirements and the structural piping integrity of the flow circulation system are satisfied.

• Journal of Korean Society of Water and Wastewater
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• v.18 no.2
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• pp.121-127
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• 2004

It is necessary to supply external carbon source for enhancement of biological nitrogen removal from domestic wastewater with low influent C/N ratio. Sulfide was chosen as a cost effective electron donor and reaction stoichiometry for autotrophic denitrification was investigated by conducting bench-scale experiments in this study. Higher sulfur to nitrogen (S/N) ratio than the calculated value from theoretical reaction stoichiometry was required when the anoxic reactor was operated at open condition because dissolved oxygen introduced by surface aeration reacted with sulfide with ease. In addition, higher sulfate production and lower yield of microorganism could be observed under the same condition. It was possible to obtain reliable reaction stoichiometry for autotrophic denitrification by establishing pure anoxic condition. Linear relationship between bacterial growth and consumption of nitrate, sulfide, alkalinity, and sulfate production enabled to derive a relatively correct reaction stoichiometry for autotrophic denitrification when sulfide was used as an electron donor.

• Nuclear Engineering and Technology
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• v.40 no.2
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• pp.133-138
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• 2008

The use of Supercritical Fluids(SCF) has been proposed for numerous power cycle designs as part of the Generation IV advanced reactor designs, and can provide for higher thermal efficiency. One particular area of interest involves the behavior of SCF during a blowdown or depressurization process. Currently, no data are available in the open literature at supercritical conditions to characterize this phenomenon. A preliminary computational analysis, using a homogeneous equilibrium model when a second phase appears in the process, has shown the complexity of behavior that can occur. Depending on the initial thermodynamic state of the SCF, critical flow phenomena can be characterized in three different ways; the flow can remain in single phase(high temperature), a second phase can appear through vaporization(high pressure low temperature) or condensation(high pressure, intermediate temperature). An experimental facility has been built at the University of Wisconsin to study SCF depressurization through several diameter breaks. The preliminary results obtained show that the experimental data can be predicted with good agreement by the model for all the different initial conditions.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.339-350
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• 2016

Based on the resonance integral (RI) tables produced by the NJOY program, the conventional subgroup method usually ignores both the resonance elastic scattering and the resonance interference effects. In this paper, on one hand, to correct the resonance elastic scattering effect, RI tables are regenerated by using the Monte Carlo code, OpenMC, which employs the Doppler broadening rejection correction method for the resonance elastic scattering. On the other hand, a fast resonance interference factor method is proposed to efficiently handle the resonance interference effect. Encouraging conclusions have been indicated by the numerical results. (1) For a hot full power pressurized water reactor fuel pin-cell, an error of about +200 percent mille could be introduced by neglecting the resonance elastic scattering effect. By contrast, the approach employed in this paper can eliminate the error. (2) The fast resonance interference factor method possesses higher precision and higher efficiency than the conventional Bondarenko iteration method. Correspondingly, if the fast resonance interference factor method proposed in this paper is employed, the $k_{inf}$ can be improved by ~100 percent mille with a speedup of about 4.56.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1165-1171
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• 2017

The traditional form of parallelism in Monte Carlo particle transport simulations, wherein each individual particle history is considered a unit of work, does not lend itself well to data-level parallelism. Event-based algorithms, which were originally used for simulations on vector processors, may offer a path toward better utilizing data-level parallelism in modern computer architectures. In this study, a simple model is developed for estimating the efficiency of the event-based particle transport algorithm under two sets of assumptions. Data collected from simulations of four reactor problems using OpenMC was then used in conjunction with the models to calculate the speedup due to vectorization as a function of the size of the particle bank and the vector width. When each event type is assumed to have constant execution time, the achievable speedup is directly related to the particle bank size. We observed that the bank size generally needs to be at least 20 times greater than vector size to achieve vector efficiency greater than 90%. When the execution times for events are allowed to vary, the vector speedup is also limited by differences in the execution time for events being carried out in a single event-iteration.

• Nuclear Engineering and Technology
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• v.42 no.4
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• pp.434-441
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• 2010

The activation products of aluminum and argon are key radionuclides for operational and environmental radiological safety during the normal operation of open-tank-in-pool type research reactors using aluminum-clad fuels. Their activities measured in the primary coolant and pool surface water of HANARO have been consistent. We estimated their sources from the measured activities and then compared these values with their production rates obtained by a core calculation. For each aluminum activation product, an equivalent aluminum thickness (EAT) in which its production rate is identical to its release rate into the coolant is determined. For the argon activation calculation, the saturated argon concentration in the water at the temperature of the pool surface is assumed. The EATs are 5680, 266 and 1.2 nm, respectively, for Na-24, Mg-27 and Al-28, which are much larger than the flight lengths of the respective recoil nuclides. These values coincide with the water solubility levels and with the half-lives. The EAT for Na-24 is similar to the average oxide layer thickness (OLT) of fuel cladding as well; hence, the majority of them in the oxide layer may be released to the coolant. However, while the average OLT clearly increases with the fuel burn-up during an operation cycle, its effect on the pool-top radiation is not distinguishable. The source of Ar-41 is in good agreement with the calculated reaction rate of Ar-40 dissolved in the coolant.