• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3194-3201
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• 2023

In water-cooled power reactor, hydrogen is generated in case of steam zirconium reaction during severe accident condition and later on in addition to hydrogen; CO is also generated during molten corium concrete interaction after reactor pressure vessel failure. Passive Autocatalytic Recombiners (PARs) are provided in the containment for hydrogen management. The performance of the PARs in presence of hydrogen and carbon monoxide along with air has been evaluated. Depending on the conditions, CO may either react with oxygen to form carbon dioxide (CO2) or act as catalyst poison, reducing the catalyst activity and hence the hydrogen conversion efficiency. CFD analysis has been carried out to determine the effect of CO on catalyst plate temperature for 2 & 4% v/v H₂ and 1-4% v/v CO with air at the recombiner inlet for a reported experiment. The results of CFD simulations have been compared with the reported experimental data for the model validation. The reaction at the recombiner plate is modelled based on diffusion theory. The developed CFD model has been used to predict the maximum catalyst temperature and outlet species concentration for different inlet velocity and temperatures of the mixture gas. The obtained results were used to fit a correlation for obtaining removal rate of carbon monoxide inside PAR as a function of inlet velocity and concentrations.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3888-3896
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• 2022

Nuclear data is the basis of reactor physics analysis. This paper aim at studying the influence of major evaluated nuclear data libraries, CENDL-3.2, ENDF/B-VIII.0, JEFF-3.3, and JENDL-4.0u, on the neutronics modelling of CEFR start-up tests. Results show these four libraries have a good performance and consistency in the modelling CEFR start-up tests. The JEFF-3.3 results exhibit only an 8 pcm k_eff difference with the measurement. The difference in criticality is decomposed by nuclide, which shows the large overestimation of CENDL-3.2 is mainly from the cross-section of ⁵²Cr. Except for few cases, the calculation results are within 1σ of measurement uncertainty in control rod worth, sodium void reactivity, temperature reactivity, and subassembly swap reactivity. In the evaluation of axial and radial reaction distribution, there are about 65% of relative errors that are less than 5% and 82% of relative errors that are less than 10%.

• Korean Chemical Engineering Research
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• v.61 no.2
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• pp.328-339
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• 2023

Methanol steam reforming (MSR) is a promising method for hydrogen supplying as a critical step in hydrogen fuel cell commercialization in mobile applications. Modelling and understanding of the reactor behavior is an attractive research field to develop an efficient reformer. Three-layer feed-forward artificial neural network (ANN) and Box-Behnken design (BBD) were used to modelling of MSR process using the Cu-SiO₂ aerogel catalyst. Furthermore, impacts of the basic operational variables and their mutual interactions were studied. The results showed that the most affecting parameters were the reaction temperature (56%) and its quadratic term (20.5%). In addition, it was also found that the interaction between temperature and Steam/Methanol ratio is important on the MSR performance. These models precisely predict MSR performance and have great agreement with experimental results. However, on the basis of statistical criteria the ANN technique showed the greater modelling ability as compared with statistical BBD approach.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.803-810
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• 2022

The small modular liquid-metal fast reactor (SMFR) is an important component of advanced nuclear systems. SMFRs exhibit relatively low breeding capability and constraint space for control rod installation. Consequently, control rods are deeply inserted at beginning and are withdrawn gradually to compensate for large burnup reactivity loss in a long lifetime. This paper is committed to investigating the impact of control rod compensation operation on core neutronics characteristics. This paper presents a whole core fine depletion model of long lifetime SMFR using OpenMC and the influence of depletion chains is verified. Three control rod position schemes to simulate the compensation process are compared. The results show that the fine simulation of the control rod compensation process impacts significantly the fuel burnup distribution and absorber consumption. A control rod equivalent position scheme proposed in this work is an optimal option in the trade-off between computation time and accuracy. The control position is crucial for accurate power distribution and void feedback coefficients in SMFRs. The results in this paper also show that the pin level power distribution is important due to the heterogeneous distribution in SMFRs. The fuel burnup distribution at the end of core life impacts the worth of control rods.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.579-585
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• 1996

A CATHENA model for the reactor regulating system is developed and tested independently. A CATHENA plant model is created by combining this model with the reference CATHENA model which has been developed to analyze a loss-of-coolant accident (LOCA) for the Wolsong 2 generating station. This model is intended to provide a trip coverage analysis capability. The CATHENA reactor regulating system model includes the demand power routine. the light water zone control absorbers, mechanical control absorbers and adjusters. The CATHENA model is tested for steady state at 103% full power. A postulated accident transient (small LOCA) was also tested. The results show that the control routines in CATHENA were set up properly.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1687-1707
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• 2023

This paper presents a molten salt reactor (MSR) design parameter sensitivity study using a nodal dynamic modelling methodology with explicitly modified point kinetics equation and Mann's model for heat transfer. Six parameters that can impact MSR safety are evaluated. A MATLAB-Simulink model inspired by Thorcon's 550MWth MSR is used for parameter evaluations. A safety envelope was formed to encapsulate power, maximum and minimum temperature, and temperature-induced reactivity feedback. The parameters are perturbed by ±30%. The parameters were then ranked by their subsequent impact on the considered safety envelope, which ranks acceptable parameter uncertainty. The model is openly available on GitHub.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.368-372
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• 2016

Background: Methodologies for a series of radiological consequence assessments show a distinctive difference according to the design principles of the original nuclear suppliers and their technical standards to be imposed. This is due to the uncertainties of the accidental source term, radionuclide behavior in the environment, and subsequent radiological dose. Both types of PWR and PHWR are operated in Korea. However, technical standards for evaluating atmospheric dispersion have been enacted based on the U.S. NRC's positions regardless of the reactor types. For this reason, it might cause a controversy between the licensor and licensee of a nuclear power plant. Materials and Methods: It was modelled under the framework of the NRC Regulatory Guide 1.145 for light-water reactors, reflecting the features of heavy-water reactors as specified in the Canadian National Standard and the modelling features in MACCS2, such as atmospheric diffusion coefficient, ground deposition, surface roughness, radioactive plume depletion, and exposure from ground deposition. Results and Discussion: An integrated accident consequence assessment code, ACCESS (Accident Consequence Assessment Code for Evaluating Site Suitability), was developed by taking into account the unique regulatory positions for reactor types under the framework of the current Korean technical standards. Field tracer experiments and hand calculations have been carried out for validation and verification of the models. Conclusion: The modelling approaches of ACCESS and its features are introduced, and its applicative results for a hypothetical accidental scenario are comprehensively discussed. In an applicative study, the predicted results by the light-water reactor assessment model were higher than those by other models in terms of total doses.

• Korean Chemical Engineering Research
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• v.46 no.5
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• pp.931-937
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• 2008

Rigorous multiscale modelling and simulation of the MTR for WGSR was carried out to accurately predict the behavior of process variables and the reactor performance. The MTR consists of 4 fixed bed tube reactors packed with heterogeneous catalysts, as well as surrounding shell part for the cooling purpose. Considering that fluid flow field and reaction kinetics give a great influence on the reactor performance, employing multiscale methodology encompassing Computational Fluid Dynamics (CFD) and process modeling was natural and, in a sense, inevitable conclusion. Inlet and outlet temperature of the reactant fluid at the tube side was $345^{\circ}C$ and $390^{\circ}C$, respectively and the CO conversion at the exit of the tube side with these conditions approached to about 0.89. At the shell side, the inlet and outlet temperature of the cooling fluid, which flows counter-currently to tube flow, was $190^{\circ}C$ and $240^{\circ}C$. From this heat exchange, the energy saving was achieved for the flow at shell side and temperature of the tube side was properly controlled to obtain high CO conversion. The simulation results from this research were accurately comparable to the experimental data from various papers.

• Structural Engineering and Mechanics
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• v.19 no.1
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• pp.97-106
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• 2005

The standard practice is to seismically qualify the safety related equipment and structural components used in the nuclear power plants. Among several qualification approaches the qualification by the analysis using finite element (FE) method is the most common approach used in practice. However the predictions by the FE model for a structure is known to show significant deviations from the dynamic behaviour of 'as installed' structure in many cases. Considering such limitation, few researchers have advocated re-qualification of such structures after installation at site to enhance the confidence in qualification vis-$\grave{a}$-vis plant safety. For such an exercise the validation of FE model with experimental modal data is important. A validated FE model can be obtained by the Model Updating methods in conjugation with the in-situ experimental modal data. Such a model can then be used for qualification. Seismic analysis using the updated FE model and its advantage has been presented through an example of an in-core component - a perforated horizontal tube of a nuclear reactor.

• Journal of Energy Engineering
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• v.6 no.1
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• pp.87-95
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• 1997

A chemical heat pump based on the reversible reactions between metal chlorides and ammonia gas is attractive alternative to compression system and liquid absorption systems in cooling and refrigerating fields. The advantages of chemical heat pump are no regulatory constants due to CFC refrigerants, utilization of gas, industrial waste heat, electricity, fuel oil etc. as heat sources and wide applications to energy storage system, large-scale energy managements for industrial process. The scale-up of chemical heat pump from laboratory prototype to pilot plants necessitates the interpretation of system performance and evaluation of dynamic behavior in the chemical reactor. This study contains the prediction of performance of chemical refrigerator according to operating condition, the dynamic simulations through reactor modelling, which is used for the calculation of reactive medium temperature and the conversion variation with reactor cooling temperature, and the effect survey of block parameters on the power of refrigerator.