• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.933-940
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• 2024

A macroscopic cross section generation model was developed for the conceptual horizontal, compact high temperature gas reactor (HC-HTGR). Because there are many sources of spectral effects in the design and analysis of the core, conventional LWR methods have limitations for accurate simulation of the HC-HTGR using a neutron diffusion core neutronics simulator. Several super-cell model configurations were investigated to consider the spectral effect of neighboring cells. A new history variable was introduced for the existing library format to more accurately account for the history effect from neighboring nodes and reactivity control drums. The macroscopic cross section library was validated through comparison with cross sections generated using full core Monte Carlo models and single cell cross section for both 3D core steady-state problems and 2D and 3D depletion problems. Core calculations were then performed with the AGREE HTR neutronics and thermal-fluid core simulator using super-cell cross sections. With the new history variable, the super-cell cross sections were in good agreement with the full core cross sections even for problems with significant spectrum change during fuel shuffling and depletion.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1549-1559
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• 2022

As an accident tolerant fuel (ATF) concept, oxide dispersion strengthened Zircaloy-4 (ODS Zry-4) cladding has been developed to enhance the mechanical properties of cladding using laser processing technology. In this study, a simulation technique was established to investigate the mechanical properties and effects of Y₂O₃ particles for the ODS Zry-4. A 3D representative volume element (RVE) model was developed considering the parameters of the size, shape, distribution and volume fraction (VF) of the Y₂O₃ particles. From the 3D RVE model, the Young's modulus, coefficient of thermal expansion (CTE) and creep strain rate of the ODS Zry-4 were effectively calculated. It was observed that the VF of Y₂O₃ particles had a significant effect on the aforementioned mechanical properties. In addition, the predicted properties of ODS Zry-4 were applied to a simulation model to investigate cladding deformation under a transient condition. The ODS Zry-4 cladding showed better performance, such as a delay in large deformation compared to Zry-4 cladding, which was also found experimentally. Accordingly, it is expected that the simulation approach developed here can be efficiently employed to predict more properties and to provide useful information with which to improve ODS Zry-4.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2709-2716
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• 2020

Compositions of large nuclear cores (e.g. boiling water reactors) are highly heterogeneous in terms of fuel composition, control rod insertions and flow regimes. For this reason, they usually lack high order of symmetry (e.g. 1/4, 1/8) making it difficult to estimate their neutronic parameters for large spaces of possible loading patterns. A detailed hyperparameter optimization technique (a combination of manual and Gaussian process search) is used to train and optimize deep neural networks for the prediction of three neutronic parameters for the Ringhals-1 BWR unit: power peaking factors (PPF), control rod bank level, and cycle length. Simulation data is generated based on half-symmetry using PARCS core simulator by shuffling a total of 196 assemblies. The results demonstrate a promising performance by the deep networks as acceptable mean absolute error values are found for the global maximum PPF (~0.2) and for the radially and axially averaged PPF (~0.05). The mean difference between targets and predictions for the control rod level is about 5% insertion depth. Lastly, cycle length labels are predicted with 82% accuracy. The results also demonstrate that 10,000 samples are adequate to capture about 80% of the high-dimensional space, with minor improvements found for larger number of samples. The promising findings of this work prove the ability of deep neural networks to resolve high dimensionality issues of large cores in the nuclear area.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.1
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• pp.87-111
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• 2021

Nuclear power plants (NPPs) produce radioactive waste and decommissioning this waste entails additional cost; determining these costs for various types and specifications of radioactive waste can be challenging. The purpose of this study is to identify major determinants of the decommissioning cost and their impact on NPPs. To this end, data from defunct NPPs were gathered and 2SLS (Two Stage Least Squares) regression models were developed to investigate the major contributors depending on the reactor types, viz. PWR (Pressurized Water Reactors) and BWR (Boiling Water Reactors). Additionally, cost estimations and the Monte Carlo simulation were performed as part of performance validation. Our study established that the decommissioning costs are primarily influenced by the level of radioactivity in the decommissioned waste, which can be realized from operational factors like operation period, overall efficiency, and plant capacity, as well as from duration of decommissioning and labour cost. While our study provides an improved statistical approach to recognize these factors, we acknowledge that our models have limitations in forecasting accurately which we envisage to bolster in future studies by identifying more substantive factors.

• Journal of Radiation Protection and Research
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• v.48 no.1
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• pp.52-57
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• 2023

Background: Korea Atomic Energy Research Institute (KAERI) operates several nuclear research facilities licensed by Nuclear Safety and Security Commission (NSSC). The emergency preparedness requirements, GSR Part 7, by International Atomic Energy Agency (IAEA) request protection strategy based on the hazard assessment that is not applied in Korea. Materials and Methods: In developing the protection strategy, it is important to consider an accident scenario and its consequence. KAERI has tried the hazard assessment based on a hypothesis accident scenario for the major nuclear facilities. During the assessment, the safety analysis report of the related facilities was reviewed, the simulation using MELCOR, MACCS2 code was implemented based on a considered accident scenario of each facility, and the international guidance was considered. Results and Discussion: The results of the optimized protective actions were 300 m evacuation and 800 m sheltering for the High-Flux Advanced Neutron Application Reactor (HANARO), the evacuation to radius 50 m, the sheltering 400 m for post-irradiation examination facility (PIEF), 100 m evacuation or sheltering for HANARO fuel fabrication plant (HFFP) facility. Conclusion: The results of the optimized protective actions and its distances for the KAERI facilities for the maximum postulated accidents were considered in establishing the emergency plan and procedures and implementing an emergency exercise for the KAERI facilities.

• Tunnel and Underground Space
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• v.30 no.4
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• pp.306-319
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• 2020

Characterizations of Excavation Damage Zone (EDZ), which is hydro-mechanical degrading the host rock, are the important issues on the geological repository for the spent nuclear fuel. In the DECOVALEX 2019 project, Task G aimed to model the fractured rock numerically, describe the hydro-mechanical behavior of EDZ, and predict the change of the hydraulic factor during the lifetime of the geological repository. Task G prepared two-dimensional fractured rock model to compare the characteristics of each simulation tools in Work Package 1, validated the extended three-dimensional model using the TAS04 in-situ interference tests from Äspö Hard Rock Laboratory in Work Package 2, and applied the thermal and glacial loads to monitor the long-term hydro-mechanical response on the fractured rock in Work Package 3. Each modelling team adopted both Finite Element Method (FEM) and Discrete Element Method (DEM) to simulate the hydro-mechanical behavior of the fracture rock, and added the various approaches to describe the EDZ and fracture geometry which are appropriate to each simulation method. Therefore, this research can introduce a variety of numerical approaches and considerations to model the geological repository for the spent nuclear fuel in the crystalline fractured rock.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.478-487
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• 2018

Pyroprocessing technology has been regarded as a promising solution for recycling spent fuel in nuclear power plants. The Korea Atomic Energy Research Institute has been studying the current status of equipment and facilities for pyroprocessing and found that existing facilities are manually operated; therefore, their applications have been limited to laboratory scale because of low productivity and safety concerns. To extend the pyroprocessing technology to a commercial scale, the facility, including all the processing equipment and the material-handling devices, should be enhanced in view of automation. In an automated pyroprocessing facility, a supervised control system is needed to handle and manage material flow and associated operations. This article provides a preliminary design of the supervising system for pyroprocessing. In particular, a manufacturing execution system intended for an automated pyroprocessing facility, named Pyroprocessing Execution System, is proposed, by which the overall production process is automated via systematic collaboration with a planning system and a control system. Moreover, a simulation-based prototype system is presented to illustrate the operability of the proposed Pyroprocessing Execution System, and a simulation study to demonstrate the interoperability of the material-handling equipment with processing equipment is also provided.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.445-450
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• 1997

Self-Powered Neutron Detectors(SPNDs) are currently used to estimate the power generation distribution and fuel burn-up in several nuclear power reactors in Korea. In this paper, Monte Carlo simulation is accomplished to calculate the escape probability of beta particle as a function of their birth position fur the typical geometry of rhodium-based SPNDs. Also, a simple numerical method calculates the initial generation rate of beta particles and the change of generation rate due to rhodium burn-up. Using the simulation and the numerical method, the burn-up profile of rhodium density and the neutron sensitivity are calculated as a function of burn-up time in the reactor. The sensitivity of the SPNDs decreases non-linearly due to the high absorption cross-section and the non-uniform burn-up of rhodium in the emitter rod. In addition, for improvement of some properties of rhodium-based SPNDs which are currently used, this paper presents a new material. The method used here can be applied to the analysis of other types of SPNDs and will be useful in the optimum design of new SPNDs for long term usage.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.731-737
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• 2018

In this research paper, the thermal transport in thorium dioxide is investigated by using nonequilibrium molecular dynamics. The thermal conductivity of bulk thorium dioxide was measured to be 20.8 W/m-K, confirming reported values, and the phonon mean free path was estimated to be between 7 and 8.5 nm at 300 K. It was observed that the thermal conductivity of thorium dioxide shows a strong dependency on temperature; the highest thermal conductivity was estimated to be 77.3 W/m-K at 100 K, and the lowest thermal conductivity was estimated to be 4.3 W/m-K at 1200 K. In addition, by simulating thorium dioxide structures with different lengths at different temperatures, it was identified that short wavelength phonons dominate thermal transport in thorium dioxide at high temperatures, resulting in decreased intrinsic phonon mean free paths and minimal effect of boundary scattering while long wavelength phonons dominate the thermal transport in thorium dioxide at low temperatures.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.2
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• pp.195-202
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• 2018

As a rule, geological disposal is considered a safe method for final disposal of high-level radioactive waste. However, some long-lived fission products like $^{99}Tc$ and $^{129}I$ contained in spent nuclear fuel are highly mobile as less sorbing anionic species in the subsurface environment and can mainly cause exposure dose to the ecosystem by emission of beta rays in the hundreds of keV range. Therefore, if these two nuclides can be separated and converted with high efficiency into radioactively unharmful nuclides, this would have a positive effect on disposal safety. One candidate method is to transmute these two nuclides in nuclear reactors into short-lived nuclides or into stable nuclides. For this purpose, it is necessary to evaluate which reactor type is more efficient in burning these two nuclides. In this study, the simulation results of nuclear transmutation of $^{99}Tc$ and $^{129}I$ in light water reactor (PWR), heavy water reactor (CANDU) and fast neutron reactor (SFR, MET-1000) are compared and discussed.