• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.729-738
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• 2009

Given the evolution of High-Temperature Gas-cooled Reactor(HTGR) designs, the source terms for licensing must be developed. There are three potential source terms: fission products, actinides in the fuel and tritium in the coolant. It is necessary to provide first an inventory of the source terms under normal operations. An analysis of source terms has yet to be performed for HTGRs. The previous code, which can estimate the inventory of the source terms for LWRs, cannot be used for HTGRs because the general data of a typical neutron cross-section and flux has not been developed. Thus, this paper uses a combination of the MCNP, ORIGEN, and MONTETEBURNS codes for an estimation of the source terms. A method in which the HTR-10 core is constructed using the unit lattice of a body-centered cubic is developed for core modeling. Based on this modeling method by MCNP, the generation of fission products, actinides and tritium with an increase in the burnup ratio is simulated. The model developed by MCNP appears feasible through a comparison with models developed in previous studies. Continuous fuel management is divided into five periods for the feeding and discharging of fuel pebbles. This discrete fuel management scheme is employed using the MONTEBURNS code. Finally, the work is investigated for 22 isotope fission products of nuclides, 22 actinides in the core, and tritium in the coolant. The activities are mainly distributed within the range of $10^{15}{\sim}10^{17}$ Bq in the equilibrium core of HTR-10. The results appear to be highly probable, and they would be informative when the spent fuel of HTGRs is taken into account. The tritium inventory in the primary coolant is also taken into account without a helium purification system. This article can lay a foundation for future work on analyses of source terms as a platform for safety assessment in HTGRs.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2743-2759
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• 2020

A detailed computational fluid dynamics (CFD) simulation analysis model was developed using ANSYS CFX 16.1 and analyzed to simulate the basic design and internal flow characteristics of a 180 MW small modular reactor (SMR) with a natural circulation flow system. To analyze the natural circulation phenomena without a pump for the initial flow generation inside the reactor, the flow characteristics were evaluated for each output assuming various initial powers relative to the critical condition. The eddy phenomenon and the flow imbalance phenomenon at each output were confirmed, and a flow leveling structure under the core was proposed for an optimization of the internal natural circulation flow. In the steady-state analysis, the temperature distribution and heat transfer speed at each position considering an increase in the output power of the core were calculated, and the conceptual design of the SMR had a sufficient thermal margin (31.4 K). A transient model with the output ranging from 0% to 100% was analyzed, and the obtained values were close to the T_hot and T_cold temperature difference value estimated in the conceptual design of the SMR. The K-factor was calculated from the flow analysis data of the CFX model and applied to an analysis model in RELAP5/MOD3.3, the optimal analysis system code for nuclear power plants. The CFX analysis results and RELAP analysis results were evaluated in terms of the internal flow characteristics per core output. The two codes, which model the same nuclear power plant, have different flow analysis schemes but can be used complementarily. In particular, it will be useful to carry out detailed studies of the timing of the steam generator intervention when an SMR is activated. The thermal and hydraulic characteristics of the models that applied porous media to the core & steam generators and the models that embodied the entire detail shape were compared and analyzed. Although there were differences in the ability to analyze detailed flow characteristics at some low powers, it was confirmed that there was no significant difference in the thermal hydraulic characteristics' analysis of the SMR system's conceptual design.

• Nuclear Engineering and Technology
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• v.37 no.1
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• pp.25-78
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• 2005

As a key and core knowledge for the design of various types of nuclear reactors, the discipline of reactor physics has been advanced continually in the past six decades and has led to a very sophisticated fabric of analysis methods and computer codes in use today. Notwithstanding, the discipline faces interesting challenges from next-generation nuclear reactors and innovative new fuel designs in the coming. After presenting a brief overview of important tasks and steps involved in the nuclear design and analysis of a reactor, this article focuses on the currently-used design and analysis methods, issues and limitations, and current activities to resolve them as follows: (1) Derivation of the multi group transport equations and the multi group diffusion equations, with representative solution methods thereof. (2) Elements of modem (now almost three decades old) diffusion nodal methods. (3) Limitations of nodal methods such as transverse integration, flux reconstruction, and analysis of UO2-MOX mixed cores. Homogenization and related issues. (4) Description of the analytic function expansion nodal (AFEN) method. (5) Ongoing efforts for three-dimensional whole-core heterogeneous transport calculations and acceleration methods. (6) Elements of spatial kinetics calculation methods and coupled neutronics and thermal-hydraulics transient analysis. (7) Identification of future research and development areas in advanced reactors and Generation-IV reactors, in particular, in very high temperature gas reactor (VHTR) cores.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.4
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• pp.405-413
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• 2015

In the US, the number of cases of subterranean water contamination from tritium leaking through a damaged buried nuclear power plant pipe continues to increase, and the degradation of the buried metal piping is emerging as a major issue. A pipe blocked from corrosion and/or degradation can lead to loss of cooling capacity in safety-related piping resulting in critical issues related to the safety and integrity of nuclear power plant operation. The ASME Boiler and Pressure Vessel Codes Committee (BPVC) has recently approved Code Case N-755 that describes the requirements for the use of polyethylene (PE) pipe for the construction of Section III, Division 1 Class 3 buried piping systems for service water applications in nuclear power plants. This paper contains tensile and slow crack growth (SCG) test results for high-density polyethylene (HDPE) pipe welds under the environmental conditions of a nuclear power plant. Based on these tests, the fracture surface of the PENT specimen was analyzed, and the fracture mechanisms of each fracture area were determined. Finally, by using 3D finite element analysis, limit loads of HDPE related to premature failure were verified.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.11 no.9
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• pp.3517-3523
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• 2010

Single exposure dual X-ray imaging can be used to separate soft and dense-material images for medical and industrial applications. This study keep focusing baggage inspection system(BIS) specifically. New detector modules for single exposure dual X-ray imaging are consisted of low energy detector (LED) and high energy detector (HED). First, the optimized thickness of copper filter coupled HED to separate low energy and high energy was simulated by the given X-ray energy (140 kVp, 1 mA) using Monte Carlo simulation codes, MCNPX. So as a result of simulation, the copper filter thickness is 0.7 mm. For the design of PIN photodiode, ATLAS device simulation tool was used. 16 channels PIN photodiode of 1.5 mm ${\times}$ 3.2 mm for Dual X-ray imaging detector was fabricated in the process of ETRI. And its dark current and quantum efficiency, terminal capacitance were measured. It was proven that the Lanex Fast B coupled HED were a sufficient candidate to replace the CsI(Tl) commerced in dual X-ray system, since these give a strong signal, overcoming system noise. Finally dual X-ray image was acquired through correction of the LED X-ray Image and the HED X-ray Image.

• KSII Transactions on Internet and Information Systems (TIIS)
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• v.13 no.2
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• pp.1100-1123
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• 2019

A code-based cryptosystem can resist quantum-computing attacks. However, an original system based on the Goppa code has a large key size, which makes it unpractical in embedded devices with limited sources. Many special error-correcting codes have recently been developed to reduce the key size, and yet these systems are easily broken through side channel attacks, particularly differential power analysis (DPA) attacks, when they are applied to hardware devices. To address this problem, a higher-order masking scheme for a McEliece cryptosystem based on the quasi-dyadic moderate density parity check (QD-MDPC) code has been proposed. The proposed scheme has a small key size and is able to resist DPA attacks. In this paper, a novel McEliece cryptosystem based on the QD-MDPC code is demonstrated. The key size of this novel cryptosystem is reduced by 78 times, which meets the requirements of embedded devices. Further, based on the novel cryptosystem, a higher-order masking scheme was developed by constructing an extension Ishai-Sahai-Wagne (ISW) masking scheme. The authenticity and integrity analysis verify that the proposed scheme has higher security than conventional approaches. Finally, a side channel attack experiment was also conducted to verify that the novel masking system is able to defend against high-order DPA attacks on hardware devices. Based on the experimental validation, it can be concluded that the proposed higher-order masking scheme can be applied as an advanced protection solution for devices with limited resources.

• Journal of the Korea Institute of Information Security & Cryptology
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• v.34 no.2
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• pp.157-166
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• 2024

Probability of decryption failure of a public key cryptography based on LWE(learning with errors) is determined by its architecture and parameter settings. Since large decryption failure probability leads to attacks[1] on scheme as well as degradation of performance, TiGER[2], a Ring-LWE(R)-based KEM proposed for the first round of KpqC, tried to reduce the decryption failure probability by using error correction code Xef and D2 encoding method. However, D'Anvers et al. has shown that the commonly assumed independence of each bit error is not established since in the case of an encryption scheme based on Ring-LWE(R) using an error correction code, there is error dependency which is not negligible[3]. In this paper, since TiGER does not consider the error dependency, we calcualte the decryption failure probability of TiGER by considering the error dependency. In addition, we found that the bit error probability is incorrectly calculated in TiGER, so we present the correct calculation.