• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2989-2998
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• 2022

Line contact structures, such as the contact between graphite brick and graphite tenon, widely exist in high-temperature gas-cooled reactors. Due to the stress concentration effect, the line contact area is one of the dangerous positions prone to failure in the nuclear reactor core. In this paper, the failure mechanism of line contact structures composed of IG11 nuclear graphite column and brick were investigated by means of experiment and finite element simulation. It was found that the failure process mainly includes three stages: firstly, the damage accumulation in nuclear graphite material led to the characteristic yielding of the line contact structure, but no macroscopic failure can be observed at this stage; secondly, the stresses near the contact area met Mohr failure criterion, and a crack initiated and propagated laterally in the contact zone, that is, local macroscopic failure occurred at this stage; finally, a second crack initiated in the contact area and developed in to a Y-shape, resulting in the final failure of the structure. This study lays a foundation for the structural design and safety assessment of high-temperature gas-cooled reactors.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.763-764
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• 2004

• Nuclear Engineering and Technology
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• 제54권4호
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• pp.1394-1405
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• 2022

The importance of ensuring the inherent safety and security has been more emphasized in recent years to demonstrate the integrity of nuclear facilities under external human-induced events (e.g. aircraft crashes). This work suggests a simulation methodology to effectively evaluate the impact of a commercial aircraft engine onto a dry storage facility. A full-scale engine model was developed and verified by Riera force-time history analysis. A reinforced concrete (RC) structure of a dry storage facility was also developed and material behavior of concrete was incorporated using three constitutive models namely: Continuous Surface Cap, Winfrith, and Karagozian & Case for comparison. Strain-based erosion limits for concrete were suitably defined and the local responses were then compared and analyzed with empirical formulas according to variations in impact velocity. The proposed methodology reasonably predicted such local damage modes of RC structure from the engine missile, and the analysis results agreed well with the calculations of empirical formulas. This research is expected to be helpful in reviewing the dry storage facility design and in the probabilistic risk assessment considering diverse impact scenarios.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.1211-1212
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• 2005

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1661-1668
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• 2020

A physical protection system (PPS) is used for the protection of critical facilities. This paper proposes a structure analytic hierarchy approach (SAHA) for the hierarchical evaluation of the PPS effectiveness in critical infrastructure. SAHA is based on the traditional analysis methods "estimate of adversary sequence interruption, EASI". A community algorithm is used in the building of the SAHA model. SAHA is applied to cluster the associated protection elements for the topological design of complicated PPS with graphical vertexes equivalent to protection elements.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.546-555
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• 2019

Since the first nuclear reactor first critical, nuclear systems has gone through four generations of history, and the fourth generation nuclear system will be truly realized in the near future. The notions of SVBR and lead-bismuth eutectic alloy coolant put forward by Russia were well received by the international nuclear science community. Lead-bismuth eutectic alloy with the ability of the better neutron economy, the low melting point, the high boiling point, the chemical inertness to water and air and other features, which was considered the most promising coolant for the 4th generation nuclear reactors. This study mainly focuses on the structural design optimization of the 4th-generation reactor coolant pump, including analysis of external characteristics, inner flow, and transient characteristic. It was found that: the reactor coolant pump with a central symmetrical dual-outlet volute structure has better radial-direction balance, the pump without guide vane has better hydraulic performance, and the pump with guide vanes has worse torsional vibration and pressure pulsation. This study serves as experience accumulation and technical support for the development of the 4th generation nuclear energy system.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2147-2155
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• 2022

Nuclear safety-related underground liquid storage tanks, such as those used to store fuel for emergency diesel generators, are critical components for safety of hundreds of existing nuclear power plants (NPP) worldwide. Since most of those NPP will continue to operate for decades, a beyond design base (BDB) seismic screening of safety-related underground tanks in those NPP is beneficial and essential to public safety. The analytical methodology for buried tank subjected to seismic effect, including a BDB seismic evaluation, needs to consider both soil-structure and fluid-structure interaction effects. Comprehensive analysis of such a soil-structure-fluid system is costly and time consuming, often subjected to availability of state-of-art finite element tools. Simple, but practically and reasonably accurate techniques for seismic evaluation of underground liquid storage tanks have not been established. In this study, a mechanics based solution is proposed for the evaluation of a cylindrical underground liquid storage tank using hand calculation methods. For validation, a practical example of two underground diesel fuel tanks in an existing nuclear power plant is presented and application of the proposed method is confirmed by using published results of the computer-aided System for Analysis of Soil Structural Interaction (SASSI). The proposed approach provides an easy to use tool for BDB seismic assessment prior to making decision of applying more costly technique by owner of the nuclear facility.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2019-2028
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• 2024

Advanced nuclear reactors, especially the newly developed small and micro-reactors have complex neutron spectrum, which makes the deterministic reactor core calculations sensitive to the energy group structure of few-group cross-sections. To avoid significantly increasing the cost of energy discretization in the core calculation, two energy group structures with 31 groups and 33 groups were adopted for typical thermal and fast reactor cores, respectively. Then, an adaptive scheme of group division for reactor cores with a medium neutron spectrum was proposed. The works were based on the full spectrum nuclear reactor analysis code SARAX/TULIP. An equivalent one-dimensional model of the core was proposed to capture the key neutron spectrum features of the reactor core. Such features were used to adaptively determine a few-group structure for the following reactor core calculations. Then, the neutron spectrum in different zones with more details was calculated. With this spectrum, the cross-sections were condensed into the determined energy groups. Three tests based on different neutron spectrum were calculated to verify the schemes. The results show that using the adaptive energy group division scheme, the following core calculation can meet the accuracy requirement of different reactors with different neutron spectra.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1997년도 추계학술대회 논문집 학회본부
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• pp.588-590
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• 1997

In this paper, a highly reliable communication network for DCS in nuclear power plant is designed. The structure and characteristics of DCS in nuclear power plant is briefly explained. The features needed for a communication network for DCS in nuclear power plant is described. According to the abovo features, the layer structure of the communication network is determined and each layer is designed in detail.

• Nuclear Engineering and Technology
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• 제38권7호
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• pp.681-688
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• 2006

This study investigates the dynamic response characteristics of a structure impacted by a high speed projectile. The impact of a 300 kg projectile on a water storage tank is simulated by the general purpose computer codes ANSYS and LS-DYNA. Several methods to simulate the impact are considered and their results are compared. Based upon this, an alternative impact analysis method that is equivalent to an explicit dynamic analysis is proposed. The effect of fluid on the responses of the tank is also addressed.