• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.135-136
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.75-76
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.66.1-66
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• 1999

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2792-2800
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• 2022

In the core of the WWR-K reactor, a long-term irradiation of tristructural isotopic (TRISO)-coated fuel particles (CFPs) with a UO₂ kernel was carried out under high-temperature gas-cooled reactor (HTGR)-like operating conditions. The temperature of this TRISO fuel during irradiation varied in the range of 950-1100 ℃. A fission per initial metal atom (FIMA) of uranium burnup of 9.9% was reached. The release of gaseous fission products was measured in-pile. The release-to-birth ratio (R/B) for the fission product isotopes was calculated. Aspects of fuel safety while achieving deep fuel burnup are important and relevant, including maintaining the integrity of the fuel coatings. The main mechanisms of fuel failure are kernel migration, silicon carbide corrosion by palladium, and gas pressure increase inside the CFP. The formation of gaseous fission products and carbon monoxide leads to an increase in the internal pressure in the CFP, which is a dominant failure mechanism of the coatings under this level of burnup. Irradiated fuel compacts were subjected to electric dissociation to isolate the CFPs from the fuel compacts. In addition, nondestructive methods, such as X-ray radiography and gamma spectrometry, were used. The predicted R/B ratio was evaluated using the fission gas release model developed in the high-temperature test reactor (HTTR) project. In the model, both the through-coatings of failed CFPs and as-fabricated uranium contamination were assumed to be sources of the fission gas. The obtained R/B ratio for gaseous fission products allows the finalization and validation of the model for the release of fission products from the CFPs and fuel compacts. The success of the integrity of TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated. A low fuel failure fraction and R/B ratios indicated good performance and reliability of the studied TRISO fuel.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4465-4471
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• 2023

Hands-on education and training is a key part of fixing and developing technology knowledge and is an inherent part of many engineering and scientific curricula. However, access to large complex training facilities, such as nuclear reactor, could be limited by various factors, such as unavailability of those facilities in the region, high traveling costs or harmonization of the schedules of hands-on E&T with theoretical lectures and with the operational schedule of the facility. To handle the issue, several success stories have been reached with the introduction of the Internet Reactor Labs (IRL). The Internet Reactor Labs can strongly contribute to accessibility of training at research reactors and can contribute to improvements in their utilization. The paper describes the development of the Internet Reactor Lab at the VR-1 reactor of the Czech Technical University in Prague. Contrary to single-purpose IRLs, it presents various modalities of online teaching and training in experimental reactor physics and reactor operation in general as well as outreach activities that have been developed in recent years.

• Nuclear Engineering and Technology
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• v.37 no.5
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• pp.433-446
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• 2005

The development of accelerator science and technology has been accommodating ever increasing demand from scientific community of the beam energy and intensity of proton beams. The use of high-powered proton beams has extended from the traditional application of nuclear and high-energy physics to other applications, including spallation neutron source replacing nuclear reactor, nuclear actinide transmutation, energy amplification reactors. This article attempts to review development of proton accelerator, both linear and circular, and issues related to the proton beam energy, intensity as well as its output power. For related accelerator physics and technical review, one should refer to the recent article in the Reviews of Modem Physics [1]

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2315-2324
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• 2023

Space nuclear reactors are becoming popular in deep space exploration owing to their advantages of high-power density and stability. Following the fourth-generation nuclear reactor technology, a conceptual design of the dual drum-controlled space molten salt reactor (D²-SMSR) is proposed. The reactor concept uses molten salt as fuel and heat pipes for cooling. A new reactivity control strategy that combines control drums and safety drums was adopted. Critical physical characteristics such as neutron energy spectrum, neutron flux distribution, power distribution and burnup depth were calculated. Flow and heat transfer characteristics such as natural convection, velocity and temperature distribution of the D²-SMSR under low gravity conditions were analyzed. The reactivity control effect of the dual-drums strategy was evaluated. Results showed that the D²-SMSR with a fast spectrum could operate for 10 years at the full power of 40 kWth. The D²-SMSR has a high heat transfer coefficient between molten salt and heat pipe, which means that the core has a good heat-exchange performance. The new reactivity control strategy can achieve shutdown with one safety drum or three control drums, ensuring high-security standards. The present study can provide a theoretical reference for the design of space nuclear reactors.

• Proceedings of the Korean Vacuum Society Conference
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• 2010.08a
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• pp.314-314
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• 2010

For the optimal design of a tokamak-type reactor, self-consistent determination of a radial build of reactor systems is important and the radial build has to be determined by considering the plasma physics and engineering constraints which inter-relate various reactor systems. In a low aspect ratio (LAR) tokamak reactor with a superconducting toroidal field (TF) coil, the shield should provide sufficient protection for the superconducting TF coil and the shield plays a key role in determining the size of a reactor. To determine the radial build of a reactor, neutronic effects such as tritium breeding in the blanket, nuclear heating, and radiation damage to toroidal field (TF) coil has to be included in the systems analysis. In this work, the outboard blanket only is considered where tritium self-sufficiency is possible by using an inboard neutron reflector instead of breeding blanket. The reflecting shield should provide not only protection for the superconducting TF coil but also improved neutron economy for the tritium breeding in outboard blanket. Tungsten carbide, metal hydride such as titanium hydride and zirconium hydride can be used for improved shielding performance and thus smaller shield thickness. With the use of advanced technology in the shield, conceptual design of a compact superconducting LAR reactor with aspect ratio of less than 2 will be presented as a viable power plant.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1822-1827
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• 2019

Mechanical responses and failure behaviors of advanced C/C composite tube are very important for structural component design in nuclear reactor. In this study, an experimental investigation was conducted to study mechanical properties of C/C composite tube. Quasi-static compression loading was applied to a type of advanced composite tube to determine the response of the quasi-static load displacement curve during progressive damage. Acoustic emissions (AE) signals were captured and analyzed to characterize the crack formation and crack development. In addition, the crack propagation of the specimens was monitored by imaging technique and failure mode of the specimen was analyzed. FEM is appled to simulate the stress distribution. Results show that advanced C/C composite tube exhibits considerable energy absorption capability and stability in load-carrying capacity.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3409-3416
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• 2023

Licensing the next-generation of nuclear reactor designs requires extensive use of Modeling and Simulation (M&S) to investigate system response to many operational conditions, identify possible accidental scenarios and predict their evolution to undesirable consequences that are to be prevented or mitigated via the deployment of adequate safety barriers. Deep Learning (DL) and Artificial Intelligence (AI) can support M&S computationally by providing surrogates of the complex multi-physics high-fidelity models used for design. However, DL and AI are, generally, low-fidelity 'black-box' models that do not assure any structure based on physical laws and constraints, and may, thus, lack interpretability and accuracy of the results. This poses limitations on their credibility and doubts about their adoption for the safety assessment and licensing of novel reactor designs. In this regard, Physics Informed Neural Networks (PINNs) are receiving growing attention for their ability to integrate fundamental physics laws and domain knowledge in the neural networks, thus assuring credible generalization capabilities and credible predictions. This paper presents the use of PINNs as surrogate models for accidental scenarios simulation in Nuclear Power Plants (NPPs). A case study of a Loss of Heat Sink (LOHS) accidental scenario in a Nuclear Battery (NB), a unique class of transportable, plug-and-play microreactors, is considered. A PINN is developed and compared with a Deep Neural Network (DNN). The results show the advantages of PINNs in providing accurate solutions, avoiding overfitting, underfitting and intrinsically ensuring physics-consistent results.