• Journal of computational fluids engineering
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• v.14 no.2
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• pp.59-67
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• 2009

An accurate prediction of the bypass flow is of great importance in the VHTR core design concerning the fuel thermal margin. Nevertheless, there has not been much effort in evaluating the amount and the distribution of the core bypass flow. In order to evaluate the behavior and the distribution of the coolant flow, a unit-cell experiment was carried out. Unit-cell is the regular triangular section which is formed by connecting the centers of three hexagonal blocks. Various conditions such as the inlet mass flow rate, block combinations and the size of bypass gap were examined in the experiment. CFD analysis was carried out to analyze detailed characteristics of the flow distribution. Commercial CFD code FLUENT 6.3 was validated by comparing with the experimental results. In addition, SST model and standard k-$\varepsilon$ model were validated. The results of CFD simulation show good agreements with the experimental results. SST model shows better agreement than standard k-$\varepsilon$ model. Results showed that block combinations and the size of the bypass gap have an influence on the bypass flow ratio but the inlet mass flow rate does not.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.304-317
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• 2016

An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan), a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a Cockcrofte-Walton type accelerator, which generates the external neutron source by deuteriu-metritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.

• 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.

• Korean Journal of Metals and Materials
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• v.50 no.10
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• pp.721-728
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• 2012

This paper evaluates the reliability of creep rupture life under service conditions of Alloy 617, which is considered as one of the candidate materials for use in a very high temperature reactor (VHTR) system. A Z-parameter, which represents the deviation of creep rupture data from the master curve, was used for the reliability analysis of the creep rupture data of Alloy 617. A Service-condition Creep Rupture Interference (SCRI) model, which can consider both the scattering of the creep rupture data and the fluctuations of temperature and stress under any service conditions, was also used for evaluating the reliability of creep rupture life. The statistical analysis showed that the scattering of creep rupture data based on Z-parameter was supported by normal distribution. The values of reliability decreased rapidly with increasing amplitudes of temperature and stress fluctuations. The results established that the reliability decreased with an increasing service time.

• Korean Journal of Metals and Materials
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• v.47 no.10
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• pp.613-620
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• 2009

The Kachanov-Rabotnov (K-R) creep model was proposed to accurately model the long-term creep curves above $10^5$ hours of Alloy 617. To this end, a series of creep data was obtained from creep tests conducted under different stress levels at $950^{\circ}C$. Using these data, the creep constants used in the K-R model and the modified K-R model were determined by a nonlinear least square fitting (NLSF) method, respectively. The K-R model yielded poor correspondence with the experimental curves, but the modified K-R model provided good agreement with the curves. Log-log plots of ${\varepsilon}^{\ast}$-stress and ${\varepsilon}^{\ast}$-time to rupture showed good linear relationships. Constants in the modified K-R model were obtained as ${\lambda}$=2.78, and $k=1.24$, and they showed behavior close to stress independency. Using these constants, long-term creep curves above $10^5$ hours obtained from short-term creep data can be modeled by implementing the modified K-R model.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.832-845
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• 2024

Reactor cavity cooling systems (RCCSs) comprising passive safety features use the atmosphere as a coolant, which cannot be lost. However, their drawback is that they are easily affected by atmospheric disturbances. To realize the commercial application of the two types of passive RCCSs, namely RCCSs based on atmospheric radiation and atmospheric natural circulation, their safety must be evaluated, that is, they must be able to remove heat from the reactor pressure vessel (RPV) surface at all times and under any condition other than under normal operating conditions. These include both expected and unexpected natural phenomena and accidents. Moreover, they must be able to eliminate the heat leakage emitted from the RPV surface during normal operation. However, utilizing all of the heat emitted from the RPV surface increases the degree of waste heat utilization. This study aims to understand the characteristics and degree of passive safety features for heat removal by comparing RCCSs based on atmospheric radiation and atmospheric natural circulation under the same conditions. It was concluded that the proposed RCCS based on atmospheric radiation has an advantage in that the temperature of the RPV could be stably maintained against disturbances in the ambient air.

• Journal of the Korean Ceramic Society
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• v.50 no.1
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• pp.37-42
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• 2013

Tristructural-isotropic (TRISO)-coated particles were fabricated by a fluidized-bed chemical vapor deposition (FBCVD) method for use in a very high temperature gas-cooled reactor (VHTR). ZrC as a constituent layer of TRISO coating layers was deposited by a chloride process using $ZrCl_4$ and $CH_4$ source gases in a temperature range of $1400^{\circ}C$ and $1550^{\circ}C$. The change in the microstructure of ZrC depending on the deposition temperature and its effect on the hardness were evaluated. As the deposition temperature increased to $1500^{\circ}C$, the grain size of the ZrC increased and the hardness of the ZrC decreased according to the Hall-Petch relationship. However, at $1550^{\circ}C$, the ZrC layer was highly non-stoichiometric and carbon-rich and did not obey the Hall-Petch relationship in spite of the decrease of the grain size. A considerable amount of pyrolytic carbon at the grain boundaries of the ZrC as well as coarse granular pyrolytic carbon were locally distributed in the ZrC layer deposited at $1550^{\circ}C$. Therefore, the hardness decreased largely due to the formation of a large amount of pyrolytic carbon in the ZrC layer.

• Transactions of the KSME C: Technology and Education
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• v.3 no.4
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• pp.299-305
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• 2015

Nuclear hydrogen production technology is being developed for the future energy supply system. The sulfur-iodine thermo-chemical hydrogen production process directly splits water by using of the heat generated from very high temperature gas-cooled reactor, a typical Generation IV nuclear system. Nuclear hydrogen key technologies are composed of VHTR simulation technology at elevated temperature, computational tools, TRISO fuel, and sulfur iodine hydrogen production technology. Key technology for nuclear hydrogen production system were developed and demonstrated in a laboratory scale test facility. Technical challenges for the commercial hydrogen production system were discussed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.2
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• pp.34-41
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• 2011

Wrought nickel-base superalloys are being considered as the structural materials in very-high temperature gas-cooled reactors. To understand the effects of impurities, especially oxygen, in helium coolant on the mechanical properties of Alloy 617, creep tests were performed in high temperature flowing He environments with varying $O_2$ contents at 800, 900, and $1000^{\circ}C$. Also, creep life in static He was measured to simulate the pseudo-inert environment. Creep life was the longest in static He, while the shortest in flowing helium. In static He, impurities like $O_2$ and moisture were quickly consumed by oxidation in the early stage of creep test, which prevented further oxidation during creep test. Without oxidation, microstructural change detrimental to creep such as decarburization and internal oxidation were prevented, which resulted in longer creep life. On the other hand, in flowing He environment, surface oxides were not stable enough to act as diffusion barriers for oxidation. Therefore, extensive decarburization and internal oxidation under tensile load contributed to premature failure resulting in short creep life. Limited test in flowing He+200ppm $O_2$ resulted in even shorter creep life. The oxidation samples showed extensive spallation which resulted in severe decarburization and internal oxidation in those environments. Further test and analysis are underway to clarify the relationship between oxidation and creep resistance.

• Journal of Power System Engineering
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• v.18 no.5
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• pp.122-128
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• 2014

Generally, the mechanical components and structures are joined by many welding techniques, and therefore the welded joints are inevitable in the construction of structures. The Alloy 617 was initially developed for high temperature applications above $800^{\circ}C$. It is often considered for use in aircraft and gas turbines, chemical manufacturing components, and power generation structures. Especially, the Alloy 617 is the primary candidate for construction of intermediate heat exchanger (IHX) on a very high temperature reactor (VHTR) system. In the present paper, the low cycle fatigue (LCF) life of Alloy 617 base metal (BM) and the gas tungsten arc welded (GTAWed) weld joints (WJ) are evaluated by using the previous experimental results under strain controlled LCF tests. The LCF tests have been performed at room temperature with total strain ranges of 0.6, 0.9, 1.2 and 1.5%. The LCF lives for the BM and WJ have been evaluated from the Coffin-Manson and strain energy based life methods. For both the BM and WJ, the LCF lives predicted by both Coffin-Manson and strain energy based life methods was found to well coincide with the experimental data.