• The Korean Journal of Mycology
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• v.40 no.1
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• pp.49-53
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• 2012

Hot-water extracted natural media were made from raw materials for mycelial culture of Pleurotus eryngii. Poplar sawdust, wheat bran and rice bran were used as substrates for hot water extraction. The mixed substrates of poplar sawdust, wheat bran, and rice bran with 50 : 20 : 30 (v/v/v, PWR523) and 50 : 30 : 20 (v/v/v, PWR532) were optimal for mycelial growth of P. eryngii, respectively. The hot-water extracted natural media from PWR523 and PWR532 showed a rapid mycelial growth and spawn running compared to PDA. There was no significant difference in mushroom yield when the mycelium grown on the hot-water extracted natural media was used as the inoculum source for producing fruit body.

• Korean Journal of Materials Research
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• v.25 no.4
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• pp.183-190
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• 2015

Fatigue crack growth rate tests were conducted as a function of temperature, dissolved hydrogen (DH) level, and frequency in a simulated PWR environment. Fatigue crack growth rates increased slightly with increasing temperature in air. However, the fatigue crack growth rate did not change with increasing temperature in PWR water conditions. The DH levels did not affect the measured crack growth rate under the given test conditions. At $316^{\circ}C$, oxides were observed on the fatigue crack surface, where the size of the oxide particles was about $0.2{\mu}m$ at 5 ppb. Fatigue crack growth rate increased slightly with decreasing frequency within the frequency range of 0.1 Hz and 10 Hz in PWR water conditions; however, crack growth rate increased considerably at 0.01 Hz. The decrease of the fatigue crack growth rate in PWR water condition is attributed to crack closure resulting from the formation of oxides near the crack tips at a rather fast loading frequency of 10 Hz.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.479-489
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• 2019

Low-cycle fatigue (LCF) tests were performed for Alloy 690 and 316 SS in a simulated pressurized water reactor (PWR) environment. Alloy 690 showed about twice longer LCF life than 316 SS at the test condition of 0.4% amplitude at strain rate of 0.004%/s. Observation of the oxide layers formed on the fatigue crack surface showed that Cr and Ni rich oxide was formed for Alloy 690, while Fe and Cr rich oxide for 316 SS as an inner layer. Electrochemical analysis revealed that the oxide layers formed on the LCF crack surface of Alloy 690 had higher impedance and less defect density than those of 316 SS, which resulted in longer LCF life of Alloy 690 than 316 SS in a simulated PWR environment.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.851-859
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• 2021

Small PWR can be used for power generation and heating. Considering that small PWR has the characteristics of flexible operating conditions and complex operating environment, the controller designed based on single power level is difficult to achieve the ideal control of small PWR in the whole range of core power range. To solve this problem, a flexible switching controller based on fuzzy controller and LQG/LTR controller is designed. Firstly, a core fuzzy multi-model suitable for full power range is established. Then, T-S fuzzy rules are designed to realize the flexible switching between fuzzy controller and LQG/LTR controller. Finally, based on the core power feedback principle, the core flexible switching control system of small PWR is established and simulated. The results show that the flexible switching controller can effectively control the core power of small PWR and the control effect has the advantages of both fuzzy controller and LQG/LTR controller.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.668-674
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• 2017

Experimental verification of boron species in fuel CRUD (Chalk River Unidentified Deposit) would provide essential and important information about the root cause of CRUD-induced power shifts (CIPS). To date, only bonaccordite and elemental boron were reported to exist in fuel CRUD in CIPS-troubled pressurized water reactor (PWR) cores and lithium tetraborate to exist in simulated PWR fuel CRUD from some autoclave tests. We have reevaluated previous analysis of similar threadlike crystals along with examining some similar threadlike crystals from CRUD samples collected from a PWR cycle that had no indications of CIPS. These threadlike crystals have a typical [Ni]/[Fe] atomic ratio of ~2 and similar crystal morphology as the one (bonaccordite) reported previously. In addition to electron diffraction study, we have applied electron energy loss spectroscopy to determine boron content in such a crystal and found a good agreement with that of bonaccordite. Surprisingly, such crystals seem to appear also on corroded surfaces of Alloy 600 that was exposed to simulated PWR primary water with a dissolved hydrogen level of $5mL\;H_2/kg\;H_2O$, but absent when exposed under $75mL\;H_2/kg\;H_2O$ condition. It remains to be verified as to what extent and in which chemical environment this phase would be formed in PWR primary systems.

• Journal of Radiation Protection and Research
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• v.42 no.2
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• pp.106-113
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• 2017

Background: Interfacing-system loss-of-coolant-accident (ISLOCA) has been identified as the most hazardous accident scenario in the typical PWR plants. The present study as an effort to improve the knowledge of the source term to the environment during ISLOCA focuses on an improvement of the estimation method. Materials and Methods: The improvement was performed to take into account an effect of broken pipeline and auxiliary building structures relevant to ISLOCA. An estimation of the source term to the environment was for the OPR-1000 plants by MELOCR code version 1.8.6. Results and Discussion: The key features of the source term showed that the massive amount of fission products departed from the beginning of core degradation to the vessel breach. Conclusion: The release amount of fission products may be affected by the broken pipeline and the auxiliary building structure associated with release pathway.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.178-183
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• 2004

For developing fatigue design curve of cast stainless steels that would be used in piping material of domestic nuclear power plants, a low-cycle fatigue test rig was built. It is capable of performing tests in pressurized high temperature water environment of PWR. Cylindrical specimens of CF8M were used for the strain-controlled environmental fatigue tests. Fatigue life was measured in terms of the number of cycles with the variation of strain amplitude at 0.04%/s strain rates. The disparity between target length and measured length of specimens was corrected by using finite element method. The corrected test results showed similar fatigue life trend with another previous results.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.308-308
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• 1998

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2880-2886
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• 2020

A small leak occurring on the surface of a fuel rod due to damage exposes UO₂ to a steam atmosphere. During this time, fission gas trapped inside the fuel rod leaks out, and the gas leakage can be increased due to UO₂ oxidation. Numerous studies have focused on the steam oxidation and its thermodynamic calculation in UO₂. However, the thermodynamic calculation of the UO₂ oxidation in a pressurized water reactor (PWR) environment has not been studied extensively. Moreover, the kinetics of the oxidation of UO₂ pellet also has not been investigated. Therefore, in this study, the thermodynamics of UO₂ oxidation under steam injection due to a damaged fuel rod in a PWR environment is studied. In addition, the diminishing radius of the UO₂ pellet with time in the PWR environment was calculated through an experiment simulating the initial time of steam injection at the puncture.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.301-301
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• 2003