• Journal of Physiology & Pathology in Korean Medicine
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• v.16 no.2
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• pp.220-225
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• 2002

To determine the anti-stress effect of KCG(加味天麻鉤藤飮) extract on sprague-dawley rats. we conducted a research about the change of weight, activity, reactivity, c-fos protein, cytotoxicity against PC12 cell line and heal shock protein. 1) KCG extract siginificantly inhibited the decrease of body weight induced by stress, compared with the control group. 2) KCG extract had no siginificant effect in the activity and reactivity of rats between the control and the experimental groups. 3) KCG extract siginificantly restrained c-fos protein manifestation, compared with the control group. 4) KCG extract siginificantly restrained heat shock protein, compared with the control group. These results suggested that KCG might be usefully anti-stress effect.

• Journal of the Korean Institute of Telematics and Electronics
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• v.4 no.3
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• pp.19-27
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• 1967

In the reactor control system, it is one of important part to know the accurate reactivity variability in order to keep the safe operation of the reactor which has the high flux neutron density, a short neutron life time and a large sudden reactivity insertion. For this problem, the reactivity measurement study is described by the analog computing method using the characteristics of the reactor kinetic equation.

• Proceedings of the KIEE Conference
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• 2007.07a
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• pp.1626-1627
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• 2007

This paper deals with the design and manufacturing of power cabinet for reactor power control system(PCS). The PCS provides the control signals and motive power to operate the CEDMs(Control Element Drive Mechanism). The CEDM is raise and lower the CEAs(Control Element Assemblies) in the reactor core. The CEAs are constructed with the Boron-10 isotope which has a high microscopic cross section of absorption for thermal neutrons. This characteristic causes the addition of negative reactivity when a CEA is inserted and positive reactivity when it is withdrawn from the reactor core.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1715-1720
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• 2019

Jozef Stefan Institute (JSI), TRIGA Mark II reactor employs the homogeneous mixture of uranium and zirconium hydride fuel type. Since its upgrade, a series of fresh fuel steady state experimental benchmarks have been conducted. The benchmark results have provided data for testing computational neutronics codes which are important for reactor design and safety analysis. In this work, we investigated the JSI TRIGA Mark II reactor neutronics characteristics: the effective multiplication factor and two safety parameters, namely the control rod worth and the fuel temperature reactivity coefficient using SuperMC. The modeling and real-time cross section generation methods of SuperMC were evaluated in the investigation. The calculation analysis indicated the following: the effective multiplication factor was influenced by the different cross section data libraries; the control rod worth evaluation was better with Monte Carlo codes; the experimental fuel temperature reactivity coefficient was smaller than calculated results due to change in water temperature. All the results were in good agreement with the experimental values. Hence, SuperMC could be used for the designing and benchmarking of other TRIGA Mark II reactors.

• Proceedings of the KIEE Conference
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• 2006.07d
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• pp.1919-1920
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• 2006

This paper deals with the design review of a power converter topologies for CEDMCS (Control Element Drive Mechanism Control System). The CEDMCS provides the control signals and motive power to operate the CEDMS. The CEDM's raise and lower the CEAs (Control Element Assemblies) in the reactor core. The CEAs are constructed with the Boron-10 isotope which has a high microscopic cross section of absorption for thermal neutrons. This characteristic causes the addition of negative reactivity when a CEA is inserted and positive reactivity when it is withdrawn from the reactor core.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 1990.04a
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• pp.44-49
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• 1990

This study is a seismic analysia of absorber rod in KMRR Reactivity Control Mechanism. The model being studied i8 two coaxial tubes (control absorber rod and flow tube) immersed in the water and partially coupled (overlap) by water Hap. The hydrodynamic mass effects by the water in each surrounding conditions are considered in the model. The natural frequencies, stresses and displacements of the system due to Safe Shutdown Earthquake are computed in the cases of in-phase modes and out-of-phase modes of two coaxial tubes. The results show that maximum stresses are well below the allowable limit and maximum displacements at the ends of both tubes in out-of-phase modes are so huck that the tubes contact each other in the overlap zone.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.163-168
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• 1996

Simulations of Wolsong-1 Phase-B commissioning measurements have been performed, as part of the program to validate WIMS-AECL lattice cell calculations for application to CANDU reactor simulations in RFSP. A required component of these simulations is the calculation of incremental cross sections representing reactivity control devices in the reactor. The incremental cross section properties of the Wolsong-1 adjusters, Mechanical Control Absorbers (MCA) and liquid Zone Control Units (ZCU) are based on the WIMS-AECL/MULTICELL modelling methods and the results are compared with those of WIMS-AECL/DRAGON-2 modelling methods.

• Journal of Energy Engineering
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• v.16 no.4
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• pp.155-163
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• 2007

To insure nuclear reactor safety, the reactivity of control rods should be calculated by measuring the criticality of reactor core and it is regularly performed during the annual physics test period. Also, the core criticality should be monitored during the start-up operation to avoid reactivity induced accidents. Many research works on control rod reactivity measurement and subcriticality measurement have been accomplished throughout the world for decades and recently a new method named "Modified Neutron Source Multiplication Method (MNSM)" was proposed in Japan which is known to be improved overcoming limitations of traditional Neutron Source Multiplication Method (NSM). In this study, MNSM was tested in calculation of subcriticalities and in evaluation of application validity using the educational reactor in Kyung Hee University, AGN-201. For this study, a revised nuclear data library and a neutron transport code system TRANSX - PARTISN were established. Correction factors for various control rod positions were produced using the k-effective values and the corresponding flux distributions and adjoint flux distributions. Experimental values of the core criticality were obtained using the neutron count rates of the BF3 proportional counters. The results showed that the expected reactivity worth of control rods by MNSM agreed well with the theoretical values and the correction factors contributed much for this purpose.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.41-41
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.464-464
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• 1999