• Journal of the Institute of Electronics Engineers of Korea TC
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• v.41 no.9
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• pp.55-65
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• 2004

ATM was recommended by the ITU and ATM Forum as a means of transportation for B-ISDN. At this time, due to the comprehensive mature of ATM protocol, ATM has been adapted as the backbone system for carrying Internet traffi $c^{[1,2,3,4]}$. But major conceptsregarding the ATN protocol will be used on future technology. This paper presents preventive congestion control mechanisms for detecting HTR(Hard-To Reach) code in ATM systems, in particular for an improved HTR call registration method using network signaling information will discussed. In high speed circuit switching system environments, a fast HTR control mechanism is necessary. We present research results for improving HTR call registration and control methods using network signaling information and fuzzy control mechanisms. We concluded that it showed fast congestion avoidance mechanisms with a fewer system load maximized the efficiency of network resources by restricting ineffective machine attempts.

• Nuclear Engineering and Technology
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• v.41 no.3
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• pp.307-318
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• 2009

KAERI (Korea Atomic Energy Research Institute) has developed the GAMMA+ code for a thermo-fluid and safety analysis of a VHTR (Very High Temperature Gas-Cooled Reactor). A key safety issue of the VHTR design is to demonstrate its inherent safety features for an automatic reactor power trip and power stabilization during an anticipated transient without scram (ATWS) accident such as a loss of forced cooling by a trip of the helium circulator (LOFC) or a reactivity insertion by a control rod withdrawal (CRW). This paper intends to show the ATWS assessment capability of the GAMMA+ code which can simulate the reactor power response by solving the point-kinetic equations with six-group delayed neutrons, by considering the reactivity changes due to the effects of a core temperature variation, xenon transients, and reactivity insertions. The present benchmark calculations are performed by using the safety demonstration experiments of the 10 MW high temperature gas cooled-test module (HTR-10) in China. The calculation results of the power response transients and the solid core temperature behavior are compared with the experimental data of a LOFC ATWS test and two CRW ATWS tests by using a 1mk-control rod and a 5mk-control rod, respectively. The GAMMA+ code predicts the power response transients very well for the LOFC and CRW ATWS tests in HTR-10.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.674-680
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• 2020

We used MCNP6 computer code to model HTR-10 core reactor. We used two types of fuel; UO₂ and (Th+Pu)O₂ mixture. We determined the critical height at which the reactor approached criticality in both two cases. The neutronic and burnup parameters were investigated. The results indicated that the core fueled with mixed (Th+Pu)O₂, achieved about 24% higher fuel cycle length than the UO₂ case. It also enhanced safeguard security by burning Pu isotopes. The results were compared with previously published papers and good agreements were found.

• Journal of Energy Engineering
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• v.14 no.4 s.44
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• pp.232-240
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• 2005

In this study, the IAEA Benchmark problems far HTR-10 and HTTR RCCS were assessed in order to assess the applicability of MARS code, a thermal-hydraulic safety analysis code developed for water reactors. The calculated results were compared with those or THERMIX, THANPACST2 code, and available experimental data. The calculated results showed generally good agreements with those obtained by the THERMIX code and THANPACST2 code. Deviations were analyzed to be originated from the simplification of complicated geometry and from the modeling capability of heat transfer characteristics in the HTGR components such as water cooler and air tooler. Especially, it was found that the radiation heat transfer in the reactor cavity played an important role in the after heat removal in the RCCS. Thus, it is concluded that MARS code can be successfully applied to the calculation of the RCCS cooling capability of the HTGR in this study.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.729-738
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• 2009

Given the evolution of High-Temperature Gas-cooled Reactor(HTGR) designs, the source terms for licensing must be developed. There are three potential source terms: fission products, actinides in the fuel and tritium in the coolant. It is necessary to provide first an inventory of the source terms under normal operations. An analysis of source terms has yet to be performed for HTGRs. The previous code, which can estimate the inventory of the source terms for LWRs, cannot be used for HTGRs because the general data of a typical neutron cross-section and flux has not been developed. Thus, this paper uses a combination of the MCNP, ORIGEN, and MONTETEBURNS codes for an estimation of the source terms. A method in which the HTR-10 core is constructed using the unit lattice of a body-centered cubic is developed for core modeling. Based on this modeling method by MCNP, the generation of fission products, actinides and tritium with an increase in the burnup ratio is simulated. The model developed by MCNP appears feasible through a comparison with models developed in previous studies. Continuous fuel management is divided into five periods for the feeding and discharging of fuel pebbles. This discrete fuel management scheme is employed using the MONTEBURNS code. Finally, the work is investigated for 22 isotope fission products of nuclides, 22 actinides in the core, and tritium in the coolant. The activities are mainly distributed within the range of $10^{15}{\sim}10^{17}$ Bq in the equilibrium core of HTR-10. The results appear to be highly probable, and they would be informative when the spent fuel of HTGRs is taken into account. The tritium inventory in the primary coolant is also taken into account without a helium purification system. This article can lay a foundation for future work on analyses of source terms as a platform for safety assessment in HTGRs.