• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.103-104
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.145-146
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.423-428
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• 1996

The feasibility study on burning Am-241, 243 and Cm-244 nuclides in a conventional PWR (Pressurized Water Reactor) was carried out by using the TRIFON code that was developed by the Institute of Theoretical and Experimental Physics in Russia in 1992. TRIFON code uses updated ABBN Russian nuclear cross section library. The reference reactor is the Korea nuclear power plant unit 8 (YGN 2). The burning effect of Am-241, 243 and Cm-244 nuclides was studied with UO$_2$(3.5 w/o)fuel assembly and MOX (4.44 w/o) fuel assembly. The loaded mass ratio of Am-241, 243 and Cm-244 nuclides was obtained from the mass ratio of Am-241, 243 and Cm-244 nuclides in 10 year cooling spent fuel with average discharge burnup of 33 GWD/MTU. The effective transmutation rates of Am-241, 243 and Cm-244 nuclides in UO$_2$ fuel assembly were found to be higher than those in MOX fuel assembly. The result from TRIFON code was compared to that from CASMO-3/NEM-3D code system. For more reliable calculation of transmutation for MA(Minor Actinides) more sophisticated decay chain scheme of MA should be investigated and nuclear cross section library of MA should be considerably improved.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.25-34
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• 2018

The CINDER code has about 60 years of development history, and is thus one of the world's best transmutation computing codes to date. Unfortunately, it is complex and cumbersome to use. Preparing auxiliary input files for activation computation from MCNPX output and executing them using Perl script (activation script) is the first difficulty, and separation of gamma source computing script (gamma script), which analyzes the spectra files produced by CINDER code and creates source definition format for MCNPX code, is the second difficulty. In addition, for highly nonlinear problems, multiple human interventions may increase the possibility of errors. Postprocessing such as making plots with large text outputs is also time consuming. One way to improve these limitations is to make a graphical user interface wrapper that includes all codes, such as MCNPX and CINDER, and all scripts with a visual C#.NET tool. The graphical user interface merges all the codes and provides easy postprocessing of graphics data and Microsoft office tools, such as Excel sheets, which make the CINDER code easy to use. This study describes the VCINDER code (with visual C#.NET) and gives a typical application example.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.360-367
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• 2016

A windowless spallation target can provide a neutron source and maintain neutron chain reaction for a subcritical reactor, and is a key component of China's nuclear waste transmutation of coupling accelerator and subcritical reactor. The main issue of the windowless target design is to form a stable and controllable free surface that can ensure that energy spectrum distribution is acquired for the neutron physical design when the high energy proton beam beats the lead-bismuth eutectic in the spallation target area. In this study, morphology and flow characteristics of the free surface of the windowless target were analyzed through the volume of fluid model using computational fluid dynamics simulation, and the results show that the outlet cross section size of the target is the key to form a stable and controllable free surface, as well as the outlet with an arc transition. The optimization parameter of the target design, in which the radius of outlet cross section is $60{\pm}1mm$, is verified to form a stable and controllable free surface and to reduce the formation of air bubbles. This work can function as a reference for carrying out engineering design of windowless target and for verification experiments.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.91-96
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• 1995

In this paper, research efforts to develop computer code system for analysis of the transmutation reactors at KAIST are described Especially the computer code HANCELL for assembly calculation of fast reactors is mainly described. Features and function of the code are identified md current status of the code development is provided

• Nuclear Engineering and Technology
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• v.44 no.2
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• pp.199-206
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• 2012

Using fast neutron Generation IV reactors, recycling of americium and curium may become feasible. The detrimental impact of americium on safety parameters has recently been quantified in terms of a power penalty for surviving a given set of transients in sodium fast reactors. In the present paper, a review of the physical reasons for the adverse effect of americium is provided, and different Gen-IV technologies are assessed with respect to their capability of hosting americium in the fuel.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.111-112
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.109-110
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• 2004

• Nuclear Engineering and Technology
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• v.38 no.7
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• pp.675-680
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• 2006

The neutron transmutation doping of silicon (NTD), as a method to produce a high quality semiconductor, utilizes the transmutation of a silicon element into phosphorus by neutron absorption in a silicon single crystal. In this paper, we present the design of a neutron screen for a 6' Si ingot irradiation in the NTD2 hole of HANARO. The goal of the design is to achieve an even flat axial distribution of the resistivity, or $Si^{30}(n,{\gamma})Si^{31}$ reaction rate, in the irradiated Si ingot. We used the MCNP4C code to simulate the neutron screen and to calculate the reaction rate distribution in the Si ingot. The fluctuations in the axial distribution were estimated to be within ${\pm}2.0%$ from the average for the final neutron screen design; thus, they satisfy the customers' requirement for uniform irradiation. On the other hand, we determined the optimal insertion depths of the Si ingots by varying the critical control rod position, which greatly affects the axial flux distribution.