• Journal of Energy Engineering
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• v.12 no.4
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• pp.281-288
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• 2003

The MATRA-LMR code has been developed based on a subchannel analysis method for LMR (Liquid Metal Reactor) core subassembly thermal hydraulic design and analysis. The code was improved to allow a seven assembly calculation and can account for inter-assembly heat transfer based on a lumped parameter model. This paper describes the main modifications and improvements of the code and shows reference calculation results which compared single assembly calculation with seven assembly calculation cased for driver and blanket subassemblies of the KALIMER 150 MWe breakeven conceptual design core. KAL- IMER is a pool-type sodium cooled reactor with a thermal output of 392.0 MWth, which have inherently safe, environmentally friendly, proliferation-resistant and economically viable reactor concepts.

• Journal of the Korean Society for Precision Engineering
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• v.18 no.4
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• pp.85-96
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• 2001

In this paper, part features are classified and then its coding system is constructed by the considered characteristics of features in assemble process. Analyzing the characteristics of features, code values about part features are determined. Assembly process is divided into five functions such as transporting, handing, approaching, alignment and joining, and then the detail parameters of each functions such as determined. Code values about assembly process are determined according to detail parameters. The detail parameters are kinds of available working method and assembly tools when each assembly function is going on. By the coding system, available assembly process can be grasped and perceived for the part that it is difficult to assemble.

• Proceedings of the Korean Operations and Management Science Society Conference
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• 2000.04a
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• pp.28-31
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• 2000

In this paper, part features are classified and then coding system is constructed by the considering characteristics of features in assembly process. Analyzing the characteristics of features, code values about part features are determined. Assembly process is divided into five functions such as transporting, handling, approaching, alignment, joining and then the detail parameters of each function are determined. Code values about assembly process are determined according to detail parameters. The detail parameters are the kinds of available working method and assembly tools, when each assembly function is going on. By the coding system, available assembly process can be grasped and perceived the part that Is difficult to assemble.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.85-90
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• 1995

A characteristic transport theory code CRX is developed and tested for cell and assembly calculations. Since the characteristic method treats explicitly (analytically) the streaming portion of the transport equation, CRX treats strong absorbers well and has no practical limitations placed on the geometry of the problem. To test the code, it was applied to three benchmark problems which consist of complex meshes and compared with other codes.

• ETRI Journal
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• v.25 no.6
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• pp.464-474
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• 2003

Component-based development leverages software reusability and reduces development costs. Enterprise JavaBeans (EJB) is a component model developed to reduce the complexity of software development and to facilitate reuse of components. However, EJB does not support component assembly by a plug-and-play technique due to the hard-wired composition at the code level. To cope with this problem, an architecture for EJB component assembly is defined at the abstract level and the inconsistency between the system architecture and its implementation must be eliminated at the implementation level. We propose a component-based application development tool named the COBALT assembler that supports the design and implementation of EJB component assembly by a plug-and-play technique based on the architecture style. The system architecture is first defined by the Architecture Description Language (ADL). The wrapper code and glue code are then generated for the assembly. After the consistency between the architecture and its implementation is checked, the assembled EJB components are deployed in an application server as a new composite component. We use the COBALT assembler for a shopping mall system and demonstrate that it can promote component reuse and leverage the system maintainability.

• Journal of the Institute of Electronics Engineers of Korea SP
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• v.46 no.3
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• pp.119-124
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• 2009

In this paper, AMR-WB+ audio coder is implemented, in real-time, using Intel 400MHz Xscale PXA250 with 32bit RISC processor ARM9E-J(R)core. The assembly code for ARM9E-J(R)core is developed through the serial process of C code optimization, cross compile, assembly code manual optimization and adjusting the optimized code to Embedded Visual C++ platform. C code is trimmed on Visual C++ platform. Cross compile and assembly code manual optimization are performed on CodeWarrior with ARM compiler. Through these stages the code for both ARM EVM board and PDA is implemented. The average complexities of the code are 160.75MHz on encoder and 33.05MHz on decoder. In case of static link library(SLL), the required memories are 65.21Kbyte, 32.01Kbyte and 279.81Kbyte on encoder, decoder and common sources, respectively. The implemented coder is evaluated using 16 test vectors given by 3GPP to verify the bit-exactness of the coder.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.1871-1885
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• 2019

The deterministic MOC code STREAM of the Computational Reactor Physics and Experiment (CORE) laboratory of Ulsan National Institute of Science and Technology (UNIST), was initially designed for the calculation of pressurized water reactor two- and three-dimensional assemblies and cores. Since fast reactors play an important role in the generation-IV concept, it was decided that the code should be upgraded for the analysis of fast neutron spectrum reactors. This paper presents a coupled code - TULIP/STREAM, developed for the fast reactor assembly and core calculations. The TULIP code produces self-shielded multi-group cross-sections using a one-dimensional cylindrical model. The generated cross-section library is used in the STREAM code which solves eigenvalue problems for a two-dimensional assembly and a multi-assembly whole reactor core. Multiplication factors and steady-state power distributions were compared with the reference solutions obtained by the continuous energy Monte-Carlo code MCS. With the developed code, a sensitivity study of the number of energy groups, the order of anisotropic P_N scattering, and the multi-group cross-section generation model was performed on the k_eff and power distribution. The 2D core simulation calculations show that the TULIP/STREAM code gives a k_eff error smaller than 200 pcm and the root mean square errors of the pin-wise power distributions within 2%.

• Journal of the Korean Institute of Telematics and Electronics
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• v.16 no.3
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• pp.36-48
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• 1979

HP 2100 S computer의 disc operating system 을 사용한 본cross-assembler는 alpha-minicomputer 의 assembly language program을 source 입력으로 하여 이 alpha-computer 에 의한 실행 (execution)을 목적으로 16진수 code의 등가 object program 을 출력토록 설계되어 있다.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2803-2815
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• 2021

This paper addresses the uncertainty quantification and sensitivity analysis of a depleted light-water fuel assembly of the Turkey Point-3 benchmark. The uncertainty of the fuel assembly decay heat and isotopic densities is quantified with respect to three different groups of diverse parameters: nuclear data, assembly design, and reactor core operation. The uncertainty propagation is conducted using a two-step analysis code system comprising the lattice code STREAM, nodal code RAST-K, and spent nuclear fuel module SNF through the random sampling of microscopic cross-sections, fuel rod sizes, number densities, reactor core total power, and temperature distributions. Overall, the statistical analysis of the calculated samples demonstrates that the decay heat uncertainty decreases with the cooling time. The nuclear data and assembly design parameters are proven to be the largest contributors to the decay heat uncertainty, whereas the reactor core power and inlet coolant temperature have a minor effect. The majority of the decay heat uncertainties are delivered by a small number of isotopes such as ²⁴¹Am, ¹³⁷Ba, ²⁴⁴Cm, ²³⁸Pu, and ⁹⁰Y.