• Journal of Radiation Protection and Research
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• v.36 no.1
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• pp.35-43
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• 2011

The PHENIX Experiment is the largest of the four experiments that have taken data at the Relativistic Heavy Ion Collider. PHENIX, the Pioneering High Energy Nuclear Interaction eXperiment, is designed specifically to measure direct probes of the collisions such as electrons, muons, and photons. The primary goal of PHENIX is to discover and study a new state of matter called the Quark-Gluon Plasma. Among many particles, muons coming from W-boson decay gives us key information to analyze the spin of proton. Resistive plate chambers are proposed as a suitable solution as a muon trigger because of their fast response and good time resolution, flexibility in signal readout, robustness and the relatively low cost of production. The RPC detectors for upgrade were assembled and their performances were evaluated. The procedure to make the detectors better was optimized and described in detail in this thesis. The code based on ROOT was written and by using this the performance of the detectors made was evaluated, and all of the modules for north muon arm met the criteria and installation at PHENIX completed in November 2009. As RPC detectors that we made showed fast response, capacity of covering wide area with a resonable price and good spatial resolution, this will give the opportunity for applications, such as diagnosis and customs inspection system.

• Nuclear industry
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• no.5_6 s.7
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• pp.34-42
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• 1982

• Nuclear Engineering and Technology
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• v.17 no.1
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• pp.16-24
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• 1985

A computer model is developed capable of simulating the transient behavior of a pool-type liquid metal-cooled fast breeder reactor (LMFBR). The model, SIMFARP, is a fast running computer code which may be used to simulate the loss of power to any pump(s), a complete loss-of-forced cooling, and the natural circulation behavior. Eight governing equations are derived and a Runge-Kutta algorithm is applied to integrate the eight differential equations. The developed computer program is applied to two cases; loss of electric power to any pump(s), and loss of all external electric supply power without scram in Super-Phenix-I.

• Nuclear Engineering and Technology
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• v.18 no.2
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• pp.78-84
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• 1986

An equation of material number density sensitivity coefficient is derived using first-order perturbation theory. The beginning of cycle of Super-Phenix I is taken as the reference system for this study. Effective multiplication factor of the reference system is defined as system response function and fuel enrichment and fuel effective density are chosen for the variation of reference input data since they are described by material number density which is a component of Boltzmann operator. The nuclear computational code system (KAERI-26 group cross section library/1DX/2DB/PERT-V) is employed for this calculation. Sensitivity coefficient of fuel enrichment on effective multiplication factor is 4.576 and sensitivity coefficient of effective fuel density on effective multiplication factor is 0.0756. This work shows that sensitivity methodology is lesser timeconsuming and gives more informations on important design parameters in comparison with the direct iterative calulation through large computer codes.

• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.267-275
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• 1999

This paper deals with the industrial problem of thermal striping damage on the French prototype fast breeder reactor, Phenix and it was studied in coordination with the research program of IAEA. The thermomechanical and fracture mechanics evaluation procedure of thermal striping damage on the tee-junction of the secondary piping using Green's function method and standard FEM is presented. The thermohydraulic(T/H) loading condition used in the present analysis is the random type thermal loads computed by T/H analysis on the turbulent mixing of the two flows with different temperatures. The thermomechanical fatigue damage was evaluated according to ASME code section 111 subsection NH. The results of the fatigue analysis showed that fatigue failure would occur at the welded joint within 90,000 hours of operation. The assessment for the fracture behavior of the welded joint showed that the crack would be initiated at an early stage in the operation. It took 42,698.9 hours for the crack to propagate up to 5 mm along the thickness direction. After then, however, the instability analysis, using tearing modulus, showed that the crack would be arrested, which was in agreement with the actual observation of the crack. An efficient analysis procedure using Green's function approach for the crack propagation problem under random type load was proposed in this study. The analysis results showed good agreement with those of the practical observations.

• Journal of Mechanical Science and Technology
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• v.19 no.5
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• pp.1206-1215
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• 2005

The thermal stratification phenomena, frequently occurring in the component of nuclear power plant system such as pressurizer surge line, steam generator inlet nozzle, safety injection system (SIS), and chemical and volume control system (CVCS), can cause through-wall cracks, thermal fatigue, unexpected piping displacement and dislocation, and pipe support damage. The phenomenon is one of the unaccounted load in the design stage. However, the load have been found to be serious as nuclear power plant operation experience accumulates. In particular, the thermal stratification by the turbulent penetration or valve leak in the SIS and SCS pipe line can lead these safety systems to failure by the thermal fatigue. Therefore in this study an 1/10 scaledowned experimental rig had been designed and installed. And a series of experimental works had been executed to measure the temperature distribution (thermal stratification) in these systems by the turbulent penetration, valve leak, and heat transfer through valve. The results provide very valuable informations such as turbulent penetration depth, the possibility of thermal stratification by the heat transfer through valve, etc. Also the results are expected to be useful to understand the thermal stratification in these systems, establish the thermal strati­fication criteria and validate the calculation results by CFD Codes such as Fluent, Phenix, CFX.

• Nuclear Engineering and Technology
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• v.16 no.2
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• pp.70-79
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• 1984

With the help of the nuclear computational system for a large LMFBR (KAERI-26 group cross section library/1DX/2DB), the reactivity coefficients for the diagrid expansion and the pad expansion at the beginning of cycle of the equilibrium core of SUPER-PHENIX I are calculated and reviewed. the core is described using R-Z geometry model, and a two-dimensional multigroup diffusion theory is used. For reference cases, reactivity calculations for radial and axial uniform expansion are performed, and also calculated are reactivity variations due to changes in material density and core volume. The reactivity coefficient for the diagrid expansion is calculated to be -0.553pcm/mil. The temperature coefficient corresponding to the above value is -1.0766pcm/$^{\circ}C$ and is well in accord with the French datum of -1.09pcm/$^{\circ}C$ within 1.2% difference. With the use of 4he calculational method for the diagrid expansion effect, reactivity calculations for the pad expansion bringing about nonuniform expansion are performed, which show that the calculational method is very useful in the analysis of the pad expansion effect. The reactivity coefficients for the pad expansion are calculated to be -0.2743 pcm/mil and -0.2786pcm1mi1 for the averaged expansion model and for the integrated pancake model, respectively. Under the assumption of the free expanding core the temperature reactivity coefficients for each model are obtained to be -0.5766pcm/$^{\circ}C$ and -0.5858pcm/$^{\circ}C$, both of which agree with the French datum of -0.574pcm/$^{\circ}C$ within 2% difference.