[en] In order to assess the uncertainties of radioactivity released to the environment resulting from analysis of combined particulate samples collected from multiple gaseous effluent release points, a series of generic methodology was developed. As a case study, the range of total errors entailed by measuring combined particulate samples at an operating Korean Standard Nuclear Power plant was predicted and both merits and limitations of each adjustment methodology were analyzed. It turns out that the errors are to be maximized when the release point with highest ratio of effluent release rate to sampling flow rate is dominant. In principle, the activity measurement should be made on a composite of all filters collected from each release point, however the measurement of combined samples from all release points can be considered as an alternative especially in the aspect of cost-effectiveness of effluent control. In order to verify the feasibility of measuring combined samples, a series of licensing conditions must be documented based upon the results of case-by-case safety analysis

[en] The classification of accidents for a specific design of the PGSFR(Prototype Gen-IV Sodium-cooled Fast Reactor) was completed. The design data used in the safety analysis was summarized as a document. Based on the design data, safety analysis of design basis events such as transient over power, loss of flow, loss of heat sink, PHTS pipe break, and station black out were implemented. Furthermore, safety analysis of design extension conditions such as ATWS and multi-failure event etc. was also conducted. The final report of the Level-1 PSA was prepared in a specific design of PGSFR, and PSA related with the seismic margin and the fire/flood was carried out. Evaluation of in-vessel source term was conducted in case of whole pins failures of PGSFR. In order to improve the accuracy of evaluation for mechanical behavior characteristic of reactor core, START code was developed by coupling with ISFRA code and SAS4A/SASSYS-1 code. The START code can evaluate the degree of reactor core damage through the elaborate reactor core and fuel model of the SAS4A/SASSYS-1 code. The behavior of the radioactive material eventually released into the environment was evaluated using the transport model of the fission products. Level-2 PSA analysis was also carried out by START code for the PGSFR. The severe accident analysis of PGSFR was carried out using the newly developed SAS4A code, and the molten fuel, the flow in the molten fuel, internal and external ejection phenomena, and core reactivity behaviors was evaluated. The analysis results showed that severe accident could be early terminated. In order to verify a severe accident model, the reactivity insertion results of TREAT M series was used. The comparative verification was conducted with the analysis results of SAS4A code with metal fuel model, and the results showed a good prediction of molten fuel behavior and reactivity

[en] The wire spacer has important roles to avoid collisions between adjacent rods, to mitigate a vortex induced vibration, and to enhance convective heat transfer by wire spacer induced secondary flow. Many experimental and numerical works has been conducted to understand the thermal-hydraulics of the wire-wrapped fuel bundles. There has been enormous growth in computing capability. Recently, a huge increase of computer power allows to three-dimensional simulation of thermal-hydraulics of wire-wrapped fuel bundles. In this study, the geometry optimization methodology with RANS based in-house CFD (Computational Fluid Dynamics) code has been successfully developed in air condition. In order to apply the developed methodology to fuel assembly, GGI (General Grid Interface) function is developed for in-house CFD code. Furthermore, three-dimensional flow fields calculated with in-house CFD code are compared with those calculated with general purpose commercial CFD solver, CFX. The geometry optimization methodology with RANS based in-house CFD code has been successfully developed in air condition. In order to apply the developed methodology to fuel assembly, GGI function is developed for in-house CFD code as same as CFX. Even though both analyses are conducted with same computational meshes, numerical error due to GGI function locally occurred in only CFX solver around rod surface and boundary region between inner fluid region and outer fluid region.

[en] A study was performed on the behavior of low head safety injection water in the upper plenum during Large Break Loss of Coolant Accident for Kori Unit 1 using TRAC-M/F77, which is multi-dimensional best estimate thermal-hydraulic computer code. The results showed that TRAC-M/F77 well predicted overall LBLOCA transient, especially low head safety injection water penetration into the lower plenum through the low power channel and establishment of circulating flow path between upper plenum and core, which is the important thermal-hydraulic behavior during LBLOCA for the Upper Plenum Injection plant

[en] For the application of dynamic neutron radiography to the two-phase flow research using HANARO, several experimental items to which the radiography technique is beneficial were identified through the review of the outputs from the related researches and the discussions with experts. Also, the investigation of the equipments including the beam port, camera and converter was made and a hardware and a software for image processing were equipped. It was confirmed that the calibration curve for the attenuation of neutron beam in fluid which is required for the two-phase flow experiment could be obtained by the computer code calculation. Based on the investigation results on the equipment and the results from the measurement of BNCT beam characteristics, a high speed camera and an image intensifier will be purchased. Then, the high speed dynamic neutron radiography facility for two-phase flow experiments will be fully equipped

[en] In resolving the safety issue of sump clogging due to debris generated by the type of high-energy line break known as a GSI-191 event, determination of the debris transport fraction is very important in the sizing of the sump screen area. In general method for evaluation of debris transport fraction, the mean fluid velocity distribution within the containment floor on recirculation transport mode during LOCA by CFD analysis is combined fundamental transport properties, such as tumbling velocities, of various types of debris by experimental research to identify the debris transport fraction. In the determination of the debris floor transport, it is also advised that the turbulent kinetic energy effect (TKE) as well as mean flow velocity on turbulent flooding flow is considered by previous experimental researches. However, the quantification results pertaining to the debris floor transport by TKE were not published in the literature. In the present study, the debris floor transport on flooding flow is evaluated with and without consideration of turbulence effect. To do this, experiments involving tumbling velocities measurements of the surrogate debris and supplementary CFD analyses are performed to verify the turbulence effect on debris transport. From these findings, the turbulence effect on the degree of debris floor tumbling augmentation was found to be represented by the algebraic sum of the mean horizontal velocity and the horizontal fluctuating velocity deduced from the turbulent kinetic energy

[en] The fuel assembly of the SFR system consists of long and thin wire-wrapped fuel bundles and a hexagonal duct, in which wire-wrapped fuel bundles in the hexagonal duct has triangular array. The main purpose of a wire spacer is to avoid collisions between adjacent rods. Furthermore, a vortex induced vibration can be mitigated by wire spacers. The wire spacer can enhances a convective heat transfer due to the secondary flow by helically wrapped wires. In this study, complicated and separated flow phenomena in the 7-pin fuel assembly without wire spacer and with wire spacer were captured by a RANS (Reynolds-Averaged Navier-Stokes) flow simulation with the SST (Shear Stress Transport) turbulence model, and by the vortex structure identification technique based on the critical point theory. The wire effect on three-dimensional flow field and heat transfer characteristics in a helically wrapped 7-pin fuel assembly mock-up of the SFR have been investigated through a numerical analysis using the commercial CFD code, CFX. Complicated and separated flow phenomena in the 7-pin fuel assembly without wire spacer and with wire spacer were captured by the RANS flow simulation with the SST turbulence model. It is concluded that the wire spacers locally induce a tangential flow by up to about 13 % of the axial velocity. The tangential flow in the corner and edge sub-channels is much stronger than that in the interior sub-channels. The flow with a high tangential velocity is periodically rotating in a period of wire lead pitch. The cross flow due to the wire spacer can achieve to enhance heat transfer characteristics up to about 50 %

[en] It is evident that the flow blockage is basically a local phenomenon, and the main issue to investigate is the thermal-hydraulic behavior of the region downstream from the obstacle because it determines the clad temperature peak. For this reason, a local detailed CFD analysis has been carried out in order to assess the impacts of a flow blockage. The flow blockage events are classified into two types, internal and external blockage, depending on their locations. The objective of this paper is to investigate the influence caused by a flow blockage. A CFD analysis using fully resolved RANS simulations has been carried on the fluid flow and heat transfer in the case of a flow blockage for fuel assemblies in a PGSFR. A fuel assembly with 91 pins instead of all 217 pins was considered for this study. Two main effects can be distinguished in a flow blockage: a locally lower mass flow rate in the wake/recirculation region downstream of the blockage, and the peak temperature behind the blockage. Both of them are closely related. The recirculation region exists within a short distance downstream from the blockage, and it has an effect on the cladding integrity. The maximum cladding temperature is about 1000 .deg. C and is located in the central pins of the blockage region. It could lead to a rupture of the cladding. From these analysis results, the axial blockage size may have a significant impact on the clad integrity

[en] Critical heat flux experiment was carried out using freon R-134a in a vertical round tube and the fluid-to-fluid modeling techniques are applied. The experimental range covers all the application ranges of critical heat flux correlations developed for both PWR and PHWR. The critical heat flux appears to increase linearly with inlet subcooling. For the constant inlet subcooling, increases in mass flow rate cause an increase in critical heat flux. The freon data are scaled to the water critical heat flux using both Katto and Ahmad's methods. Both methods under-predicted in experimental range when compared with the Groeneveld's 1995 look-up table. It was found that the Katto's method predicts better than the Ahmad's

[en] The MARS code uses the Henry-Fauske critical flow model as a user option. The Henry-Fauske critical flow model of the MARS code is assessed using the data of two-phase critical flow experiments with noncondensable gas performed at KAERI. The simulation results of steady-state two-phase critical flow experiments show that they agree with the measured critical flow rates within 6% root-mean-square error. However, the calculation results of pressure and temperature distributions along the test section show a little differences with the experimental data. The simulation results of transient two-phase critical flow experiments show that the calculation results do not predict well the variation of the critical flow rate and the pressure at the initial stage. Also the calculation results show that the discharged water is vaporized after exiting the test section region without the noncondensable gas injected, but it is vaporized while passing the test section with the noncondensable gas injected