[en] Safety management may be classified into three dimensions: risk management, accident management, and emergency management. This paper addresses the recent trends of safety management in nuclear industry, focussing on risk management and accident management

[en] A transport theory code CRX-Hex based on characteristic methods with a general geometric tracking routine is developed for the heterogeneous hexagonal geometry. With the general geometric tracking routine, the formulation of the characteristic method is not changed. To test the code, it was applied to two benchmark problems which consist of complex meshes and compared with other codes (HELIOS, TWOHEX)

[en] In the detail reaction-rate measurements in a critical assembly using the foil activation method, the measured activations of detector foils have inevitably errors caused by detector foil self-shielding effect. If neutron flux could be approximated to Westcott flux; i. e. well thermalized Mawellian distribution, these activations of detector foil could be corrected to represent the unperturbated flux at any detected position in the cell with using Westcott option and reaction-rate option of the lattice code, WIMS-AECL. provided much information about neutron spectrum of test lattice cell as well as the correction factors themselves. The results could be verified by another lattice calculations

[en] This paper concerns the cladding residual stresses in a reactor vessel induced during cooling from the manufacturing temperature down to room temperature. Finite element results show that very large stress gradients are present at the interface corner and such stress singularity might lead to local yielding or cladding-base metal debonding

[en] Flow induced vibration experiments were done to determine the vibration magnitude of tubes in the economizer tube lane region of a steam generator. The objective of this experiment is to demonstrate that the tubes are not susceptible to harmful levels of vibration at 100% of full power flow and to quantify the remaining design margin at 120% and 140% of full power flow

[en] A preliminary conceptual design of the DUPIC fuel fabrication plant with production capacity of 400 MTHE/yr is presented. Capital and operating costs are also included. The levelized unit fabrication cost (LUC) for a reference mode was estimated at 509/kgHE, and sensitivity of some variable parameters to this reference was analysed

[en] Individual internal monitoring program by in-vivo measurement technique at the Korea Atomic Energy Research Institute includes the capability for the assessment of uranium and americium lung burdens. This capability is an important part of the health and safety program. This article addresses the lung burden assessment portion of our in vivo measurement capabilities

[en] In this study the RETRAN-03 code was modified to analyze the thermal-hydraulic transients under three-dimensional ship motions for the application to the future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations are successfully simulated at various conditions

[en] In the present the mean temperature and the temperature fluctuation of non-isothermal parallel liquid sodium jets were measured and analyzed changing the temperature difference and mean velocity of the hot and cold sodium. The sampling frequency and sampling time were 420Hz and three seconds, respectively. The wave-form characteristics were discussed in regard to the peak-to-peak amplitudes and the periods provided by a wave analysis. And also the correlations of the temperature fluctuation in rms value and the peak amplitude are derived. The overall mean accuracy ratios of the correlations are 1.07 and 1.08 with a standard deviation of 0.17 and 0.15, respectively

[en] The nonlinear dynamic analyses were performed by newly developing an appropriate impact modelling for the evaluation of the CANFLEX fuel bundle structural integrity during the refuelling period. The initial load under the refuelling condition is considered as initial velocity at impact incident, and the impact of one bundle contacted another bundle for at short time is studied by performing several dynamic analysis method. The impact analysis shows to predict an appropriate velocity and acceleration profile according to load time history for two bundles impact