• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.18 no.2
• /
• pp.37-42
• /
• 2022

An influence analysis on multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout is performed to compare the plant responses according to the number of ruptured u-tubes under the assumption of a total of 10 ruptured u-tubes. In all calculation cases, the transient behaviour of major thermal-hydraulic parameters, such as the discharge flow rate through the ruptured u-tubes, reactor header pressure, and void fraction in the fuel channels is found to be overall similar to that of the base case having a single SG with 10 u-tubes ruptured. Additionally, as the conditions of low-flow coolant with high void fraction in the broken loop continued, causing the degradation of decay heat removal, the peak cladding temperature (PCT) would be expected to exceed the limit criteria for ensuring nuclear fuel integrity. However, despite the same total number of ruptured u-tubes, because of the different connection configuration between the SG and pressurizer, a difference is foud in time between the pressurizer low-level signal and reactor header low-pressure signal, affecting the time to trip the reactor and to reach the PCT limit. The present study is expected to provide the technical basis for the accident management strategy for mSGTR transient conditions of CANDU-6 plants.

• Proceedings of the KSME Conference
• /
• 2001.11b
• /
• pp.499-504
• /
• 2001

A validation of a 3D CFD model for predicting local subcooling of moderator in the vicinity of calandria tubes in a CANDU reactor is performed. The small scale moderator experiments performed at Sheridan Park Experimental Laboratory(SPEL) in Ontario, Canada[1] is used for the validation. Also a comparison is made between previous CFD analyses based on 2DMOTH and PHOENICS, and the current model analysis for the same SPEL experiment. For the current model, a set of grid structures for the same geometry as the experimental test section is generated and the momentum, heat and continuity equations are solved by CFX-4.3, a CFD code developed by AEA technology. The matrix of calandria tubes is simplified by the porous media approach. The standard $k-\varepsilon$ turbulence model associated with logarithmic wall treatment and SIMPLEC algorithm on the body fitted grid are used and buoyancy effects are accounted for by the Boussinesq approximation. For the test conditions simulated in this study, the flow pattern identified is a buoyancy-dominated flow, which is generated by the interaction between the dominant buoyancy force by heating and inertial momentum forces by the inlet jets. As a result, the current CFD moderator analysis model predicts the moderator temperature reasonably, and the maximum error against the experimental data is kept at less than $2.0^{\circ}C$ over the whole domain. The simulated velocity field matches with the visualization of SPEL experiments quite well.

• 제어로봇시스템학회:학술대회논문집
• /
• 2003.10a
• /
• pp.1220-1224
• /
• 2003

In this paper, we measured the radiation dose of a fuel handling machine of the CANDU type Wolsong nuclear reactor directly during operation, in spite of the high radiation level. In this paper we will describe the sensor development, measurement techniques, and results of our study. For this study, we used specially developed semiconductor sensors and matching dosimetry techniques for the mixed radiation field. MOSFET dosimeters with a thin oxide, that are tuned to a high dose, were used to measure the ionizing radiation dose. Silicon diode dosimeters with an optimum area to thickness ratio were used for the radiation damage measurements. The sensors are able to distinguish neutrons from gamma/X-rays. To measure the radiation dose, electronic sensor modules were installed on two locations of the fuel handling machine. The measurements were performed throughout one reactor maintenance cycle. The resultant annual cumulative dose of gamma/X-rays on the two spots of the fuel handling machine were 18.47 Mrad and 76.50 Mrad, and those of the neutrons were 17.51 krad and 60.67 krad. The measured radiation level is high enough to degrade certain cable insulation materials that may result in electrical insulation failure.

• Nuclear Engineering and Technology
• /
• v.48 no.3
• /
• pp.642-649
• /
• 2016

Important safety parameters such as the fuel temperature coefficient (FTC) and the power coefficient of reactivity (PCR) of the CANada Deuterium Uranium (CANDU-6) reactor have been evaluated using the Monte Carlo method. For accurate analysis of the parameters, the Doppler broadening rejection correction scheme was implemented in the MCNPX code to account for the thermal motion of the heavy uranium-238 nucleus in the neutron-U scattering reactions. In this work, a standard fuel lattice has been modeled and the fuel is depleted using MCNPX. The FTC value is evaluated for several burnup points including the mid-burnup representing a near-equilibrium core. The Doppler effect has been evaluated using several cross-section libraries such as ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1.1, and JENDL-4.0. The PCR value is also evaluated at mid-burnup conditions to characterize the safety features of an equilibrium CANDU-6 reactor. To improve the reliability of the Monte Carlo calculations, we considered a huge number of neutron histories in this work and the standard deviation of the k-infinity values is only 0.5-1 pcm.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1995.10a
• /
• pp.429-436
• /
• 1995

This paper describes a methodology for determination of representative CANDU feeder geometry and the pressure drops between inlet/outlet header and fuel channel in the primary loop. A code, MEDOC, was developed based on this methodology and helps perform a calculation of equivalent feeder geometry for a selected channel group on the basis of feeder geometry data (fluid volume, mass flow rate, loss factor) and given property data pressure, quality, density) at inlet/outlet header. The equivalent feeder geometry calculated based on this methodology will be useful fur the transient thermohydraulic analysis of the primary heat transport system for the CANDU heavy water-cooled pressure tube reactor.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.05a
• /
• pp.77-82
• /
• 1997

The CANDU-6 transition core refuelled from 37-element fuel to CANFLEX-NU fuel has been evaluated by an 100full power day time-dependent fuel-management simulation to find the core compatibility with the CANFLEX fuel loading. The simulation calculations for the transition core were carried out with the RFSP code, provided by the cell averaged fuel properties obtained from the POWDERPUFS-V code. The simulation results were compared with those of the current 37-element fuel loading only. The results show that the CANFLEX-NU fuel bundles will be compatible with the CANDU-6 reactor because the core physics characteristics of CANFLEX-NU fuel are very similar to those of the 37-element fuel bundle.

• Nuclear Engineering and Technology
• /
• v.44 no.7
• /
• pp.799-806
• /
• 2012

In recent years, a new algorithm has been introduced to perform the regional overpower protection (ROP) detector layout optimization for $CANDU^{(R)}$ reactors. This algorithm is called DETPLASA. This algorithm has been shown to successfully come up with a detector layout which meets the target trip set point (TSP) value. Knowing that these ROP detectors are placed in a number of safety channels, one expects that there is an optimal placement of the candidate detectors into these channels. The objective of the present paper is to show that a slight improvement to the TSP value can be realized by optimizing the channelization of these ROP detectors. Depending on the size of the ROP system, based on numerical experiments performed in this study, the range of additional TSP improvement is from 0.16%FP (full power) to 0.56%FP.

• Journal of the Computational Structural Engineering Institute of Korea
• /
• v.22 no.5
• /
• pp.381-390
• /
• 2009

In this paper, a complementary analysis for the structural safety evaluation of the spent nuclear fuel disposal canister developed for the Canadian Deuterium and Uranium(CANDU) reactor for about 10,000 years long term deposition at a 500m deep granitic bedrock repository has been performed. However this developed structural model of the spent nuclear fuel disposal canister which has 33 spent nuclear fuel baskets and whose diameter is 122cm is too heavy to handle without any structural safety problem. Hence a lighter structural model of the spent nuclear fuel disposal canister which is easy to handle has been required to develop very much. There are two methods to reduce the weight of the CANDU canister model. The one is to alleviate severe design conditions such as external loads and safety factor. The other is to optimize the cross section shape of the canister by reducing the spent nuclear fuel basket number. Hence, in this paper a complementary analysis to alleviate such severe design conditions is carried out and simultaneously structural analyses to optimize the cross section shape of the canister by reducing the spent nuclear fuel basket number below 33 are carried out by varying the external load and the canister diameter for the reduction of the canister weight. The complementary analysis results show that the diameter of canister can be shortened below 122cm to reduce the weight of the spent nuclear fuel disposal canister.

• Journal of the Korean Society of Visualization
• /
• v.12 no.1
• /
• pp.18-24
• /
• 2014

The Korea Atomic Energy Research Institute (KAERI) has a scaled-down moderator test program to simulate the CANDU6 moderator circulation phenomena during steady state operation and accident conditions. In the present work a preliminary experiment using a 1/8 scaled-down moderator tank has been performed to identify the potential problems of the flow visualization and measurement in the scaled-down moderator test facility. With a transparent moderator tank model, a velocity field is measured with a Particle Image Velocimetry (PIV) technique under an isothermal state. The flow patterns from the inlet nozzles to the top region of the tank are investigated using PIV for a 1/8 scale moderator tank.

• Journal of Mechanical Science and Technology
• /
• v.20 no.7
• /
• pp.1030-1042
• /
• 2006

An experimental study has been performed to investigate the off-take phenomena at the header-feeder systems (horizontal header pipe with multiple feeder branch pipes) in a CANDU (CANadian Deuterium Uranium) reactor with the branch orientation varies ${\pm}36^{\circ}\;or\;{\pm}72^{\circ}$. In order to evaluate the applicability of the conventional correlations used in the safety analysis code, RELAP5-Mod3, the test facility is designed with the 1/2 scale of the. CANDU 6. It was found that the data set for the top, bottom and side branches are in a good agreement with the correlations used. However, for the specific angled branches, ${\pm}36^{\circ}\;and\;{\pm}72^{\circ}$, the onsets of off-take data and quality data showed large deviation with the conventional model inside RELAP5-MOD3. Furthermore, based on the uncertainty analysis, the conventional 2.5 power law needs to be modified. The present experimental data set can be useful for the construction of the general correlation considering the arbitrary branch orientation.