• 제어로봇시스템학회:학술대회논문집
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• 2005.06a
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• pp.1445-1449
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• 2005

The ram assembly is important equipment in fueling machine of PHWR(Pressurized Heavy Water Reactor) plant where fuel replacement is possible while the plant is in service. Troubles in the ram assembly can cause lots of difficulties in power plant operation. The ram assembly is typically composed of the B-ram, the L-Ram and the C-Ram. The B-ram is focused in this paper because it plays the most important role in the ram assembly. Among the ram fault phenomena, ram stuck phenomena are the most frequent cases in the B-ram, which has a ball screw mechanism driven by a hydraulic motor. Ram stuck phenomena are due to ball wear and damage in ball nut that increase in proportion to the number of fuel replacement. It is required to predict ram stuck phenomena before they occur. In this paper, a method is proposed for predicting ram stuck phenomena using a discrete wavelet transform. The discrete wavelet transform provides information on both the time and frequency characteristics of the input signals. The proposed method uses the frequency bandwidths of coefficients of discrete wavelet decompositions and detail coefficients of discrete wavelet transform to predict ram stuck phenomena. The signal used in this paper is a torque-related signal such as a hydraulic service outlet pressure signal in a hydraulic driving system or a current signal in a DC motor driving system. Finally, the validity of the proposed method is shown via experiment using ball nut characteristic test equipment that simulates ram stuck phenomena due to increased ball friction in ball nut.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.9
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• pp.1273-1279
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• 2010

Pressure tubes at the CANada Deuterium Uranium (CANDU) nuclear power plants are periodically inspected in accordance with the CSA N285.4 code. If flaws that do not satisfy the criteria given in CSA N285.4 are detected, the code permits a fitness-for-service assessment to determine the acceptability of the flawed pressure tubes. In this paper, the Wolsong In-service Evaluation System (WIES) is introduced; this system has been developed for the assessment of the flawed pressure tubes and is based on CSA N285.8. Since the system evaluates the integrity of flawed pressure tubes exactly and promptly during an in-service inspection, it will help in operating the Wolsong nuclear power plants without prolonging the outage period.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.24 no.4
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• pp.449-455
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• 2011

Recently, international collaborative research which was organized at Bhabha Atomic Research Centre in India, was conducted to develop for pressure capacity and nonlinear behavior of PHWR 1/4 scale nuclear containment building between experimental test and numerical code. In this paper, a nonlinear finite element analysis was carried out in order to predict ultimate pressure capacity and nonlinear behavior of the 1/4 scale containment building. The 1/4 scale containment building is consisted of basemat, cylinder wall, dome and 4-buttress. For the finite element analysis, commercial program ABAQUS was used. Finite element models including concrete, rebar and tendon have been developed for assessment of ultimate pressure capacity and failure mode for nuclear containment building. From the analysis results, first crack of the concrete, the yielding of the rebar and ultimate capacity pressure occurred at $1.6P_d$(design pressure), $3.36P_d$ and $4.0P_d$, respectively.

• Journal of the Korean Society for Nondestructive Testing
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• v.24 no.4
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• pp.307-314
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• 2004

A circumferential guided wave method was developed to detect the axial crack on the bent feeder pipe. Dispersion curves of circumferential guided waves were calculated as a function of curvature of the pipe. In the case of thin plate, i.e. infinite curvature, as the frequency increases, the $S_0$ and $A_0$ mode coincide and eventually become Rayleigh wave mode. In the case of pipe, however, as the curvature increases, the lowest modes do not coincide even in the high frequencies. Based on the analysis, a rocking technique using angle beam transducer was applied to detect an axial defect in the bent region of PHWR feeder pipe. Based on the analysis of experimenal data for artificial notches, the vibration modes of each signal were identified. It was found that the notches with the depth of )0% of wall thickness can be detected with the method.

• Applied Chemistry for Engineering
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• v.7 no.6
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• pp.1164-1173
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• 1996

Treating methods and characteristics of waste from a nuclear fuel powder conversion plant were studied. To recovery or treat a trace uranium in liquid waste, the ammonium uranyl carbonate(AUC) filtrate must be heated for $CO_2$ expelling, essentially. Uranium content of final treated waste solution from fuel powder processes for a heavy water reactor(HWR) could be lowered to 1 ppm by the lime treatment after the ammonium di-uranate(ADU) precipitation by simple heating. Otherwise, in case of the waste from fuel powder processes for a pressurized light water reactor(PWR), it is result in 0.8 ppm as a form of uranium peroxide such as $UO_4{\cdot}2NH_4F$ compounds. Optimum condition was found at $101^{\circ}C$ by the simple heating method in case of HWR powder process waste. And in case of PWR powder process waste, optimum condition could be obtained by precipitating with adding hydrogen peroxide and adjusting at pH 9.5 with ammonia gas at $60^{\circ}C$ after heating the waste In order to expelling $CO_2$. As the characteristics of recovered uranium compounds, median particle size of ADU was increased with pH increasing in case of HWP waste. Also, in case of uranium proxide compound recovered from PWR waste, the property of $U_3O_8$ power obtained after thermal treatment in air atmosphere was similar to that of the powder prepared from AUC conversion plant.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.375-387
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• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.

• Journal of the Korean Society for Nondestructive Testing
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• v.23 no.4
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• pp.334-341
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• 2003

When the pressure tubes(f are in contact with the calandria tube(CT) in the pressurized heavy water reactor(PHWR), the temperature difference between inner and outer wall of W results in a thermal diffusion of hydrogen (deuterium) and hydride blisters are formed on the outer surface of PT. Because the hydride blisters and zirconium matrix are acoustically continuous, it is not easy to distinguish the blisters from the matrix with conventional ultrasonic method. An ultrasonic velocity ratio method was developed to detect small hydride blisters on the zirconium pressure tube. Hydride blisters were grown in the PT specimen using a steady state thermal diffusion device. The flight times of longitudinal echo and reflected shear echo from the outer surface were measured accurately. The velocity ratio of the longitudinal wave to the shear wave was calculated and displayed using contour plot. Compared to the conventional flight time method of longitudinal wave, the velocity ratio method shows superior sensitivity to detect smaller blisters as well as better images for the blister shapes. Detectable limit of the outer shape of the hydride blisters was conservatively estimated as $500{\mu}m$, with the same specifications of ultrasonic transducer used in the actual PHWR pressure tube inspection.