• Proceedings of the KSME Conference
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• 2004.11a
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• pp.383-387
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• 2004

The Fuel Test Loop (FTL) consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). The test condition in IPS such as pressure, temperature and quality of the main cooling water, can be controlled by the OPS. The FTL has been developed to be able to irradiate three pins to the core irradiation hole (IR1 hole) by considering for its utility and user's irradiation requirement. The IPS vessel assembly (IVA) consists of IPS head, outer pressure vessel, inner pressure vessel, inner assembly and test fuel carrier. The IVA is approximately 5.6 m long and fits within a 74 mm in diameter envelope over the full height of the chimney. Above the top of the chimney, the head of the IPS is enlarged to allow the closure flanges and pipe work connections. IVA was designed to test the CANDU and PWR nuclear fuel pin together. Specially, wished to minimize interference by nuclear fuel change in design and synthesize these items and shape design for IVA.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.60 no.10
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• pp.1804-1810
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• 2011

It is necessary to adopt some logical techniques and methods of alarm processing for a large complex plant such as nuclear power plants in order to present the occurred alarm messages properly and concisely. Among such alarm processing techniques, the alarm suppressing function is a strong tool to avoid alarm flooding during the sudden transients of plant output power such as turbine trips, reactor trips and other incidents. Unless any suppression or representation technologies are used in an alarm message listing system, it cannot provide quick assistance to plant operators or supervisors during plant upsets because too many alarm messages are presented in an alarm list window. This paper presents the key suppression methods and analysis processes developed for implementing a suppressed alarm message listing function of an integrated alarm system called LogACTs which has been applied to a CANDU nuclear power plant. A simulation testing of the suppressing function conducted with the real plant alarm message list data has demonstrated an effective performance of the developed logics with the high suppression rate.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.1-19
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• 2020

Based on the current high-level radioactive waste management basic plan and the analysis results of spent nuclear fuel characteristics, such as dimensions and decay heat, an improved geological disposal concept for spent nuclear fuel from domestic nuclear power plants was proposed in this study. To this end, disposal container concepts for spent nuclear fuel from two types of reactors, pressurized water reactor (PWR) and Canada deuterium uranium (CANDU), considering the dimensions and interim storage method, were derived. In addition, considering the cooling time of the spent nuclear fuel at the time of disposal, according to the current basic plan-based scenarios, the amount of decay heat capacity for a disposal container was determined. Furthermore, improved disposal concepts for each disposal container were proposed, and analyses were conducted to determine whether the design requirements for the temperature limit were satisfied. Then, the disposal efficiencies of these disposal concepts were compared with those of the existing disposal concepts. The results indicated that the disposal area was reduced by approximately 20%, and the disposal density was increased by more than 20%.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.23 no.2
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• pp.235-243
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• 2010

Three-dimensional structural analysis system for nuclear containment building is presented in this paper. This system includes high-performance plate/shell elements as finite element library. It also adopts numerical modeling technique for unbonded tendon as well as bonded tendon in prestressed concrete structures. This system is constructed by connecting several in-house program to a commercial program DIANA, and then is capable of performing nonlinear analysis for ultimate pressure capacity of nuclear containment building. Finally, three-dimensional structural analysis of CANDU-type containment building is carried out in order to test the reliability of this system. These numerical results are compared with reference values, which obtained from axisymmetric structural analysis.

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.14 no.5
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• pp.1-11
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• 2000

A control system using a programmable logic controller (PLC) has been developed for a dual tower dryer (DTD) in a canadian deuterium natural uranium reactor (CADND) type nuclear power plant. This system will replace the existing DTD control system which was implemented with mechanical timers and relays. The new control system makes it possible for an operator to perform more precise time and dew point control for the DTD, thanks to the high efficiency and flexibility of the PLC. The operational cost for the control system is much reduced compared to the existing system.

• Nuclear Engineering and Technology
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• v.44 no.4
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• pp.355-362
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• 2012

The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict Pressure Tube (PT) diametral creep employing the previously measured PT diameters and operating conditions. There are twelve bundles in a fuel channel, and for each bundle a linear model was developed by using the dependent variables, such as the fast neutron fluences and the bundle coolant temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3, and 4 of the Wolsung nuclear plant in Korea were used to develop the BPLM. The data from the remaining 10 channels were used to test the developed BPLM. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from Units 2, 3, and 4. Two error components for the BPLM, which are the epistemic error and the aleatory error, were generated. The diametral creep prediction and two error components will be used for the generation of the regional overpower trip setpoint at the corresponding effective full power days. The root mean square (RMS) errors were also generated and compared to those from the current prediction method. The RMS errors were found to be less than the previous errors.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2000.11a
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• pp.115-120
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• 2000

The analysis of flux distribution under stead-state in large power reactors with assymetry reactivity insertions requires the use of three-dimensional diffusion calculations. For the purpose, consistently formulated modern nodal methods based on higher order interface techniques have become popular tools for flux distributions in large commercial nuclear reactors. Among the earlier developments, the nodal Green's function method obtains its nodal interface equation from the transverse-integrated integral diffusion equation using a finite-medium Green's function. In this method, the outgoing current from a node surface is formulated as a response of the incoming currents and the spatially integrated neutron source within the same node. The well-known nodal expansion method is also based on an interface partial current formulation. Nodal methods high-level interface variables, i.e., interface net current and flux, may be more computationally efficient than the nodal Green's function method because they have one fewer unknown per interface. The Analytic Nodal Method(ANM), which can be classified as an interface net current technique and, was faster in solving some standard benchmark problems than the other two methods.(omitted)

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.147-157
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• 2004

The production of $^{14}C$ occurs in the Moderator(MOD), Primary Heat Transport System (PHTS), Annulus Gas System(AGS) and Fuel in the CANDU reactor. Among the four systems, The MOD system is the largest contributor to $^{14}C$ production(approximately 94.8%). $^{14}C$ is distributed of $^{14}CO_2$, $H_2^{14}CO_3$, $H^{14}{CO_3}^-$ and $^{14}{CO_3}^{2-}$ species as a function of the pH of water. Of these species, $H_2^{14}CO_3$ and $H^{14}{CO_3}^-$ form are predominant because the pH of MOD system is ＞ 5. In this paper, adsorption-desorption characteristics of bicarbonate ion (${HCO_3}^-$) by IRN 150 resin was investigated. ${HCO_3}^-$ ion existed in neutral condition(app. pH 7)was reacted with ion exchange resin (IRN-150) and saturated with it. Then $NaNO_3$ and $Na_3PO_4$ solutions selected as extraction materials were used to make an investigation into feasibility of ${HCO_3}^-$ extraction from resin saturated with ${HCO_3}^-$. Desorption of $CO^{2+}$ and $Cs^+$ ion by $Na^+$ ion was not occurred, and desorption of ${HCO_3}^-$ ion by ${NO_3}^-$ and ${PO_4}^{3-}$ was occurred slowly. Also, the status of ion exchange which is used in Wolsong NPPs and generation of spent resin yearly were surveyed.

• Journal of Radiation Protection and Research
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• v.22 no.2
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• pp.119-125
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• 1997

PHWR achieves high neutron economy by adopting heavy water as its moderator and coolant. On the other hand it permits much tritium generation, compared to LWR, due to the neutron capture reaction of deuterium in heavy water. Meanwhile in the reactor core, $^3He formed as the result of-decay of tritium, captures a thermal neutron and transforms to tritium again. The existing calculation models on tritium generation in PHWR neglect the contribution of $^3He$ in both moderator and coolant due to its relatively low solubility. However the neutron capture cross-section of $^3He$ is almost $1.6{\times}10^7$ times as large as that of deuterium. That means that the dissolved amount of 0.03 ppm of $^3He$ in heavy water is enough to generate the same amount of tritium as that generated by the deuterium of total heavy water in the system. This study dealt with the contribution of $^3He$ to tritium generation. As a sample case, the contribution of $^3He$ to the tritium generation in Wolsong #1 was evaluated and compared to the measured values. According to the result of this study, it is concluded that $^3He$ in coolant contributes very much to the tritium generation but that in moderator shows negligible effects due to the low solubility and $^4He$ cover gas. At the beginning of the plant operation, the contribution of $^3He$ is slightly greater than the measured value but agrees well with the measured as the operating time increases.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.2
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• pp.125-133
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• 2004

Fission products of DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) fuel, irradiated in HANARO research reactor with 61 ㎾/m of maximum linear power and 1,770 ㎿d/tU of average burn-up, was characterized by EPMA(Electron Probe Micro Analyzer). In order to find accurate characterization, the analysis results by EPMA of fresh simulated DUPIC fuel containing fission products as chemicals were compared with that of wet chemical analysis. The metallic precipitates observed at the center of the fresh simulated DUPIC fuel were about 1 $\mu\textrm{m}$ in size and their major components by EPMA were Mo-53.89 at.%, Ru-37.40 at.%, and Pd+Rh-8.71 at.%. Established procedure through the fresh simulated DUPIC fuel was applied to the irradiated DUPIC fuel. Observed size of metallic precipitates were 2∼2.5 $\mu\textrm{m}$ and their compositions were Mo-47.34 at.%, Ru-46 at.%, and Pd+Rh-6.65 at.%. What are uncommon things for this experiment, special treatment for improving the conductivity was attempted to the specimen and the conditions of exact irradiation of electron beam to small metallic precipitate were suggested.