• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3581-3590
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• 2023

High-temperature design evaluations were conducted on Type 316L stainless steel piping for a 700 ℃ large-capacity thermal energy storage verification test loop (TESET) under construction at KAERI. The hot leg piping with sodium coolant at 700 ℃ connects the main components of the loop heater, hot storage tank, and air-to-sodium heat exchanger. Currently, the design rules of ASME B31.1 and RCC-MRx provide design procedures for high-temperature piping in the creep range for Type 316L stainless steel. However, the design material properties around 700 ℃ are not available in those rules. Therefore, a number of material tests, including creep tests at various temperatures, were conducted to determine the insufficient material properties and relevant design coefficients so that high-temperature design on the 700 ℃ piping may be possible. It was shown that Type 316L stainless steel can be used in a 700 ℃ high-temperature piping system of Generation IV reactor systems or a renewable energy systems, such as thermal energy storage systems, for a limited operation time.

• Journal of the Korean Society for Nondestructive Testing
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• v.36 no.1
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• pp.60-67
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• 2016

A pressurized heavy water reactor (PHWR) designed to refuel in service produces the energy required by nuclear fission. The fuel channel consists of components such as a pressure tube which directly contacts the fuel and is a passage for the reactor coolant, a calandria tube which contacts the moderator and is rolled joint with calandria, and a spacer which is not to contact the pressure tube and a calandria tube. As the fuel channel is one of the most important equipments, it requires accurate and periodic inspections to assess the integrity of a reactor in accordance with CSA N285.4. A fuel channel inspection system is developed to inspect fuel channels during in-service inspection in Wolsong unit. In this paper, the results and considerations of a field test are presented in order to show the effectiveness of the developed fuel channel inspection system.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2008.04a
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• pp.641-646
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• 2008

Long and slender body with or without flexible supports under severe operating condition can be unstabilized even by the small cross flow. Turbulent flow mixer, which actually increases thermal-hydraulic performance of the nuclear fuel by boosting turbulence, disturbs the flow field around the fuel rod and affects dynamic behavior of the nuclear fuel rods. Few studies on this problem can be found in the literature because these effects depend on the specific natures of the support and the design of the system. This work shows how the dynamics of a multi-span fuel rod can be affected by the turbulent flow, which is discretely activated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was established. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.

• Nuclear Engineering and Technology
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• v.31 no.1
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• pp.68-79
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• 1999

The YGN Units 3&4 plant conditions during shutdown operation were reviewed to identify the possible event scenarios following the loss of shutdown cooling (SDC) event. For the five cases of typical reactor coolant system (RCS) configurations under the worst event sequence, such as unavailable secondary cooling and no RCS inventory makeup, the thermal hydraulic analyses were performed using the RELAP5/MOD3.2 code to investigate the plant behavior following the event. The thermal hydraulic analyses include the estimation of time to boil, time to core uncovery, and time to core heat up to determine the containment closure time to prevent the uncontrolled release of fission products to atmosphere. The result indicates that the containment closure is recommended to be achieved within 42 minutes after the loss of SDC for the steam generator (SG) inlet plenum manway open case or the large cold leg open case under the worst event sequence. The containment closure time is significantly dependent on the elevation and size of the opening and the SG secondary water level condition. It is also found that the containment closure needs to be initiated before the boiling time to ensure the survivability of the workers in the containment. These results will provide useful information to operators to cope with the loss of SDC event.

• 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 Korean Nuclear Society Conference
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• 1998.05a
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• pp.417-422
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• 1998

he Reactor Coolant Gas Vent System(RCGVS) design for Ulchin Nuclear Power Plant Units 3&4(UCN 3&4) has been improved from the Yonggwang Nuclear Power Plant Units 3&4(YGN 3&4) based on the evaluation results for depressurization capability tests performed at YGN 3&4. There has been a series of plant safety analyses for Natural Circulation Cooldown(NCC) event and thermo-dynamic analyses with RELAP5 code for the steam blowdown Phenomena in order to optimize the orifice size of UCN 3&4 RCGVS. Baesd on these analyses results, the RCGVS orifice size for UCN 3&4 has been reduced to 9/32 inch from the l1/32 inch for YGN 3&4. The depressurization capability tests, which were performed at UCN 3 in order to verify the FSAR NCC analysis results, show that the RCGVS depressurization rates are being within the acceptable ranges. Therefore, it is concluded that the orificed flow path of UCN 3&4 RCGVS is adequately designed, and can provide the safety-grade depressurization capability required for a safe plant operation.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2757-2772
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• 2023

The fluid retention effect of the Volume Control Tank (VCT) leads to a long time delay in Reactor Coolant System (RCS) concentration during the boration process. A coupled model combining a lumped-parameter sub-model and a computational fluid dynamics sub-model is currently used to investigate the concentration dynamic characteristic of RCS during the boration process. This model is validated by comparison with experimental data, and the predicted results show excellent agreement with experimental data. We provide detailed fields in VCT and concentration variations of RCS to study the interaction between mixing in VCT and the transient responses of RCS. Moreover, the impacts of the inlet flow rate, inlet nozzle diameter, original concentration, and replenishing temperature of VCT on the RCS concentration characteristic are studied. The inlet flow rate and nozzle diameter of VCT remarkably affect the RCS concentration characteristic. Too-large or too-small inlet flow rates and nozzle diameters will lead to unacceptable long delays. In this work, the optimal inlet flow rate and nozzle diameter of VCT are 5 m³/h and 58.8 mm, respectively. Besides, the impacts of the original concentration and replenishing temperature of VCT are negligible under normal operating conditions.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2454-2465
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• 2023

Film boiling may lead to burnout of the heating element. Even though burnout does not occur, the heating element is subject to deformation because it is not sufficiently strong to withstand external loads. In particular, the ballooning and rupture of a tube under film boiling are important phenomena in the field of nuclear reactor safety. If the tube-type cladding of nuclear fuel ruptures owing to high internal pressure and thermal load, radioactive materials inside the cladding are released to the coolant. Therefore, predicting the ballooning and rupture is important. This study presents numerical simulations to predict the ballooning behavior and rupture time of a horizontal tube at high internal pressure under saturated film boiling. To do so, a multi-step coupled simulation of conjugated film boiling heat transfer and ballooning using creep model is adopted. The numerical methods and models are validated against experimental values. Two different nonuniform heat flux distributions and four different internal pressures are considered. The three-step simulation is enough to obtain a convergent result. However, the single-step simulation also successfully predicts the rupture time. This is because the film boiling heat transfer characteristics are slightly affected by the tube geometry related to creep ballooning.

• 한국태양에너지학회:학술대회논문집
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• 2011.04a
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• pp.204-209
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• 2011

According to the environment report of UN, korea was classified as potable water shortage countries. Approximately 71% of the Earth's surface is covered by ocean. However, it is difficult to use for industry of residential purpose without a certain processing. The development of solar and waste-heat used absorption desalination technology have been examined as a viable option for supplying clean energy. In this study, the modelling of the main devices for solar and waste-heat used and adsorption desalination system was introduced. The design is divided into three parts. First, the evaporator for the vaporization of the top water is designed, and then the reactor for the adsorption and release of the steam is designed, followed by the condenser for the condensation of the fresh water is designed. In addition, new features based on the energy balance are also included to design absorption desalination system. In this basicresearch, One-bed(reactor) adsorption desalination plant that employ a low-temperature solar and waste energy was proposed and experimentally studied. The specific water yield is measured experimentally with respect to the time controlling parameters such as heat source temperatures, coolant temperatures, system switching and half-cycle operational times.

• Nuclear Engineering and Technology
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• v.34 no.1
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• pp.42-59
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• 2002

In order to transmute long-lived radioactive nuclides such as transuranics(TRU), Tc-99, and I- l29 in LWR spent fuel, a preliminary conceptual design study has been performed for the accelerator driven subcritical reactor system, called HYPER(Hybrid Power Extraction Reactor) The core has a hybrid neutron energy spectrum: fast and thermal neutrons for the transmutation of TRU and fission products, respectively. TRU is loaded into the HYPER core as a TRU-Zr metal form because a metal type fuel has very good compatibility with the pyre- chemical process which retains the self-protection of transuranics at all times. On the other hand, Tc-99 and I-129 are loaded as pure technetium metal and sodium iodide, respectively. Pb-Bi is chosen as a primary coolant because Pb-Bi can be a good spallation target and produce a very hard neutron energy spectrum. As a result, the HYPER system does not have any independent spallation target system. 9Cr-2WVTa is used as a window material because an advanced ferritic/martensitic steel is known to have a good performance under a highly corrosive and radiation environment. The support ratios of the HYPER system are about 4∼5 for TRU, Tc-99, and I-129. Therefore, a radiologically clean nuclear power, i.e. zero net production of TRU, Tc-99 and I-129 can be achieved by combining 4 ∼5 LWRs with one HYPER system. In addition, the HYPER system, having good proliferation resistance and high nuclear waste transmutation capability, is believed to provide a breakthrough to the spent fuel problems the nuclear industry is faced with.