• Nuclear Engineering and Technology
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• v.16 no.3
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• pp.169-179
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• 1984

An efficient procedure of evaluating the fuel cladding failures occurring in the normal operations of typical PWR's has been investigated through the analysis of fission product(FP) activities in the reactor coolant using an analytical model, FIPREL code. Performed by this code is an extensive study on the sensivities of FP activities to such physical parameters as enrichment, turnup, and operation temperature of failed fuel rod as well as the effective failure size quantified in terms of the magnitude of gap release coefficient. The results of study are generally in agreement with those by PROFIP method. In the presence of tramp uranium the portion of activities released from failed rod is separated by an iterative calculation based on the activity ratios of fission nuclides chemically more stable than iodines. Obtained are the linear power density and the number of failed rods, the effective failure size, and the mass of tramp uranium. The operation experiences of 4 cycles of Kori Unit 1 are analyzed and the results show that the model is highly reliable for the survey and evaluation of fuel rod conditions during reactor operations.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.1
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• pp.51-58
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• 2021

Chemical decontamination of primary systems in a nuclear power plant (NPP) prior to commencing the main decommissioning activities is required to reduce radiation exposure during its process. The entire process is repeated until the desired decontamination factor is obtained. To achieve improved decontamination factors over a shorter time with fewer cycles, the appropriate flow characteristics are required. In addition, to prepare an operating procedure that is adaptable to various conditions and situations, the transient analysis results would be required for operator action and system impact assessment. In this study, the flow characteristics in the steady-state and transient conditions for the chemical decontamination operations of the Kori-1 NPP were analyzed and compared via the MARS-KS code simulation. Loss of residual heat removal (RHR) and steam generator tube rupture (SGTR) simulations were conducted for the postulated abnormal events. Loss of RHR results showed the reactor coolant system (RCS) temperature increase, which can damage the reactor coolant pump (RCP)s by its cavitation. The SGTR results indicated a void formation in the RCS interior by the decrease in pressurizer (PZR) pressure, which can cause surface exposure and tripping of the RCPs unless proper actions are taken before the required pressure limit is achieved.

• Nuclear Engineering and Technology
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• v.13 no.1
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• pp.22-35
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• 1981

The U$_3$O$_{8}$ requirements are estimated for the high, intermediate, and low growth projections of nuclear power in Korea. To each projection, four illustrative reactor-mix strategies and four fuel cycle options are applied for estimating the requirements. The reactor types considered are PWR, PHWR. and FBR. The fuel cycles considered are once-through cycle, U/Pu recycle, and improved once-through cycle. Also the amount of Pu-fissile recovered from U recycle is estimated. The maximum cumulative (to the year 2000) requirements of U$_3$O$_{8}$ occupy about 4 to 5 percent of the WOCA requirements and are about 23 times larger than the U$_3$O$_{8}$ resources in Korea. For the high nuclear power growth projection, the cumulative amount of Pu-fissile recovered from U recycle is sufficient for the startup of 2 units of 1200 MWe fast reactors by the year 2000. 2000.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.12
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• pp.2581-2589
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• 2002

In this research, fretting tests were conducted in air to investigate the wear characteristics of fuel cladding materials with the fretting parameters such as normal load, slip amplitude, frequency and the number of cycles. A high frequency fretting wear tester was designed for this experiment by KAERI. After the experiments, the wear volume and the shape of wear contour were measured by the surface roughness tester. Tribologically transformed structures(TTS) were analysed by means of optical and scanning electron microscopes to identify the main wear mechanisms. The results of this study showed that the wear volume were increased with increasing slip amplitude, and the shape of wear contour was transformed V-type to W-type. Also, it was found that the critical slip amplitude was 168${\mu}{\textrm}{m}$. These phenomena mean that wear mechanism transformed partial slip to gross slip to accelerate wear volume. The wear depth increased with an increase of friction coefficient due to increase of normal load and frequency. The fretting wear mechanisms were believed that, after adhesion and surface plastic deformation occurred by relative sliding motion on the contact between two specimens, TTS creation was induced by surface strain hardening and wear debris were detached from the contact surface which were produced by the micro crack propagation and creation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.4
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• pp.339-350
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• 2007

Status on the spent nuclear fuel management policy and R&D plan of the major countries is surveyed. Also the prospect of the future R&D plan is suggested. Recently so-called fuel cycle nations, which have the reprocess policy of the spent fuel, announced new spent fuel management policy based on the advanced fuel cycle technology. The policy is focused to transmute highly radioactive material and material having a very long half-life, and to recycle the Pu and U contained in the spent fuel. In this way the radio-foxily of the spent fuel as well as the amount of the high level waste to be eventually disposed can greatly be reduced. Most of countries selected the wet process as a primary option for the treatment of the spent fuel since the advanced wet process, which is based on the existing PUREX process, looks more feasible as compared with the dry process. The wet process, however, is much more sensitive in terms of proliferation-resistance compared with the dry process. The pure Pu can easily be obtained by simply modifying the process. On the other hand the pure Pu can not be extracted in the dry process based on the high temperature molten salt process such as a pyroprocess. Even though the pyroprocess technology is very premature, it has a great merit. Thus it is necessary for Korea to have a long term strategy for pursuing a spent fuel treatment technology with a proliferation resistance and a great merit for the GEN-IV fuel cycles. Pyroprocess is one of the best candidates to satisfy these purposes.

• Korean Chemical Engineering Research
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• v.52 no.3
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• pp.279-288
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• 2014

Nuclear energy is expected to meet the growing energy demand while avoiding CO2 emission. However, the problem of accumulating spent fuel from current nuclear power plants which is mainly composed of uranium oxides should be addressed. One of the most practical solutions is to reduce the spent oxide fuel and recycle it. Next-generation fuel cycles demand innovative features such as a reduction of the environmental load, improved safety, efficient recycling of resources, and feasible economics. Pyroprocessing based on molten salt electrolysis is one of the key technologies for reducing the amount of spent nuclear fuel and destroying toxic waste products, such as the long-life fission products. The oxide reduction process based on the electrochemical reduction in a LiCl-$Li_2O$ electrolyte has been developed for the volume reduction of PWR (Pressurized Water Reactor) spent fuels and for providing metal feeds for the electrorefining process. To speed up the electrochemical reduction process, the influences of the feed form for the cathode and the type of anode shroud on the reduction rate were investigated.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.185-190
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• 1996

RFSP is a computer program to do fuel management calculations for CANDU reactors. Its main function is to calculate neutron flux and power distributions using two-energy-group, three dimensional neutron diffusion theory. However, up to now the treatment has not been true two-group but actually "one-and-half groups". In other words, the previous (1.5-group) version of RFSP lumps the fast fission term into the thermal fission term. This is based on the POWDERPUFS-V Westcott convention. Also, there is no up-scattering term or bundle power over cell flux (H1 factor) for the fast group. While POWDERPUFS-V provides only 1.5 group properties, true two-group cross sections for the design and analysis of CAUDU reactors can be obtained from WIMS-AECL. To treat the full two-group properties, the previous RFSP version was modified by adding the fast fission, up-scatter terms, and H1 factor. This two-group version of RFSP is a convenient tool to accept lattice properties from any advanced lattice code (e.g. WIMS-AECL DRAGON, HELIOS...) and to apply to advanced fuel cycles. In this study, the modification to implement the true two-group treatment was performed only in the subroutines of the ＊SIMULATE module of RFSP. This module is the appropriate one to modify first, since it is used for the tracking of reactor operating histories. The modified two-group RFSP was evaluated with true two-group cross sections from WIMS-AECL. Some tests were performed to verify the modified two-group RFSP and to evaluate the effects of fast fission and up-scatter for three core conditions and four cases corresponding to each condition. The comparisons show that the two-group results are quite reasonable and serve as a verification of the modifications made to RFSP. To assess the long-term impact of the full 2-group treatment, it is necessary to simulate a long period (several months) of reactor history. It will also be necessary to implement the full two-group treatment of reactivity devices and assess the reactivity-device worths.ce worths.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.15-28
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• 2019

In accordance with the decommissioning plan for the Kori Unit 1 NPP, the reactor coolant system will be chemically decontaminated as soon as possible after permanent shutdown. This study developed the chemical decontamination process though the development project of decontamination technology of reactor coolant system and dismantled equipment for NPP decommissioning, which has been carried out since 2014. In this study, Oxidation/reduction process was conducted using system decontamination process development equipment of lab scale and was divided into unit and continuous processes. The optimal process time was derived from the unit process, and decontamination agent and the number of process were derived through the continuous processes. Through the unit process, the oxidation process took 5 hours and the reduction process took 4 hours. As optimum decontamination agent, the oxidizing agent was $200mg{\cdot}L^{-1}$ Permanganic acid + $200mg{\cdot}L^{-1}$ Nitric acid and the reducing agent was $2000mg{\cdot}L^{-1}$ Oxalic acid. In the case of the number of processes, all oxide films were removed during the two-cycle chemical decontamination process of STS304 and SA508. In the case of Alloy600, all oxide films were removed when chemical decontamination was performed for three cycles or more.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.10
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• pp.673-679
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• 2016

Because of the growing interest in supercritical carbon dioxide power cycle technology owing to its potential enhancement in compactness and efficiency, supercritical carbon dioxide cycles have been studied in the fields of nuclear power, concentrated solar power (CSP), and fossil fuel power generation. This study introduces the current status of the research project on the supercritical carbon dioxide power cycle by Korea Institute of Energy Research (KIER). During the first phase of the project, the un-recuperated supercritical Brayton cycle test loop was built and tested. In phase two, researchers are designing and building a supercritical carbon dioxide dual Brayton cycle, which utilizes two turbines and two recuperators. Under the simulation condition considered in this study, it was confirmed that the design parameter has an optimal value for maximizing the net power in the supercritical carbon dioxide dual cycle.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.39 no.10
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• pp.787-793
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• 2015

In response to the growing interest in supercritical carbon dioxide ($S-CO_2$) power cycle technology because of its potential enhancement in compactness and efficiency, the $S-CO_2$ cycles have been studied intensively in the fields of nuclear power, concentrated solar power (CSP), and fossil fuel power generation. Despite this interest, there are relatively few studies on waste heat recovery applications. In this study, the $S-CO_2$ cycle that has a split flow with preheating was modeled and simulated. The variation in the power was investigated with respect to the changes in the value of a design parameter. Under the simulation conditions considered in this study, it was confirmed that the design parameter has an optimal value that can maximize the power in the $S-CO_2$ power cycle that has a split flow with preheating.