• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.347-360
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• 2013

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.385-390
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• 2005

The effect of the Wolsong Tritium Removal Facility on the change of tritium concentration in the soil water was assessed by introducing a dynamic compartment model. For the mathematical modeling, the tritium in the environment was thought to come from two different sources. Three global tritium cycling models were compared with the natural background concentration. The dynamic compartment model was used to model the behavior of the tritium from the nuclear power plants at the Wolsong site. The source term for the dynamic compartment model was calculated with the dry and wet deposition rates. The area around the Wolsong nuclear power plants was represented by the compartments. The mechanisms considered in deriving the transfer coefficients between the compartments were evaporation, runoff, infiltration, hydrodynamic dispersion, and groundwater flow. We predicted what the change of the tritium concentration around the Wolsong nuclear power plants would be after future operation of the tritium removal facility to show the applicability of the model. The results showed that the operation of the tritium removal facility would reduce the tritium concentration in topsoil water quickly.

• Nuclear Engineering and Technology
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• v.32 no.3
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• pp.274-282
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• 2000

The simulation of water balance was conducted for suggested four alternative multi-layer cover design of near-surface radioactive waste disposal facility under domestic climate condition. The analysis was also conducted for the most favorable one out of four alternative cover design under conservative scenarios. Until 100 years after closure of disposal vault, the infiltration flux for the most favorable cover design was negligible even under doubling of the ambient precipitation condition. When the degradation of asphalt and geomembrane after 100 years of closure was considered, the infiltration flux significantly increased almost to the design criteria of cover system in I' Aube disposal facility. And it was found that the hydraulic conductivity of bentonite/sand as a bottom barrier should be no greater than 1$\times$10$^{-7}$ cm/sec recommended by U.S. EPA.

• Journal of the Korean Society of Systems Engineering
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• v.17 no.1
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• pp.76-84
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• 2021

Radioactive waste generated from nuclear power plant decommissioning shall be strictly managed so that radioactive materials above the allowable limit are not leaked into the environment. Radioactive wastes shall be classified and treated for management based on characteristics such as the type of waste, physicochemical properties, nuclide concentration and radioactivity. Waste characterization report shall be prepared and submitted to the disposal facility operator to ensure that the treated waste is suitable for disposal. The disposal facility operator shall review the waste Characterization report and visit the nuclear power plant decommissioning site to ensure that the wastes are processed step by step according to the plan. The waste Characterization report may be used as input data to evaluate disposal facility safety. Domestic and foreign data are collected and reviewed to confirm the entire processes from waste generation to delivery. This paper proposes the method to prepare the waste Characterization report which contains data and information on waste characteristics, treatment facilities & method and packaging method & container.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1869-1877
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• 2021

This paper presents the neutron dose rate analysis of the new CONSTOR® RBMK-1500/M2 storage cask intended for the spent nuclear fuel storage at Ignalina Nuclear Power Plant in Lithuania. These casks are designed to be stored in a new "closed" type interim storage facility, with the capacity to store up to 202 CONSTOR® RBMK-1500/M2 casks. In 2016 y, the "hot trials" of this new facility were conducted and 10 CONSTOR® RBMK-1500/M2 casks loaded with the spent nuclear fuel were transported to the dedicated storage places in this facility. During "hot trials", the dose rate measurements of the CONSTOR® RBMK-1500/M2 casks were performed as the dose rate is one of the critical parameter to control and it must be below design (and safety) criteria. Therefore, having the actual data of the spent nuclear fuel characteristics, the neutron dose rate modeling of the CONSTOR® RBMK-1500/M2 cask loaded with this particular fuel was also performed. Neutron dose rate modeling was performed using MCNP 5 computer code with very detailed geometrical representation of the cask and the fuel. The obtained modeling results were compared with the measurement results and it was revealed, that modeling results are generally in good agreement with the measurements.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3505-3516
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• 2021

During nuclear facility decommissioning, workers are continuously exposed to high-level radiation. Hence, adequate path planning is critical to protect workers from unnecessary radiation exposure. This work discusses recent development in radioactive path planning and the algorithms recommended for the task. Specifically, we review the conventional methods for nuclear decommissioning path planning, analyze the techniques utilized in developing algorithms, and enumerate the decision factors that should be considered to optimize path planning algorithms. As a major contribution, we present the quantitative performance comparison of different algorithms utilized in solving path planning problems in nuclear decommissioning and highlight their merits and drawbacks. Also, we discuss techniques and critical consideration necessary for efficient application of robots and robotic path planning algorithms in nuclear facility decommissioning. Moreover, we analyze the influence of obstacles and the environmental/radioactive source dynamics on algorithms' efficiency. Finally, we recommend future research focus and highlight critical improvements required for the existing approaches towards a safer and cost-effective nuclear-decommissioning project.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.991-992
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• 2004

• Nuclear Engineering and Technology
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• v.39 no.6
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• pp.747-752
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• 2007

The Proton Engineering Frontier Project (PEFP) has designed and developed a proton linear accelerator facility operating at 100 MeV - 20 mA. The radiological effects of such a nuclear facility on the environment are important in terms of radiation safety. This study estimated the production rates of radionuclides in the soil around the accelerator facility using MCNPX. The groundwater migration of the radioisotopes was also calculated using the Concentration Model. Several spallation reactions have occurred due to leaked neutrons, leading to the release of various radionuclides into the soil. The total activity of the induced radionuclides is approximately $2.98{\times}10^{-4}Bq/cm^3$ at the point of saturation. $^{45}Ca$ had the highest production rate with a specific activity of $1.78{\times}10^{-4}Bq/cm^3$ over the course of one year. $^3H$ and $^{22}Na$ are usually considered the most important radioisotopes at nuclear facilities. However, only a small amount of tritium was produced around this facility, as the energy of most neutrons is below the threshold of the predominant reactions for producing tritium: $^{16}O(n,\;X)^3H$ and $^{28}Si(n,X)^3H$ (approximately 20 MeV). The dose level of drinking water from $^{22}Na$ was $1.48{\times}10^{-5}$ pCi/ml/yr, which was less than the annual intake limit in the regulations.

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.387-398
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• 2003

For reference human intrusion scenarios constructed in previous study, a probabilistic safety assessment to derive the radionuclide concentration limits for the low- and intermediate- level radioactive waste disposal facility is conducted. Statistical approach by the Latin Hypercube Sampling method is introduced and new assumptions about the disposal facility system are examined and discussed. In our previous study of deterministic approach, the post construction scenarios appeared as most limiting scenario to derive the radionuclide concentration limits. Whereas, in this statistical approach, the post drilling and the post construction scenarios are mutually competing for the scenario selection according to which radionuclides are more important in safety assessment context. Introduction of new assumption shows that the post drilling scenario can play an important role as the limiting scenario instead of the post-construction scenario. When we compare the concentration limits between the previous and this study, concentrations of radionuclides such as Nb-94, Cs-137 and alpha-emitting radionuclides show elevated values than the case of the previous study. Remaining radionuclides such as Sr-90, Tc-99 I-129, Ni-59 and Ni-63 show lower values than the case of the previous study.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.989-995
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• 2018

Argonne National Laboratory of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine, and its commissioning process is underway. The facility will be used for researches, producing medical isotopes, and training young nuclear specialists. The neutron source facility is designed with a provision to include a cryogenically cooled moderator system-a cold neutron source (CNS). This CNS provides low-energy neutrons, which will be used in the scattering experiment and material structures analysis. Cold neutron guides, coated with reflective material for the low-energy neutrons, will be used to transport the cold neutrons to the experimental site. The cold neutron guides would keep the cold neutrons within certain energy and angular space concentrated inside, while most of the gamma rays and high-energy neutrons are not affected by the cold neutron guides. For the KIPT design, the cold neutron guides need to extend several meters outside the main shield of the facility, and curved guides will also be used to remove the gamma and high-energy neutron. The neutron guides should be installed inside a shield structure to ensure an acceptable biological dose in the facility hall. Heavy concrete is the selected shielding material because of its acceptable performance and cost. Shield design analysis was carried out for the CNS guide hall. MCNPX was used as the major computation tool for the design analysis, with neutron and gamma dose calculated separately. Weight windows variance reduction technique was also used in the shield design. The goal of the shield design is to keep the total radiation dose below the $5.0{\mu}Sv/hr$ guideline outside the shield boundary. After a series of iterative MCNPX calculations, the shield configuration and parameters of CNS guide hall were determined and presented in this article.