• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3183-3193
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• 2023

The water radiolysis in-core at light water reactors (LWRs) produces various radicals with other ionic species/molecules and radioactive nitrogen species in the reactor coolant. Nitrogen species can exist in many different chemical forms and recirculate in water and steam, and consequently contribute to what extent the environmental safety at nuclear power plants. Therefore, a clear understanding of formation kinetics and chemical behaviors of nitrogen species under irradiation is crucial for better insight into the characteristics of major radioactive species released to the main steam or relevant coolant systems and eventually development of advanced processes/methodologies to enhance the environmental safety at nuclear power plants. This paper thus focuses on basic principles on electrochemical interaction kinetics of radiolytic molecules and various nitrogen species in high temperature water, fundamental approaches for calculating thermodynamic values to predict their stability and domain in LWRs, and the effect of nitrogen species on crevice chemistry/corrosion and intergranular stress corrosion cracking (IGSCC) susceptibility of structure materials in high temperature water.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4338-4347
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• 2022

The CUPID code was developed and is continuously updated in KAERI. Verification and validation (V&V) is mainly done for light water reactors (LWRs). This paper describes a benchmarking of the detailed mesh level compared with sub-channel level for application to pressurized heavy water reactors (PHWRs), even though component scale comparison for the PHWR moderator system was done once before. We completed a sub-channel level comparison between the CUPID code and the ASSERT code and a CUPID code analysis. Because the ASSERT code has already been validated with numerous experiments, benchmarking with the ASSERT code will offer us more trust on the CUPID code. The target channel has high power and thus high pressure deformation. The high power channel tends to have a high possibility of critical heat flux (CHF), because a high void fraction and quality in channel exit region appear. In this research, after determining the reference grid and T/H model, we compared the sub-channel level results of the CUPID code with those of the ASSERT code.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2315-2324
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• 2023

Space nuclear reactors are becoming popular in deep space exploration owing to their advantages of high-power density and stability. Following the fourth-generation nuclear reactor technology, a conceptual design of the dual drum-controlled space molten salt reactor (D²-SMSR) is proposed. The reactor concept uses molten salt as fuel and heat pipes for cooling. A new reactivity control strategy that combines control drums and safety drums was adopted. Critical physical characteristics such as neutron energy spectrum, neutron flux distribution, power distribution and burnup depth were calculated. Flow and heat transfer characteristics such as natural convection, velocity and temperature distribution of the D²-SMSR under low gravity conditions were analyzed. The reactivity control effect of the dual-drums strategy was evaluated. Results showed that the D²-SMSR with a fast spectrum could operate for 10 years at the full power of 40 kWth. The D²-SMSR has a high heat transfer coefficient between molten salt and heat pipe, which means that the core has a good heat-exchange performance. The new reactivity control strategy can achieve shutdown with one safety drum or three control drums, ensuring high-security standards. The present study can provide a theoretical reference for the design of space nuclear reactors.

• Journal of Radiation Industry
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• v.18 no.2
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• pp.117-126
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• 2024

This study aims to identify high exposure tasks among the tasks performed in domestic nuclear power plants as a basis for developing training programs to improve the efficiency of workers' work. To this end, we first analyzed the exposure status of radiation work in domestic nuclear power plants. Radiation tasks in nuclear power plants were categorized, collective doses were investigated, and the collective doses were calculated based on the collective doses, and representative high exposure tasks were identified. We found that the collective and individual doses in domestic nuclear power plants are continuously decreasing, but there is an imbalance of exposure among workers. In terms of work classification, nuclear power plants are managed in 236 work codes based on light water reactors and 181 work codes based on heavy water reactors, depending on the work equipment and location. Among the total work codes, 23 codes have an annual average dose exceeding 10 μSv, and based on this, 10 representative high exposure tasks were derived. The representative high exposure tasks were selected as S/G nozzle dam work, S/G debris removal work, nuclear instrumentation system, S/G eddy current detection work, and insulation work. The results of this study are expected to serve as an important basis for reducing the exposure of workers in nuclear power plants and improving work efficiency.

• Nuclear Engineering and Technology
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• v.41 no.4
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• pp.493-520
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• 2009

During the last four decades, 16 Pressurized Water Reactors (PWR) and 4 Pressurized Heavy Water Reactors (PHWR) have been constructed and operated in Korea, and nuclear fuel technology has been developed to a self-reliant state. At first, the PWR fuel design and manufacturing technology was acquired through international cooperation with a foreign partner. Then, the PWR fuel R&D by Korea Atomic Energy Research Institute (KAERI) has improved fuel technology to a self-reliant state in terms of fuel elements, which includes a new cladding material, a large-grained $UO_2$ pellet, a high performance spacer grid, a fuel rod performance code, and fuel assembly test facility. The MOX fuel performance analysis code was developed and validated using the in-reactor test data. MOX fuel test rods were fabricated and their irradiation test was completed by an international program. At the same time, the PWR fuel development by Korea Nuclear Fuel (KNF) has produced new fuel assemblies such as PLUS7 and ACE7. During this process, the design and test technology of fuel assemblies was developed to a self-reliant state. The PHWR fuel manufacturing technology was developed and manufacturing facility was set up by KAERI, independently from the foreign technology. Then, the advanced PHWR fuel, CANFLEX(CANDU Flexible Fuelling), was developed, and an irradiation test was completed in a PHWR. The development of the CANFLEX fuel included a new design of fuel rods and bundles.. The nuclear fuel technology in Korea has been steadily developed in many national R&D programs, and this advanced fuel technology is expected to contribute to a worldwide nuclear renaissance that can create solutions to global warming.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2442-2451
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• 2020

The safety level of Sodium Fast Reactors is directly related with the sodium void reactivity. A low-void effect design has been proposed within the Horizon2020 ESFR-SMART project thanks to the introduction of a sodium plenum above the active core. In order to assess the impact of this core conception on transient analysis, a map with the spatial distribution of sodium void worth can be computed and fed into a point-kinetics-based transient code. Due to the spatial correlations between neighboring zones, the global effect of voiding two different axial or radial regions is not necessarily the sum of both individual contributions. Neglecting those correlations in the void worth map and consequently in the transient analysis may lead to an unrealistic prediction of the transient sequences. In this work, a method based on sensitivity analysis and similarity assessment is proposed for predicting those correlations. The method proved to be able to establish correlations between axial slices of a sub-assembly and was checked against realistic sodium void propagation patterns.

• Nuclear Engineering and Technology
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• v.33 no.4
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• pp.365-374
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• 2001

In order to examine the in-reactor performance of very-high-density dispersion fuels for high flux performance research reactors, U-l0wt.%Mo microplates containing centrifugally atomized powder were irradiated at low temperature. The U-l0wt.%Mo dispersion fuels show stable in- reactor irradiation behaviors even at high burn-up, similar to U$_3$Si$_2$ dispersion fuels. The atomized U-l0wt.%Mo fuel particles have a fine and a relatively uniform fission gas bubble size distribution. Moreover, only one of third of the area of the atomized fuel cross-sections at 70a1.% burn-up shows fission gas bubble-free zones, This appears to be the result of segregation into high Mo and low Mo.

• International Journal of Fuzzy Logic and Intelligent Systems
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• v.3 no.1
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• pp.112-120
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• 2003

This paper reports the development work carried out in respect of a proposed application of Neural Network approach for the Korean Next generation Reactor (KNGR) now referred as APR-1400. The emphasis is on establishing the methodology and the approach to be adopted towards realizing this application in the next generation reactors. Keeping in view the advantages and limitation of Artificial Neural Network Approach, the role of ANN has been limited to plant status or to be more precise plant transient monitoring. The simulation work carried out so far and the results obtained shows that artificial neural network approach caters to the requirements of plant status monitoring and qualifies to be incorporated as a part of proposed operator support systems of the referenced nuclear power plant.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.26-32
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• 2016

Research and development (R&D) methodology for the practical use of accident tolerant fuel (ATF) in commercial light water reactors is discussed in the present review. The identification and quantification of the R&D-metrics and the attribute of candidate ATF-concepts, recognition of the gap between the present R&D status and the targeted practical use, prioritization of the R&D, and technology screening schemes are important for achieving a common understanding on technology screening process among stakeholders in the near term and in developing an efficient R&D track toward practical use. Technology readiness levels and attribute guides are considered to be proper indices for these evaluations. In the midterm, the selected ATF-concepts will be developed toward the technology readiness level-5, at which stage the performance of the prototype fuel rods and the practicality of industrial scale fuel manufacturing will be verified and validated. Regarding the screened-out concepts, which are recognized to have attractive potentials, the fundamental R&D should be continued in the midterm to find ways of addressing showstoppers.

• Nuclear Engineering and Technology
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• v.38 no.7
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• pp.591-618
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• 2006

An integrated R&D program is being conducted to study, qualify, and in some cases, develop materials with required properties for the reactor systems being developed as part the U.S. Department of Energy's Generation IV Reactor Program. The goal of the program is to ensure that the materials research and development (R&D) needed to support Gen IV applications will comprise a comprehensive and integrated effort to identify and provide the materials data and its interpretation needed for the design and construction of the selected advanced reactor concepts. The major materials issues for the five primary systems that have been considered within the U.S. Gen IV Reactor Program-very high temperature gas-cooled, supercritical water-cooled, gas-cooled fast spectrum, lead-cooled fast spectrum, and sodium-cooled fast spectrum reactors-are described along with the R&D that has been identified to address them.