NUCLEAR HUMAN RESOURCE PROJECTION UP TO 2030 IN KOREA

Min, Byung-Joo,Lee, Man-Ki,Nam, Kee-Yung,Jeong, Ki-Ho Korean Nuclear Society 2011 Nuclear Engineering and Technology Vol.43 No.4

The prospects for growth of the nuclear power industry in Korea have improved remarkably as the demand for energy increases in stride with economic development. Meanwhile, as nuclear energy development is enhanced, nuclear technology has also improved evolutionarily and innovatively in the areas of reactor design and safety measures. As nuclear technology development in Korea advances, more human resources are required. Accordingly, the need for a well-managed program of human resource development (HRD) aimed at assuring needed capacities, skills, and knowledge and maintaining valuable human resources through education and training in various nuclear-related fields has been recognized. A well-defined and object-oriented human resource development and management (HRD&M) is to be developed in order to balance between the dynamics of supply and demand of the workforce in the nuclear industry. The HRD&M schemes include a broad base of disciplines, education, sciences, and technologies within a framework of national sustainable development goals, which are generally considered to include economics, environment, and social concerns. In this study, the projection methodology considering a variety of economic, social, and environmental factors was developed. Using the developed methodology, medium- and long-term nuclear human resources projections up to 2030 were conducted in compliance with the national nuclear technology development programmes and plans.

Conceptual design of a high neutron flux research reactor core with low enriched uranium fuel and low plutonium production

Rahimi, Ghasem,Nematollahi, MohammadReza,Hadad, Kamal,Rabiee, Ataollah Korean Nuclear Society 2020 Nuclear Engineering and Technology Vol.52 No.3

Research reactors for radioisotope production, fuel and material testing and research activities are designed, constructed and operated based on the society's needs. In this study, neutronic and thermal hydraulic design of a high neutron flux research reactor core for radioisotope production is presented. Main parameters including core excess reactivity, reactivity variations, power and flux distribution during the cycle, axial and radial power peaking factors (PPF), Pu₂₃₉ production and minimum DNBR are calculated by nuclear deterministic codes. Core calculations performed by deterministic codes are validated with Monte Carlo code. Comparison of the neutronic parameters obtained from deterministic and Monte Carlo codes indicates good agreement. Finally, subchannel analysis performed for the hot channel to evaluate the maximum fuel and clad temperatures. The results show that the average thermal neutron flux at the beginning of cycle (BOC) is 1.0811 × 10¹⁴ n/㎠-s and at the end of cycle (EOC) is 1.229 × 10¹⁴ n/㎠-s. Total Plutonium (Pu₂₃₉) production at the EOC evaluated to be 0.9487 Kg with 83.64% grade when LEU (UO₂ with 3.7% enrichment) used as fuel. This designed reactor which uses LEU fuel and has high neutron flux and low plutonium production could be used for peaceful nuclear activities based on nuclear non-proliferation treaty concepts.

Development of a Traceability Analysis Method Based on Case Grammar for NPP Requirement Documents Written in Korean Language

Yoo Yeong Jae,Seong Poong Hyun,Kim Man Cheol Korean Nuclear Society 2004 Nuclear Engineering and Technology Vol.36 No.4

Software inspection is widely believed to be an effective method for software verification and validation (V&V). However, software inspection is labor-intensive and, since it uses little technology, software inspection is viewed upon as unsuitable for a more technology-oriented development environment. Nevertheless, software inspection is gaining in popularity. KAIST Nuclear I&C and Information Engineering Laboratory (NICIEL) has developed software management and inspection support tools, collectively named "SIS-RT. "SIS-RT is designed to partially automate the software inspection processes. SIS-RT supports the analyses of traceability between a given set of specification documents. To make SIS-RT compatible for documents written in Korean, certain techniques in natural language processing have been studied [9]. Among the techniques considered, case grammar is most suitable for analyses of the Korean language [3]. In this paper, we propose a methodology that uses a case grammar approach to analyze the traceability between documents written in Korean. A discussion regarding some examples of such an analysis will follow.

A SMALL MODULAR REACTOR DESIGN FOR MULTIPLE ENERGY APPLICATIONS: HTR50S

Yan, X.,Tachibana, Y.,Ohashi, H.,Sato, H.,Tazawa, Y.,Kunitomi, K. Korean Nuclear Society 2013 Nuclear Engineering and Technology Vol.45 No.3

HTR50S is a small modular reactor system based on HTGR. It is designed for a triad of applications to be implemented in successive stages. In the first stage, a base plant for heat and power is constructed of the fuel proven in JAEA's $950^{\circ}C$, 30MWt test reactor HTTR and a conventional steam turbine to minimize development risk. While the outlet temperature is lowered to $750^{\circ}C$ for the steam turbine, thermal power is raised to 50MWt by enabling 40% greater power density in 20% taller core than the HTTR. However the fuel temperature limit and reactor pressure vessel diameter are kept. In second stage, a new fuel that is currently under development at JAEA will allow the core outlet temperature to be raised to $900^{\circ}C$ for the purpose of demonstrating more efficient gas turbine power generation and high temperature heat supply. The third stage adds a demonstration of nuclear-heated hydrogen production by a thermochemical process. A licensing approach to coupling high temperature industrial process to nuclear reactor will be developed. The low initial risk and the high longer-term potential for performance expansion attract development of the HTR50S as a multipurpose industrial or distributed energy source.

Uncertainty quantification based on similarity analysis of reactor physics benchmark experiments for SFR using TRU metallic fuel

YuGwon Jo,Jaewoon Yoo,Jong-Hyuk Won,Jae-Yong Lim Korean Nuclear Society 2024 Nuclear Engineering and Technology Vol.56 No.9

One of the issues in the development of the sodium-cooled fast reactor (SFR) using transuranic (TRU) metallic fuel is the absence of criticality benchmark experiment that faithfully mocks up the nuclear characteristics of the target design for validation of the reactor core design code and its uncertainty quantification (UQ). This study aims to quantify the criticality uncertainty of a typical TRU burner with metallic fuel by using the standard upper safety limit (USL) estimation framework based on the similarity analysis of existing benchmark experiments but elaborated in two aspects:1) application of two-sided rather than one-sided tolerance interval and 2) inclusion of additional uncertainty to account for fission products and minor actinides not included in the benchmark experiments. To conduct the similarity analysis and evaluate the nuclear-data induced uncertainty, existing, well-verified computing codes were integrated, including the nuclear data sampling code SANDY, the nuclear data processing code NJOY, and the continuous-energy Monte Carlo code McCARD. Finally, using the SFR benchmark database comprising both publicly available and proprietary benchmark experiments, the criticality uncertainty of the TRU core model with metallic fuel was evaluated.

Path planning in nuclear facility decommissioning: Research status, challenges, and opportunities

Adibeli, Justina Onyinyechukwu,Liu, Yong-kuo,Ayodeji, Abiodun,Awodi, Ngbede Junior Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.11

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.

Review on sodium corrosion evolution of nuclear-grade 316 stainless steel for sodium-cooled fast reactor applications

Dai, Yaonan,Zheng, Xiaotao,Ding, Peishan Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.11

Sodium-cooled fast reactor (SFR) is the preferred technology of the generation-IV fast neutron reactor, and its core body mainly uses nuclear-grade 316 stainless steel. In order to prolong the design life of SFRs to 60 years and more, it is necessary to summarize and analyze the anti-corrosion effect of nuclear grade 316 stainless steel in high temperature sodium environment. The research on sodium corrosion of nuclear grade 316 stainless steel is mainly composed of several important factors, including the microstructure of stainless steel (ferrite layer, degradation layer, etc.), the trace chemical elements of stainless steel (Cr, Ni and Mo, etc) and liquid impurity elements in sodium (O, C and N, etc), carburization and mechanical properties of stainless steel, etc. Through summarizing and constructing the sodium corrosion rate equations of nuclear grade 316 stainless steel, the stainless steel loss of thickness can be predicted. By analyzing the effects of temperature, oxygen content in sodium and velocity of sodium on corrosion rate, the basis for establishing integrity evaluation standard of SFR core components with sodium corrosion is provided.

Ammonium uranate hydrate wet reconversion process for the production of nuclear-grade UO₂ powder from uranyl nitrate hexahydrate solution

Byungkuk Lee,Seungchul Yang,Dongyong Kwak,Hyunkwang Jo,Youngwoo Lee,Youngmoon Bae,Jayhyung Lee Korean Nuclear Society 2023 Nuclear Engineering and Technology Vol.55 No.6

The existing wet reconversion processes for the recovery of scraps generated in manufacturing of nuclear fuel are complex and require several unit operation steps. In this study, it is attempted to simplify the recovery process of high-quality fuel-grade UO₂ powder. A novel wet reconversion process for uranyl nitrate hexahydrate solution is suggested by using a newly developed pulsed fluidized bed reactor, and the resultant chemical characteristics are evaluated for the intermediate ammonium uranate hydrate product and subsequently converted UO₂ powder, as well as the compliance with nuclear fuel specifications and advantages over existing wet processes. The UO₂ powder obtained by the suggested process improved fuel pellet properties compared to those derived from the existing wet conversion processes. Powder performance tests revealed that the produced UO₂ powder satisfies all specifications required for fuel pellets, including the sintered density, increase in re-sintered density, and grain size. Therefore, the processes described herein can aid realizing a simplified manufacturing process for nuclear-grade UO₂ powders that can be used for nuclear power generation.

An inter-comparison between ENDF/B-VIII.0-NECP-Atlas and ENDF/B-VIII.0-NJOY results for criticality safety benchmarks and benchmarks on the reactivity temperature coefficient

Kabach, Ouadie,Chetaine, Abdelouahed,Benchrif, Abdelfettah,Amsil, Hamid Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.8

Since the nuclear data forms a vital component in reactor physics computations, the nuclear community needs processing codes as tools for translating the Evaluated Nuclear Data Files (ENDF) to simulate nuclear-related problems such as an ACE format that is used for MCNP. Errors, inaccuracies or discrepancies in library processing may lead to a calculation that disagrees with the experimentally measured benchmark. This paper provides an overview of the processing and preparation of ENDF/B-VIII.0 incident neutron data with NECP-Atlas and NJOY codes for implementation in the MCNP code. The resulting libraries are statistically inter-compared and tested by conducting benchmark calculations, as the mutualcomparison is a source of strong feedback for further improvements in processing procedures. The database of the benchmark experiments is based on a selection taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP handbook) and those proposed by Russell D. Mosteller. In general, there is quite good agreement between the NECP-Atlas1.2 and NJOY21^(1.0.0.json) results with no substantial differences, if the correct input parameters are used.

Corrigendum to "Li4SiO4 slurry conditions and sintering temperature for fabricating Li4SiO4 pebbles as tritium breeder for nuclear-fusion reactors" [Nucl. Eng. Technol. 55 (2023) 2966]

Young ah Park,Ji Won Yoo,Yi-Hyun Park,Young soo Yoon Korean Nuclear Society 2024 Nuclear Engineering and Technology Vol.56 No.5