Results 41 to 50 of about 176 (127)

Comparison of Blind and Open Calculation Results for Top-Slot Break LOCA in Fourth ATLAS Domestic Standard Problem

Energies, 2022
The advanced thermal-hydraulic test loop for accident simulation (ATLAS) was developed and operated at The Korea Atomic Energy Research Institute. The ATLAS is operated to simulate accidents in pressurized water reactors (PWRs).
Jongrok Kim, Byoung-Uhn Bae, Yusun Park, Jae Bong Lee, Nam-Hyun Choi, Seok Cho, Kyoung-Ho Kang   +6 more
doaj   +1 more source

Probabilistic Failure Analysis of SiC‐CMC/Nb Fuel Claddings Under Loss‐Of‐Coolant Accident Conditions

International Journal of Energy Research, Volume 2026, Issue 1, 2026.
A numerical framework leveraging finite element analysis is established to examine the thermomechanical behavior and burst safety of a novel accident‐tolerant fuel (ATF) cladding design. This design incorporates a thin niobium (Nb) liner positioned between two layers of silicon carbide ceramic matrix composites (SiC‐CMC).
Jabir Ubaid, Maithah Alaleeli, Saeed A. Alameri, Andreas Schiffer, Rajashik Paul   +4 more
wiley   +1 more source

Transmuting the Arid: Fission‐Powered Visions—Unlocking Ghana’s Aquatic Destiny Through Nuclear Desalination

Science and Technology of Nuclear Installations, Volume 2026, Issue 1, 2026.
Water scarcity poses a mounting constraint on Ghana’s socioeconomic development, intensified by climate‐driven hydrological variability, rapid urbanization, and industrial expansion. This study presents a Ghana‐specific techno‐economic and policy‐integrated assessment of nuclear‐powered seawater desalination, moving beyond generic applications of ...
Felix Ameyaw, Etornam Akaho, Seth Debrah, Doddy Kastanya   +3 more
wiley   +1 more source

Burst Safety Analysis of Strengthened FeCrAl Claddings in APR1400 Fuel Rods Subject to LOCA Conditions

Energy Science &Engineering, Volume 13, Issue 2, Page 728-741, February 2025.
A computational framework is developed to evaluate the thermo‐mechanical performance of APR1400 fuel rods with FeCrAl claddings under normal and LOCA conditions. Results show that reducing the pellet‐cladding gap enhances heat conduction and burst safety, while an 80% strength enhancement in FeCrAl claddings improves burst safety by up to 80%.
Jabir Ubaid, Maithah Alaleeli, Yongsun Yi, Saeed A. Alameri, Andreas Schiffer   +4 more
wiley   +1 more source

Development of advanced containment thermal-hydraulic analysis methodology based on three-field model codes for APR1400

Nuclear Engineering and Technology
Advanced ME (Mass and Energy) release analysis methodology and PT (Pressure and Temperature) analysis methodology were developed using advanced thermal-hydraulic computer codes of three-field model for the Gen-III reactor, APR1400.
Seong-ho Jee, Yong-Ju Cho, Kum-Ho Han, Jisu Kim, Sung-yong Kim, Je-woo Cho, Jehee Lee, Yeon-Jun Choo, Jeong-ju Kim, Chan-eok Park, Seok-jeong Park, Sun-Hong Yoon, Soon-Joon Hong   +12 more
doaj   +1 more source

Estimating Allowable Operator Action Time for Mitigating Multiple‐Failure Accidents in Korean Advanced Power Reactor

International Journal of Energy Research, Volume 2024, Issue 1, 2024.
In the case of multiple‐failure accidents in light water reactors, it is imperative to demonstrate that a significant fuel damage can be prevented via prearranged mitigation measures performed by operators. However, the time elapsed from the initiation of the event to the implementation of mitigation actions significantly influences the progress and ...
Jia Yu, Han Sol Park, Yeon-Gun Lee, Geng Chen   +3 more
wiley   +1 more source

A Concept of Online Refueling TRISO‐Fueled and Salt‐Cooled Reactor

International Journal of Energy Research, Volume 2024, Issue 1, 2024.
This paper introduces a novel concept of a TRISO‐fueled salt‐cooled reactor (TFSCR). The core contains circulating pipes filled with molten salt carrying TRISO particles. The reactor achieves online refueling by slowly circulating the molten salt through the pipeline.
Xiaoyong Feng, Hyun Chul Lee, Ozguer Omer Ersin   +2 more
wiley   +1 more source

Steam Condensation Heat Transfer in the Presence of Noncondensable Gases (NCGs) in Nuclear Power Plants (NPPs): A Comprehensive Review of Fundamentals, Current Status, and Prospects for Future Research

International Journal of Energy Research, Volume 2024, Issue 1, 2024.
Efficient steam condensation is crucial for safe nuclear power plant (NPP) operations, preventing pressure buildup, overheating, and the release of radioactive materials. However, the presence of noncondensable gases (NCGs), such as air, nitrogen, hydrogen, or helium, can hinder the condensation process by creating a thermal resistance layer that ...
Samah A. Albdour, Yacine Addad, Nourah Alyammahi, Imran Afgan, Amin Shokrollahi   +4 more
wiley   +1 more source

Verification and Validation of CFD analysis for 1/7th scale model of APR1400 for CVAP

Nuclear Engineering and Technology
This paper presents a methodology for analyzing the internal flow of the APR1400 Reactor Vessel Internals (RVI) using Computational Fluid Dynamics (CFD).
Giwon La, Kunwoo Yi, Doyoung Ko, Kyuhyung Kim   +3 more
doaj   +1 more source

Design Change and Operational Consideration of the HVAC System during Nuclear Power Plant Decommissioning

Science and Technology of Nuclear Installations, Volume 2024, Issue 1, 2024.
The heating, ventilation, and air conditioning (HVAC) system plays a crucial role in ensuring the safety of workers and preventing the release of gaseous radioactive materials into the environment during the decommissioning of a nuclear power plant (NPP). To establish an HVAC operation strategy, decommissioning phases were divided into four stages, and
Ho-Jin Jeon, Chang-Lak Kim, Raffaella Testoni   +2 more
wiley   +1 more source