Results 51 to 60 of about 5,597 (253)

Development and Verification of Activated Source Term Library Based on ENDF/B‐VIII.0

International Journal of Energy Research, Volume 2025, Issue 1, 2025.
Developed by the Oak Ridge National Laboratory (ORNL) in the United States, the ORIGEN code for computing irradiation and decay processes is one of the most widely used point depletion programs. However, the nuclear data that accompanies it, crucial for its application, has been measured and evaluated a long time ago.
Xirui Zhang, Dacai Zhang, Hao Luo, Wenxiang Fang, Te Sheng, Ganglin Yu, Chaofan Sun   +6 more
wiley   +1 more source

Fretting Wear and Fatigue Life Analysis of Fuel Bundles Subjected to Turbulent Axial Flow in CEFR

Science and Technology of Nuclear Installations, Volume 2019, Issue 1, 2019., 2019
In a fast spectrum reactor, the fuel rod bundle is mainly positioned radially by the wire which can make contact with the adjacent fuel rods, and then it is inevitable that flow‐induced vibration (FIV) will cause fretting wear and vibration fatigue of the fuel cladding at the contact position.
Yafeng Shu, Jianjun Wu, Yongwei Yang, Zelong Zhao, Arkady Serikov   +4 more
wiley   +1 more source

Recycling option search for a 600-MWe sodium-cooled transmutation fast reactor

Nuclear Engineering and Technology, 2015
Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER.
Yong Kyo Lee, Myung Hyun Kim
doaj   +1 more source

Large-eddy simulations of turbulent flow for grid-to-rod fretting in nuclear reactors

, 2013
The grid-to-rod fretting (GTRF) problem in pressurized water reactors is a flow-induced vibration problem that results in wear and failure of the fuel rods in nuclear assemblies.
Bakosi, J., Christon, M. A., Lowrie, R. B., Nourgaliev, R. R., Pritchett-Sheats, L. A.   +4 more
core   +1 more source

Analysis of carbon‐14 discharges from Korean nuclear power plants

Energy Science &Engineering, Volume 12, Issue 9, Page 3576-3588, September 2024.
Comparison of radioactive effluent discharges from Korean nuclear power plants before and after carbon‐14 discharge monitoring in Korean pressurized water reactors. Abstract The routine practice accompanying the operation of nuclear facilities involves the discharge of radioactive effluents from nuclear power plants (NPPs). Regulation of this discharge
Hwapyoung Kim, Tae Young Kong, Seongjun Kim, Jinho Son, Changju Song, Jiung Kim, Hee Geun Kim   +6 more
wiley   +1 more source

Analysis of Control Rod Drop Accidents for the Canadian SCWR Using Coupled 3-Dimensional Neutron Kinetics and Thermal Hydraulics

Science and Technology of Nuclear Installations, 2018
The Canadian Supercritical Water-cooled Reactor (SCWR), a GEN IV reactor design, is a hybrid design of the well-established CANDU™ and Boiling Water Reactor with water above its thermodynamic critical point.
Frederic Salaun, David R. Novog
doaj   +1 more source

A Lattice-Based Monte Carlo Evaluation of Canada Deuterium Uranium-6 Safety Parameters

Nuclear Engineering and Technology, 2016
Important safety parameters such as the fuel temperature coefficient (FTC) and the power coefficient of reactivity (PCR) of the CANada Deuterium Uranium (CANDU-6) reactor have been evaluated using the Monte Carlo method.
Yonghee Kim, Donny Hartanto, Woosong Kim
doaj   +1 more source

Critical heat flux in a CANDU end shield – Influence of shielding ball diameter

Nuclear Engineering and Technology, 2022
Experiments were performed to measure the critical heat flux (CHF) on a vertical surface abutting a coarse packed bed of spherical particles. This geometry is representative of a CANDU reactor calandria tubesheet facing the end shield cavity during the ...
Justin Spencer
doaj   +1 more source

Nickel‐based alloy corrosion in CANDU steam generators: E–pH diagrams of the Ni–NH3–H2O and Ni–CH3COO−–H2O ternary systems

Materials and Corrosion, Volume 75, Issue 4, Page 529-542, April 2024.
The simulated E–H diagrams of nickel could be used to predict the corrosion of steam generator tubes of Canadian Deuterium Uranium reactors. Abstract Nuclear power plant steam generator (SG) tubes contain nickel as the main alloying element and there is concern about their corrosion.
Mohammad Amin Razmjoo Khollari, Hamid Zebardast, Edouard Asselin   +2 more
wiley   +1 more source

Proposal and Validation of a Diagnosis Method of Fuel Failures in Pressurized Water Reactor Nuclear Power Plants

Science and Technology of Nuclear Installations, Volume 2024, Issue 1, 2024.
Based on research into the diagnosis methods for fuel failures in pressurized water reactor nuclear power plants and an analysis of existing problems, this paper proposes an enhanced fuel failure diagnosis method, addressing three key improvements: identification of the occurrence of fuel failure, diagnosis of the number of failed fuel rods, and the ...
Weifeng Lyu, Wenwang Ran, Runchun Guo, Yaoyi Gao, Jun Xiong, Arkady Serikov   +5 more
wiley   +1 more source