Results 11 to 20 of about 362 (143)

Advanced Presentation of BETHSY 6.2TC Test Results Calculated by RELAP5 and TRACE [PDF]

Science and Technology of Nuclear Installations, 2012
Today most software applications come with a graphical user interface, including U.S. Nuclear Regulatory Commission TRAC/RELAP Advanced Computational Engine (TRACE) best-estimate reactor system code.
Andrej Prošek, Ovidiu-Adrian Berar
doaj   +3 more sources

Study on the Density Wave Instability in Natural Circulation System under Rolling Conditions

International Journal of Energy Research, Volume 2023, Issue 1, 2023., 2023
In the ocean environment, the thermal fluid system installed on the ship moves with the waves. This unsteady transient motion imposes additional body forces on the liquid. Thereby, the thermohydraulic properties of the thermal fluid system are changed, especially for natural circulation systems with a low driving force.
Houjun Gong, Xiaoxi Zhang, Nan Gui, Yanping Huang, Xingtuan Yang, Shengyao Jiang, Rafael M. Santos   +6 more
wiley   +1 more source

Methods for Predicting the Minimum Temperature of the Outage Loop and the Maximum Power Caused by the Low‐Temperature Coolant

Science and Technology of Nuclear Installations, Volume 2023, Issue 1, 2023., 2023
When the feedwater valve at the outage loop of the floating nuclear power plant leaks, thermal stratification occurs in the steam generator. It causes lower water temperature in the outage loop. The extent of hazard of this phenomenon cannot be directly determined by the existing measurement parameters, which poses a threat to the operational safety of
Xinxin Liu, Lei Yu, Jianli Hao, Xiaolong Wang, Han Zhang   +4 more
wiley   +1 more source

ELSMOR European Project: Experimental Results on an Innovative Decay Heat Removal System Based on a Plate‐Type Heat Exchanger

Science and Technology of Nuclear Installations, Volume 2023, Issue 1, 2023., 2023
This paper summarises the results of an experimental campaign carried out at SIET on the ELSMOR facility built in 2022 to validate a decay heat removal system for the E‐SMR. Based on the passive mechanisms of natural circulation, the system aims to dissipate the reactor decay heat to a water pool, using two heat exchangers: a plate‐type one coupling ...
Roberta Ferri, Andrea Achilli, Cinzia Congiu, Stefano Marcianò, Stefano Gandolfi, Mattia Marengoni, Alberto Bersani, Alessandro Passerin D’Entreves, Doddy Kastanya   +8 more
wiley   +1 more source

Development and Assessment of the COBRA/RELAP5 Code. [PDF]

Journal of Nuclear Science and Technology, 1997
The COBRA/RELAP5 code has been developed to combine the realistic three-dimensional reactor vessel model of COBRA-TF with RELAP5/MOD3.2, thus to produce an advanced system analysis code with a multidimensional thermal-hydraulic module. This paper provides the integration scheme of the two codes. The results of developmental assesments are also provided,
JEONG, Jae-Jun, SIM, Suk Ku, LEE, Sang Yong   +2 more
openaire   +1 more source

Simulations of Core Damage Progression for TMI‐2 Severe Accident Using CINEMA Computer Code

Science and Technology of Nuclear Installations, Volume 2023, Issue 1, 2023., 2023
As an integrated computer code development for severe accident sequence analysis in Korea, CINEMA has been developing from an initiation event to a containment failure. The CINEMA computer code is composed of CSPACE, SACAP, and SIRIUS, which are capable of simulating core melt progression with thermal hydraulic analysis of the RCS (reactor coolant ...
Rae-Joon Park, Donggun Son, Jun Ho Bae, Sung Won Bae, Bub-Dong Chung, Kwang Soon Ha, Han Zhang   +6 more
wiley   +1 more source

Experimental study and safety analysis on the heating surfaces in the 660 MW supercritical CFB boiler under sudden electricity failure

Energy Science &Engineering, Volume 10, Issue 7, Page 2088-2105, July 2022., 2022
In this project, the necessity of equipping with an emergency water supply system in a supercritical circulating fluidized bed (CFB) boiler has long been a controversial topic in the industry. And a boiler electricity failure experiment was accomplished, a mathematical model established and a calculation program developed.
Yinlong Li, Beibei Xie, Lingfeng Bi, Haoyu Yang, Chao Nie, Hao Qing, Dong Yang   +6 more
wiley   +1 more source

The Concept of the Heat Removal System of a High‐Flux Research Reactor

Science and Technology of Nuclear Installations, Volume 2022, Issue 1, 2022., 2022
Achieving high neutron fluxes in research pressurized water reactors is directly related to the intensity of the coolant flow through the core and the pressure in it, which provides an increased saturation temperature and a margin to critical heat flux.
Vitaly Uzikov, Ildar Suleimanov, Irina Uzikova, Alejandro Clausse   +3 more
wiley   +1 more source

Development of an MPS Code for Corium Behavior Analysis: 3D Alloy Melting

Science and Technology of Nuclear Installations, Volume 2022, Issue 1, 2022., 2022
The moving particle semi‐implicit (MPS) method as a Lagrangian method is attracting increasing attention in severe accident analysis. In this paper, we developed an MPS code for the corium behavior analysis with several additional models added: an improved heat transfer model to improve the calculation between different materials, an enthalpy‐based ...
Lijun Jian, Peng Yu, Jie Pei, Xiao Zeng, Yidan Yuan, Leon Cizelj   +5 more
wiley   +1 more source

A Data‐Driven Fault Prediction Method for Nuclear Power Systems Based on End‐to‐End Deep Learning Framework

Science and Technology of Nuclear Installations, Volume 2022, Issue 1, 2022., 2022
With the increase in system complexity and operational performance requirements, nuclear energy systems are developing in the direction of intelligence and unmanned, which also requires a higher demand for its safety so that intelligent fault diagnosis and prediction have become a technology that nuclear power plants need to develop at present.
Lu Chao, Chunbing Wang, Shuai Chen, Qizhi Duan, Hongyun Xie, Afaque Shams   +5 more
wiley   +1 more source