Summary of thermal-hydraulic calculations for a pressurized water reactor [PDF]
, 1980 The results of two transients involving the loss of a steam generator in a single-pass, steam generator, pressurized water reactor have been analyzed using a state-of-the-art, thermal-hydraulic computer code. Computed results include the formation of a steam bubble in the core while the pressurizer is solid.
Bolstad, J.W., Haarman, R.A.
openaire +2 more sources
TRIGA-2000 Research Reactor Thermal-hydraulic Analysis using RELAP/SCDAPSIM/MOD3.4
International Journal of Technology, 2017 Any events presumed to risk the safety of a nuclear reactor should be analyzed. In a research reactor, the applicability of best estimate thermal-hydraulic codes has been assessed for safety analysis purposes.
Anhar Riza Antariksawan +4 more
doaj +1 more source
Post-test simulation of a PLOFA transient test in the CIRCE-HERO facility [PDF]
, 2019 CIRCE is a lead–bismuth eutectic alloy (LBE) pool facility aimed to simulate the primary system of a heavy liquid metal (HLM) cooled pool-type fast reactor.
Caruso G. +4 more
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Analysis of heat balance of piston hydraulic damper system installed on thrust bearing
Zhongguo Jianchuan Yanjiu, 2018 [Objectives] This paper aims to analyze the thermal equilibrium performance of a piston hydraulic damper in a stable working state so as to solve the problem in which the oil circuit of the hydraulic damping system of a longitudinal vibration-reduction ...
CHEN Fan +3 more
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SUBSALS: a subchannel thermal-hydraulic code for IRT type fuel analysis
Acta Polytechnica CTU Proceedings, 2022 The LVR-15 research reactor is operated with a tube type fuel IRT-4M. Due to fuel’s unique concentric square annular shape with, coolant flow is subject to significant pressure driven crossflow.
Tomáš Adámek
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Increasing the thermal efficiency of water heating boilers by improving design parameters [PDF]
E3S Web of Conferences, 2023 Currently, research is being conducted all over the world to create clearly defined operating, technological and design parameters that ensure the continuity of hydrodynamic and thermal processes, which will improve the energy efficiency of heat supply ...
Arifjanov Aybek, Kurbonov Kozim
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Code improvement and model validation for Asco-II Nuclear Power Plant model using a coupled 3D neutron kinetics/thermal-hydraulic code [PDF]
, 2016 This paper provides a Best Estimate validation calculation with a coupled thermal–hydraulic and 3D neutron kinetic model for Ascó-II Nuclear Power Plant. Common NRC codes have been used for its purpose.
Batet Miracle, Lluís +3 more
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Development and validation of reactor nuclear design code CORCA-3D
Nuclear Engineering and Technology, 2019 The advanced node core code CORCA-3D is one of the independent developed codes of NPIC for the nuclear reactor core design. CORCA-3D code can calculate the few-group cross section, solve the 3D diffusion equations, consider the thermal-hydraulic feedback,
Ping An +5 more
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Thermal-hydraulic analysis of light water reactors under different steady-state operating conditions, Part 1: Boiling water reactor [PDF]
Nuclear Technology and Radiation Protection, 2022 The steady-state thermal-hydraulic analysis of the core of the Boiling Water Reactor (BWR/6) at nominal operating conditions is presented in this paper. The BWR/6 is produced by General Electric USA.
Hutli Ezddin, Kridan Ramadan
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Advanced Exergy Analysis in the Dynamic Framework for Assessing Building Thermal Systems [PDF]
, 2019 This work applies the Dynamic Advanced Exergy Analysis (DAEA) to a heating and domestic hot water (DHW) facility supplied by a Stirling engine and a condensing boiler. For the first time, an advanced exergy analysis using dynamic conditions is applied to
Picallo-Perez, Ana +3 more
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