Flow Transient in Primary Coolant System During Reactor Coolant Pump Start-Up Period

2013 ◽  
Author(s):  
Hong Gao ◽  
Feng Gao ◽  
Xianchao Zhao ◽  
Jie Chen ◽  
Xuewu Cao
Keyword(s):  
Primary Coolant ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Start Up ◽  
Coolant System

Analysis of reactor coolant pump transient performance in primary coolant system during start-up period

2013 ◽  
Vol 54 ◽  
pp. 202-208 ◽  
Author(s):  
Hong Gao ◽  
Feng Gao ◽  
Xianchao Zhao ◽  
Jie Chen ◽  
Xuewu Cao
Keyword(s):  
Transient Performance ◽  
Primary Coolant ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Start Up ◽  
Coolant System

Research on Modular Modeling Method of Reactor Coolant System Based on THEATRe

2014 ◽  
Author(s):  
Bo Shi ◽  
Zhao-Fei Tian
Keyword(s):  
Real Time ◽  
Data Exchange ◽  
Reactor Core ◽  
Simulation Software ◽  
Modeling Method ◽  
Time Step ◽  
Coolant Pump ◽  
Modular Modeling ◽  
Reactor Coolant ◽  
Coolant System

At present, research on the reactor coolant system is less yet, though modular modeling method has been widely used in the second-loop system of reactor. This paper takes the reactor coolant system of Qinshan-1 nuclear power plant as the object of study, analyses and researches on modular modeling method of reactor coolant system based on THEATRe, which is a large Thermal-Hydraulic real time simulation software developed by GSE Company and adopts NMNP (Nodal Momentum Nodal Pressure) solving method. This research establishes the modular model of the reactor coolant system equipments (including reactor core, main coolant pump, pressurizer, steam generator) using the THEATRe code. Due to each module is wrote into through different input cards, they can be solved by using their own matrix of velocity-pressure to guarantee the independence of the numerical calculation for different modular modules. THEATRe code does not have its own TDV like relap-5, meanwhile it also needs to ensure the pressurizer module can play a role in the multi-pressure node system. So this paper modifies solving method of the THEATRe source code to get suitable pressure boundary and flux boundary for RCS equipment modular module, and selects reasonable time step and data exchange frequency to achieve the data exchange of boundary pressure, flux and enthalpy among the equipment modules, which lays the foundation of establishing the real-time modular simulation model of the reactor coolant system in the future.

scholarly journals Decontamination of a Contaminated RCP Shaft Using the SP-HyBRID Process

Processes ◽  
2021 ◽  
Vol 9 (10) ◽  
pp. 1725
Author(s):  
Hee-Chul Eun ◽  
Na-On Chang ◽  
Wang-Kyu Choi ◽  
Sang-Yoon Park ◽  
Seon-Byeong Kim ◽  
...  
Keyword(s):  
Decontamination Factor ◽  
Hybrid Process ◽  
Waste Generation ◽  
Nuclear Facility ◽  
Experimental Conditions ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Pump Shaft ◽  
Shaft Surface ◽  
Coolant System

It is very important to minimize the waste generation for decontamination of the reactor coolant system in a nuclear facility. As an alternative to commercial decontamination technologies, an inorganic acid chemical decontamination (SP-HyBRID) process can be effectively applied to the decontamination because it can significantly reduce the waste generation. In this study, the decontamination of a contaminated reactor coolant pump shaft from a nuclear facility was conducted using the SP-HyBRID process. First, equipment for a mock-up test of the decontamination was prepared. Detailed experimental conditions for the decontamination were determined through the mock-up test. Under the detailed conditions, the contaminated shaft was successfully decontaminated. The dose rate on the shaft surface was greatly reduced from 1400 to 0.9 μSv/h, and the decontamination factor showed a very high value (>1500).

Benchmark Calculations of Pressurizer Model for Maanshan Nuclear Power Plant Using TRACE Code

2008 ◽  
Author(s):  
Yi-Hsiang Cheng ◽  
Chunkuan Shih ◽  
Jong-Rong Wang ◽  
Hao-Tzu Lin
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Start Up ◽  
Simulation Results ◽  
Net Load ◽  
Coolant System

Pressurizer plays an important role in controlling the pressure of the primary coolant system in pressurized water reactor (PWR) power plants. An accurate modelling of the pressurizer is needed to determine the pressure histories of the primary coolant system, and thus to successfully simulate overall PWR power plant behavior during transients. The purpose of this study is to develop a pressurizer model, and to assess its pressure transients using the TRACE code version 5.0. The benchmark of the pressurizer model was performed by comparing the simulation results with those from the tests at the Maanshan nuclear power plant. Four start-up tests of the Maanshan nuclear power plant are collected and simulated: 1) turbine trip test from 100% power; 2) large-load reduction at 100% power; 3) net-load trip at 100% power; and 4) net-load trip at 50% power. The simulation results are in reasonable agreement with the start-up tests, and thus the pressurizer model built in this study is successfully verified and validated.

Numerical and Experimental Research on the Fluid-Induced Forces of Clearance Flow in Canned Motor Reactor Coolant Pump

2019 ◽  
Vol 141 (6) ◽  
Author(s):  
Rui Xu ◽  
Yun Long ◽  
Yaoyu Hu ◽  
Junlian Yin ◽  
Dezhong Wang
Keyword(s):  
Large Mass ◽  
Stiffness Coefficient ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Various Boundary Conditions ◽  
Clearance Flow ◽  
Mass Coefficient ◽  
Cfd Method ◽  
Canned Motor

Reactor coolant pump (RCP) is one of the most important equipment of the coolant loop in a pressurized water reactor system. Its safety relies on the characteristics of the rotordynamic system. For a canned motor RCP, the liquid coolant fills up the clearance between the metal shields of the rotor and stator inside the canned motor, forming a long clearance flow. The fluid-induced forces of the clearance flow in canned motor RCP and their effects on the rotordynamic characteristics of the pump are numerically and experimentally analyzed in this work. A transient computational fluid dynamics (CFD) method has been used to investigate the fluid-induced force of the clearance. A vertical experiment rig has also been established for the purpose of measuring the fluid-induced forces. Fluid-induced forces of clearance flow with various whirl frequencies and various boundary conditions are obtained through the CFD method and the experiment. Results show that clearance flow brings large mass coefficient into the rotordynamic system and the direct stiffness coefficient is negative under the normal operating condition. The rotordynamic stability of canned motor RCP does not deteriorate despite the existence of significant cross-coupled stiffness coefficient from the fluid-induced forces of the clearance flow.

Stand-Alone Containment Analysis of the PHÉBUS FPT-1 Test With the ASTEC and the MELCOR Codes

2016 ◽  
Author(s):  
Bruno Gonfiotti ◽  
Sandro Paci
Keyword(s):  
State Of The Art ◽  
External Environment ◽  
Fission Products ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Complex Task ◽  
Primary Coolant ◽  
Control Volumes ◽  
Coolant System

The estimation of Fission Products (FPs) release from the containment system of a nuclear plant to the external environment during a Severe Accident (SA) is a quite complex task. In the last 30–40 years several efforts were made to understand and to investigate the different phenomena occurring in such a kind of accidents in the primary coolant system and in the containment. These researches moved along two tracks: understanding of involved phenomenologies through the execution of different experiments, and creation of numerical codes capable to simulate such phenomena. These codes are continuously developed to reflect the actual SA state-of-the-art, but it is necessary to continuously check that modifications and improvements are able to increase the quality of the obtained results. For this purpose, a continuous verification and validation work should be carried out. Therefore, the aim of the present work is to re-analyze the Phébus FPT-1 test employing the ASTEC (F) and MELCOR (USA) codes. The analysis focuses on the stand-alone containment aspects of the test, and three different modellisations of the containment vessel have been developed showing that at least 15/20 Control Volumes (CVs) are necessary for the spatial schematization to correctly predict thermal-hydraulics and the aerosol behavior. Furthermore, the paper summarizes the main thermal-hydraulic results, and presents different sensitivity analyses carried out on the aerosols and FPs behavior.

Sign in / Sign up

Export Citation Format

Share Document