Transient simulation code development of primary coolant system of Chinese Experimental Fast Reactor

2013 ◽  
Vol 53 ◽  
pp. 158-169 ◽  
Author(s):  
Manman Cui ◽  
Yun Guo ◽  
Zhijian Zhang
Keyword(s):  
Fast Reactor ◽  
Transient Simulation ◽  
Primary Coolant ◽  
Simulation Code ◽  
Coolant System ◽  
Code Development

Analysis code development for the direct reactor auxiliary cooling system of the pool-type sodium-cooled fast reactor

Kerntechnik ◽  
2018 ◽  
Vol 83 (3) ◽  
pp. 232-236 ◽  
Author(s):  
D. L. Zhang ◽  
P. Song ◽  
S. Wang ◽  
X. Wang ◽  
J. Chen ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Cooling System ◽  
Pool Type ◽  
Code Development

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.

Thermal Fatigue of Pwr Coolant System Loop Drain Lines Due to Outflow Activities

2021 ◽  
Author(s):  
Yue Zou ◽  
Brian Derreberry
Keyword(s):  
Thermal Fatigue ◽  
Operating Experience ◽  
Fatigue Cracking ◽  
Screening Methods ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Thermal Transients ◽  
Piping Systems ◽  
Class 1 ◽  
Coolant System

Abstract Thermal cycling induced fatigue is widely recognized as one of the major contributors to the damage of nuclear plant piping systems, especially at locations where turbulent mixing of flows with different temperature occurs. Thermal fatigue caused by swirl penetration interaction with normally stagnant water layers has been identified as a mechanism that can lead to cracking in dead-ended branch lines attached to pressurized water reactor (PWR) primary coolant system. EPRI has developed screening methods, derived from extensive testing and analysis, to determine which lines are potentially affected as well as evaluation methods to perform evaluations of this thermal fatigue mechanism for the U.S. PWR plants. However, recent industry operating experience (OE) indicate that the model used to predict thermal fatigue due to swirl penetration is not fully understood. In addition, cumulative effects from other thermal transients, such as outflow activities, may also contribute to the failure of the RCS branch lines. In this paper, we report direct OE from one of our PWR units where thermal fatigue cracking is observed at the RCS loop drain line close to the welded region of the elbow. A conservative analytical approach that takes into account the influence of thermal stratification, in accordance with ASME Class 1 piping stress method, is also proposed to evaluate the severity of fatigue damage to the RCS drain line, as a result of transients from outflow activities. Finally, recommendations are made for future operation and inspection based on results of the evaluation.

Modeling of Porous Crud Layer Based on Effects of Diffusion and Water Radiolysis on Anomalous Porous Crud Deposition on Fuel Pin Surfaces in PWRs

2002 ◽  
Author(s):  
Hyun-Jong Joe ◽  
Barclay G. Jones
Keyword(s):  
Negative Impact ◽  
Nucleate Boiling ◽  
Water Radiolysis ◽  
Trace Impurities ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Fuel Pin ◽  
Negative Effect ◽  
Water Reactors ◽  
Coolant System

Many studies have been undertaken to understand crud formation on the upper spans of fuel pin clad surfaces, which is called axial offset anomaly (AOA), is observed in pressurized water reactors (PWR) as a result of sub-cooled nucleate boiling. Separately, researchers have considered the effect of water radiolysis in the primary coolant of PWR. This study examines the effects of radiolysis of liquid water, which aggressively participate in general cladding corrosion and solutes within the primary coolant system, in the terms of pH, temperature, and Linear Energy Transfer (LET). It also discusses the effect of mass transfer, especially diffusion, on the concentration distribution of the radiolytic products, H2 and O2, in the porous crud layer. Finally it covers the effects of chemical reactions of boric acid (H3BO3), which has a negative impact on the mechanisms of water recombination with hydrogen, lithium hydroxide (LiOH), which has a negative effect on water decomposition, dissolved hydrogen (DH), and some trace impurities.

A Study and Development of Primary Circuit Simulation Code for China Experimental Fast Reactor

2013 ◽  
Author(s):  
Jin Wang ◽  
Donghui Zhang ◽  
Wenjun Hu ◽  
Lixia Ren
Keyword(s):  
Fast Reactor ◽  
Nuclear Energy ◽  
Circuit Simulation ◽  
Thermal Power ◽  
Physical Models ◽  
Primary Circuit ◽  
Cooling Channel ◽  
Simulation Code ◽  
Wide Range ◽  
China Experimental Fast Reactor

A fast reactor is one of recommended candidates of Generation IV nuclear energy systems, which would meet wide requirements such as sustainability, safety and economics for nuclear energy development. To be the China’s first fast reactor, China Experimental Fast Reactor (CEFR) typical technical options are following: 65 MW thermal power and 20 MW electric power, three circuits of sodium-sodium-water, integrated pool type structure for the primary circuit. To establish modular simulation system for sodium fast reactor, the code which simulated the thermal-hydraulic behavior of primary circuit was developed. The physical models include reactor core, reactor vessel cooling channel, pumps, protection vessel, intermediate heat exchangers, ionization chamber cooling channel, cold sodium pool, hot sodium pool, inlet plenum, and pipes, etc. The code could compute coolant pressures, flow rates, and temperatures in the primary circuit. This module was designed for analysis of a wide range of transients. Although based on CEFR, it can treat an arbitrary arrangement of components.

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