Disital calculation of transient performance of the primary coolant system in a water reactor

1959 ◽  
Vol 77 (6) ◽  
pp. 1036-1043
Author(s):  
D. G. Lewis
Keyword(s):  
Transient Performance ◽  
Primary Coolant ◽  
Water Reactor ◽  
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

ASME Subcommittee XI Initiatives for the Detection and Correction of Boric Acid Corrosion of Pressure Retaining Low Alloy Steel Components in Pressurized Water Reactor Plants

2005 ◽  
Author(s):  
Jerry McClanahan
Keyword(s):  
Boric Acid ◽  
Alloy Steel ◽  
Nuclear Power ◽  
Pressure Vessels ◽  
Low Alloy Steel ◽  
Pressurized Water Reactor ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Water Reactor ◽  
Coolant System

In pressurized water reactor power plants sodium pentaborate is commonly used as a soluble chemical neutron absorber to control the reactivity level within the reactor core. This chemical solution of boron in the reactor primary cooling water forms a mild boric acid. The main reactor coolant system pressure vessels, i.e., the nuclear reactor and the steam generators, are protected internally from this corrosive solution by a corrosion resistant cladding. The nuclear power industry has recorded instances of leakage from the primary coolant system resulting in corrosion of these pressure vessels on their outer surfaces. In 2002 a primary coolant leak resulted in the nearly complete loss of the pressure retaining capability of the reactor vessel closure head at a domestic nuclear power plant. In response to this event the ASME Subcommittee XI has undertaken to establish inspection rules for pressurized water reactor plants for the detection of primary coolant leakage and the resulting corrosion of low alloy steel components. This paper discusses the history of this issue and the development of inspection rules designed to ensure the early detection of primary coolant system leakage and the possible resulting corrosion prior to loss of the pressure retaining capability of primary coolant system components.

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.

Development of a Thermal Fatigue Screening and Evaluation Methodology for Pressurized Water Reactor Plants

2004 ◽  
Author(s):  
J. D. Keller ◽  
A. J. Bilanin ◽  
S. T. Rosinski
Keyword(s):  
Thermal Cycling ◽  
Development Program ◽  
Fatigue Cracking ◽  
Evaluation Methodology ◽  
Experimental Investigations ◽  
Pressurized Water Reactor ◽  
Primary Coolant ◽  
Modeling Tools ◽  
Pressurized Water ◽  
Water Reactor

Thermal cycling has been identified as a mechanism that can potentially lead to fatigue cracking in un-isolable branch lines attached to pressurized water reactor (PWR) primary coolant piping. A significant research and development program has been undertaken to understand the mechanisms causing thermal cycling and to develop models for predicting the thermal-hydraulic boundary conditions for use in piping structural and fatigue analysis. A combination of first-principles engineering modeling and scaled experimental investigations has been used to formulate improved thermal cycling modeling tools. This paper will provide an overview of the model development program, a summary of the supporting test program, and a description of the thermal cycling model structure. Benchmarking of the thermal cycling model against several PWR plant configurations is presented, demonstrating favorable comparison with cases where thermal stratification and cycling has been previously observed.

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.

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