scholarly journals Research, engineering, and construction report, Engineering Analysis Division. Thermal analysis: LOFT primary coolant pump inlet nozzle, thermal analysis Class 1 review

1977 ◽  
Author(s):  
W.C. Kettenacker
Keyword(s):  
Thermal Analysis ◽  
Engineering Analysis ◽  
Primary Coolant ◽  
Coolant Pump ◽  
Class 1 ◽  
Pump Inlet

Thermal Analysis of the FSP-1 Experiment in the Advanced Test Reactor

2016 ◽  
Author(s):  
Grant L. Hawkes ◽  
Nicolas E. Woolstenhulme
Keyword(s):  
Thermal Analysis ◽  
Flow Instability ◽  
National Laboratory ◽  
Safety Evaluation ◽  
Thermal Safety ◽  
Primary Coolant ◽  
Coolant Pump ◽  
Safety Margins ◽  
Enriched Uranium ◽  
Test Reactor

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Three different types of fuel plates with matching pairs for a total of six plates were analyzed. These three types of plates are: full burn, intermediate power, and thick meat. A thermal analysis has been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A thermal safety evaluation was performed to demonstrate that the FSP-1 irradiation experiment complies with the thermal-hydraulic safety requirements of the ATR Safety Analysis Report (SAR). The ATR SAR requires that minimum safety margins to critical heat flux and flow instability be met in the case of a loss of commercial power with primary coolant pump coast-down to emergency flow. The thermal safety evaluation was performed at 26 MW NE lobe power to encompass the expected range of operating power during a standard cycle. Additional safety evaluations of reactivity insertion events, loss of coolant event, and free convection cooling in the reactor and in the canal are used to determine the response of the experiment to these events and conditions. This paper reports and shows that each safety evaluation complies with each safety requirement of the ATR SAR.

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.

Thermal Analysis of Canned Induction Motor for Coolant Pump Considering the Eddy Current Loss in Cans

2010 ◽  
Vol 670 ◽  
pp. 466-476 ◽  
Author(s):  
Jian Li ◽  
Jung Tae Song ◽  
Yun Hyun Cho
Keyword(s):  
Thermal Analysis ◽  
Finite Element ◽  
Induction Motor ◽  
Eddy Current ◽  
Three Dimensional ◽  
Eddy Current Loss ◽  
Time Step ◽  
Coolant Pump ◽  
Current Loss ◽  
Canned Motor

This paper describes thermal analysis of canned induction motor for coolant pump considering eddy current loss. The electromagnetic field of a canned motor was analyzed by using the time-step finite element method, and the eddy loss was obtained. Equivalent circuit considering can loss was developed and the equitation to calculate can loss was derived from theory of conventional motor. Using the loss from electromagnetic analysis as heat source of temperature field, thermal analysis was conducted by three dimensional finite element analyses. The simulation results show good agreement with experiment data, which indicates that this method has good accuracy and reliability for dealing with thermal behavior of canned motor.

Reactor Coolant Pump Flywheel Overspeed Evaluation

1974 ◽  
Vol 96 (4) ◽  
pp. 279-285 ◽  
Author(s):  
P. C. Riccardella ◽  
W. H. Bamford
Keyword(s):  
Fracture Mechanics ◽  
Ductile Failure ◽  
Light Water Reactor ◽  
Scale Model ◽  
Intensity Factors ◽  
Test Program ◽  
Primary Coolant ◽  
Coolant Pump ◽  
Reactor Primary Coolant ◽  
Experimental Effort

The overspeed capability of the large steel flywheels used on light water reactor primary coolant pumps has been evaluated through a combined analytical and experimental effort. Limiting speeds of the prototype flywheel design were calculated for the ductile failure mode using the principles of Section III of the ASME Boiler and Pressure Vessel Code, and for the brittle fracture mode using a fracture mechanics approach in which stress intensity factors were determined from finite element computer analysis. The accuracy of the analytical approach was verified by a scale model test program which demonstrates excellent agreement between experiment and analysis. The results of the evaluation are presented in this paper, and they illustrate the kinds of things which can be accomplished through application of modern fracture mechanics technology, including plasticity considerations, to the solution of hardware problems of real engineering interest.

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
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