scholarly journals LOFT primary coolant pump separate-effects tests

1975 ◽  
Author(s):  
F S Felicione
Keyword(s):  
Primary Coolant ◽  
Coolant Pump

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.

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.

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

Feasibility Study on the Natural Circulation Cooling Procedure at the Loop Unbalanced Condition

2017 ◽  
Author(s):  
Wataru Sakuma ◽  
Shinya Miyata ◽  
Manabu Maruyama ◽  
Junto Ogawa
Keyword(s):  
Large Scale ◽  
Natural Circulation ◽  
Test Facility ◽  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Scale Test ◽  
Safety Enhancement

In typical pressurized water reactor (PWR) plant, in case that one steam generator (SG) is dried out and cannot be credited for the primary cooldown, at least one reactor coolant pump (RCP) has to be operated in order to homogenize the primary coolant temperature distribution among loops when the plant is cooled down to the cold shutdown state. For example, an accident such as steam line break (SLB) and feedwater line break (FLB) leads to this situation. If the natural circulation condition is established due to unavailability of all the RCPs, the natural circulation in the primary loop connected to the affected SG would be interrupted in the plant cooldown phase. In this situation, the continuous cooldown disturbs the smooth depressurization because it leads to void generation at the top of the affected SG tube where the high temperature coolant is left. In addition, there is a possibility that all RCPs cannot be operated in case of the earthquake or the fire if the RCPs are not earthquake-proof and fire-resistant. Therefore the establishment of the cooldown procedure without RCPs operation under the temperature unbalanced condition among the primary loops can contribute to the safety enhancement for typical PWR plants. The several experiments have been already performed to observe the natural circulation phenomena under the temperature unbalanced condition. It has been reported that the plant can be continuously cooled down with smooth depressurization by stepwise cooling manner using MSRVs of the intact SGs. In this study, Mitsubishi Heavy Industries, Ltd. (MHI) performed the transient analyses to simulate the natural circulation cooldown test under the temperature unbalanced condition among loops performed by Large Scale Test Facility (JAEA ROSA/LSTF) using M-RELAP5, which was a modified plant system transient code by MHI based on RELAP5-3D. Based on the analysis results, the thermal hydraulic phenomena of natural circulation cooldown under the temperature unbalanced condition were investigated. As a result, the mechanism of natural circulation interruption was clarified, and this paper shows the outline of the cooldown procedure under the temperature unbalanced condition which could be applied to the PWR plants.

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
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