scholarly journals Safety Design Strategy for the Advanced Test Reactor Primary Coolant Pump and Motor Replacement Project

2011 ◽  
Author(s):  
Noel Duckwitz
Keyword(s):  
Design Strategy ◽  
Primary Coolant ◽  
Coolant Pump ◽  
Safety Design ◽  
Test Reactor ◽  
Reactor Primary Coolant

Safety Design Strategy for the Versatile Test Reactor

2019 ◽  
Author(s):  
J. Andrus ◽  
M. Bucknor ◽  
D. Gerstner ◽  
D. Grabaskas
Keyword(s):  
Design Strategy ◽  
Safety Design ◽  
Test Reactor

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.

Thermal Safety Margin Calculation of the MP-2 Experiment in the Advanced Test Reactor

2020 ◽  
Author(s):  
Grant L. Hawkes
Keyword(s):  
Finite Element ◽  
Flow Instability ◽  
National Laboratory ◽  
Thermal Analyses ◽  
Safety Margin ◽  
Heat Rate ◽  
Element Model ◽  
Nucleate Boiling ◽  
Primary Coolant ◽  
Test Reactor

Abstract The Mini-Plate 2 (MP-2) irradiation test is a fueled experiment designed for irradiation in multiple test locations in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). The experiment is a drop-in test where small aluminum-clad fuel plate samples (mini plates) are cooled directly by the ATR Primary Coolant System (PCS) water. The MP-2 fuel plate experiment will be irradiated in several different irradiation locations of the ATR. This fueled experiment contains aluminum-clad fuel mini plates consisting of monolithic U-Mo. Four different types of fuel plates with fuel meat thickness and cladding are part of the MP-2 test. A thermal analysis has been performed on the MP-2 experiment. A method for calculating Departure from Nucleate Boiling Ratio (DNBR) and Flow Instability Ratio (FIR) during a reactivity transient using the commercial finite element and heat transfer code ABAQUS is discussed. At the start of an ATR cycle the heat generation rate of the fueled experiment is high and the heat rate multiplier from the outer shim control cylinders is low, while the reverse is true at the end of the ATR cycle. Thermal analyses at 10-day increments during the cycle calculate the DNBR and FIR during a reactivity transient. This technique calculates DNBR for the fuel plate surfaces and FIR for all water components for each finite element surface and node at various times during the ATR cycle. Heat rates vary with time during the transient calculations that are provided by a detailed physics analysis. Oxide growth on the fuel plates is also incorporated. Results from the transient calculations are displayed with the ABAQUS post processor. By calculating these parameters at each location in the finite element model, conservatism is replaced with accuracy. This allows for a greater margin for the thermal hydraulic safety parameters.

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