Recapturing Net Positive Suction Head Margins in Boiling Water Reactor Emergency Core Cooling Systems

2017 ◽  
Author(s):  
Alan J. Bilanin ◽  
Andrew E. Kaufman ◽  
Warren J. Bilanin
Keyword(s):  
Cooling System ◽  
Cooling Water ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Cooling Systems ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling ◽  
Reactor Pressure

Boiling Water Reactor pressure suppression pools have stringent housekeeping requirements, as well as restrictions on amounts and types of insulation and debris that can be present in the containment, to guarantee that suction strainers that allow cooling water to be supplied to the reactor during a Loss of Coolant Accident remain operational. By introducing “good debris” into the cooling water, many of these requirements/restrictions can be relaxed without sacrificing operational readiness of the cooling system.

Large-Break Loss-of-Coolant Accident Testing and Simulation for 200-MWe Simplified Boiling Water Reactor

2002 ◽  
Author(s):  
S. T. Revankar ◽  
Y. Xu ◽  
H. J. Yoon ◽  
M. Ishii
Keyword(s):  
Safety System ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Integral Test ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Main Steam ◽  
Core Cooling ◽  
Safety Systems

The performance of the safety systems of a new design of the 200-MWe simplified boiling water reactor during a large-break, loss-of-coolant accident transient was investigated through code modeling and integral system testing. The accident considered was a break in the main steam line which is the major design basis accident. RELAP5/MOD3 best estimate reactor thermalhydraulic code was used and its applicability to the reactor safety system evaluation was examined. The integral tests were perfomed to assess the safety systems and the response of the emergency core cooling systems to accident conditions in a scaled facility called PUMA. The details of the safety system behavior are presented. The integral test simulations examined code applicability at the scaled facility level as well as prototype key safety system performance.

Bounding Acoustic Load Development for Boiling Water Reactor Shroud and Shroud Supports

2020 ◽  
Author(s):  
Matthew Walter ◽  
Minghao Qin ◽  
Daniel Sommerville
Keyword(s):  
Pressure Vessel ◽  
Reactor Core ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Structural Evaluation ◽  
Loss Of Coolant Accident ◽  
Specific Analysis ◽  
Reactor Pressure

Abstract As part of the license basis of a nuclear boiling water reactor pressure vessel, a sudden loss of coolant accident (LOCA) event needs to be analyzed. One of the loads that results from this event is a sudden depressurization of the recirculation line. This leads to an acoustic wave that propagates through the reactor coolant and impacts several structures inside the reactor pressure vessel (RPV). The authors have previously published a PVP paper (PVP2015-45769) which provides a survey of LOCA acoustic loads on boiling water reactor core shrouds. Acoustic loads are required for structural evaluation of core shrouds; therefore, a defensible load is required. The previous research compiled plant-specific data that was available at the time. Since then, additional data has become available which will add to the robustness of the bounding load methodology that was developed. Investigations are also made regarding the shroud support to RPV weld, which was neglected from the previous study. This will allow a practitioner a convenient method to calculate bounding acoustic loads on all shroud and shroud support welds in the absence of a plant-specific analysis.

scholarly journals Simulation Experiment of Five Percent Small Break Loss-of-Coolant Accident of Boiling Water Reactor

1983 ◽  
Vol 20 (2) ◽  
pp. 89-104 ◽  
Author(s):  
Kanji TASAKA ◽  
Yasuo KOIZUMI ◽  
Yoshinari ANODA ◽  
Hiroshige KUMAMARU ◽  
Masayoshi SHIBA
Keyword(s):  
Simulation Experiment ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

ROSA-III Double-Ended Break Test Series for a Loss-of-Coolant Accident in a Boiling Water Reactor

1985 ◽  
Vol 68 (1) ◽  
pp. 77-93 ◽  
Author(s):  
K. Tasaka ◽  
M. Suzuki ◽  
Y. Anoda ◽  
Y. Koizumi ◽  
T. Yonomoto ◽  
...  
Keyword(s):  
Boiling Water ◽  
Test Series ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Boiling Water Reactor Jet Pump Acoustic Loads Resulting From a Recirculation Outlet Line Break Loss of Coolant Accident: Some Sensitivity Studies

2010 ◽  
Author(s):  
Daniel V. Sommerville ◽  
Raju Ananth
Keyword(s):  
Boiling Water ◽  
Boiling Water Reactor ◽  
Jet Pump ◽  
Element Analysis ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Acoustic Load ◽  
Jet Pumps ◽  
Sensitivity Studies

Acoustic loads caused by a Recirculation Line Break Loss of Coolant Accident are a required design basis event that must be considered for stress analyses of Boiling Water Reactor internal components such as Jet Pumps. This event causes acoustic and fluid loads on BWR internals. These loads must also be considered for fracture mechanics evaluations performed to assess allowable operating periods for flaws detected during in-service inspections. Acoustic loads methods generally utilized in the past have been 1-D or simplified 2-D models of the domain of interest. These models generally do not enable accurate assessment of the variation of acoustic loading on jet pumps away from the break plane. Previous methods conservatively apply the acoustic loading predicted for the jet pump nearest the break for all jet pumps. Insight can be gained and lower loading may be justifiable, for jet pumps away from the break, using methods that enable accurate acoustic load predictions for all jet pumps in the BWR annulus. This paper presents the results of Recirculation Line Break acoustic loads analyses of a typical BWR and investigates the variation of acoustic loading between all jet pumps in the annulus. The paper also presents the results of preliminary sensitivity studies performed to identify which geometric characteristics of the BWR annulus have the most significant effect on the resulting acoustic load predictions. The analyses documented in this paper are performed using acoustic finite element analysis.

Design and Testing of Vacuum Breaker Check Valve for Simplified Boiling Water Reactor

2002 ◽  
Author(s):  
M. Ishii ◽  
Y. Xu ◽  
S. T. Revankar
Keyword(s):  
Check Valve ◽  
Mechanical Valve ◽  
Reactor System ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Design Calculations ◽  
Design And Testing

A new design of the vacuum breaker check valve was developed to replace the mechanical valve in a simplified boiling water reactor. Scaling and design calculations were performed to obtain the geometry of new passive hydraulic vacuum breaker check valve. In order to check the valve performance, a RELAP5 model of the simplified boiling water reactor system with the new valve was developed. The valve was implemented in an integral facility, PUMA and was tested for large break loss of coolant accident.

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