Mechanistic model for predicting two-phase behavior in CANDU reactor pumps
Conference
·
· Trans. Am. Nucl. Soc.; (United States)
OSTI ID:5746222
The behavior of the primary coolant pumps driven in two-phase flow plays an important role in determining the magnitude and direction of the core flow. This critically dominates the core heat transfer mechanisms. Loss-of-coolant accident (LOCA) conditions and a dryout in the reactor core may force steam to be carried through the pumps. Therefore, the flow through the pump under these conditions cannot be treated as a single-phase flow. Review of the existing thermohydraulic transient codes and pump models revealed that there are instability problems associated with the pump modeling. This is due to the methodology of modeling, the pump interaction with primary system fluid, and particularly to the two-phase flow degradation in the pump. Therefore, the need arises to develop an improved, reliably accurate model to describe the interaction between primary system fluid and the centrifugal pump during severe transients in CANDU reactors. It is strongly believed that the described model will enhance the capabilities of CANDU transient thermohydraulic codes in predicting the reactor's coolant pump behavior.
- Research Organization:
- Univ. of Sherbrooke, Canada
- OSTI ID:
- 5746222
- Report Number(s):
- CONF-870601-
- Conference Information:
- Journal Name: Trans. Am. Nucl. Soc.; (United States) Journal Volume: 54
- Country of Publication:
- United States
- Language:
- English
Similar Records
Modeling of drift effects in CANDU reactors
Review and analysis of state-of-the-art of multiphase pump technology. Technical Report No. 1
A transient two-phase velocity difference model for drift calculation in CANDU thermohydraulic codes
Conference
·
Tue Dec 31 23:00:00 EST 1985
· Trans. Am. Nucl. Soc.; (United States)
·
OSTI ID:6891895
Review and analysis of state-of-the-art of multiphase pump technology. Technical Report No. 1
Technical Report
·
Sat Jan 31 23:00:00 EST 1976
·
OSTI ID:7313549
A transient two-phase velocity difference model for drift calculation in CANDU thermohydraulic codes
Journal Article
·
Sun Nov 30 23:00:00 EST 1986
· Nucl. Technol.; (United States)
·
OSTI ID:6754623
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210400 -- Power Reactors
Nonbreeding
Otherwise Moderated or Unmoderated
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
42 ENGINEERING
420400 -- Engineering-- Heat Transfer & Fluid Flow
ACCIDENTS
CANDU TYPE REACTORS
COOLING SYSTEMS
DRYOUT
ENERGY SYSTEMS
ENERGY TRANSFER
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HEAVY WATER MODERATED REACTORS
HYDRAULICS
LOSS OF COOLANT
MECHANICS
PRESSURE TUBE REACTORS
PRIMARY COOLANT CIRCUITS
PUMPS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
THERMAL REACTORS
TWO-PHASE FLOW
210400 -- Power Reactors
Nonbreeding
Otherwise Moderated or Unmoderated
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
42 ENGINEERING
420400 -- Engineering-- Heat Transfer & Fluid Flow
ACCIDENTS
CANDU TYPE REACTORS
COOLING SYSTEMS
DRYOUT
ENERGY SYSTEMS
ENERGY TRANSFER
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HEAVY WATER MODERATED REACTORS
HYDRAULICS
LOSS OF COOLANT
MECHANICS
PRESSURE TUBE REACTORS
PRIMARY COOLANT CIRCUITS
PUMPS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
THERMAL REACTORS
TWO-PHASE FLOW