slide1
Download
Skip this Video
Download Presentation
Thermal-Hydraulic Transient Analysis of the Missouri University Research Reactor (MURR) TRTR Annual Meeting September 17-20, 2007

Loading in 2 Seconds...

play fullscreen
1 / 41

Thermal-Hydraulic Transient Analysis of the Missouri University Research Reactor MURR TRTR Annual Meeting September 17- - PowerPoint PPT Presentation


  • 184 Views
  • Uploaded on

Thermal-Hydraulic Transient Analysis of the Missouri University Research Reactor (MURR) TRTR Annual Meeting September 17-20, 2007. Dr. Robert C. Nelson 1 , J. Charles McKibben 2 , Dr. Kiratadas Kutikkad 2 , and Leslie P. Foyto 2 1 RRSAS 2 MURR. Introduction.

loader
I am the owner, or an agent authorized to act on behalf of the owner, of the copyrighted work described.
capcha
Download Presentation

PowerPoint Slideshow about 'Thermal-Hydraulic Transient Analysis of the Missouri University Research Reactor MURR TRTR Annual Meeting September 17-' - storm


An Image/Link below is provided (as is) to download presentation

Download Policy: Content on the Website is provided to you AS IS for your information and personal use and may not be sold / licensed / shared on other websites without getting consent from its author.While downloading, if for some reason you are not able to download a presentation, the publisher may have deleted the file from their server.


- - - - - - - - - - - - - - - - - - - - - - - - - - E N D - - - - - - - - - - - - - - - - - - - - - - - - - -
Presentation Transcript
slide1

Thermal-Hydraulic Transient Analysis of the Missouri University Research Reactor (MURR)TRTR Annual MeetingSeptember 17-20, 2007

Dr. Robert C. Nelson1, J. Charles McKibben2, Dr. Kiratadas Kutikkad2, and Leslie P. Foyto2

1 RRSAS 2 MURR

introduction
Introduction
  • Potential for fuel damage has been evaluated for various accidents and transients postulated for the MURR.
  • Criterion for ‘no fuel damage’ is that fuel plate peak temperatures do not approach the minimum fuel plate blister temperature of 900 °F (484 °C).
  • Model includes full core – 8 fuel elements; each with all 24 fuel plates and 25 coolant channels.
  • The model developed and analyses for the LOCA and LOF transients will be discussed.
murr model
MURR Model
  • Fuel region is cooled by a pressurized primary coolant system.
  • Pool coolant system is separate from primary coolant system.
  • Pressurized primary coolant system is located in the reactor pool allowing direct heat transfer during normal operation and a transition to natural convection under accident conditions.
  • Reflector region, control blade region, and center test hole are cooled by pool water (forced flow transitioning to natural convection).
calculated accidents
Calculated Accidents
  • Loss of Primary Coolant
    • Cold Leg
    • Hot Leg
  • Loss of Primary Flow
loss of coolant accident loca
Loss of Coolant Accident (LOCA)
  • Historically the most serious accident considered.
  • Initiated in theory by the double-ended rupture in a section of main coolant piping.
  • Mitigated by use of engineered Safety Features (ESF) – Anti-Siphon System.
  • These features are demonstrated in the following schematic of the in-pool portion of the primary coolant system.
slide11

PT 943

PS 938

V527C

top level loca results
Top Level LOCA Results
  • Peak Steady State Temperature – 272.1 °F (133.4 °C) [centerline of plate number-1]
  • Hot Leg Break – 282.1 °F (138.4 °C) [centerline of plate number-1 at 0.2 seconds]
  • Cold Leg Break – 311.7 °F (155.4 °C) [centerline of plate number-3 at 0.5 seconds]
  • No challenge to fuel plate blister temperature of 900 °F (482 °C)

NORMAL REACTOR OPERATING CONDITIONS AND CONSERVATIVE ASSUMPTIONS WHEN THE LOCA INITIATES

1 Pressure above atmosphere

slide16
Centerline Temperature of the 24 Fuel Plates (Section 4) During the First 20 Seconds of the Cold Line LOCA
slide17
Liquid Fraction of the 25 Individual Coolant Channels (Volume 1) During the First 600 Seconds of the Cold Leg LOCA
slide18
Liquid Fraction of the 25 Individual Coolant Channels (Volume 2) During the First 600 Seconds of the Cold Leg LOCA
slide19
Liquid Fraction of the 25 Individual Coolant Channels (Volume 3) During the First 600 Seconds of the Cold Leg LOCA
slide20
Liquid Fraction of the 25 Individual Coolant Channels (Volume 4) During the First 600 Seconds of the Cold Leg LOCA
slide21
Centerline Temperature of the 24 Fuel Plates (Section 1) During the First 20 Seconds of the Cold Leg LOCA
slide22
Centerline Temperature of the 24 Fuel Plates (Section 2) During the First 20 Seconds of the Cold Leg LOCA
slide23
Centerline Temperature of the 24 Fuel Plates (Section 3) During the First 20 Seconds of the Cold Leg LOCA
slide24
Temperature of the 25 Individual Coolant Channels (Volume 1) During the First 20 Seconds of the Cold Leg LOCA
slide25
Temperature of the 25 Individual Coolant Channels (Volume 4) During the First 20 Seconds of the Cold Leg LOCA
slide29
Centerline Temperature of the 24 Fuel Plates (Section 3) During the First 40 Seconds of the Hot Leg LOCA
slide30
Temperature of the 25 Individual Coolant Channels (Volume 1) During the First 200 Seconds of the Hot Leg LOCA
loca conclusions
LOCA Conclusions
  • None of the postulated scenarios results in uncovering of the core or core damage, including the most serious cold line break.
  • Post LOCA, decay heat can safely be dissipated to the reactor pool with no core damage.
  • Sufficient redundant safety features exist to prevent core damage as a result of a double-ended rupture of the largest diameter primary coolant piping without any additional protective system.
  • Model does not include two small check valves that allow make-up water from the reactor pool into the primary coolant system inverted loop.
lof accident
LOF Accident
  • Loss of Flow (LOF) accident initiation
    • Loss of Facility or pump power
    • Inadvertent closure of coolant loop isolation valves
    • Inadvertent loss of pressurizer pressure
    • Locked rotor in a coolant circulation pump
    • Failure of a coolant circulation pump coupling
slide36
Downward, Upward, and Net Coolant Flow Through the Core During the First 12 Seconds of the LOF Accident
slide37
Downward, Upward, and Net Coolant Flow Through the Core During the First 100 Seconds of the LOF Accident
slide38
Coolant Flow Through the 25 Individual Coolant Channels During the First 60 Seconds of the LOF Accident
summary
Summary
  • A new detailed RELAP5 model has been developed to evaluate the thermal-hydraulic characteristics of the MURR during normal and accident conditions.
  • Evaluation of LOCA and LOF accidents, including individual fuel plate and coolant channel temperatures, demonstrate that fuel clad integrity is not challenged.
  • Future efforts will evaluate potential effects of the new LEU fuel element design, which includes wider coolant channels, and an expansion of the model for consideration of fuel elements with variable burn-up histories.
ad