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Thermal-Hydraulic Transient Analysis of the Missouri University Research Reactor (MURR) TRTR Annual Meeting September 17-20, 2007 PowerPoint PPT Presentation


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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.

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

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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

  • 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

  • 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).


Detailed RELAP5 Model of the MURR Primary Coolant Loop


Detailed RELAP5 Model of the MURR Bulk Reactor Pool and Pool Coolant Loop


MURR 24 Plate Core Model


Fuel Plate Power Distribution


Calculated Accidents

  • Loss of Primary Coolant

    • Cold Leg

    • Hot Leg

  • Loss of Primary Flow


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.


PT 943

PS 938

V527C


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


Cold Leg LOCA


Junction Flow Rates During The First 20 Seconds of the Cold Leg LOCA


Coolant Flow Through the 25 Individual Channels During the First 20 Seconds of the Cold Leg LOCA


Centerline Temperature of the 24 Fuel Plates (Section 4) During the First 20 Seconds of the Cold Line LOCA


Liquid Fraction of the 25 Individual Coolant Channels (Volume 1) During the First 600 Seconds of the Cold Leg LOCA


Liquid Fraction of the 25 Individual Coolant Channels (Volume 2) During the First 600 Seconds of the Cold Leg LOCA


Liquid Fraction of the 25 Individual Coolant Channels (Volume 3) During the First 600 Seconds of the Cold Leg LOCA


Liquid Fraction of the 25 Individual Coolant Channels (Volume 4) During the First 600 Seconds of the Cold Leg LOCA


Centerline Temperature of the 24 Fuel Plates (Section 1) During the First 20 Seconds of the Cold Leg LOCA


Centerline Temperature of the 24 Fuel Plates (Section 2) During the First 20 Seconds of the Cold Leg LOCA


Centerline Temperature of the 24 Fuel Plates (Section 3) During the First 20 Seconds of the Cold Leg LOCA


Temperature of the 25 Individual Coolant Channels (Volume 1) During the First 20 Seconds of the Cold Leg LOCA


Temperature of the 25 Individual Coolant Channels (Volume 4) During the First 20 Seconds of the Cold Leg LOCA


Coolant Flow Through the 25 Individual Channels During the First 20 Seconds of the Cold Leg LOCA


Volume Pressures During the First 20 Seconds of the Cold Leg LOCA


Hot Leg LOCA


Centerline Temperature of the 24 Fuel Plates (Section 3) During the First 40 Seconds of the Hot Leg LOCA


Temperature of the 25 Individual Coolant Channels (Volume 1) During the First 200 Seconds of the Hot Leg LOCA


Cold Leg LOCA Channel Temperatures (Volume 1)


Hot Leg LOCA Channel Temperatures (Volume 1)


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

  • 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


Downward, Upward, and Net Coolant Flow Through the Core During the First 12 Seconds of the LOF Accident


Downward, Upward, and Net Coolant Flow Through the Core During the First 100 Seconds of the LOF Accident


Coolant Flow Through the 25 Individual Coolant Channels During the First 60 Seconds of the LOF Accident


Centerline Temperature of the 24 Fuel PlatesDuring the First 60 Seconds of the LOF Accident


Temperature of the 25 Individual Coolant ChannelsDuring the First 60 Seconds of the LOF Accident


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.


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