Pressure vessel fracture studies pertaining to the PWR thermal-shock issue: experiment TSE-7
Thermal-shock experiment TSE-7 was conducted for the purpose of investigating the behavior of surface flaws under pressurized-water reactor (PWR) overcooling-accident conditions. This experiment was the eighth in a series of thermal-shock experiments conducted for this purpose with large steel cylinders (A 508, class-2 chemistry; 991-mm OD x 152-mm wall x 1.2-m length) as a part of the Heavy-Section Steel Technology (HSST) Program. The initial flaw for TSE-7 was a shallow, semielliptical, inner-surface, axially oriented, sharp crack located at midlength of the test cylinder. The thermal shock was applied to the inner surface only, and this was accomplished by effectively dunking the test cylinder, initially at approx.93/sup 0/C, into a large volume of liquid nitrogen. The specific purpose of TSE-7 was to determine whether, in agreement with analysis, a short and shallow surface flaw, in the absence of cladding, would extend on the surface to effectively become a very long flaw as a result of severe thermal-shock loading. During the experiment, there were three major initiation-arrest events. The first event consisted of some radial propagation and very extensive surface extension, with many bifurcations taking place. The second and third events consisted primarily of radial propagation. A fourth initiation event was prevented by warm prestressing. These results were in good agreement with predictions. 50 refs., 77 figs., 13 tabs.
- Research Organization:
- Oak Ridge National Lab., TN (USA)
- DOE Contract Number:
- AC05-84OR21400
- OSTI ID:
- 5461277
- Report Number(s):
- NUREG/CR-4304; ORNL-6177; ON: TI85017300
- Country of Publication:
- United States
- Language:
- English
Similar Records
Behavior of short flaws during thermal shock: thermal-shock experiment TSE-7
Large-scale thermal-shock experiments with clad and unclad steel cylinders
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
36 MATERIALS SCIENCE
PRESSURE VESSELS
FRACTURE MECHANICS
PWR TYPE REACTORS
LOSS OF COOLANT
STEEL-ASTM-A508
CHEMICAL COMPOSITION
CRACK PROPAGATION
CRACKS
HEAT TRANSFER
RESEARCH PROGRAMS
STRESS INTENSITY FACTORS
THERMAL SHOCK
ACCIDENTS
ALLOYS
CONTAINERS
ENERGY TRANSFER
IRON ALLOYS
IRON BASE ALLOYS
MECHANICS
REACTOR ACCIDENTS
REACTORS
STEELS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
360103 - Metals & Alloys- Mechanical Properties