A probabilistic approach to the evaluation of the PTS [pressurized thermal shock] issue
Description
The pressurized-thermal-shock (PTS) issue is concerned with the possibility of failure of pressurized-water-reactor (PWR) pressure vessels under a very specific set of conditions. These conditions include: (1) the occurrence of reactor transients that subject the vessel to severe thermal shock as well as the normal pressure loading, (2) the existence of sharp, crack-like defects (flaws) at the inner surface of the vessel wall, and (3) high enough fast neutron fluence and concentrations of copper and nickel in the vessel wall to result in a extensive radiation-included reduction in the fracture toughness of the vessel material. Under these conditions, the mechanism for vessel failure involves propagation of the flaws through the vessel wall, in which case adequate containment of coolant for the core might not be possible. The portion of the vessel of concern is the so-called beltline region because, it is directly opposite the core (high influence rate), it is adjacent to the coolant downcomer (potential for thermal shock), and coolant leakage in this area would tend to uncover the core. This document discusses the behavior of flaws in reactor pressure vessels under pressure and thermal-shock loading conditions
Availability note (English)
MF available from INIS under the Report Number; NTIS, PC A03/MF A01 as DE90010717; OSTI; INIS; US Govt. Printing Office Dep.Files
21068174.pdf
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Additional details
Publishing Information
- Imprint Pagination
- 29 p.
- Report number
- CONF-9006173--1
Conference
- Title
- US-USSR exchange meeting on reactor safety.
- Dates
- 25-29 Jun 1990.
- Place
- Moscow (USSR).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 21068174
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S36: MATERIALS SCIENCE; S99: GENERAL AND MISCELLANEOUS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- CALVERT CLIFFS-1 REACTOR; COPPER; CRACKS; FAILURES; FRACTURE MECHANICS; HEAT TRANSFER; HUMAN FACTORS; HYDRAULICS; MATHEMATICAL MODELS; NEUTRON FLUENCE; NICKEL; OCONEE-1 REACTOR; PRESSURE DEPENDENCE; PRESSURE VESSELS; PWR TYPE REACTORS; REACTOR MATERIALS; REACTOR SAFETY; REGULATIONS; ROBINSON-2 REACTOR; SENSITIVITY ANALYSIS; TEMPERATURE DEPENDENCE; THERMAL SHOCK; WELDED JOINTS
- Descriptors DEC
- CONTAINERS; ELEMENTS; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; JOINTS; LAW; MATERIALS; MECHANICS; METALS; POWER REACTORS; REACTORS; SAFETY; THERMAL REACTORS; TRANSITION ELEMENTS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Contract/Grant/Project number
- Contract AC05-84OR21400