STRUCTURE AND SOUNDNESS
A reactor pressure vessel in a nuclear power plant springs a leak. Water used to cool the nuclear fuel escapes. The fuel overheats, causing localized melting of the thick vessel wall and a discharge of radioactivity into the containment building.
Although such a scenario has never occurred in the United States, preventing it has been a prime concern of the Laboratory's Heavy Section Steel Technology (HSST) Program for 25 years.
The HSST program was established in response to a November 1965 letter by William Manly of the Advisory Committee on Reactor Safeguards to the Atomic Energy Commission, which recommended a more sophisticated approach to evaluation of the structural soundness of pressure vessels. In March 1967 the HSST program, sponsored by the AEC's Division of Reactor Development and Technology, came into being under Laboratory management. Its first director was F. J. Witt; successors have been Grady Whitman, Claud Pugh, Bill Corwin, and Bill Pennell. Today the HSST program, which continues as a major effort in the Engineering Technology and Metals and Ceramics divisions, is sponsored by the U.S. Nuclear Regulatory Commission (NRC).
Using large-scale testing procedures, the HSST program demonstrated that the thick steel walls of new reactor pressure vessels possess enough ductilitythe ability to accommodate stresses caused by pressurization, heating, and coolingto prevent vessel failure. In the late 1960s, the program also initiated a fracture toughness data base for reactor vessel materials. This information, detailing the ability of materials to resist cracking, is essential to all fracture-margin assessments for reactor pressure vessels.
During manufacture of steel plates for vessel walls, flaws may develop and spread into cracks as the walls become brittle. "Thermal shock" may occur when the heated walls of a vessel are suddenly subjected to cold water as a result of loss of pressure and the operation of safety injection systems to cool the nuclear fuel. In the late 1970s, ORNL researchers led by Dick Cheverton discovered that thermal shock, combined with repressurization (during emergency cooling, for example), could drive a crack through the vessel wall under postulated conditions.
More recently, Laboratory researchers have turned their attention to the problem of vessel aging. Over many years, as the vessel interior is bombarded by neutrons from nuclear reactions in the fuel, the walls tend to lose their ductility. Such radiation-induced embrittlement can occur in older pressurized-water reactors and, to a lesser extent, in boiling-water plants.
For older nuclear power plants, radiation-induced embrittlement is an issue that must be addressed if plant operating licenses are to be renewed. The HSST program provides the NRC with guidance on this issue by estimating the probability that a reactor vessel will fail over a specific operating time. The embrittlement rate in each reactor vessel is monitored, and operating limits are imposed by NRC regulations and regulatory guides that the Laboratory has helped to establish.
Today, the HSST program continues to investigate the properties of materials for pressure vessels to develop and evaluate ways to predict fracture, fatigue, and creep. It also conducts vessel and material tests to assess the validity of the predictions, which help to set and update national codes, standards, and regulations. HSST researchers intend to carry on the tradition of the past 25 years by providing the NRC with information that will help the agency respond to the new challenges of reactor safety.
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