News & Views, Volume 46 | Assessing Prestress Losses in a Nuclear Containment Structure for License Renewal
By: Eric Kjolsing
Nuclear power plants around the world are approaching the end of their original 40-year design life. Efforts are underway to extend the operating license for these plants to 60 years or beyond. As part of the license extension, it must be demonstrated that the reactor containment building remains able to safely perform its intended functions for the extended duration of operation. Many of these containment buildings utilize a post-tensioned concrete design where the tendons are grouted after tensioning. Since these grouted tendons cannot be re-tensioned, an assessment for the loss in prestress beyond the original design life must be performed.
This article describes a methodology to assess the structural performance of a containment structure over time as a function of confidence in the tendon losses and is split into three parts:
- A description of the methodology
- A representative probabilistic assessment
- Representative analysis results
As part of the general design criteria for nuclear power plants, the primary structures and systems of the plant must be designed to handle postulated accident events, including the dynamic effects of postulated pipe ruptures. For a Boiling Water Reactor, analyzed events include various accident conditions in the recirculation piping, including a Loss of Coolant Accident (LOCA). One postulated LOCA event is assumed to be an instantaneous double-ended guillotine break of the recirculation line. This event causes several loads to be imparted on the reactor vessel, attached piping, and reactor internal components. [Some loads such as jet impingement, annulus pressurization, and pipe whip impart loads on the outside of the reactor vessel and the attached piping.][ Other loads, including flow-induced drag and acoustic loads, transmit loads inside the vessel on critical components such as jet pumps, core shroud, and the shroud support structure.] Figure 1 shows the pipe and resulting loads.
suction pipe-to-elbow stainless steel weld was identified in both units of a nuclear power plant as depicted in Figure 1.
While the 2018 Spring outage season is mostly behind us, we all know a key element in being able to provide safe, reliable, clean and economic power to energy consumers is how successfully plant outages are accomplished. 
Nuclear plant workers accrue most of their radiation exposure during refueling outages, when many plant systems are opened for corrective and preventive maintenance. The total refueling outage radiation exposure can be 100-200 person-Rem at a typical Boiling Water Reactor (BWR), and 30-100 person-Rem at a typical Pressurized Water Reactor (PWR). Accrued refueling outage radiation exposure values can be significantly greater than these values depending upon radiation fields, outage work scope, and emergent work. Outage radiation exposure is one metric used by a plant to determine outage success and by industry regulators in assessing the overall performance of a plant. Plants with high personnel radiation exposure tend to be those plants with more equipment problems and more unscheduled shutdowns; consequently, they may be subjected to increased regulatory oversight. 
As all of us who work with nuclear energy know the US nuclear industry is engaged in a multi-year effort to generate power more efficiently, economically and safely. A key goal includes a significant reduction in operating expenses. This initiative is termed “Delivering the Nuclear Promise” (DNP) and is supported by nuclear utilities, vendors such as Structural Integrity, the Nuclear Energy Institute (NEI), Institute 