Robust LBB Analyses for Atucha II Nuclear Plant

Abstract

The Atucha II nuclear power plant is a unique pressurized heavy water reactor being constructed in Argentina. The original plant design was by KWU in the 1970’s using the then German methodology of break preclusion, which assumed that the largest break-opening area would be 10-percent of the cross-sectional area of the largest pipe diameter. That philosophy was used for the design of the emergency core cooling system in the 1970’s. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. The construction is progressing with initial start-up in 2011. Since the 10-percent of the cross-sectional area is a smaller ECCS design requirement than the normally assumed double-ended-guillotine break, the safety evaluation of the plant for beyond design basis seismic loading of the nuclear plant was a regulatory requirement. This overview paper describes a Robust LBB Evaluation that was conducted in great detail to assess the safety aspects of the piping system under beyond design basis seismic loading and the implications to the ECCS. Key aspects involved: • Static and dynamic material property testing, • Determination of weld residual stresses, • Determination of crack sizes that might evolve by worst case SCC growth rates under weld residual stresses and normal operating stresses, • Determination of leakage rates as a function of time with the upper-bounding crack growth rates, • Development of seismic hazard curves for the site, • Development of FE models of the containment building and primary NSSS system within the building, • Determination of normal operating stresses, SSE stresses and 10−6 seismic stresses using worst case soil foundation assumptions, • Evaluation of flaw behavior for circumferential cracks using the shapes from the natural crack growth. • Evaluation of margins on the critical flaw size and times to leakage, and • Standard LBB analyses, as well as Transition Break Size evaluations. The key result from this effort was that even with all the normal operating plus 10−6 seismic event loading, the pipe system behaved more like it was displacement-controlled than load-controlled. The displacement-controlled behavior made the pipe much more flaw tolerant, and it was found that a DEGB was not possible because the flaw could never reach the critical flaw size without greatly surpassing the leakage and water make-up capacity of the plant. Since there are many details in this multi-year effort, only the key points will be summarized in this paper while other details will be the topics of other papers.

The Huge Missing Factor in Leak-Before-Break Analysis: How a Circumferential Through-Wall-Crack in a Pipe System Changes the Flexibility and Reduces the Applied Moments

“…A very large FE model was created for the entire NSSS of the Argentina Atucha II nuclear plant including the containment building, reactor pressure vessel, steam generator, pump, primary pipe loop, supports between the building and the components and piping 1 , see Figure 3. The details can be found in References [Wilkowski et al 2011, Uddin et al 2015, Uddin et al 2014]. The loading was pressure, dead-weight, thermal expansion, and seismic excitation from the basemat of the containment building and then naturally transmitted though the supports to the components/piping.…”

“…Figure 4 Natural crack shape for large crack growth from surface crack to through-wall crack (Wilkowski et al 2011), (legend is in years of crack growth)…”

“…In the nuclear industry the leakage of subcooled water is determined by a number of software codes, one of which is called SQUIRT (Paul et al 1994, SQUIRT 2009). SQUIRT was developed for the US NRC, validated by many leak-rate tests, and was also used in the Atucha II evaluation in Reference [Wilkowski et al 2011]. The crack was inserted in the FE model of the whole pipe system to determine the crack-opening area under normal operating loads, so it accounted for any restraint of the induced bending from the pressure loads.…”

“…In the design and analyses of those pipe system tests, finite element (FE) models of the entire pipe system were used and large margins were found due to a number of factors, of which not all of them were recognized until about 2011. In 2011 (Wilkowski et al 2011), a full FE model was created for the Argentinian Atucha II PHWR plant including; the containment building, reactor pressure vessel (RPV), steam generator (SG), pumps, main coolant line, surge lines, pressurizer and main steam lines and all the supports from the components/pipes to the building. These difficult and tedious efforts provided valuable insights into the system behavior under stress.…”

Leak-before-break: Global perspectives and procedures

“…A recent work undertaken to assess the safety aspects of Atucha II nuclear power plant piping system under beyond design basis seismic loading [67]- [68] shown surprising results. It was found important margins due to the LBB assumption that all stresses are load-controlled while the pipe system behaved more like it was displacement-controlled.…”