Regulatory Guidance on Replacement Steam Generators - An Experimental Study of the Effects of Flat Bar Supports on Streamwise Fluidelastic Instability in Nuclear Steam Generators (R682.1)

1.0 Background

Flow-induced vibrations of tubes in steam generator U-bends can lead to a loss of tube integrity due to fretting wear at the tube supports and/or from tube-to-tube impacting. Since these tubes represent the pressure boundary between the irradiated primary side coolant and the secondary side flows, their failure constitutes a breach of containment, and therefore such tube leaks cannot be tolerated. Until the recent operating experience of steam generator tube degradation at San Onofre Nuclear Power Plant, streamwise (or more frequently referred to as in-plane) fluid-elastic instability (FEI) was considered not to be a problem. Indeed, the ASME and JSME Codes do not require consideration of streamwise FEI. While design guidelines against such damaging vibrations are in place, tubes failures still occur and the danger increases as the time in service of steam generators grows longer, and when replacement steam generators do not precisely replicate the original designs.

The steam generator (SG) design guidelines are based largely on empirical data from laboratory experiments and design code verification is strongly dependent on service experience. Thus, when new or replacement steam generators (RSGs) are designed, any departure in design from previously proven hardware creates risk. The best proof of this is the permanent shutdown of the San Onofre Nuclear Generating Station (SONGS) in 2012 after operating brand new replacement steam generators for only two years. In fact, one of the units developed large numbers of tube failures after only 11 months of service. These new SGs replaced the original steam generators which had given about 25 years of service. The failures were due to tube-to-tube impacting from in-plane fluid-elastic instability in the U-bends, something which had never before been seen in steam generators in service. Indeed, the current practice was to ignore this as a possible failure mode. The failure of RSGs resulted in a disaster for the nuclear industry and, numerous critiques from the Office of Inspector General pointed to US Nuclear Regulatory Commission (US NRC). Neither Southern California Edison, as the plant operator, nor the US NRC, as the national regulator, saw the risk associated with the RSG design and operation. It is clear that the current design guidelines are inadequate to deal with advanced designs which specify hardware which is beyond our operating experience. Independent research study is required to address needs of a national regulator and to improve our understanding of the effects of such parameters as tube pattern and pitch ratio, tube-to-support clearances and void fraction on tube response to flow-induced vibrations.

2.0 Objectives

CANDU utilities are in process of replacing aging steam generators (48 replacement steam generators to be design, manufactured and installed by 2028. The replacement steam generators differ from the original steam generators already licensed. The purpose of this project is to develop this required knowledge and to provide essential information for ensuring that the CNSC staff can make informed assessments of vendor’s designs. The key deliverable of the research project is a guide for a regulatory assessment of the design and operation of replacement steam generators for CANDU reactors. 

3.0 Scope of Work

The proponent must develop modelling and complete an experimental laboratory study of flow-induced vibrations of tubes in a scaled sectional model of a CANDU type nuclear steam generator under conditions in which the current understanding and available data for design and evaluation is inadequate. Steam-water or a suitable refrigerant must be used to achieve two-phase flow conditions rather than an air-water mixture. Particular attention should be paid to consider the situations of high void fractions (around 90%) in which service failures generally occur. Tube failures in steam generator service may occur in the regions of the U-bends where the void fraction is highest and therefore, the velocities are the highest and the tube damping is least. It is these areas which have not been studied adequately, especially with regard to stream-wise fluid-elastic instability. Extensive measurements must be performed to determine tube response, stability and damping as a function of mass flux and void fraction. Experiments must be carried out to study the effect of flat bar supports on both transverse and streamwise stability of the tubes, including the effects of tube-to-support clearance.

CNSC staff are aware that such experiments conducted in two phase flows at very high void fractions could be challenging.

4.0 Tasks to be Performed

1. 1.  Literature survey and review of existing knowledge in the area of stream-wise vibrations in steam generators and heat exchangers.

   2.  Review operating experience with replacement steam generators already in operation along with design changes made with RSG vs original SG design. 

   3.  Identify critical flow and structural parameters of importance for initiation and propagation of streamwise vibrations

   4.  Identify most vulnerable array geometry for initiation of streamwise vibrations and critical parameters (tube pitch, array geometry, circulation ratio, tube support clearances, void fraction, …)

   5.  Develop structural scaling from CANDU steam generator to experimental test bundle

   6.  Propose preliminary design for experimental setup

   7.  Develop instrumentation to measure critical parameters, tube frequency response, damping ratio, tube rms amplitude response, and thermal hydraulics parameters

      4.8 Define experimental test matrices

      4.9 Propose model for streamwise flow induced vibration

      4.10 Perform benchmark experiments with data for tube response, stability threshold and damping for a range of void fraction from 60% to about 95%

1. 1. Analyse experimental results and model predictions to validate model.

   2. Re-evaluate test plan if necessary. Perform experiments to study effects of flat bar supports on stability. Quantify the effect of flat bar supports with specified clearance

   3. Perform set of experiments on the effect of clearance on stability

Prepare report with complete experimental results, analysis, and implications for steam generator design. Develop guidelines for regulatory assessment of design and performance of replacement steam generators.