Steam Generator Tube Rupture Sequences Research Articles

Realistic evaluation of source terms for steam generator tube rupture accidents

In the existing Probabilistic Safety Assessment (PSA) model, all the Steam Generator Tube Rupture (SGTR) sequences have been assumed to have identical fission product release characteristics without considering the sequences of no source term release resulting in a conservative evaluation. Since the Korean Nuclear Safety Act was revised in 2016, a realistic method that can reasonably evaluate the risk became needed to satisfy the PSA targets. Therefore, a new method is proposed to improve the Level 2 PSA model (decomposition event tree, containment event tree, source term category logic diagram) that can identify the SGTR sequences without fission product release to the environment. The sequences are classified by dividing the Source Term Category (STC) of SGTRs into SGTR_CON (containment) and SGTR_ENV (environment). If fission product is trapped inside the containment building and not released to the environment, it is classified as SGTR_CON. Otherwise it is classified as SGTR_ENV. About 72% of the existing STC for SGTR is treated as SGTR_CON, which does not release fission product to the environment. If SGTR STCs are divided into SGTR-CON and SGTR-ENV, then a more reasonable source term evaluation of SGTR sequences is expected.

Key Findings from the Artist Project on Aerosol Retention in a Dry Steam Generator

A steam generator tube rupture (SGTR) event with a stuck-open safety relief valve constitutes one of the most serious accident sequences in pressurized water reactors (PWRs) because it may create an open path for radioactive aerosol release into the environment. The release may be mitigated by the deposition of fission product particles on a steam generators (SGs) dry tubes and structures or by scrubbing in the secondary coolant. However, the absence of empirical data, the complexity of the geometry, and the controlling processes have, until recently, made any quantification of retention difficult to justify. As a result, past risk assessment studies typically took little or no credit for aerosol retention in SGTR sequences. To provide these missing data, the Paul Scherrer Institute (PSI) initiated the Aerosol Trapping In Steam GeneraTor (ARTIST) Project, which aimed to thoroughly investigate various aspects of aerosol removal in the secondary side of a breached steam generator. Between 2003 and 2011, the PSI has led the ARTIST Project, which involved intense collaboration between nearly 20 international partners. This summary paper presents key findings of experimental and analytical work conducted at the PSI within the ARTIST program.

Evaluation of the offsite dose contribution to the global risk in a Steam Generator Tube Rupture scenario

The current main figure of merit for risk based decision making process based on Probabilistic Safety Assessment level 1 is usually related with the fuel failure (i.e., Peak Cladding Temperature (PCT)>1477.15K). In this approach, the core damage is the first and necessary step in a potential radiological release, being the containment failure the second one. Nevertheless, SGTR sequences in PWR plants are able to release large quantities of radioactive products without previous core damage or containment failure. For that reason, it seems necessary to analyze which sequences exceed the allowed offsite dose criteria prior to the core damage criterion.The aim of this analysis has been to evaluate the risk contribution due to the offsite dose and the core damage in case of Steam Generator Tube Rupture (SGTR) sequences at full power in a 3-loop Pressurized Water Reactor (PWR) Westinghouse-design. The study has been performed with SCAIS/MAAP and RADTRAD codes. For that purpose, this analysis unfolds the SGTR Dynamic Event Tree for both the core damage and the offsite dose risk metrics. The results indicate that dose criteria complement the PCT criterion and allow quantifying both risk contributions in SGTR sequences.

Hydrodynamics of Jets From Guillotine Steam Generator Tube Rupture: Modeling, Analytical Results, Computational Fluid Dynamics Calculation, and Comparison With Experimental Data

In this work we study the hydrodynamics of characteristic gas jets resulting from guillotine breaks of steam generator tube rupture sequences (SGTR) in pressurized nuclear power reactors. As an initial step towards describing an “in-bundle” gas jet, a hydrodynamic model of free gas jets emerging from a guillotine break under prototypical SGTR conditions has been developed. First we have studied the jet characteristic for an isolated tube; the analytical model estimates variables such as trajectories, centerline velocities, velocity distribution, and Reynolds stresses. We have performed model comparisons with experimental data for different experimental conditions with different mass flow rates, and we have found good agreement of the model with the experimental results. Additionally, an “ad hoc” expression has been derived for the centerline jet velocity, which has been experimentally confirmed. Consistently with the experimental data and the computational fluid dynamics (CFD) calculations the analytical model predicts no outflow near the jet center. As a complementary issue, we have performed CFD calculations for a guillotine tube rupture when the tube is surrounded by several rows of neighboring tubes, in this case the jet trajectories are affected by the Coanda effect near the tubes.

Characterization of a gas elliptic jet in a nuclear safety scenario

This paper summarizes the major insights gained into the aerodynamics of a subsonic gas jet expanding across a bundle of tubes from an axial-type tube breach. This scenario is highly relevant in nuclear safety since it determines the potential retention of radioactive aerosol particles during a severe accident Steam Generator Tube Rupture (SGTR) sequence. A scaled-down mock-up with representative dimensions of a real shell and tube heat exchanger of a nuclear power plant was used in the experiments. Two dimensional (2D) PIV technique was used to characterize the flow field in the space between the breach and the neighbor tubes in the Reynolds range investigated (ReD=0.8–2.7×105). The free jet case (i.e. jet expansion in a “tube-free” space) was used as a reference to compare the “in-bundle” jet case. By comparing with previous results obtained with an axisymmetric radial tube breach, the effect of the type of breach in the nearby flow field across the bundle has been also analyzed. Results show that an axial tube breach generates an elliptic jet with “quasi top-hat” velocity profile. The presence of tubes distorts the jet shape, reduce its penetration and its gas entrainment with respect to a “tube-free” space. The type of breach drastically affects the jet evolution in the surroundings of the breach. Particularly, different velocity profiles and jet flow entrainments have been found for the same upstream conditions. Results are being applied in the development of aerosol retention models of severe accident SGTR sequences.

De-agglomeration mechanisms of TiO2 aerosol agglomerates in PWR steam generator tube rupture conditions

A steam generator tube rupture (SGTR) in a pressurized water reactor (PWR) might be a major source of accidental release of radioactive aerosols into the environment during severe accident due to its potential to by-pass the reactor containment. In the ARTIST program, tests were carried out at flow conditions typical to SGTR events to determine the retention of dry aerosol particles inside a steam generator tube. The experiments with TiO2 agglomerates showed that for high velocities in the range of 100–350 m/s, the average particle size at the outlet of the tube was significantly smaller than at the inlet due to particle de-agglomeration. Earlier, particle de-agglomeration has not been considered significant in nuclear reactor severe accidents. However, the tests in ARTIST program have shown that there is a possibility that TiO2 aerosol particles de-agglomerate inside a tube and in the expansion zone after the tube exit under SGTR conditions. In this investigation, we measured TiO2 aerosol de-agglomeration in the tube with very high flow velocities with two different TiO2 aerosols. The de-agglomeration was determined by measuring the size of the agglomerates at the inlet and outlet of the test section. The test section was composed of tubes with three different lengths, 0.20, 2.0 and 4.0 m, followed by an expansion zone. The main results were: (i) the de-agglomerate process was relatively insensitive to the initial particle size distribution, (ii) the agglomerates were observed to de-agglomerate in all the tubes, and the resulting particle size distributions were similar for both TiO2 aerosols, (iii) at high flow rates, increasing the gas mass flow rate did not produce further de-agglomeration, and (iv) the agglomerates did not de-agglomerate to primary particles. Instead, after de-agglomeration the particles had a median outer diameter Dc = 1.1 μm. Based on analysis using computational fluid dynamics (CFDs), the de-agglomeration was caused by the turbulent shear stresses due to the fluid velocity difference across the agglomerates in the viscous subrange of turbulence. It has to be noted that the particles used in this investigation were TiO2 agglomerates, and not prototypical nuclear aerosols with significantly different characteristics. Therefore, the results of this investigation cannot be directly used to determine whether the nuclear aerosol particles may de-agglomerate in SGTR sequences. However, this investigation highlights the possibility of particle de-agglomeration under SGTR conditions, and identifies the mechanism of the de-agglomeration inside the broken tube and when the aerosol is discharged to an open space.

Major Challenges to Modeling Aerosol Retention Near a Tube Breach During Steam Generator Tube Rupture Sequences

Postulated accident sequences of a pressurized water reactor, consisting of steam generator tube ruptures (SGTRs) in combination with a melting core, have been demonstrated to represent a dominant contribution to the overall public risk. However, it should be expected that even in the absence of any water in the secondary side of the steam generator (“dry” SGTR scenario), some radioactivity retention takes place as a result of the interaction of the carrier gas with internal structures. The region near the tube breach becomes a key region because it behaves as a sink for the radioactive particles entering the secondary side, and consequently, it changes size distribution of aerosols flowing toward upper structures.This paper identifies major issues that should be addressed to accurately estimate aerosol retention in the field near a tube breach during dry SGTR scenarios. By developing a simple Lagrangian model based on the filter-concept approach (ARISG-I), the specific aspects of fluid dynamics and aerosol physics involved have been explored and the major knowledge gaps highlighted.Inertial impaction and turbulent deposition have been demonstrated to be major particle removal mechanisms. Their respective collection efficiencies have been derived by gathering and correlating separate effect data on particle deposition on cylinders in a crossflow configuration. Comparisons of model predictions to experimental data taken in a mock-up facility of the break stage under similar conditions to those anticipated in dry SGTR scenarios have been set. The substantial discrepancies found and their analysis have provided insights into the significance of drawbacks of model fundamentals, the inaccuracy of specific equations of deposition mechanisms, and most importantly, the lack of consideration of key phenomena that hinder aerosol retention.According to this analysis the main areas where research is needed are: gas jet behavior across the tube bank; particle resuspension, erosion, and/or bouncing; and particle inertial impaction and turbulent deposition under foreseen conditions.

Aerosol Retention Near the Tube Breach During Steam Generator Tube Rupture Sequences

Steam generator tube rupture sequences are identified as major contributors to the risk assessments of pressurized water reactors. Despite very low probability, they involve a direct pathway for radioactivity release into the environment. Nonetheless, fission products could be partially retained in the secondary side of the steam generator, even in the absence of water. This paper summarizes the main results of a bench-scale experimental program focused on the aerosol retention near the tube breach at the secondary side of a dry steam generator. The major variables investigated were the breach configuration (i.e., type, orientation, and location) and the gas mass flow rate. The results showed that near the breach, aerosol retention is low (<20%), and it generally decreases when the gas mass flow rate increases. Discussion of the experimental results suggested that certain phenomena, such as fragmentation and/or resuspension, as well as particle nature could have a large effect on the scenario studied, and they should be considered as potential issues and/or variables to be explored in future work.

Steam generator tube rupture (SGTR) scenarios

The steam generator tube rupture (SGTR) scenarios project was carried out in the EU 5th framework programme in the field of nuclear safety during years 2000–2002. The first objective of the project was to generate a comprehensive database on fission product retention in a steam generator. The second objective was to verify and develop predictive models to support accident management interventions in steam generator tube rupture sequences, which either directly lead to severe accident conditions or are induced by other sequences leading to severe accidents. The models developed for fission product retention were to be included in severe accident codes. In addition, it was shown that existing models for turbulent deposition, which is the dominating deposition mechanism in dry conditions and at high flow rates, contain large uncertainties. The results of the project are applicable to various pressurised water reactors, including vertical steam generators (western PWR) and horizontal steam generators (VVER).