Steam Generator Tubes Research Articles

Parametric sensitivity analysis of liquid metal helical coil once-through tube steam generator

Lead-bismuth Fast Reactor (LFR) has attracted attentions due to its intrinsic characteristics, such as sustainability, safety, and economics. Helical coil once-through tube steam generator (HCOTSG) is a proposed form of steam generator. Due to its unique advantages, HCOTSG is widely used in various reactor power systems, including LMFR. In this paper, FLUENT CFD commercial code is adopted to simulate heat transfer from the primary side (liquid metal) and secondary side (water/steam). The robustness of the method is validated by comparing it with relevant experimental research, and the maximum error is less than 25%. Based on this method, 9 models with different geometrical parameters were established to analyze the effects of geometrical parameters on the performance of liquid metal HCOTSG. The wall heat flux and comprehensive performance factor (En) of different models is compared. The optimal geometric arrangement scheme for the working conditions is recommended. The geometrical parameters of the tube side and shell side are analyzed as sensitivity parameters. Among the established geometric models, the HCOTSG has the best comprehensive performance when Pr /D is 0.167 and Pa /D is 0.137 under the working conditions in this paper. This study provides a numerical simulation method on structural design optimization of liquid metal HCOTSG.

Evaluation of the Potential for Containment Bypass due to Steam Generator Tube Rupture in VVER-1000/V320 Reactor during Extended SBO sequence using SCDAP/RELAP5 code

A severe accident-induced of a Steam Generator (SG) tube releases radioactivity from the Reactor Coolant System (RCS) into the SG secondary coolant system from where it may escape to the environment through the pressure relief valves and an environmental release in this manner is called “Containment Bypass”. This study aims to evaluate the potential for “Containment Bypass” in VVER/V320 reactor during extended Station Blackout (SBO) scenarios that challenge the tubes by primarily involving a natural circulation of superheated steam inside the piping loop and then induce creep rupture tube failure. Assessments are made of SCDAP/RELAP5 code capabilities for predicting the plant behavior during an SBO event and estimates are made of the uncertainties associated with the SCDAP/RELAP5 predictions for key fluid and components condition and for the SG tube failure margins.

Comprehensive investigation of the corrosion state and surface properties of the stainless steel tubes of steam generators

Evaluating the water chemistry in the primary circuit and the effect of chemical econtamination
 of the heat exchanger tubes performed by the AP-CITROX (AP: alkaline permanganate; CITROX: citric and oxalic acid) procedure at Paks NPP (Hungary), a project dealing with the comprehensive investigation of the general corrosion state of the steam generators (SGs) has been initiated. Owing to the fact that there is no investigation method available for the in-situ monitoring of the inner surfaces of heat exchanger tubes, a research program based on sampling as well as on ex-situ electrochemical (voltammetric) and surface analytical measurements (SEM-EDX, CEMS, XRD, XPS) was developed and elaborated. In the time period of 2000-2008 - within the frame of the above project - 45 stainless steel specimens, cut out from various locations of the steam generators of the Paks NPP were investigated. Based on the measured corrosion characteristics (corrosion rate, thickness and chemical composition of the protective oxide-layer) it was found that these parameters are strongly dependent on the decontamination history of steam generators. The present work gives a brief overview on the general corrosion state of the heat exchanger tubes of SGs, concerning the long-term effects of the AP-CITROX procedure on the chemical composition and structure of the protective oxide-layer.

Effect of surface scratch depth on microstructure change and stress corrosion cracking behavior of alloy 690TT steam generator tube

Samples with six different scratch depths were prepared on Alloy 690TT tube, and the scratch induced microstructure change and stress corrosion cracking (SCC) behavior were studied in detail. The scratched area could be divided into four regions based on the microstructure, residual strain and micro-mechanical property. With the increasing of scratch depth, the density of slip steps increased at scratched area. After a 1000 h corrosion test, both the number and the length of cracks increased with the increasing of scratch depth, indicating the increasing of SCC sensitivity. In addition, the plastic accumulation zone had the highest SCC sensitivity.

CFD simulation and flow-induced vibration evaluation of PWR steam generator U tubes

For new types of steam generators (SG) for pressurized water reactors (PWRs), the flow-induced vibration (FIV) of U tubes is one of the most important potential failure modes. The presence of numerous U tubes complicates the FIV process and makes it difficult to model. Although a global media model with a coarse mesh has been used to simplify the flow field of SG tube bundles, precise outputs cannot be obtained to evaluate the FIV, so a more accurate model needs to be developed. Support failure, support clearance, and offset are the key factors and are believed to significantly affect the FIV, but it is unclear how to properly account for this factor and its quantitative influence. In this work, a more accurate CFD simulation of two-phase flow and heat transfer on the SG secondary side is conducted by combining the local solid model with the global porous media model. In addition, comprehensive finite element model (FEM) models are built to study the quantitative influence of sensitive factors on FIV. The results show that the newly designed SG satisfied the FIV requirements. The maximum flow-elastic instability (FEI) coefficient increases as the support failure increases. The turbulence buffeting stress amplitude is lower than the fatigue stress limit. The most significant fretting wear occurs when the support clearance is near 2 times the nominal clearance, and the U tube offsets are beneficial to reducing the fretting wear.

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.

Analysis of steam generator tube rupture accident for OPR 1000 nuclear power plant

Evaluation of radionuclide releases to the environment from operating nuclear power plants has increased in importance in the Republic of Korea following the 2015 amendment to its Nuclear Safety Act that includes not only design basis accidents but also severe accident management. Containment bypass accidents are among the crucial scenarios that should be considered under the new circumstances due to their high consequences, even though the probability of such accidents is low. In this study, analysis of a thermally induced steam generator tube rupture (TI-SGTR) accident caused by a severe accident has been conducted with MELCOR. To increase the reliability of the analysis results, qualification processes were performed in the process of producing the input. In addition, computational fluid dynamics analysis was performed to simulate the natural circulation phenomena in the reactor coolant system (RCS) during the accident and major flow parameters were obtained for the target plant. Thermal-hydraulic and fission product behavior was found during the accident for the base case scenario, and uncertainty band for CsOH release fraction to the environment were presented through the uncertainty analysis on the natural circulation flow parameters. Sensitivity calculations were also performed on uncertain parameters such as the number of ruptured tubes and the depressurization time of the secondary system, and the CsOH release fraction to the environment without passing through the containment were evaluated.

Prediction of the Loss of Feed Water Fault Signatures Using Machine Learning Techniques

Fault diagnosis occurrence and its precise prediction in nuclear power plants are extremely important in avoiding disastrous consequences. The inherent limitations of the current fault diagnosis methods make machine learning techniques and their hybrid methodologies possible solutions to remedy this challenge. This study sought to develop, examine, compare, and contrast three robust machine learning methodologies of adaptive neurofuzzy inference system, long short-term memory, and radial basis function network by modeling the loss of feed water event using RELAP5. The performance indices of residual plots, mean absolute percentage error, root mean squared error, and coefficient of determination were used to determine the most suitable algorithms for accurately diagnosing the loss of feed water transient signatures. The study found out that the adaptive neurofuzzy inference system model outperformed the other schemes when predicting the temperature of the steam generator tubes, the radial basis function network scheme was best suited in forecasting the mass flow rate at the core inlet, while the long short-term memory algorithm was best suited for the estimation of the severities of the loss of the feed water fault.

Effects of Zinc Addition on the Corrosion Behavior of Pre-Filmed Alloy 690 in Borated and Lithiated Water at 330 °C.

The purpose of this work is to quantify the effects of dissolved zinc cations on corrosion and release rates from a pre-filmed Alloy 690 steam generator tubing material that was subsequently exposed to water containing zinc. The corrosion tests were performed in circulating 2 ppm Li and 1000 ppm B water without and with 60 ppb zinc at 330 °C. Gravimetric analyses and oxide characterization revealed that the corrosion rates, release rates, and oxide thicknesses decreased by subsequent exposure of the pre-filmed Alloy 690 to zinc. These benefits are attributed to the formation of a chromium-rich inner oxide layer incorporating zinc.

A Bayesian updating of crack distributions in steam generator tubes

The structural failure of steam generator tubes is a common problem that can a ect the availability and safety of nuclear power plants. To minimize the probability of occurrence of failure, it is needed to implement maintenance strategies such as periodic nondestructive inspections of tubes. Thus, a tube is repaired or plugged whenever it has detected a crack which a threshold size is overtaken. In general, uncertainties and errors in crack sizes are associated with the nondestructive inspections. These uncertainties and errors should be appropriately characterized to estimate the actual crack distribution. This work proposes a Bayesian approach for updating crack distributions, which in turn allows computing the failure probability of steam generator tubes at current and future times. The failure criterion is based on plastic collapse phenomenon, and the failure probability is computed by using the Monte-Carlo simulation. The failure probability at current and future times is in good agreement with the ones presented in the literature.

Analysis of steam generator tube rupture accidents for the development of mitigation strategies

We analyzed mitigation strategies for steam generator tube rupture (SGTR) accidents using MARS code under both full-power and low-power and shutdown (LPSD) conditions. In general, there are two approaches to mitigating SGTR accidents: supplementing the reactor coolant inventory using safety injection systems and depressurizing the reactor coolant system (RCS) by cooling it down using the intact steam generator. These mitigation strategies were compared from the viewpoint of break flow from the ruptured steam generator tube, the core integrity, and the possibility of the main steam safety valves opening, which is associated with the potential release of radiation. The “cooldown strategy” is recommended for break flow control, whereas the “RCS make-up strategy” is better for RCS inventory control. Under full power, neither mitigation strategy made a significant difference except for on the break flow while, in LPSD modes, the RCS cooldown strategy resulted in lower break and discharge flows, and thus less radiation release. As a result, using the cooldown strategy for an SGTR under LPSD conditions is recommended. These results can be used as a fundamental guide for mitigation strategies for SGTR accidents according to the operational mode.

Analysis of dependencies among performance shaping factors in human reliability analysis based on a system dynamics approach

Performance shaping factors (PSFs) describe the influence of given contexts on human performance and are used to quantify human error probabilities (HEPs). HEPs are calculated within human reliability analysis, which is a part of probabilistic risk assessment. One challenge lies in the fact that the dependencies among PSFs are neglected, which leads to the poor estimation of HEPs. The objective of this research is to apply a system dynamics approach to the modeling and analysis of the dependencies of PSFs within the standardized plant analysis of risk–human reliability analysis method. Dependencies are built on the causalities of PSFs and are determined by the mutual information theory and the analytic hierarchy process.The method is demonstrated on three human failure events. The first one is the failure of initiating reactor coolant system depressurization. The second one is the failure of throttling the high-pressure injection to reduce reactor coolant system pressure. The third one is the failure of diagnosing the steam generator tube rupture. The sensitive analysis results show that our model is reasonable and robust when dependencies are taken into account.

The operational assessment of steam generator tubes based on a stochastic crack growth model

In pressurized water reactor plants, the degradation of steam generator tubes has been one of the main economic and safety problems. Steam generator tubes are subjected to the degradation mechanism of stress corrosion cracking, in which an unstable crack growth may lead to their structural failure. In general, deterministic criteria provide the least conservative assessment in order to plug or sleeve the cracked tubes. As the crack growth rate under stress corrosion cracking is of high random nature, the structural integrity assessment of steam generator tubes must be performed taking into account the failure probability at current and next inspections. In this paper, a stochastic model is applied for performing the operational assessment of steam generator tubes for any time interval. The results obtained with this model are compared with others, such as Monte-Carlo simulation and non-evolutionary statistical model. The proposed methodology is able to provide efficiently a less conservative failure probability. The plausibility of stochastic models in the structural integrity assessment of steam generator tubes is reached in favor of their operational life extension.

DFT study of the fouling deposition process in the steam generator by simulating the adsorption of Fe2+ on Fe3O4 (0 0 1)

In order to reveal the fouling problem on the outer surface of the steam generator (SG) tube in the secondary circuit condition of pressurized water reactor (PWR) nuclear power plant, based on the density functional theory (DFT) method, the Cambridge sequential total energy program package (CASTEP) is used to simulate seven kinds of highly symmetric adsorption structure models of termination with tetrahedral Fe (A termination) and termination with octahedral Fe (B termination) on Fe3O4 (0 0 1) surface. The adsorption energies and stable adsorption conformations are calculated. The results show that the most stable adsorption structures of the Fe2+/Fe3O4 (0 0 1) configurations are Fe2+ above Fe-O bond of B layer termination (Fe3O4(001) A-b). During the adsorption, the Fe-Fe, Fe-O bond length, and Fe-Fe-O bond angle of (0 0 1) surface change, and the atomic positions parallel and perpendicular to (0 0 1) surface change correspondingly. The change happened to the surface layer is the most drastic one. The calculation of charge population, the density of states (DOS), and electron local function of Fe2+/Fe3O4 (0 0 1) optimal adsorption configuration show that there is electron transfer between Fe2+ and Fe3O4 (0 0 1), and the adsorption type is chemisorption. Among them, Fe (Fe2+)-Fe (Fe3O4) forms a metal bond, and Fe (Fe2+)-O (Fe3O4) forms the ionic bond. The results illustrate the interaction between free Fe2+ and Fe3O4 is the reason of the nucleation and agglomeration of Fe3O4 scale and it provides the foundation for the further research on Fe3O4 scale deposition.

Bubble Behaviors in a Fuel Assembly of Accelerator Driven System

The accelerator driven system (ADS) is a nuclear reactor that can burn spent nuclear fuel with inherent safety features and operational flexibility. China is developing an ADS research device named China Initiative Accelerator Driven System (CiADS). The steam generator is one of the most important parts in the primary circuit of CiADS. The rupture of steam generator tubes may cause a significant impact on the reactor core due to potential risks of steam bubble intrusion into the core region. When the steam generator tube-rupture accident occurs, the primary coolant turns to a two-phase flow, which may cause variations of flow conditions through fuel assemblies. This paper adopted the open-source computational fluid dynamics calculation software OpenFOAM, and performed numerical simulations of a two-phase flow scenario in the ADS fuel assembly with the volume-of-fluid method. By changing boundary conditions at the assembly inlet, different scenarios of steam bubble intrusion were simulated and compared, based on our preliminarily conclusion that steam bubbles evenly distributed at the assembly inlet tend to accumulate at the periphery region, which partly confirmed the necessity of building experimental facilities for relevant two-phase flow research.

Effects of Advanced Amines on Magnetite Deposition of Steam Generator Tubes in Secondary System

Ethanolamine (ETA) is widely used to control the pH value in the secondary water of pressurized water reactors. However, it is necessary to consider other advanced amines which can replace ETA due to its serious human hazards and environmental treatment problems. The purpose of this study is to contemplate the effects of three advanced amines (ETA, 3-methoxypropylamine (MPA), and dimethylamine (DMA)) on the magnetite deposition behavior of a thermally treated (TT) Alloy 690 tubes by using a steam generator (SG) tube fouling loop in simulated secondary water. All particles were identified as a magnetite and were polyhedral or spherical in shape. When using ETA, MPA, and DMA, the average porosity of the deposits was about 34.7%, 33.0%, and 24.6%, respectively. The amount of deposits was largest when ETA was added, and it decreased by 41% when adding MPA and 55% when adding DMA. The mechanism of magnetite deposition was discussed in terms of zeta potentials of both the magnetite particles and the Alloy 690TT surface and magnetite solubility depending on the amines. To compare the potential for replacing ETA with other advanced amines, the various factors such as SG integrity, human hazards, and environmental treatment problems were discussed.

Achieving 50% weight reduction of offshore steam bottoming cycles

Adding a bottoming cycle to the gas turbines powering offshore oil and gas production plants allows additional power to be produced from recovered excess heat. Hence, the power demand of the platform can be met by burning less natural gas, and the CO2 emissions reduced by up to 25%. However, the weight of the current bottoming cycles must come down to enable widespread implementation. This work presents a thorough weight minimization of a steam bottoming cycle utilizing gas turbine exhaust heat. Unconventional, but feasible designs of heat exchangers, ductwork and structural components are considered along with materials switching. Overall weight reductions of 38% and 52% were achieved for a 16 MW and a 12 MW offshore bottoming cycle respectively when compared to a 16 MW reference system. Key factors in achieving the weight reduction were the use of small steam generator tubes with an inner diameter of only 10 mm, improved condenser design and the use of aluminium structural framework replacing steel. By more than halving the weight of the bottoming cycle, it's implementation potential on offshore platforms has been greatly improved and can move the oil and gas industry towards significantly reduced CO2 emissions.

The Fouling Behavior of Steam Generator Tube at Different Positions in the High-Temperature Water

The fouling behavior of a steam generator (SG) tube was investigated at different positions after 500 h of immersion in high-temperature water. A triple-layer structure of fouling appeared at both the crevice position and the free span position, namely, the large, dispersedly distributed deposition layer on the top; the small and faceted outer layer; and the relatively continuous inner layer. There was no obvious positional effect on the thickness of the inner layer. However, in the crevice position, the density of the deposited particle and the thickness of the outer layer was much higher than those of the free span position. The tube support plate (TSP) made of 410 stainless steel contributed significantly to the fouling behavior of the SG tube in the crevice between the SG tube and the TSP.

Computational study of Fe3O4 adsorption behaviour on the secondary side of the heat exchange tube in the steam generator

Based on the scaling problem on the outer surface of the steam generator tube in the secondary circuit condition of pressurizedwaterreactor, the molecular dynamic was used to calculate the adsorption characteristics of Fe3O4. The trajectory, mean squared displacement, radial distribution function and adsorption energy of Fe3O4 agglomeration and deposition process were further analyzed through variation of temperature and concentration. When the temperature was between 510 K and 530 K, the diffusion coefficient of Fe2+ and OH− could reach the minimum value of 2.255 × 10-5 cm2/s and 2.638 × 10-5 cm2/s, respectively. During variation of Fe2+ concentration, the size of agglomerate particles increased continuously, and the diffusion coefficient of H2O can reach the minimum value of 3.78 × 10-5 cm2/s. Fe2+ with positive charges can effectively agglomerate with H2O and OH−, forming the tetrahedral and octahedral structures of Fe3O4 crystal. At this time, a small number of OH− were adsorbed at the interface of the matrix, forming the double electrical layer and affecting the deposition of agglomerated Fe3O4. Apart from this, high temperature is beneficial to inhibit scaling deposition in part. Likewise, low concentration of the dissolved Fe2+ and OH− can also lead to weaker deposition of agglomerated Fe3O4 on the outer scaling layer of the secondary side in the steam generator.

Analysis of the Accelerator-Driven System Fuel Assembly during the Steam Generator Tube Rupture Accident.

China is developing an ADS (Accelerator-Driven System) research device named the China initiative accelerator-driven system (CiADS). When performing a safety analysis of this new proposed design, the core behavior during the steam generator tube rupture (SGTR) accident has to be investigated. The purpose of our research in this paper is to investigate the impact from different heating conditions and inlet steam contents on steam bubble and coolant temperature distributions in ADS fuel assemblies during a postulated SGTR accident by performing necessary computational fluid dynamics (CFD) simulations. In this research, the open source CFD calculation software OpenFOAM, together with the two-phase VOF (Volume of Fluid) model were used to simulate the steam bubble behavior in heavy liquid metal flow. The model was validated with experimental results published in the open literature. Based on our simulation results, it can be noticed that steam bubbles will accumulate at the periphery region of fuel assemblies, and the maximum temperature in fuel assembly will not overwhelm its working limit during the postulated SGTR accident when the steam content at assembly inlet is less than 15%.