Steam Generator Tubes Research Articles

Non-uniform boiling heat transfer characteristics and calculation evaluation in U-tube steam generator tube bundle

U-tube steam generator, as one of the core equipment in nuclear power plants, has lateral non-uniform heating on the secondary side, which complicates the progress of heat transfer on the secondary side of the evaporator by causing fluid cross mixing. However, there is relatively little research on the properties of transverse non-uniform heat transfer in the secondary side tube bundle of the evaporator under natural circulation conditions with the free liquid surface. Therefore, to conduct comparative experiments and studies between uniform and non-uniform heating, an experimental setup with the natural circulation of the free liquid surface and the ability of transverse non-uniform heating was constructed, considering the characteristics mentioned above for the secondary side of the evaporator. The results indicate that higher proportions of gas-phase distribution are frequently found in the tube bundle on the high-power side, and corresponding positions usually have higher water temperature, heat transfer capacity, and heat transfer coefficient. Besides, four heat transfer correlations that are suitable for pool boiling are estimated, and the results show that the Rohsenow correlation has the best applicability to compute the heat transfer coefficient for uniform heating experiments while performing worse in non-uniform heating experiments. The main reason is the lack of consideration for the factors of fluid cross mixing in traditional correlations.

Wear work-rate measurement for different steam generator tube layouts in two-phase cross flow

Tube-to-support contact damage is directly related to the wear work-rate in steam generator tube bundles. Generally, the work-rate is related to two main factors: the contact force between the tube and the supports, and the sliding distance during this contact. This work presents an experimental study of the wear work-rate of the APR1400 steam generator tube bundle array layouts. The three array layouts, rotated square, rotated triangle and normal triangle, are tested in single phase and two-phase flow. The test apparatus was designed to accurately measure the work-rate of the three arrays. The approach followed in this work is to appropriately measure the contact force and sliding distance during the impact, which leads to an accurate calculation of the work-rate. Based on the results of this study, array layout effect on the work-rate was analyzed by statistical analysis. The measurements were performed for different flow velocities, covering a wide range to include post- and pre-critical flow velocity of each array. Results showed a substantial difference in the mean work-rate between the rotated square array and the triangular arrays. Furthermore, the larger gap size between the vibrating tube and the supports resulted into higher work-rates for all void fractions.

Numerical analysis and structural optimization of a DLC coated capacitance wire-mesh sensor for measuring liquid metal-gas two-phase distribution

The detection of liquid metal-gas two-phase distribution in lead-cooled fast reactor once steam generator tube rupture occurs is vital to the security assessment of the reactor core. In this paper, considering the high temperature, corrosiveness and high conductivity of the liquid metal, a diamond-like carbon coated capacitance wire-mesh sensor is proposed to detect the two-phase distribution in liquid metal. Based on the software COMSOL, a numerical model is firstly established that is quantificationally verified by experiment to study and optimize the performance of the coated WMS. The measurement accuracy of the sensor is investigated by ranging the AC excitation frequency and structure parameters of the system. The results show that the liquid metal has blocking effect on the small bubbles and the sensor has great measurement accuracy when the transverse and vertical interval are 2 mm while the exitation frequecy, coat thickness and electrode diameter show little effect. This research could guide the measurement in liquid metal-gas two-phase flow.

Investigation of the impact-sliding fretting behaviour of 690 alloy tube using the finite element method

Abstract Flow induced vibration inevitably leads to fretting damage behavior on the surface of steam generator tubes. Impact-sliding fretting wear indicates that alloy tube surface experiences a dynamic impact and a sliding shear behavior simultaneously. Finite element analysis was conducted to investigate the dynamic mechanical response of the Inconel 690 alloy tube, which is influenced by different impact-sliding fretting parameters under frictionless condition. Results showed that the effects of sliding frequency and amplitude on the contact stress, elastic-plastic strain and energy dissipation of the fretting interface were not directly proportional. Increasing the impact amplitude would enhance this dynamic behavior effect.

Dealloying of Monel 400 in acetate solutions: Predicting susceptibility using Thermo-Kinetic diagrams

Monel 400 steam generator tubes in CANDU® reactors are reported to be susceptible to dealloying corrosion, but the degradation mechanism and kinetics remain unclear. We construct the Thermo-Kinetic diagrams (TK diagrams) for Monel 400, and commercially pure nickel and copper alloys in 0.1 M sodium acetate solutions at 25 and 85 ℃ to predict the conditions of susceptibility to dealloying. Subsequently, coupons of Monel 400 were immersed at various pHs at both temperatures to validate the accuracy of the predictions. After 168 h, the coupons showed selective dissolution of nickel at pH=6 at 85 ℃, with a dealloying depth of ∼2.5 µm on the polished surfaces and a maximum depth of ∼60 µm on unpolished edges. The dissolution of nickel in the form of acetate complexes appears to trigger dealloying, followed by the formation of intergranular cracks. Dealloying resulted in a ligament-pore structure and nickel content of the alloy decreased from 66.5 to 17.6 wt% in the dealloyed layer. The results prove that TK diagrams are a powerful tool for predicting the corrosion mechanism of alloys under complex conditions.

Evaluations of radionuclide activity releases into environment during loss of coolant accidents using the ASTEC code in pressurized water reactors within design basis and design extension conditions

The work described in this paper was carried out within the R2CA (Reduction of Radiological Consequences of design basis and extension Accidents) project, funded in HORIZON 2020 and coordinated by IRSN (France). An increase of the level of Nuclear Power Plant (NPP) safety by consolidated and more realistic evaluations of the Radiological Consequences (RC) of Design Basis Accidents (DBA) and a strengthening of the assessments of the NPP safety levels by considering accidental situations more severe than those integrated in plant designs (i.e belonging to Design Extension Conditions domain) were the two main motivations behind this project.More specifically, the project aims at consolidating and/or refining the assessments of the radiological consequences of explicit accidental scenarios within Design Basis Accidents (DBA) and Design Extension Conditions (DEC-A conditions without significant fuel degradation) in Light Water Reactors (LWR) through the improvements of existing code predictability; the upgrading of calculation chains and methodologies; the development/refinements of models. Within the Work Package 2 of the project, coordinated by TRACTEBEL and dedicated to calculation methodologies, existing methodologies or calculation chains and simulation tools have been applied to run a first batch of calculations dealing with different reactor types: PWR (Pressurized Water Reactor), BWR (Boiling Water Reactor), VVER (Water-Water Power Reactor) and EPR (European Pressurized Reactor). Loss Of Coolant (LOCA) and Steam Generator Tube Rupture (SGTR) accidents have been selected for the exercise and bounding scenarios of the DBA and DEC-A domains have been analysed. The results of this first set of calculations will be used as a reference to quantify the gains obtained by the updated methodologies/simulation tools developed within the project.This paper describes the results of the first batch of calculations, performed with the ASTEC integral code, simulating LOCA scenarios (DBA and DEC-A categories) in a PWR 900 MWe with a focus on the predicted number of failed fuel rods and Source Term (ST) in the environment governing the RC. Limitations of the used approach are outlined, as well as the needs for further upgrading the calculation chains are proposed in the light of the improvement that was planned within the project in Work Package 3 (WP3: LOCA – Loss of Cooling Accidents), coordinated by IRSN and dedicated to the improvement of code models dealing with LOCA scenarios.

Design and optimization analysis of a new double-layer tube type heat exchanger for lead-bismuth reactors

Double-layer heat tubes have been designed to effectively reduce the occurrence of heat pipe rupture accidents. However, inter-tube thermal contact resistance can decrease heat transfer efficiency, thus hampering the heat dissipation in the primary loop system of lead-bismuth reactors. Therefore, optimizing the design of double-layer heat tubes is necessary. This work focuses on the double-layer heat exchanger of a lead-bismuth reactor and utilizes gallium-based graphene nanofluids as a thermal interface material to fill the gap between the heat tubes. Furthermore, the impact of the length, wall thickness, outer diameter, and spacing of heat tubes on the heat transfer performance of the double-layer heat exchanger with and without the nanofluids has been analyzed. The study aims to optimize the JF factor and cost-effectiveness ratio (CER). Genetic algorithms are employed to optimize and evaluate the heat transfer performance of the main heat exchanger based on the four aforementioned parameters. Consequently, a new design scheme is obtained for the double-layer heat exchanger, which increases the optimized overall heat transfer coefficient of the main heat exchanger by 5.79%, pressure drop in the primary loop by 2.32%, JF factor by 5%, and CER by 24.62%. These results demonstrate that the gallium-based graphene nanofluids can effectively enhance the heat transfer performance of the double-layer heat exchanger while reducing the likelihood of steam generator tube rupture accidents.

Safety evaluation of Multiple Steam Generator Tube rupture accident using the best estimate plus uncertainty approach

Following the disaster in Fukushima Nuclear Power Plant (NPP) in 2011, the awareness of securing the safety of NPP under extreme events has been raised. For this purpose, the concept of Design Extension Conditions (DECs) was introduced, to enhance the plant’s capability to withstand events that are more severe than Design Basis Accidents (DBA), such as Multiple Steam Generator Tube Rupture (MSGTR) accident, given its rapid progression and significant potential for large release of radioactive materials to the environment. It is worth noting that MSGTR has never occurred during the commercial operation of nuclear power plant, therefore the knowledge of consequences of the event and potential mitigation strategy is limited. In this study, a postulated MSGTR (simultaneous rupture of 5 U-tubes in a single SG) accident is analysed using the best-estimate thermal hydraulics code RELAP5/SCDAPSIM/MOD3.4. The main focus of this work is evaluation of appropriate operator mitigation actions under various inherent uncertainties. To this end, the best estimate plus uncertainty (BEPU) approach is adopted, whereby RELAP5/SCDAPSIM/MOD3.4 was coupled to the statistical analysis software, DAKOTA, to conduct the uncertainty quantification (UQ). Using BEPU analysis, an ensemble of system responses under various epistemic and aleatory uncertainties is obtained and used to validate the efficacy of operator actions in bringing the plant to a safe shutdown condition.

Comparative Analysis of Bursting Pressure Prediction Methods for Steam Generator Tube With Volume Defect

Abstract This paper compares the difference and accuracy of bursting pressure prediction based on the flow stress σf prediction method, plastic collapse prediction method, and ductile damage model prediction method in Inconel 690 steam generator tube (SGT) with volume defect. The tensile and smooth tube bursting tests determine the parameters required for the three prediction methods. The three methods predict the bursting pressures for four deep volume defects in SGT. The results are compared and analyzed with the experimental data. The results show that the ductile damage model prediction method is the best to predict the SGT bursting pressure error with volume defects simulating the structure's deformation and damage failure process.

Effect of surface hemispherical dent depth on the microstructure evolution and stress corrosion cracking of heat transfer tubes in steam generator

The deformation mechanism and SCC behavior of alloy 690TT with dents are investigated. Dents lead to stress concentration and microstructure changes, increasing stress corrosion cracking (SCC) sensitivity. The deformation mechanism at dents is slip and twinning. The dent shoulder zone experiences oxidation due to the residual tensile stress, whereas the dent bottom zone induces SCC due to the stress concentration, exhibiting a symmetrical distribution of “splayed character”. There is a critical threshold exists for dent depth that induces SCC, and the length of SCC cracks is positively correlated with dent depth in a power function.

Coupled AC2-CFD simulations for a high-pressure core melt accident scenario

Abstract Even though very unlikely to occur, severe accident scenarios in nuclear power plants have to be analyzed. During high-pressure core meltdown scenarios in a pressurized water reactor the primary circuit should fail first. Previous analyses found that a free convection flow within the vertical steam generator (SG) tubes with a simultaneous stratified gas counterflow in the hot legs could arise. This phenomenon leads to higher thermal loads on individual SG tubes which might then fail leading to a containment bypass and the release of radioactive material into the environment. Lumped parameter system codes used for safety analyses do not provide the models necessary to simulate phenomena like mixing in three-dimensional flows and could not consider local turbulence effects. Computational fluid dynamic (CFD) codes provide such capabilities but are much more computationally expensive. Coupling of a system code with a CFD code can therefore be used to simulate such phenomena. The advantages of both approaches can be maximized by splitting up the simulation domain between the codes, depending on the expected flow conditions. The system code AC2 coupled with the CFD code OpenFOAM was used to simulate part of the severe accident transient. Free convection in the hot leg and the U-tubes of the vertical SG was observed in case of high-pressure severe accident sequences. The thermal load of individual SG tubes has been estimated from the results. These loads can be used as inputs for structural-mechanical analyses to estimate which part of the primary circuit would fail first.

Analytical model of eddy current bobbin coil probe responses at support plates in steam generator and heat exchanger tubes

The steam generators (SGs) of CANadian Deuterium Uranium nuclear reactors require periodic inspections to ensure their safe operation. Eddy current (EC) testing is the primary method by which the SG tubes are inspected. Many conditions in the SG tubes affect the EC response, such as fretting, pitting, cracking, as well as tube expansion, and the presence of tubesheet and support structures. When two or more of these parameters overlap, the EC signals from degradation induced wall losses are frequently difficult to interpret because these wall loss signals are often distorted due to interference with background signal variations. In this paper, a novel analytical model that describes the impedance of a bobbin coil with a plate encircling an arbitrary number of cylindrical conductors is developed. The model is validated against finite element method modelling and experiment. This model can be used to simulate the main probe response, while inspecting SG tubes near the tubesheet and support structures and thereby, provides the potential to separate out smaller flaw responses.

An XPS and TEM study of the composition and structure of native oxides on the inner surface of as-received Ni base alloy steam generator tubes

Ni based alloys steam generator (SG) tubes in pressurized water reactors may suffer generalized corrosion and Ni release in primary water. The processes of corrosion and oxidation are strongly dependent on the surface properties. In particular, the chemical composition and structure of the native oxide films present at the surface of the SG tubes are crucial and improving the passive properties of the surface could contribute to reduce the corrosion damage and Ni release. The native oxides on the inner surface of two as-received Ni base alloy SG tubes were investigated using two main techniques, namely X-ray photoemission spectroscopy and transmission electron microscopy. The chemical elements present in the native film for the two tubes were identified and the structure and composition of the native film at the nanoscale was determined. A duplex oxide structure was identified for both samples, but with differences in the chemical composition of the inner and outer oxide layer. Chemical heterogeneities such as alumina, Ti oxides and Cr oxides were also detected; they are considered to play a role in the evolution of the surface exposed to primary water. The two techniques used allowed a multi-scale description of the surface and led to complementary results.

Full scope scaling analysis and recommendations for PWR SGTR scenarios

Scaling studies were conducted in the field of nuclear thermal-hydraulics to analyze the scaling distortions in a Steam Generator Tube Rupture (SGTR) scenario. Test 4 of the NEA/OECD/ROSA-2 Large-Scale Test Facility (LSTF) was chosen as the basis for the analysis of scaling distortions. Firstly, a post test calculation of the experiment was performed with the RELAP5 system code along with a phenomenological assessment. This model was then scaled over a wide range of sizes to investigate the possible scaling distortions by means of the Power-to-Volume Scaling Tool (PVST) which is an in-house tool developed and validated by the Universitat Politècnica de Catalunya (UPC) for scaling nodalizations. The paper demonstrates that the main scaling distortion observed in Test 4 is attributed to the lack of preservation of the L/D ratio in the SGTR assembly nozzle. This rationale is quite important in the design of break units in experimental facilities to preserve the friction and the mass flows of the reference (prototype) plant and scenario. Subsequently, the paper presents a series of guidelines for minimizing scaling distortions and reports the results obtained by applying them, the proportional scaling distortion was reduced by more than tenfold. Additionally, the paper analyzes the RELAP5 models involved in the conservation of momentum equation for the SGTR assembly nozzle, and their potential scaling distortions.

Evaluation of fracture resistance behavior of Alloy 690 material upto 1000°C

Alloy 690 is a nickel-based alloy with a nominal composition of 30% Cr and 10% Fe (wt%). It is denoted as Ni-30Cr-10Fe. This alloy is currently being investigated by various researchers for its use in nuclear industry as a possible replacement of alloy 600 steam generator tubes because of its excellent resistance to inter-granular stress corrosion cracking in water reactor environment. Alloy 690 is mainly used in vitrification melters for immobilization of high level waste in nuclear recycle plant. The temperature of water reactor environment is considerably lower compared to that of vitrification application, where temperature can reach an order of 1000 to 1100˚C. The mechanical and fracture properties of alloy 690 at the elevated temperature is important from the point of view of design and safety analysis of vitrification melter components and it is not available in literature. For this purpose, single edged notched tensile specimens have been fabricated from alloy 690 plates of 10 mm thickness. The fracture specimens have been tested in the temperature range of 25˚C to 1000˚C, which is the operating range of vitrification melters. The crack initiation toughness was obtained using modified blunting line equation in order to account for material strain hardening effect at a given temperature. From the fracture test results, it was observed that the initiation toughness is lowest at 700˚C. To understand this behavior, scanning electron microscopy was carried out on the fracture surface of the tested specimens. The fracture surface of the specimen tested at 700˚C showed presence of brittle phases in the microstructure, which are responsible for lower value of fracture toughness at this condition. The results of this work shall be useful for design and analysis of vitrification melter components.

Time-dependent change in occurrence rate of steam generator tube leak in sodium-cooled fast reactors—Phenix and BN-600

Time-dependent change in occurrence rate of steam generator tube leak in sodium-cooled fast reactors—Phenix and BN-600

Parametric analysis and its effect on machinability for Incoloy 800 with copper electrode

In Electric Spark Machining, choice of manufacturing constraints is a necessary feature. This Electrical Spark Machining is an advanced process in all manufacturing processes. For the manufacturing of sophisticated parts, which require accurate machining that time Electric Discharge Machining used. A super alloy, heat-treated tool steel, ceramics, heat resistance steels etc. which are tough to machining. For this, we use Electric Discharge Machining which also known as Electric Spark Machining. Incoloy alloy set in the group of super stainless steels. The main element of this Incoloy alloy is nickel-chromium-iron with other extra copper, molybdenum metal. High solidity at high peak temperature and super resist of erosion is the main feature of Incoloy alloy. Vast application of Incoloy 800 is like Heat exchangers, Carburizing equipments, Heating elements and nuclear steam generator tubing. In our research, we use hollow, flat, convex and concave electrode profile for machining of Incoloy 800.MRR and TWR are the parameters we analyze after the experiment.

Iodine source term assessment as result of iodine spiking and mass transfer phenomena during a SGTR transient using MELCOR 2.2 and CATHARE 2 codes

The European project denominated Reduction of Radiological Consequences of design basis and design extension Accidents (R2CA) was launched in September 2019, with a very broad participation: 11 countries, 17 participating organisations, including international organisations, utilities, regulators, technical support organisations, researchers and developers and was coordinated by IRSN. The main goal of the project was to assess the conservatisms in the radiological releases calculations in nuclear power plant (NPP) studies.The work here presented is focused on developing new calculation methodologies and updating computer code models to carry out more detailed assessments of source terms resulting from a steam generator tube rupture (SGTR) accident in the Design Extension conditions A (DEC-A) domain. For this purpose, participants in Work Package 2 (WP2) of R2CA developed and implemented simplified models in the Severe Accident (SA) and Thermal-Hydraulic (TH) codes that take into account iodine spiking and iodine transport phenomena within primary and secondary circuit. These methodologies will not only provide a better estimation of the radiological consequences, but should also serve to improve accident management procedures, innovative instrumentation development and early detection tools.The new approaches are tested in a SGTR + Steam Line Break Outside Containment scenario without significant fuel degradation in a three-loop Western 1000 MWe PWR. During the transient, operators are also assumed to implement emergency operating procedures (EOPs) in line with Westinghouse EOPs to limit the release of radioactivity and control the evolution of plant parameters.The calculations are performed by Tractebel and CIEMAT with the American SA tool MELCOR and Bel-V with the French TH code CATHARE.The results highlighted some limitations of implemented models in predicting iodine behaviour as well as small discrepancies in the TH evolution of the transient, both of which were analysed and discussed in detail.

Conjugate heat transfer characteristics numerical simulation on steam generator tube of lead-cooled fast reactor

The safety assessment of the Steam Generator Tube Rupture (SGTR) accident is a crucial step in the design process of the Lead-cooled Fast Reactor (LFR). To begin with, it is necessary to simulate the steady-state heat transfer of the LFR Heat Exchanger (HX) in order to determine the initial conditions of the SGTR accident. In this study, the heat transfer characteristics of the fluids on both sides of the LFR HX are analyzed. The appropriate models are modified based on the heat transfer characteristics of liquid metal and boiling water. The accuracy of the boiling models is verified by comparing the simulation results with the outcomes of boiling heat transfer experiments. Additionally, the simulation results are compared with the results obtained from ATHLET calculation and empirical formulas. On the secondary side, the supercooled water undergoes heating until it reaches the superheated steam phase. The preheating section spans from 0 to 600 mm, the saturated boiling section ranges from 600 to 3000 mm, and the superheating section extends from 3000 to 6600 mm. Ultimately, the distributions of steady-state temperature, pressure, and vapor volume fractions of the LFR HX are determined, providing the necessary initial conditions for the occurrence of the SGTR accident.

Performance evaluation of an improved pool scrubbing system for thermally-induced steam generator tube rupture accident in OPR1000

An improved mitigation system for thermally-induced steam generator tube rupture accidents was introduced to prevent direct environmental release of fission products bypassing the containment in the OPR1000. This involves injecting bypassed steam into the containment, cooling, and decontaminating it using a water coolant tank. To evaluate its performance, a severe accident analysis was performed using the MELCOR 2.2 code for OPR1000. Simulation results show that the proposed system sufficiently prevented the release of radioactive nuclides (RNs) into the environment via containment injection. The pool scrubbing system effectively decontaminated the injected RN and consequently reduced the aerosol mass in the containment atmosphere. However, the decay heat of the collected RNs causes re-vaporization. To restrict the re-vaporization, an external water source was considered, where the decontamination performance was significantly improved, and the RNs were effectively isolated. However, due to the continuous evaporation of the feed water caused by decay heat, a substantial amount of steam is released into the containment. Despite the slight pressurization inside the containment by the injected and evaporated steam, the steam decreased the hydrogen mole fraction, thereby reducing the possibility of ignition.