Pressurized Water Reactor Components Research Articles

Determining the diffusion behavior of point defects in zirconium by a multiscale modelling approach

Zr alloys are commonly used materials for in-core components of pressurized water reactors. The evolution of defects in Zr under irradiation conditions will directly contribute to the degradation of reactor components. For a better understanding of the evolution process of defects in Zr, a multiscale simulation was performed to study the diffusion behavior of radiation-induced point defects. Because the existing potentials cannot accurately describe the energy differences of stable states self-interstitial atoms (SIAs) in Zr, an embedded atom method potential that can exactly reproduce the energy difference between metastable states and ground states, and the binding energy of divacancy of Zr, was first developed by fitting a potential to point defect properties calculated by density functional theory and other basic crystal properties. Based on the constructed potential, molecular dynamics (MD) and kinetic Monte Carlo simulations were conducted to investigate the migration of point defects. We determined several frequent migration paths of SIAs by kinetic study, which is rarely reported in previous MD studies. The SIA exhibits obvious anisotropic diffusion characteristics at low temperatures (<600 K). The most frequent migration path for SIAs is the jump between the two nearest basal octahedral (BO) sites in the basal plane, which means that SIAs diffuse faster along the basal plane than along the c axis. It is found that diffusion within the basal plane tends to be two-dimensional (2D) diffusion rather than 1D diffusion, and there is a significant correlation effect for diffusion of SIAs along the basal plane. Additionally, the existence of trivacancy-SIA complex was found in the process of divacancy migration, which can inhibit divacancy migration. Monovacancies and divacancies exhibit anisotropic diffusion characteristics in the considered temperature range Divacancies have a much faster diffusion rate than monovacancies in the present MD simulation and can easily dissociate at high temperatures (>900 K). The rapid migration of the divacancies may contribute to the formation of vacancy dislocation loops. These results are meaningful to understand the evolution process of radiation-induced defects.

Hydrogen trapping by irradiation-induced defects in 316 Lstainless steel: A combined experimental and modeling study

The irradiation-induced defects in stainless steel internal components of pressurized water reactors combined with hydrogen uptake during the oxidation process could be a key parameter in the mechanism for Irradiation-Assisted Stress Corrosion Cracking (IASCC). A heat-treated 316 L SS containing a low amount of defects was Fe3+ ions-implanted; irradiation-induced defects types and depth distributions were characterized by Transmission electron microscopy (TEM). Deuterium was then inserted in the specimens either by cathodic charging or by sample exposure to deuterated primary water. Secondary Ions Mass Spectrometry – SIMS – permitted to access the deuterium distribution at the implanted stainless steel surface. A finite difference numerical solver accounting for hydrogen diffusion/trapping coupling was used to simulate the hydrogen transport in the implanted material, taken into consideration the specific heterogeneous character of the irradiation-induced defects distribution in matter. Taking as input data the experimental defects distribution associated with Frank loops or voids, the main trapping sites for hydrogen were assigned to voids, not Frank loops. Such numerical approach was able to deal accurately with the problem of hydrogen transport in a heterogeneous material as well as to differentiate two potential trap sites contributions to the experimental deuterium distribution. In addition, according to results obtained after primary water exposure, trapping at voids was still effective at 320 °C, signature of a high binding energy of hydrogen in voids.

Measurements of Liquid AISI 304L Steel Density and Surface Tension, and Influence of Surface‐Active Elements at High Temperatures

Within the framework of the welding of steel components in pressurized water reactors, the optimization of the weld pool penetration is a major issue. The thermal‐hydraulics of the molten steel pool can be driven by Marangoni flows at the free surface. To compute these flows, liquid steel's thermophysical properties such as density ρ and surface tension σ are required. Such properties for austenitic vessel steel AISI 304L at liquid state are scarce, and even not available above 1800 °C, due to technological issues. The viscometer temperature installation (VITI) is designed to implement sessile droplet (SD) and maximum bubble pressure (MBP) techniques. These techniques allow to determine original ρ and σ data for two grades of AISI 304L steel, namely 304LP1 and 304LP4—low and high sulfur content, at various temperatures. In particular, thanks to the innovative design of MBP configuration, original surface tension data are obtained from the melting point up to 2100 °C. For the first time, experimental evidence of the Sahoo et al.'s (1988) model at high temperature is provided, emphasizing the decisive role played by surface‐active elements such as sulfur and oxygen on surface tension, whose content evolution is assessed by dedicated chemical analyses.

Cyclic Ramberg–Osgood Parameters of 316SS Weld Metal Under In-Air and PWR-Water Environments

Abstract At present, the available Ramberg–Osgood (R–O) parameters for different metals (e.g., in ASME code and other literatures) are static (generally based on a tensile curve). These static R–O parameters cannot accurately model the cyclic plasticity behavior. This work presents the cyclic R–O material hardening parameters for 316 stainless steel similar metal welds. The parameters were estimated under various conditions (in-air at room temperature, 300 °C in-air, and in-air at primary water conditions for a pressurized water reactor (PWR)). It is anticipated that the reported results would be useful for computational mechanics-based shakedown analysis and fatigue life estimation of PWR components.

Long Term Stability Evaluation of Stress Improvement Effect on Ni-Based Alloy Welds by Various Residual Stress Improvement Methods

Stress corrosion cracking (SCC) is one of serious aging degradation problems for the Alloy 600 components of pressurized water reactors (PWRs). In order to prevent SCC, various methods such as water jet peening (WJP), laser peening (LP), surface polishing have been used to introduce compressive stresses at the surfaces of the PWR components. However, it has been reported that such compressive residual stress introduced by these methods might be relaxed during the practical operation, because of high temperature environment. In this study, the hardness reduction behavior of the Alloy 600 processed by LP, Buff and WJP in the thermal aging process has been investigated to estimate the stability of the residual stress improving effect by each method, based on the fact that there is a correlation between the compressive residual stress relaxation and the decrease of hardness. The behavior of the residual stress relaxation in the processed materials in the high temperature environment has been discussed with kinetic analysis.

Evaluation of single heated channel and subchannel modeling of a nuclear once through steam generator (OTSG)

Abstract Steam generators are one of the most important components of pressurized-water reactors. This component plays the role of heat transfer and pressure boundary between primary and secondary side fluids. The Once Through Steam Generator (OTSG) is an essential component of the integrated nuclear power system. In this paper, steady-state analysis of primary and secondary fluids in the Integral Economizer Once Through Steam Generator (IEOTSG) have been presented by Single Heated Channel (SHC) and subchannel modelling. Models have been programmed by MATLAB and FORTRAN. First, SHC model has been used for this purpose (changes are considered only in the axial direction in this model). Second, the subchannel approach that considers changes in the axial and also radial directions has been applied. Results have been compared with Babcock and Wilcox (B&amp;W) 19- tube once through steam generator experimental data. Thermal- hydraulic profiles have been presented for steam generator using both of models. Accuracy and simplicity of SHC model and importance of localization of thermal-hydraulic profiles in subchannel approach have been proved.

PWR effect on crack initiation under equi-biaxial loading development of the experiment

The lifetime extension of the nuclear power stations is considered as an energy challenge worldwide. That is why, the risk analysis and the study of various effects of different factors that could potentially represent a hazard to a safe long term operation are necessary. The methodology for fatigue dimensions of the Pressurized Water Reactor components (PWR) is based on the use of design curves established from test carried out in air at 20 °C on smooth specimens by integrating safety coefficient that covers the dispersion of tests associated with the effects of structures. To formally integrate these effects, some international codes have already proposed and suggested a modification of the austenitic stainless steels fatigue curve combined with a calculation of an environmental penalty factor, namely Fen, which has to be multiplied by the usual fatigue usage factor. The aim of this paper is to present a new device “FABIME2E” developed in the CEA-LISN in collaboration with EDF and AREVA. These new tests allow quantifying accurately the effect of PWR environment on semi-structure specimen. This new device combines the structural effect like equi-biaxiality and mean strain and the environmental penalty effect with the use of PWR environment during the fatigue tests.

Effect of Si-Addition in Stainless Steel on Surface Oxidation in High Temperature Water Environments

Being used as the internal components of pressurized water reactor (PWR), austenitic stainless steels (SSs) are susceptible to irradiation-assisted stress corrosion cracking (IASCC) that is a critical issue for the long term safe operation. Radiation induced segregation (RIS) at grain boundaries in austenitic SSs, such as the depletion of Cr and enrichment of Ni and Si, is thought to contribute to IASCC. Studying the oxidation behavior of the model alloys simulating grain boundary RIS in high temperature water could provide evidence for the effect of RIS on IASCC. In this study, the model alloys 9Cr-25Ni-1.5Mo-5Si and 9Cr-25Ni-1.5Mo-3Si were used to simulate the depletion of Cr and enrichment of Ni and Si at grain boundaries of irradiated 316L SS. The oxide film properties of 316L SS and model alloys in of high temperature water environments were investigated. After the immersion in high-temperature water environments, the surface morphologies of the oxide films formed on the immersed specimens were examined using JEOL JSM-7500F field emission scanning electron microscope (FESEM). The oxide films were analyzed using INVIA laser Raman spectrometer with laser wavelength of 514.5 nm. The compositions of oxide particles were analyzed by the energy-dispersive spectrometry (EDS) attached to the FESEM. Fig.1 shows the SEM morphologies of the 316L SS, 9Cr-25Ni-1.5Mo-3Si and 9Cr-25Ni-1.5Mo-5Si alloy surfaces after immersion in pure water at 290 oC (pH at 290 oC of about 5.6) for 120h. The surfaces of all the test specimens were covered with polyhedral particles of different sizes. Larger size outer layer oxides are formed on 9Cr-25Ni-1.5Mo-3Si and 9Cr-25Ni-1.5Mo-5Si specimen surface. On the surface of 316L SS, there were only sparsely distributed large-size particles. Si addition in the alloy matrix favors the formation of the oxide layer in this environment. Fig. 2 shows the SEM morphologies of 316L SS, 9Cr-25Ni-1.5Mo-5Si and 9Cr-25Ni-1.5Mo-3Si alloys surfaces after the immersion in high temperature water of pH 9.5 (pH adjusted with monoethanolamine) at 290 oC for 24h, 48h and 120h. Compact and homogeneous oxide particles were formed on the surfaces of 316L SS and model alloys. The size of the oxide particles on model alloys was smaller than that on 316L SS. The oxide particle size of 9Cr-25Ni-1.5Mo-5Si and 9Cr-25Ni-1.5Mo-3Si was similar. The morphologies of particles became more stereoscopic with increasing immersion time. The outer layer oxide particles formed on the 316L SS and model alloys decreased with increasing solution pH value. The pH value has a significant influence on the oxidation of 316L SS and model alloys. These results show that the effect of Si in the alloy on oxide film properties is strongly dependent on the water chemistry. Si in the alloy facilitated the formation of large-sized outer layer in 290 oC water of pH 5.6, which refined the outer layer oxide particles in 290 oC water of pH 9.5. High solubility of Si-bearing oxide in high temperature water can modify the oxide film on the surface. These results show that Si in the alloy would play a role in IASCC through the modification of oxide film. Figure 1

PWR EFFECT ON CRACK INITIATION UNDER EQUI-BIAXIAL LOADING: DEVELOPMENT OF THE EXPERIMENT

The lifetime extension of the nuclear power stations is considered as an energy challenge worldwide. That is why, the risk analysis and the study of various effects of different factors that could potentially represent a hazard to a safe long term operation are necessary. These structures, often of great dimensions, are subjected during their life to complex loading combining varying mechanical loads, multiaxial, with non-zero mean values associated with temperature fluctuations and also PWR environment.Historically, the methodology for fatigue dimensions of the Pressurized Water Reactor components (PWR) (ASME, RCC-M, KTA, …) is based on the use of design curves established from tests carried out in air at 20°C on smooth specimens by integrating safety coefficients that cover, among other parameters, the dispersion of tests associated with the effects of structures.Based on more recent fatigue data (including tests at 300°C in air and PWR environment, etc…), some international codes (RCC-M, ASME and others) have proposed and suggested a modification of the austenitic stainless steels fatigue curve combined with a calculation of an environmental penalty factor, namely Fen, which has to be multiplied by the usual fatigue usage factor.The aim of this paper is to present a new device "FABIME2E" developed in the LISN in collaboration with EDF and AREVA. These new tests allow quantifying accurately the effect of PWR environment on semi-structure specimen. This new device combines the structural effect like equi-biaxiality and mean strain and the environmental penalty effect with the use of PWR environment during the fatigue tests.

Helium Effects on 316L Austenitic Stainless Steel Fracture Mechanism

The most common fracture mechanism of nuclear reactor internals is irradiation-assisted stress corrosion cracking (IASCC). Its susceptibility at relatively low dose is dominated by conventional mechanisms such as radiation-induced segregation and radiation hardening. However, the aging of the nuclear fleet combined with the increase of their life-span reveals other mechanisms that could play an important role on IASCC susceptibility. A large amount of helium (He) can be accumulated in reactor internal components of pressurized water reactors (PWR) after long term operation. This occurrence could significantly increase (or even dominate) the IASCC susceptibility at high doses. He has been homogeneously implanted in an especially designed miniaturized specimen at 300°C up to 1000 appm. Slow strain rate tests (SSRT) results in high temperature air and in simulated PWR conditions indicate that homogenized, as-implanted He does not have a significant effect on IASCC up to 1000 appm under these test conditions.

Influence of irradiation on stainless steel corrosion in PWR primary conditions

Some internal components of pressurized water reactor (PWR) are made on stainless steel. They are exposed to: a corrosive environment which is the primary medium that is to say water at high temperature (290°C-330°C) and high pressure (155 bars) with boric acid (H3BO3) , lithium hydroxide (LiOH) and dissolved hydrogen, a mechanical stress, an irradiation process due to neutron flux which can cause up to 1 displacement per atom (dpa) by year. In these conditions, these materials can be damaged by a process of IASCC (Irradiation Assisted Stress Corrosion Cracking). This degradation process has been observed since the eighties with the formation of cracks on baffle bolts under operating conditions. Oxide film rupture is the first step in this corrosion process and thereby the properties of the oxide formed on stainless steels are key parameters relative to SCC initiation. The morphology and the structure of the oxide formed on stainless steel in PWR primary medium have been detailed by several authors [1, 2, 3]. The oxide layer in divided in two parts: an internal and continuous layer rich in chromium and an external discontinuous layer composed of crystallites rich in iron. Moreover a nickel enrichment has been observed in the alloy under the oxide close to the alloy/oxide interface. In this alloy layer with a higher composition in nickel, a trapping of hydrogen has also been observed [4]. The aim of this work is to study the influence of defects created in the alloy by irradiation on the oxide properties (thickness, morphology, composition) and on oxygen transport in this oxide scale during exposition of stainless steel in PWR primary medium. In order to simulate the defects due to irradiation, ionic implantation have been performed. So samples of austenitic stainless steel (316L) have been implanted with Xenon ions at energy of 240 keV. Transmission Electron Microscopy (TEM) observations indicate an affected depth of about 80 nm with the formation of small cavities. The implanted samples have been oxidized at 325°C in hydrogenated primary water which simulates the primary medium in different devices, static autoclave and corrosion loop during different durations. The oxide film properties have been investigated by a set of techniques (SEM, TEM, SIMS) .As shown on Fig. 1, which presents implanted sample before and after 600 hours of oxidation, it has been observed a correlation between the affected area created by implantation and the morphology of the oxide formed after corrosion test. The composition of the internal layer is also linked to the defects created by implantation.

High Pressure Boiling Water Reactor

Some 400 boiling water reactors (BWRs) and pressurized water reactors (PWRs) have been in operation for several decades. The presented concept, the high pressure boiling water reactor (HP-BWR), combines the best parts and omits the troublesome components of traditional BWRs and PWRs by taking into consideration the experiences gained during their operation.One of the major benefits of the HP-BWR is that safety is improved. The design utilizes gravity-operated control rods, and there is a large space for the cross-formed control rods between fuel boxes. The bottom of the reactor vessel is smooth and without penetrations. All the pipe connections to the reactor vessel are well above the top of the reactor core, and core spray is not needed. Additionally, internal circulation pumps are used.The HP-BWR concept is also environmentally friendly: Improved thermal efficiency is achieved by feeding the turbine with ~340°C (15 MPa) steam instead of ~285°C (7 MPa), and there is less warm water release to the recipient and less uranium consumption per produced kWh, resulting in the production of less waste.Finally, the HP-BWR is cost effective and simple, operating in direct cycle mode with no need for complicated steam generators. Moisture separators and steam dryers are placed inside the reactor vessel, and additional separators and dryers can be installed inside or outside the containment. Well-proved simple dry containment or wet containment can be used.In more than half a century, an extensive regulatory licensing experience has been built from traditional BWRs and PWRs. The HP-BWR is a developed, high-performance successor of those conventional designs. Therefore, it can be expected that licensing can be accomplished in a reasonable time.Several utilities are supporting manufacturers to study concepts for future reactors. It is likely that an application to one or more electrical power companies for financial support by a manufacturer to make a detailed feasibility study of the HP-BWR would be positively treated. This could be the next step to the implementation of the HP-BWR.

Behaviour of the steam generator tubing in water with different pH values

Steam generators are crucial components of pressurized water reactors. The failure of the steam generator as a result of tube degradation by corrosion has been a major cause of Pressurized Heavy Water Reactor (PHWR) plant unavailability. Steam generator problems have caused major economic losses in terms of lost electricity production through forced unit outages and, in cases of extreme damage, as additional direct cost for large-scale repair or replacement of steam generators. The excellent performance to date of CANDU steam generators can be attributed, in part, to their design and performance characteristics, which typically involve higher recirculation ratios and lower heat fluxes and temperatures. However, the steam generator tubes are susceptible to failure by a variety of mechanisms, the vast majority of which are related to corrosion. The generalized corrosion is an undesirable process because it is accompanied by deposition of the corrosion products which affect the steam generator performances. It is very important to understand the generalized corrosion mechanism with the purpose of evaluating the quantities of corrosion products which exist in the steam generator after a determined period of operation ( IAEA, 1997). The purpose of the experimental research consists in the assessment of corrosion behaviour of the tubes material, Incoloy-800, at normal secondary circuit parameters (temperature—260 °C, pressure—5.1 MPa). The testing environment was the demineralised water without impurities, at different pH values regulated with morpholine and cyclohexylamine (all volatile treatment—AVT). The results are presented like micrographics and graphics representing weight loss of metal due to corrosion, corrosion rate, total corrosion products formed, the adherent corrosion products, released corrosion products, release rate of corrosion products and release rate of the metal.

Prediction of swelling of 18Cr10NiTi austenitic steel over a wide range of displacement rates

The internal components of pressurized water reactors of Russian types WWER-440 and WWER-1000 are constructed of annealed 18Cr10NiTi steel, a close analog to AISI 321. Void swelling of the internals is a concern for plant life extension and predictive equations are required to assess the potential of swelling in critical components such as the baffle ring or reflection shield that surrounds the WWER core. The only previously available swelling data for this steel were derived at higher than PWR-relevant displacement rates in the BOR-60 fast reactor. The swelling equation previously developed from these data does not incorporate the effect of displacement rate on swelling. Using heavy-ion irradiation at very high dpa rates (10 � 2 and 10 � 3 dpa s � 1 ) and doses (5–100 dpa) and coupling the results to available neutron data a swelling equation has been developed that specifically incorporates the effect of dpa rate on void swelling. Experimental results allow description of the swelling peak, the incubation period and the steady-state swelling rate over a wide range of irradiation temperature. For the first time it appears possible to describe both ion and neutron data on this steel within the framework of a single empirical model. Swelling maps constructed from this model permit forecasting of the behavior of the steel in WWERs under the required irradiation conditions, not only at already attained exposure doses, but more importantly to higher dose levels that will be reached following plant life extension.

Application of the miniature specimen technique to material irradiation tests and surveillance for reactor components

Yield strength, tensile strength and uniform elongation of unirradiated and irradiated reactor pressure vessel steels were determined by the modified miniature specimen test (MMST). The facility and conditions for irradiating the miniature specimens are described. The results were compared with conventional tensile test data. Empirical formulas for correcting result errors caused by the deviation of specimen thickness were established. The application of the MMST technique to radiation surveillance for pressurized water reactor components is addressed.

Experience d'electricite de France dans les techniques de surveillance des reacteurs a eau sous pression

Different techniques are used for E.D.F. pressurized water reactor components and systems monitoring. The surveillance concenrs mainly : mechanical behaviour of reactor internals, loose parts monitoring, sensors monitoring. Some results are commented. Conditions necessary for an efficient surveillance are indicated. Objectives of monitoring studies under development are presented.

Heat Transfer in Pressurized Water Reactor Components Most Often Subject to Thermal Shock

Experimental studies have been performed to quantitatively analyze heat transfer in the three pressurized water reactor (PWR) cooling system components subject to the greatest thermal loads (i.e., charging, safety, and pressurizer nozzles). Flows are so complex in these components that thermal boundary layer theory cannot be used to predict heat transfer analytically. Heat transfer coefficients and fluid temperatures have been determined for all possible flow conditions during PWR operation.