Light Water Reactor Conditions Research Articles

Enhanced Mechanical Properties and Wear Resistance of FeCrAl Alloys at ~ 300°C and Higher Temperatures

Worldwide, light water reactors (LWRs) have been using zirconium (Zr)-based alloys for the cladding of the uranium dioxide fuel for more than 6 decades. Zr alloys oxidize rapidly in the presence of water and steam at temperatures > 450°C; therefore, they do not respond well to scenarios of loss of coolant accidents. There is a global effort by nuclear materials technologists to find more robust or stronger cladding materials for LWRs. One option is to use an iron-chromium-aluminum (FeCrAl) alloy since they have excellent resistance to high temperature oxidation and superior mechanical properties at LWR operation temperatures. Results show that (1) FeCrAl alloys have better mechanical properties than Zr alloy and are orders of magnitude more resistant to creep at temperatures higher than LWR normal operation conditions. (2) FeCrAl alloys have better resistance to fretting wear than Zr alloys at the normal operation conditions of LWRs.

Stress-corrosion cracking sensitization by hydrogen upon oxidation of nickel-base alloys by water – An experiment-guided first-principles study

This work describes a viable hydrogen-induced sensitisation process toward stress corrosion cracking in chromia-forming nickel base alloys owing to oxidation by water. A mechanistic chemical understanding of the stress corrosion based on repeated cracking and re-healing of the oxide scale emerges from experiment-guided first-principles calculations. Piggybacking processes during repeated oxide scale cracking-healing cycles are understood to cause sensitisation towards the formation and growth of macroscopic cracks. Under light-water reactor conditions, the healed oxide scale comprises an outer region composed of chromium depleted nickel ferrite, and non-protective chromia, nickel oxide, and nickel hydroxide. An inner nickel iron chromite layer provides the passivating barrier that controls the oxidation rate at a steady state. Early during the re-healing of a crack in the oxide scale, water is conveyed to the alloy/scale interface by hydration/dehydration of composite nickel oxy-hydroxide Ni(OH)2∙NiO inclusions in the scale. At later stages, Ni(OH)2∙NiO serves as oxidant of preferentially chromium, while hydrogen is released as H2(g) into the primary water or becomes picked up by the alloy, possibly while assisting in the reduction of NiO to form nickel metal particles. Oxidation by water equivalents at the confining alloy/oxide scale interface favours hydrogen pick-up in the alloy that becomes increasingly detrimental owing to enrichment and inward diffusion of hydrogen along alloy grain boundaries. The pinning of alloy vacancies by hydrogen causes mitigation of outward diffusion of chromium, which in turn hampers crack re-healing while promoting alloy embrittlement.

Oxide Layers in Ni-doped FeCrAl Alloy in 320°C Radioactive Hydrogenated Water

Iron-chromium-aluminum (FeCrAl) alloys have been the interest of the international nuclear energy industry as an accident tolerant fuel cladding replacement to the Zircaloy cladding alloys due to their enhanced corrosion resistance in both normal operation and accident conditions in light water reactors. It is important to understand the corrosion behaviors of FeCrAl before deploying it to nuclear power plants. This study aims to investigate if the addition of a small amount of nickel (Ni) could stabilize the oxide layer by forming the Ni-bearing spinel oxide, such as NiFe2O4, and increase the stability of the oxide layer under irradiation conditions. Current results show that the outer oxide crystals formed on the irradiated FeCrAl-Ni sample surface became larger than that on the FeCrAl sample, which indicates that the Ni addition facilitates the growth of oxide layers and that the FeCrAl-Ni sample exhibited a protective inner oxide layer in the harshest irradiation environment, indicating that Ni improves the corrosion resistance of FeCrAl in radioactive hydrogenated water.

Initial Verification and Validation of a New CASMO5 JENDL-5 Nuclear Data Library for Typical LWR Applications

A new nuclear data library for the advanced lattice physics code CASMO5 has been prepared based on JENDL-5. In JENDL-5, the range of data, such as the number of nuclides, has been dramatically expanded for general-purpose applications. At the same time, many essential nuclides for conventional light water reactor (LWR) analysis have also been modified based on state-of-the-art evaluations. The new JENDL-5–based CASMO5 library was prepared by replacing as much of the nuclear data of the current CASMO5 ENDF/B-VII.1–based library as possible with JENDL-5. This study performs initial verification and validation of the new library under typical LWR conditions. Verifications were performed based on the Organisation for Economic Co-operation and Development/Nuclear Energy Agency burnup credit criticality safety benchmark phase III-C, and the calculated kinf and fuel compositions of the boiling water reactor fuel assembly were compared with reported benchmark results. Comparison with the MCNP6.2 result was also performed using the same benchmark model. In addition, a tank-type critical assembly critical experiment and Takahama-3 postirradiation experiment were used for validation. The results indicate that the new library performs well and is comparable to the ENDF/B-VII.1–based library in predictions of reactivity and fuel compositions for typical LWR systems.

FeCrAl fuel/clad chemical interaction in light water reactor environments

This article investigates the fuel-cladding chemical interaction (FCCI) behavior of two commercial FeCrAl alloys, APMT composition (Fe-21Cr-5Al-3Mo wt.%) and C35M (Fe-13Cr-5Al-2Mo-0.2Si-0.03Y wt.%), after neutron irradiation. “H-cup” diffusion multiples of FeCrAl alloys and ceramic UO2 fuel were irradiated at a temperature of ∼300 °C to a total estimated burnup of 26 GWd/tHM. Post-irradiation Examination results demonstrate the excellent degradation resistance of FeCrAl alloys as accident tolerant fuel (ATF) cladding materials in light water reactor conditions. The study concludes that there was no irradiation-induced defects observed in either of the two commercial FeCrAl claddings. The formation of amorphous Al/U mixed oxide was observed at the fuel-clad interface, which can serve as a tritium permeation barrier and protect against potential chemical attack from the fuel. The study attributed the formation of amorphous Al/U mixed oxide to the low temperature and limited time of neutron irradiation. APMT forms more distinct Cr and Cr-Fe intermetallic at the FeCrAl-UO2 interface than C35M due to the higher bulk Cr:Al ratio.

Behavior of FeCrAl-ODS cladding tube under loss-of-coolant accident conditions

To evaluate the behavior of an oxide-dispersion-strengthened FeCrAl (FeCrAl-ODS) cladding tube under loss-of-coolant accident (LOCA) conditions of light-water reactors (LWRs), the following two laboratory-scale LOCA-simulated tests were performed: the burst and integral thermal shock tests. Four burst and three integral thermal shock tests were performed on unirradiated, stress-relieved FeCrAl-ODS cladding tube specimens, simulating ballooning and rupture, oxidation, and quenching, which were postulated during a LOCA. The burst temperature of the FeCrAl-ODS cladding tube was 200–▪ higher than that of the Zircaloy cladding tube, and the FeCrAl-ODS cladding tube's maximum circumferential strain was smaller than or equal to that of the Zircaloy-4 cladding tube. These results indicate that the FeCrAl-ODS cladding tube has higher strength at high temperatures than the conventional Zircaloy cladding tube. The FeCrAl-ODS cladding tube did not fracture after being subjected to an axial restraint load of ∼▪, which is more than 10 times higher than the axial restraint load estimated for existing LWRs, during quenching, following isothermal oxidation at ▪ for ▪. The FeCrAl-ODS cladding tube was hardly oxidized during this isothermal oxidation condition. However, it melted after a short oxidation at ▪ and fractured after abnormal oxidation at ▪ for ▪. Based on these results, the FeCrAl-ODS cladding tube should not fracture in the time range expected during LOCAs below ▪, where no melting or abnormal oxidation occurs.

Effect of zinc injection on mitigating stress corrosion cracking initiation of structural materials in light water reactor primary water

Abstract The effect of zinc (Zn) injection on the stress corrosion cracking (SCC) initiation of an Alloy 182 weld metal and a 20% cold-worked (CW) type 316L stainless steel was investigated under simulated light water reactor conditions. Accelerated SCC initiation tests revealed that under both, boiling (BWR) and pressurized water reactor (PWR) conditions without Zn injection, Alloy 182 showed a higher SCC initiation susceptibility compared to the CW 316L stainless steel. With 40 ppb Zn injection, results indicate a tendency towards higher stress thresholds or lower crack densities for SCC initiation in both materials. In other words, Zn water chemistry seems to be able to mitigate SCC initiation. The optimized, Zn-treated oxide film is believed to be responsible for the improved SCC performance. However, no clear mitigation effects were observed when the specimens were pre-exposed to high-temperature water without Zn and then tested with Zn, which is likely related to the relatively short exposure time to the Zn-containing high-temperature water.

Physics-based model of irradiation creep for ferritic materials under fusion energy operation conditions

Irradiation creep is known to be an important process for structural materials in nuclear environments, potentially leading to creep failure at temperatures where thermal creep is generally negligible. While there is a great deal of data for irradiation creep in steels and zirconium alloys in light water reactor conditions, much less is known for first wall materials under fusion energy conditions. Lacking suitable fusion neutron sources for detailed experimentation, modeling, and simulation can help bridge the dose-rate and spectral-effects gap and produce quantifiable expectations for creep deformation of first wall materials under standard fusion conditions. In this paper, we develop a comprehensive model for irradiation creep created from merging a crystal plasticity representation of the dislocation microstructure and a defect evolution simulator that accounts for the entire cluster dimensionality space. Both approaches are linked by way of a climb velocity that captures dislocation-biased defect absorption and a dislocation strengthening term that reflects the accumulation of defect clusters in the system. We carry out our study in Fe under first wall fusion reactor conditions, characterized by a fusion neutron spectrum with average recoil energies of 20 keV and a damage dose rate of ≈3×10−7 dpa/s at temperatures between 300 and 800 K.

Wettability and CHF limits of Accident-Tolerant nuclear fuel cladding materials in light water reactor conditions

We present the results of experimental investigations aimed at evaluating the thermal–hydraulic performance of chromium-coated zircaloy, i.e., one of the most promising accident tolerant fuel (ATF) cladding material for light water nuclear reactors. Precisely, we investigate the wettability and critical heat flux (CHF) limits of chromium-coated and conventional zircaloy surfaces in prototypical reactor conditions. For both surface types, we measure the contact angle in a vapor-saturated atmosphere from ambient pressure to the operating pressure of pressurized water reactors (PWRs), i.e., ∼15 MPa. We measure the ambient-pressure steady-state flow boiling CHF with a spatially uniform heat flux. We measure the high-pressure steady-state flow boiling CHF with a cosine shape heat flux (up to 20 MPa) and with a uniform heat flux (up to 15 MPa), also exploring the effect of CRUD deposits on the chromium-coated surface. Our results reveal that the chromium surface and the bare zircaloy surface have similar wettability and both become super-hydrophilic in PWR conditions, and that there is practically no difference in the steady-state CHF limits, both at low-pressure and high-pressure conditions, also when the chromium-coated surface is covered by a CRUD deposit. However, while the chromium-coating does not improve the CHF compared to the bare zircaloy surface, it improves the post-CHF behavior. The chromium coating prevents the reaction between zircaloy and steam, which results in the formation of a brittle zirconium oxide through the surface of the cladding. We also measure the transient CHF under exponentially escalating heat flux inputs of a nano-smooth and a rough surface mimicking a chromium-coated zircaloy cladding. Interestingly, the results of the transient heat flux tests suggest that the CHF limit for very short periods (i.e., fast transients) is independent of the surface finish, being the same for a rough chromium surface or a nano-smooth surface.

Dry in-pile fracture test (DRIFT) for separate-effects validation of ceramic fuel fracture models

Fracture is an important component of nuclear fuel behavior, and significant efforts have been invested into developing fuel performance models that are capable of accurately representing fracture. Usable data on the process of fracture propagation in nuclear fuel under realistic service conditions are very limited. To address this need, a series of separate-effects experiments were developed and performed at Idaho National Laboratory's Transient Reactor Test (TREAT) facility. These experiments employ a heat sink to radially remove heat from the fuel in a manner that approximates the effect of coolant in an operating light-water reactor (LWR). The test holder for these experiments is known as the Dry In-pile Fracture Test (DRIFT). A series of experiments employing DRIFT and TREAT were performed to provide data on the extent and nature of fracture in fresh fuel at various points during a ramp to full power. Novel aspects of these experiments include the way they employ a heat sink to replicate steady-state LWR conditions, as well as the use of fiber optic sensors for in-reactor thermal instrumentation. Details on the development of this experiment, experimental conditions, and resulting data (including in situ thermal measurements and post-irradiation imaging of fracture) are provided in this paper. LWR-equivalent powers ranging from 10 to 25 kW/m were tested using this apparatus. Cracking was visible at all power levels, with increasing cracking extent as the power level increased, although there was little difference in the cracking between the two highest-power tests, which had LWR-equivalent powers of 20 and 25 kW/m.

Investigation on the interphase drag and wall friction in vertically oriented upward and downward two-phase flows under accident conditions in light water reactors

In this study, applicability of the one-dimensional two-fluid model for predicting the vapor void fraction and the frictional pressure drop in the adiabatic upward and downward two-phase flows has been studied through comparing model predictions with experimental data of Tikhonenko (1973) taken in the 33 mm diameter pipe at system pressure ranging from 2 to 9 MPa, and rated velocity of 0.2 to 2 m/s (mass flux of 170 to 1400 kg⋅m−2 s−1). The experiments reproduce the conditions of the two-phase coolant flow which are expected to occur in the light-water-reactor systems under loss-of-coolant accidents and design extension conditions. It has been found that the existing empirical models for interfacial and wall friction used in the current thermal–hydraulic simulation codes allow adequate prediction of the vapor void fraction and the frictional pressure drop in vertical pipes only for the upward direction of two phase flow. Modeling deficiencies have been identified when these models are applied to the low velocity downward flow. For the downward two-phase flow at a rated velocity of 0.2 m/s (mass flux of 170 kg⋅m−2 s−1), the Lockhart-Martinelli model underpredicts the frictional pressure gradient of the liquid phase which leads to the over-prediction of the average velocity of the liquid phase. As a result, the calculated void fraction is higher compared to the experiment. The possible ways to improve the current models of the wall friction in two-phase flows have been discussed.

Wettability and Chf Limits of Accident-Tolerant Nuclear Fuel Cladding Materials in Light Water Reactor Conditions

Wettability and Chf Limits of Accident-Tolerant Nuclear Fuel Cladding Materials in Light Water Reactor Conditions

The accommodation of lithium in bulk ZrO2

Lithium is known to accelerate the corrosion of zirconium alloys in light water reactor conditions. Identifying the mechanism by which this occurs will allow alloying additions and alternative coolant chemistries to be proposed with the aim of improved performance. Accommodation mechanisms for Li in bulk ZrO2 were investigated using density functional theory (DFT). Defects including oxygen and zirconium vacancies along with lithium, zirconium and oxygen interstitials and several small clusters were modelled. Predicted formation energies were used to construct Brouwer diagrams. These show how competing defect species concentrations change across the monoclinic and tetragonal oxide layers. The solubility of Li into ZrO2 was determined to be very low indicating that Li solution into the bulk, under equilibrium conditions, is an unlikely cause for accelerated corrosion.

Recent Advances in Protective Coatings for Accident Tolerant Zr-Based Fuel Claddings

Zirconium-based alloys have served the nuclear industry for several decades due to their acceptable properties for nuclear cores of light water reactors (LWRs). However, severe accidents in LWRs have directed research and development of accident tolerant fuel (ATF) concepts that aim to improve nuclear fuel safety during normal operation, operational transients and possible accident scenarios. This review introduces the latest results in the development of protective coatings for ATF claddings based on Zr alloys, involving their behavior under normal and accident conditions in LWRs. Great attention has been paid to the protection and oxidation mechanisms of coated claddings, as well as to the mutual interdiffusion between coatings and zirconium alloys. An overview of recent developments in barrier coatings is introduced, and possible barrier layers and structure designs for suppressing mutual diffusion are proposed.

Characterization of Austenitic Stainless Steels with Regard to Environmentally Assisted Fatigue in Simulated Light Water Reactor Conditions

A substantial amount of research effort has been applied to the field of environmentally assisted fatigue (EAF) due to the requirement to account for the EAF behaviour of metals for existing and new build nuclear power plants. We present the results of the European project INcreasing Safety in NPPs by Covering Gaps in Environmental Fatigue Assessment (INCEFA-PLUS), during which the sensitivities of strain range, environment, surface roughness, mean strain and hold times, as well as their interactions on the fatigue life of austenitic steels has been characterized. The project included a test campaign, during which more than 250 fatigue tests were performed. The tests did not reveal a significant effect of mean strain or hold time on fatigue life. An empirical model describing the fatigue life as a function of strain rate, environment and surface roughness is developed. There is evidence for statistically significant interaction effects between surface roughness and the environment, as well as between surface roughness and strain range. However, their impact on fatigue life is so small that they are not practically relevant and can in most cases be neglected. Reducing the environmental impact on fatigue life by modifying the temperature or strain rate leads to an increase of the fatigue life in agreement with predictions based on NUREG/CR-6909. A limited sub-programme on the sensitivity of hold times at elevated temperature at zero force conditions and at elevated temperature did not show the beneficial effect on fatigue life found in another study.

Investigation on the Interphase Drag and Wall Friction in Vertically Oriented Upward and Downward Two-Phase Flows Under Accident Conditions in Light Water Reactors

Investigation on the Interphase Drag and Wall Friction in Vertically Oriented Upward and Downward Two-Phase Flows Under Accident Conditions in Light Water Reactors

Nonhyperbolicity of Conservation Equations of RELAP5 Two-Fluid Model in Nuclear Reactor Safety Results (Investigation and Eigenvalue Analysis)

Abstract The RELAP5 code simulates the thermal-hydraulic characteristics of nuclear reactors by the use of a two-fluid one-dimensional, nonequilibrium, nonhomogeneous two-phase flow model. This model consists of six governing equations to describe the mass, energy, and momentum of the two fluids. The scope of this work comprises the study of the mathematical nature of the code model and to predict the accuracy of the model in the nuclear reactor safety analysis. The method of characteristics (MOC) is applied to check the nonhyperbolic nature of conservation equations for all normal and accident conditions of light water reactors (LWRs). The analysis also gives information about the soundness of the model and to identify the regions where the solutions obtained from it will be numerically convergent. The characteristics of equations of nonhyperbolic nature are complex. It implies that results thus obtained (by finite difference method) have to be interpreted very carefully in view of the sensitive nature of reactor safety analysis. The present analysis shows that governing equations of the code exhibit complex characteristics for some operating conditions thus implying nonhyperbolicity under those conditions. Results are less accurate under such conditions, so sensitivity analysis plays an important role. The sensitivity of closure relationship on the conservation equation's stability is also checked. The analysis is performed in matlab environment for three different systems, (a) pressurized water reactor (PWR), boiling water reactor (BWR), and (c) natural circulation reactor or advance heavy water reactor (AHWR). These results can also be extended to other thermal-hydraulic systems. The different values of the coefficient of closure relationship are taken for different flow regimes. It is observed that the coefficient of virtual mass (for momentum equation) has a significant effect on the hyperbolicity of the system. It is recommended that further development of the RELAP5 model be performed to identify changes that would reduce the region of complex characteristics. The importance of MOC (in nuclear reactor thermal-hydraulic safety analysis) is evident here. In addition, a detailed analysis for operating pressures range of 0.1–22.5 MPa is also performed to find out the nonhyperbolic regions of code model and realistic data of the different type of reactors is used as input of the code. It is also observed here that RELAP5 results are less accurate when system pressure exceeds 19.5 MPa.

Simulation scaling studies of reactor core two-phase flow using direct numerical simulation

Tremendous growth in supercomputing power in recent years has resulted in the emergence of high-resolution flow analysis methods as an advanced research tool to evaluate single and two-phase flow behavior. In particular, unstructured mesh-based methods have been applied to analyze flows in complex reactor core geometries, including those of light water reactors (LWR). The finite-element based code, PHASTA, is utilized to perform large-scale simulations of two-phase bubbly flows in LWR geometries. Given the large computational cost of direct numerical simulation (DNS) coupled with interface tracking methods (ITM), typical domains encompass a portion of a single subchannel. In the presented research, the state-of-the-art analysis of turbulent two-phase flows in complex LWR subchannel geometries are demonstrated at both prototypical reactor parameters as well as scaled low pressure conditions. Three different cases are studied, a high-pressure simulation in prototypical reactor subchannel geometry, a low-pressure case in prototypical geometry and a final low-pressure case in a geometry scaled up to conserve the ratio between the bubble size and the domain pitch. Utilizing advanced statistical processing tools, these simulation conditions are compared to shed light on the relevancy of two-phase flow characteristics given the significant differences between LWR and low-pressure conditions. These findings can lead to the generation of useful guiding principles when researchers need to scale the two-phase flow behavior captured at low pressure and temperature conditions to those at reactor operating conditions.

Study on dryout and rewetting during accidents including ATWS for the BWR at JAEA

Post-boiling transition (Post-BT) heat transfer is important to analyze the duration of surface dryout and the peak cladding temperature during abnormal transient and design basis accident (DBA) conditions in light water reactors. Furthermore, the Japanese new safety regulation revised after the Fukushima Daiichi accident requires the analysis of the beyond-DBA conditions including the anticipated transient without scram (ATWS), where experimental databases for the code validation are almost-non-existent especially for those with highly oscillatory power change caused by thermo-nuclear coupling. Motivated by the importance, Japan Atomic Energy Agency has conducted a series of experiments for BT and rewetting phenomena to develop mechanistic models based on physical understanding and to validate thermal hydraulic codes. So far, the experimental databases for the Post-BT heat transfer rate, deposition rates of liquid droplets, and the dryout and rewetting behavior were developed mainly for the conditions covering the anticipated operational occurrences (AOO) for BWRs. The significance of the precursory cooling just before the rewetting was identified from comparison with a theoretical model for the two-dimensional transient heat conduction in the cladding. Recently, experiments including the simulation of the ATWS were initiated using the single-tube test apparatus with simulated spacers, liquid film heat transfer visualization apparatus for visual observation, and the HIDRA facility with test sections of a full-length 4 × 4 bundle and a short-length 3 × 3 length bundle. Some promising results have already been obtained, which show effects of simulated ferrule and swirl spacers on the post-BT heat transfer and rewetting velocity, and mechanism of the rewetting by the propagation of the liquid film on the initially heated surface of the heater rod.

Shadow corrosion: Experiments and modeling

Shadow corrosion has long been an issue in BWRs (Boiling Water Reactors): the use of Ni-based grids in close connection to Zry-2 cladding induces locally an accelerated corrosion which can in some instance lead to low remaining metal thickness at the contact point.It is understood since at least 20 years that it is a galvanic coupling phenomenon. However, some important aspects are still debated, due in particular to the fact that it cannot be reproduced without irradiation. Under (neutron) irradiation and at reactor temperature, several parameters are not known accurately (for example the electrical conductivity of the coolant and of the oxide, or the Open Circuit Voltage and current of the metal parts). The role of the contact point between the grid and the cladding is also poorly understood.To improve our understanding, a photoelectrochemical cell, simulating the conditions of Light Water Reactors was designed, developed and validated. Electrochemical methods, performed in-situ, were used to study the electrochemical behavior of Zr-based and Ni-based alloys. Out of pile experiments with or without UV light were performed to measure the OCV (Open Circuit Voltage), the exchange current densities as well as the Tafel slopes (transfer coefficients) of Ni- and Zr-electrodes. The influence of the irradiation flux allowed extrapolations in order to attempt to assess what could happen under neutron irradiation.A finite element modeling based on the Butler-Volmer equation allowed to take simultaneously into account the thermal and galvanic corrosion. Thermal feedback due to the thermal insulating property of the oxide is also included. Parametric studies allowed deriving order of magnitudes of the unknown parameters. To our knowledge, this is currently the most advanced shadow corrosion modeling.