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Recurring step straining was developed to investigate the environmentally assisted cracking (EAC) initiation of a Ni-30Cr-10Fe alloy (Alloy 690) in simulated pressurized water reactor (PWR) primary water. Thanks to this technique, the roles of corrosion and straining in the EAC initiation of Alloy 690 was decoupled and typical microstructures in stress corrosion cracking, i.e. diffusion induced grain boundary migration and oxidation into the migration zone, were reproduced. Digital image correlation using indigenous surface oxide particles as markers was successfully incorporated into this technique for high-resolution in-plane strain measurement. The preliminary results show that intergranular crack initiation is correlated with the high local strain in both adjacent grains, indicating that slip transfer across the grain boundary promotes crack initiation. Recurring step straining is desirable for the mechanistic study of EAC as it makes the separate roles of chemical and mechanical processes adjustable and traceable.

In this study, we propose a simple approach to accelerate crack initiation in intergranular stress corrosion cracking (IGSCC) of Alloy 600 in a simulated pressurized water reactor (PWR) primary water environment. In addition, we perform three-dimensional crystallographic characterization of the crack initiation by electron backscatter diffraction (EBSD). Initiation of IGSCC of the alloy in the PWR primary environment was realized for a flat tensile specimen in a considerably short time through a slow strain rate test (SSRT). The accelerated initiation was attributed to the asperity of the alloy surface induced by mechano-chemical polishing with colloidal silica suspension before the SSRT was conducted. Following the SSRT, the specimen surfaces were covered with thick oxide films, which prevented EBSD measurements. The sputtering of thick oxide films enabled us to characterize cracks with EBSD, thus yielding important information on IGSCC. Finally, an approach for three-dimensional characterization of crack initiation is discussed.

The impacts of various ATF (accident tolerant fuel) claddings on neutronic performances of pressurized water reactor fuel assembly and soluble‐boron‐free small modular reactor autonomous transportable on‐demand reactor module (ATOM) are investigated. There are two ATF cladding concepts which are evaluated here: (a) Coating Zircaloy‐4 cladding with thin layer of Cr or Cr alloys; (b) High‐strength and oxidation‐resistant claddings: SS‐304 and FeCrAl. A minor modification of burnable absorber loading in the ATOM core is proposed to adopt the selected ATF cladding and maintain the core performances.

The purpose of this work was to examine the effect of tensile stress on the oxide properties of a nickel-based Alloy 600 that was exposed to simulated nuclear steam generator water at 340 °C for 1000 h. The size of the outer oxide particles increased, and the chromium content of the inner oxides decreased under tensile stress. Electrochemical measurements revealed that the charge carrier density increased, and the charge transfer resistance and film resistance were reduced under the tensile stress condition. These changes in the oxide properties are attributed to the formation of short diffusion paths such as line and surface defects due to tensile deformation.

A leading cause of fuel failures in U.S. pressurized water reactor power plants is grid-to-rod fretting. One of the major variables affecting fretting wear is the coefficient of friction (COF). Reliability of the fuel cladding is influenced by the COF and thus it is important to know how the material's COF will be affected by temperature transients. Kanthal advanced powder metallurgy technology or APMT, a FeCrAl steel alloy, is being investigated as a commercial alternative to conventional fuel cladding in a nuclear reactor due to its potentially favorable performance under accident conditions. This leaves APMT steel fretting performance to be examined. Tests were performed to examine the evolution of the COF with temperature under fretting. The contact was simulated with both a cylindrical and a rectangular specimen over a line contact area. Confocal scanning laser microscopy was used to obtain a 3D map of the surface, which was used to calculate the wear rate coefficient of the samples. The wear rate coefficient was compared to wear volume to show the effects of the COF and temperature on wear. Energy Dispersive Spectroscopy was performed to qualitatively describe the microchemical changes the material underwent in the fretted region. Nanoindentation was used to compare microstructural properties inside and outside the fretting scar after fretting at different temperatures. COF measurements showed that APMT steel's COF and wear decreases with temperature up to 350 °C.

Reactor pressure vessel (RPV) of pressurized water reactor undergoes severe neutron irradiation during normal operation and life-extension phase, thus leading to the formation of numerous nanometric irradiation-accelerated and irradiation-induced solute-enriched clusters among the ferritic matrix, including so-called late blooming phases (LBP), which was believed as Mn-Ni-Si enriched stable precipitates to emerge lately and grow rapidly and show great threat to the mechanical performance of RPV steels. However, the mechanism of nucleation and growing for LBP remains not in light, especially the effect of inherent defects such as dislocation and grain boundary on that. We investigated the atomic configurations of Fe-Cu, Fe-Mn, Fe-Ni, Fe-Mn-Ni, Fe-Mn-Cu, Fe-Ni-Cu and Fe-Mn-Ni-Cu as RPV model alloys at 600 K, implemented firstly by constructing an enough big simulation box containing edge dislocation and random-distributed alloy atoms and secondly evolving to equilibrium status after essential simulation steps with Metropolis Monte Carlo (MMC) method. The number of clusters, average solute numbers contained in each cluster and the volume atomic fraction of all clusters were characterized, as well as the structure of typical big clusters among simulation analyzed.

In-service thermal aging of CF8M cast austenitic stainless steel was investigated in materials removed from the steam generator inlet and crossover elbows of the Ringhals 2 pressurized water reactor nuclear power plant unit after approximately 92kh of full operating time. The thermal aging of these materials was investigated using the double loop electrochemical potentiokinetic reactivation method, coupled with indentation hardness measurements and microstructural characterizations, to identify correlations between the electrochemical behavior and traditional methods of investigating thermal aging embrittlement effects in cast stainless steels. While this electrochemical method can be easily employed to quantify thermal aging effects in materials aged at higher temperatures (e.g. greater than 350∘C), this study highlights the difficulties encountered when electrochemically evaluating the aging of materials exposed to nuclear power plant operating conditions.

In pressurized water reactor, fretting corrosion has become the main reason for the failure of 690TT heat exchanger tubes. The effect of temperature on the fretting corrosion behavior between 690TT tube and 405 stainless steel (SS) bar has been studied during 106 fretting cycles. The overall average coefficient of friction (COF) values descends with an increase in test temperature, while the width of worn scar becomes wider. The severest fretting corrosion happens when the test temperature is at 100 °C. The wear mechanism differs at different test temperatures, from adhesive wear at room temperature to abrasive wear and delamination at 100 °C, to abrasive wear at 200 °C. Deformation slips, high residual strain concentration, and micro-cracks are found which are disadvantageous for the further service performance of the tubes.

The effectiveness of zinc on the electrochemical behavior and oxide film formation of a 30% cold forged Alloy 690 exposed to high-temperature borated and lithiated water was investigated. The morphology and chemical composition of the formed oxide films with and without zinc injection were analyzed and compared by TEM–EDS and XPS techniques. Zinc clearly improved the electrochemical corrosion parameters of the alloy. Cathodic polarization improved the passivation of the alloy especially in presence of zinc. It was also found that actual zinc concentrations higher than 10 ppb is needed to observe a significant improvement in the corrosion resistance of the alloy in the pressurized water reactor operating conditions. Finally, further investigations of the effect of cold work on the high–temperature electrochemical performance of Alloy 690 and other iron and nickel based alloys in zinc injected solutions is recommended for the future studies.

Low-cycle fatigue (LCF) tests were performed for Alloy 690 and 316 SS in a simulated pressurized water reactor (PWR) environment. Alloy 690 showed about twice longer LCF life than 316 SS at the test condition of 0.4% amplitude at strain rate of 0.004%/s. Observation of the oxide layers formed on the fatigue crack surface showed that Cr and Ni rich oxide was formed for Alloy 690, while Fe and Cr rich oxide for 316 SS as an inner layer. Electrochemical analysis revealed that the oxide layers formed on the LCF crack surface of Alloy 690 had higher impedance and less defect density than those of 316 SS, which resulted in longer LCF life of Alloy 690 than 316 SS in a simulated PWR environment. Keywords: Environmentally-assisted fatigue, Alloy 690, 316 SS, Oxide layer, Electrochemical analysis

Double loop electrochemical potentiokinetic reactivation (DL-EPR) was applied to evaluate thermal aging of 308L stainless steel weld metal. It was found that the activation and reactivation peaks of DL-EPR curve were induced by dissolution of austenite and δ-ferrite, respectively. Before saturation of hardness, the linear relationship between reactivation ratio and hardness could be used for assessing the thermal aging-induced hardening. In the following thermal aging process, the reactivation ratio is applicable to assess the occurrence of the saturation of thermal aging-induced hardening. The results demonstrated that the DL-EPR test is applicable to assess the evolution of thermal aging.

The stress corrosion cracking behaviors of CF8A austenitic stainless steels in a simulated pressurized water reactor environment are systematically investigated by slow-strain-rate tensile tests. Numerous transgranular stress corrosion cracking (TGSCC) cracks are identified upon the fracture of CF8A austenitic stainless steel surfaces. The specimens tested at lower strain rates showed higher fractions of TGSCC, a larger number of cracks, and smaller crack opening displacement. These phenomena are caused by anodic dissolution and crack interactions. The anodic dissolution is dominant at lower strain rates, and thus the crack-tip morphologies are changed by the diffusion of metal atoms and micro-pores. The crack interactions increase rapidly with increasing number of cracks and decrease of the crack-tip radius and crack spacing ratio. In addition, the inhomogeneity of the growth of interaction cracks results in the small cracks being shielded by the preferred growth cracks. On the contrary, the main cracks are accelerated by the stagnation of small cracks, eventually becoming critical cracks that lead to specimen fracture.

High-temperature water corrosion tests were performed to investigate the corrosion behaviors of Alloy 690 TT in simulated PWR secondary water with gaseous hydrogen permeation condition. Double-layer oxide films formed on hydrogen-free and hydrogen-permeating specimens. The thickness of a Cr-rich inner oxide film on the hydrogen-free specimens has no obvious difference from the hydrogen-permeating ones. For hydrogen-free specimens, the outer oxide films were Fe and Ni enriched spinel-type oxide particles. For hydrogen-permeating specimens, outer oxide films were sheet-like or needle-like Cr-rich chromia-type oxides. In summary, the solid diffusing hydrogen affected the morphology and composition of outer oxides.

Flow-induced vibrations can result in fretting wear damage to the surfaces of clad fuel rods (tubes) against their supports in nuclear power plants. One possible solution to this problem is using oxide coatings. In this work, tangential fretting wear experiments were carried out in simulated primary cooling water such as that used in a pressurized water reactor (PWR) to study the effects of micro-arc oxidation (MAO) on the fretting wear behavior of a reactor-grade zirconium alloy known as ZIRLO™. The results indicated that the micro-arc oxidation improved the fretting wear resistance of the ZIRLO™ tube. The synergic effects of corrosion and mechanical wear on the fretting wear of the ZIRLO™ substrate and the MAO coating are discussed, and the role of the MAO coating in mitigating the fretting wear of zirconium alloy in high temperature water is revealed.

Although highly cold-worked UNS N06690 has been shown to be susceptible to stress corrosion crack growth in pressurized water reactor primary water, it is unclear whether cold-work promotes stress ...

A low pressure, superheated hydrogen-steam system has been used to accelerate the oxidation kinetics while keeping the electrochemical conditions similar to those of the primary water in a pressurized water reactor. The initiation has been investigated using a Constant Extension Rate Tensile (CERT) test. Tests were performed on flat tapered specimens made from Type 316L austenitic stainless steel with strain rates of 2×10-6 and 2×10-8 ms-1 at room temperature and at an elevated temperature of 350 °C. R = 1/6 was chosen as a more oxidizing environment and R = 6 was selected as a more reducing environment, where the parameter R represents the ratio between the oxygen partial pressure at the Ni/NiO transition and the oxygen partial pressure. Different exposures (1 day and 5 days) prior to loading were investigated post-test evaluation by scanning electron microscopy.

SMR has its easy receptivity of advanced design concepts and technology. Inherent safety features and passive safety systems realize drastic safety enhancement, and system simplification and modularization can maximize its economics. As they successfully completed the technology self-reliance program for the design and construction of pressurized water reactor (PWR) plant with American companies, Korea has pursued on technology development for SMR (small modular reactors) for the last two decades. This chapter introduces their various SMR R&D programs with their design characteristics and future prospect. Representatively, the Korea Atomic Energy Research Institute (KAERI) has developed integral type PWR—i.e., SMART to acquire the standard design approval according to Korean regulatory frame, and is making efforts to improve safety further. Sodium-cooled fast reactor (SFR) and very high temperature gas-cooled reactor (HTGR) designs are also being developed in KAERI. Besides KAERI’s research, several developments for the various types of SMR’s have also been made by the other Korean organizations such as KEPCO E&C, Seoul National University (SNU), KAIST, and UNIST, whose reactors are different in size, power, core, and application.

The International Cooperative Group on Environmentally Assisted Cracking of Water Reactor Materials coordinated an international “Round-Robin” collaboration of 10 laboratories on stress corrosion crack (SCC) initiation in three heats of cold-worked Alloy 600 in high-temperature water. The various laboratories were able to produce good quality experimental crack initiation data, and the Round-Robin exercise highlighted the benefits in generating a significant body of data when assessing the susceptibility of a material to SCC. The compilation of data from multiple laboratories using common test techniques greatly assists the understanding of the material variability and gives confidence in the reliability of the results.

Light water reactors (LWRs) have been generating electricity for over five decades for the electrical grid of more than 20 countries. Most of these reactors are built using a series of alloys and materials that have changed very little over the many decades. The most common structural materials are based on the three elements, iron (Fe), chromium (Cr) and nickel (Ni) such as stainless steels and nickel-based alloys. Most of the Fe-Cr-Ni alloys contain enough Cr to render them passive with low general corrosion in typical LWR environments. Since the chemistry of the water in the reactor is highly controlled, without aggressive impurities such as chloride or sulfate ions, localized corrosion is not a problem for the common structural alloys. The most common failure mode of the Fe-Cr-Ni alloys is environmentally assisted cracking or stress corrosion cracking (SCC), which is currently well understood and successfully mitigated, mainly by dissolving hydrogen gas into the coolant. The other fundamental material in the reactor core is the zirconium-based alloy used for cladding of the fuel rods. The traditional fuel rod alloys were Zircaloy-2 and Zircaloy-4, which were used regularly in boiling water reactor and pressurized water reactor systems, respectively. Very early in the development of the alloys (early 1950s), it was understood that a small amount (

Nuclear energy is an important ultra-low carbon electricity source. Engineering demonstration reactors, such as Shippingport for Pressurized Water Reactor technology, have been used to advance nuclear reactor technology to deployment. It may be necessary to develop similar engineering demonstration reactors for other reactor concepts that have not yet been built. This paper reviews the historical role of engineering demonstration reactors and highlights the potential for future engineering demonstrations of Fluoride Salt-Cooled High Temperature Reactors (FHRs) and Molten Chloride Fast Reactors (MCFRs). Future engineering demonstration reactors could significantly accelerate deployment of larger-scale FHRs and MCFRs by the advanced nuclear industry via cost and risk reduction. Experience in the United States shows that sustained investment in a public-private partnership, such as the current Advanced Reactor Demonstration Program, is essential for deployment of advanced reactor technologies that have never been built before. The National Reactor Innovation Center at Idaho National Laboratory and Clinch River Site near Oak Ridge National Laboratory are examples of ideal locations for these demonstrations.

Three austenitic stainless steel welds (ASSW), such as E308, ER316L and ER347, with about 10% of δ-ferrite were thermally aged at 400 °C up to 10000 h to accelerate the thermal ageing expected in pressurized water reactor (PWR) environment. Fluctuation of Fe and Cr due to spinodal decomposition is observed with width of 5–12 nm within the ferrite phase. For all 3 welds, increase in strength and decrease in tensile elongation is observed after thermal aging at room temperature and 320 °C. The reduction in fracture resistance, measured as J-R curves, is most significant for ER316L weld with decrease by about 50%. However, it is quite small (about 10% decrease) for ER347 weld even after ageing at 400 °C for 10000 h despite the large reduction in tensile elongation. Finally, the double loop-electrochemical potentiodynamic reactivation (DL-EPR) tests show the degree of sensitization (DOS) values are negligible even after the accelerated thermal ageing at 400 °C for 10000 h.

The pressurizer surge line in a pressurized water reactor (PWR) suffers from environmental fatigue damage, which may accelerate the fatigue crack growth. Currently, ASME code section XI provides a reference curve for the fatigue crack growth rate (FCGR) of stainless steel (SS) only under air conditions, and those curves under PWR water conditions are prescribed in ASME Code Case N-809. However, data for Type 347 stainless steel (347SS), which is used for pressurized surge lines in a Korea Standard Nuclear Power Plant (KSNPP), are rare, and most researches were focused on 304SS, 316SS, and cast austenitic stainless steel (CASS). In this study, the FCGRs of 347SS are evaluated using compact-tension (CT) specimens under simulated PWR water conditions with different levels of dissolved oxygen (DO), dissolved hydrogen (DH), and frequency (Hz). When the DO level increases and the frequency decreases, FCGR of 347SS is accelerated owing to more corrosion fatigue effects. These data are compared with the reference curves of the ASME XI, ASME Code Case N-809, and Japanese models. In this study, FCGRs of 347SS are slightly faster than the reference curves of the ASME Code Case N-809 and Japanese model under PWR water conditions.

Purpose The purpose of this paper is to study the electrochemical behavior of 690 alloy with corrosion products in simulated pressurized water reactor (PWR) primary water environment. Design/methodology/approach This paper opted for a laboratory study using simulation of high temperature and high pressure environment immersion testing. The electrochemical behavior was studied by potentiodynamic polarization, electrochemical impedance spectroscopy, scanning Kelvin probe microscopy (SKP). Moreover, the corrosion products were analyzed by X-ray photoelectron spectroscopy. Findings The results demonstrated that the particle majority in the 690 alloy corrosion products subsequent to high temperature and high pressure immersion testing were mainly oxides of Fe and Ni, which protected the matrix. As the immersion testing duration increased, the corrosion potential of the 690 alloy apparently increased, and the corrosion current density de'creased, while the corrosion resistance Rf increased gradually along with the density. The SKP demonstrated that the EKP increased by nearly 400 mV from −0.42 to −0.03 V following the immersion testing, indicating that the corrosion product film played an apparent protective role on the substrate. Originality/value This paper provides a theoretical basis for the corrosion behavior and inhibition mechanism of 690 alloy in PWR primary water environment.

The stress corrosion crack (SCC) initiation and propagation behavior of three heats of Alloy 600 in simulated pressurized water reactor (PWR) primary water have been studied. The SCC initiation time could be effectively detected using constant load tests with in situ direct current potential drop (DCPD) method. Materials with lower SCC initiation time also showed higher SCC crack growth rate (CGR) under constant K condition. The relationship between SCC initiation time and CGR of Alloy 600 was qualitatively built, and the SCC initiation time could be roughly predicted once the CGR was measured.

Removing the residual heat from a nuclear reactor is an important safety aspect of thermal hydraulic analysis. In this study, a typical VVER-1000 reactor residual heat removal system has been evaluated using RELAP5 thermal hydraulic loop code during cool-down. Reactor cooling down starts from hot state temperature and then continues to the cool-down stages with 130 °C and 70 °C, respectively. The second stage of cooling down is the boundary of the reactor repair condition. Main cooling pump head, steam generator (SG) water level, system pressure, the level of coolant in the pressurizer (PRZ), and the temperature of fuel element are examined in a steady state condition. PRZ level, primary circuit and secondary circuit pressure/temperature, and SG water level are evaluated during 32,000 s after cool-down scenario. By comparison, it concluded that the results of RELAP5 code are in agreement with plant experimental data and final safety analysis report (FSAR). Thus, it is proved that the studied reactor is capable to remove the residual heat generated during shutdown. Moreover, RELAP5 is properly recommended for analysis of the VVER-1000 pressurized water reactor during cool-down. Keywords: RELAP5, VVER-1000, BNPP, FSAR, Residual heat removal, RHR

Absorbing heat from the fuel rod surface, water as coolant can undergo subcooled boiling within a pressurized water reactor (PWR) fuel rod bundle. Because of the buoyancy effect, the vapor bubbles ...

The objective of this study is to characterize steam generator tube deposits collected from an operating pressurized water reactor. The deposits were mostly composed of magnetite and contained small amounts of trevorite, jacobsite, and metallic copper. The concentration of impurities within the pores of the deposits was estimated to be 2 × 106–3 × 107 for copper and 2 × 105–5 × 105 for sodium, chloride, and sulfur. The deposit density was estimated to be approximately 2.72 g/cm3. The porosity and mean pore diameter of the deposits measured by mercury intrusion porosimetry were approximately 9.82% and 196.20 nm, respectively.

Research following the earthquake and subsequent tsunami that devastated the Fukushima Daiichi nuclear power plant, causing severe damage to the reactor cores, has led to the development of accident tolerant fuel designs. Accident tolerant fuels are described as a fuel form which exhibits improved material response, when compared to traditional uranium dioxide fuel clad by a zirconium alloy, during accident (e.g., loss of coolant accident or reactivity insertion accident) while maintaining or exceeding normal reactor operational expectations. One of the primary goals of accident tolerant fuel concepts is to reduce cladding corrosion by either replacing traditional Zircaloy cladding with a new material or by applying a coating to the outer surface of the cladding with enhanced resistance to corrosion at high temperatures. However, while coating the clad may provide an increase in high temperature corrosion resistance, it also changes the mechanical state of the cladding and potentially creates alternate failure mechanisms during reactor operations or accident conditions. This research has primarily focused on the pellet-cladding interactionmechanical response of FeCrAl (Iron-Chromium-Aluminum alloy) coated cladding during light water reactor steady state and reactor startup operations. Using the MOOSE-based, finite-element fuel performance code BISON and commercial pressurized water reactor data obtained from Diablo Canyon Unit 2, a preliminary analysis was performed to evaluate the performance of FeCrAl coated cladding under steady state operation and its mechanical stability under pellet-cladding interaction during a reactor startup. This work summarizes BISON 2-D radial-axially symmetric (R-Z) and radial-circumferentially symmetric (R-θ) simulation results by comparing 20, 50, and 100-µm FeCrAl coating thicknesses as well as FeCrAl cladding to traditional Zr-4 cladding under pellet-cladding interaction condition. The results of this study will provide a feasibility analysis for steady state and transient operational response of FeCrAl coated, Zircaloy cladding. Furthermore, this work will provide a fundamental basis for the assessment of the ability of coated cladding, as an ATF candidate, to maintain or exceed current steady state reactor operating expectations.

The graphite nodule count and size distributions for boiling water reactor (BWR) and pressurized water reactor (PWR) inserts were investigated by taking samples at heights of 2160 and 1150 mm, respectively. In each cross section, two locations were taken into consideration for both the microstructural and solidification modeling. The numerical solidification modeling was performed in a two-dimensional model by considering the nucleation and growth in eutectic ductile cast iron. The microstructural results reveal that the nodule size and count distribution along the cross sections are different in each location for both inserts. Finer graphite nodules appear in the thinner sections and close to the mold walls. The coarser nodules are distributed mostly in the last solidified location. The simulation result indicates that the finer nodules are related to a higher cooling rate and a lower degree of microsegregation, whereas the coarser nodules are related to a lower cooling rate and a higher degree of microsegregation. The solidification time interval and the last solidifying locations in the BWR and PWR are also different.

Correlation of microstructure and stress corrosion cracking (SCC) initiation behaviour of the fusion boundary (FB) region in a SA508 Cl. 3-Alloy 52M weld joint in primary pressurized water reactor environment was investigated. Cr-depletion at the grain boundary and strength mismatch in the FB region are the two causes for SCC initiation in the FB region. The partially mixed zone between the dilution zone (DZ) of Alloy 52M and SA508 Cl. 3 has a retardation effect on SCC initiation. This is attributed to the decrease of strength mismatch and the alleviation of strain concentration adjacent to the grain boundary in DZ.

F6NM martensitic stainless steel was exposed to 350 °C water condition for 500, 1500, and 2500 h to simulate pressurized water reactor (PWR) condition. The characterization and corrosion behavior of the oxide film were investigated. Results indicate that the exposed steel surface formed a double-layer oxide film. The outer oxide film is Fe-rich and contains two type oxide particles. However, the inner oxide film is Cr-rich, and two oxide films, whose thicknesses increase with increasing exposure time. The oxide film reduces the corrosion behavior because the outer oxide film has many crack and pores. Finally, the mechanism and factors affecting the formation of the oxide film were investigated.

There have been recent efforts to establish methods for high-fidelity and multi-physics simulation with coupled thermal–hydraulic (T/H) and neutronics codes for the entire core of a light water reactor under accident conditions. Considering the computing power necessary for a pin-by-pin analysis of the entire core, subchannel-scale T/H analysis is considered appropriate to achieve acceptable accuracy in an optimal computational time. In the present study, the applicability of in-house code CUPID of the Korea Atomic Energy Research Institute was extended to the subchannel-scale T/H analysis. CUPID is a component-scale T/H analysis code, which uses three-dimensional two-fluid models with various closure models and incorporates a highly parallelized numerical solver. In this study, key models required for a subchannel-scale T/H analysis were implemented in CUPID. Afterward, the code was validated against four subchannel experiments under unheated and heated single-phase incompressible flow conditions. Thereafter, a subchannel-scale T/H analysis of the entire core for an Advanced Power Reactor 1400 reactor core was carried out. For the high-fidelity simulation, detailed geometrical features and individual rod power distributions were considered in this demonstration. In this study, CUPID shows its capability of reproducing key phenomena in a subchannel and dealing with the subchannel-scale whole core T/H analysis.

Under severe accident conditions of a pressurized water reactor, hydrogen will be released into the containment, forming a mixture of steam-air-hydrogen, and then transported to the surroundings. This process can be coupled with complex mass transfer behavior caused by wall condensation. Under the coupling effect of condensation and buoyancy introduced by hydrogen, the mixture gas will form different types of transport behaviors. This will be closely related to the stratification and flammability risk of high-concentration hydrogen. To quantitatively determine these phenomena and assess the potential causing reasons, experiments were performed with helium instead of hydrogen based on the COAST (condensation and species transport) facility. The tests cover a broad parameter range with steam concentrations of 8.4– 89%, helium concentrations of 10– 66% (XHe/XNon), and pressures of 0.4– 1.3 MPa. The phenomena of steam condensation induced component separation were firstly confirmed in the tests, and the component concentration threshold was clarified at XHe/XNon = 50%. The three flow patterns of steam-air- helium mixture gas was identified by Shapiro-Moffette ternary diagram. Moreover, the temperature inverse stratification caused by the high-concentration helium-air layer was discovered in the experiments, and its influence on heat transfer was clarified.

Through the early 2000’s, nuclear fuel cladding failure resulting from pellet-cladding interaction in commercial pressurized water reactor fuel was managed successfully by incorporating power ramp restrictions. Within the past decade, however, cladding failures have again begun to appear, often correlated to missing pellet surface defects, and further aggravated by modern reactor operating strategies demanding higher fuel duty operations. In response, renewed efforts in fuel behavior modeling and simulation are being pursued to improve our understanding of the conditions that can lead to missing pellet surface related cladding failure. This paper describes a methodology used to model pellet-cladding interaction related failures observed in commercial pressurized water reactors using the BISON fuel performance code developed at Idaho National Laboratory. In order to accomplish this goal, a three-step process is followed: 1) A failure-stress threshold is developed by using BISON code to model ramp test with known pellet-clad interaction related failures, 2) BISON analysis is applied to a commercial pressurized water reactor cycle startup with one rod confirmed to contain a missing pellet surface defect, and 3) an missing pellet surface sensitivity study of a cycle startup of the same commercial pressurized water reactor which contained a failed fuel rod of un-identified type. The analysis results described in this paper serve to illustrate the BISON code capabilities for modeling pressurized water reactor fuel subjected to higher duty power events.

A nonlinear model of a pressurized water reactor (PWR) fuel assembly is developed, and its performance is evaluated. The nuclear fuel assembly experiences numerous friction forces developed in the fuel rods and supporting components. Therefore, the fuel assembly deflection generates slips between the structures and develops a nonlinear behavior. The loose joints make it impossible for the rods to share a common center of curvature when the fuel assembly is deflected. In addition, when the fuel assembly is under a high temperature coolant water condition, the slip is activated because of the associated dimensional change. Thus, local and global displacements are introduced to compensate for the slip, and the relationship between the two is explained. Considering the fuel becomes soft when it is deflected, a nonlinear beam model with the varying second moment of area is proposed. The Galerkin discretization method is utilized to solve the nonlinear equation of motion, and an incremental solution is determined based on the linearized equation. Verification examples are provided, and the results are compared to the test data. It is shown that the concept of the model is reasonable and applicable. The developed model could be helpful to understand the nonlinear behavior of a fuel assembly and to predict the responses in case of an accident.

The asymmetric pressure wave caused by Loss of Coolant Accident (LOCA) transients in multi-loop type Pressurized Water Reactor (PWR) leads to great shock to the reactor core barrel. The instantaneous shock causes severe damage to the reactor vessel structure integrity. In this paper, a new method for reactor core barrel instantaneous force calculation was proposed based on the widely accepted reactor system analysis code RELAP5. The reliability of RELAP5 code for pipe blow down and pressure wave propagation simulation was assessed against Edward pipe problem firstly. Then the typical three-loop PWR system model was established utilizing Symbolic Nuclear Analysis Package (SNAP) software and the small break LOCA scenarios with different break diameters and break locations were analyzed. The maximum instantaneous asymmetric force applied on the reactor core barrel was achieved under various break conditions, providing the boundary conditions for the core barrel stress distribution and deformation calculation. Results show that the cold leg break LOCA transient leads to great instantaneous asymmetric force, while the hot leg break LOCA transient has no obvious asymmetric force generated. The core barrel instantaneous stress concentration and deformation were discovered. This research could contribute to the reliability assessment of PWR core barrel under LOCA scenarios.

Forged Type 316NG austenitic stainless steel is commonly used to fabricate primary coolant pipe in advanced pressurized water reactor systems, but it is subject to corrosion fatigue cracking in primary coolant environments. In this study, the fatigue crack growth of Type 316NG austenitic stainless steel in 325°C water was investigated. The corrosion fatigue effect in 325°C water was mainly correlated to the crack growth rate and load frequency. A maximum Fen (environment corrected factor) value up to 100 was observed at f = 0.005 Hz, ΔK = 13 MPa√m. The dissolved oxygen tended to have little influence on the fatigue crack growth. The crack path was deflected and branched. Most secondary cracks tended to be initiated from grain boundaries.

[EN] The International Atomic Energy Agency (IAEA) produced guidance on the use of Deterministic Safety Analysis (DSA) for the design and licensing of Nuclear Power Plants (NPPs) in "DSA for NPP Specific Safety Guide, No. SSG-2", which proposes four options for the application of DSA. Option 3 involves the use of Best Estimate codes and data together with an evaluation of the uncertainties, the so called BEPU methodology. Several BEPU approaches have been developed in scopes that are accepted by the regulator authorities nowadays. They normally adopt conservative assumptions on the availability of safety systems. Option 4 goes beyond by pursuing the incorporation of realistic assumption on the availability of safety systems into the DSA. This paper proposes an Extended BEPU (EBEPU) approach that integrates insights from probabilistic Safety Analysis into a typical BEPU approach. There is an aim at combining the use of well-established BEPU methods and realistic ("probabilistic") assumptions on safety system availability. This paper presents the fundamentals of the EBEPU approach and the main results obtained for an example of application that focuses on an accident scenario corresponding to the initiating event "Loss of Feed Water (LOFW)" for a typical three-loops Pressurized Water Reactor (PWR) NPP., This work has been developed partially with the support of Programa de Apoyo a la Investigacion y Desarrollo of the Universitat Politecnica de Valencia (PAID UPV).

The safety of nuclear power plants (NPPs) depends on the actions taken to prevent nuclear accidents or to limit their consequences. This paper proposes improvement of pressurized water reactor safety by introducing of new system for water injection in the primary and secondary loops. The system is envisioned as a supplementary to the regular safety systems and to be operated independently. The proposed design consists of a water source, pipes, valves and a portable fossil-fuel driven pump, and implements the “Diverse and Flexible Coping Strategy” mitigation concept recommended by the US industry following the Fukushima accident. The potential of the supplementary water storage system to cope with two major initiating events, i.e. the station blackout and large loss of coolant accident, is investigated, and its broad impact on nuclear safety is quantified. The probabilistic safety assessment is employed in the analyses. In particular, a probabilistic safety assessment model of a pressurized water reactor is modified to consider the new independent water storage system. The core damage frequency of the plant is used as a risk measure. The results reveal that the core damage frequency of the analyzed NPP decreases significantly after the introduction of the proposed water storage system. Furthermore, the criticality of some of the most important basic events related to the provision of water flow during station blackout significantly decreases.

The coupling of the Qeshm island desalination plant with different energy sources was economically evaluated by using a Desalination Economic Evaluation Program (DEEP) package. The capacity of this plant was 18,000 m3/day and its energy source was 250 MW (th) gas power plant. Six energy source plants, namely, coal, oil, gas, combined cycle (CC), pressurized water reactor (PWR), and pebble-bed modular reactor (PBMR), were coupled with this desalination plant. The estimated annual electrical costs indicated that the optimum fossil power plant was CC with 97.72 $/kWh, which was 30% higher than 67.42 and 64.63 $/kWh in PWR and PBMR power plants, respectively. The total annual water costs and the total annual costs of coupling with the desalination plant of CC, PWR, and PMMR plants were approximately 1.78, 1.49, and 0.77 $/m3, and 94.74, 49.23, and 49.1 m$, respectively. The construction costs of these power plants are approximately 131, 510 and 553 M$, respectively, On the basis of the contraction times, costs, and lifetimes of these power plants 25, 40 and 60 years for CC, PBMR and PWR, we concluded that CC, PBMR and PWR are short-term, medium-term and long-term strategy to generate electricity and couple with desalination plants, respectively.

Corrosion and particle deposition in the secondary circuits of pressurized water reactors can be mitigated by alternative water chemistries featuring film-forming amines. In the present work, the corrosion of carbon steel in secondary side water with or without octadecylamine (ODA) is studied by in situ electrochemical impedance spectroscopy, combined with weight loss/gain measurements, scanning electron microscopy and glow-discharge optical emission spectroscopy. The impedance spectra are interpreted using the mixed-conduction model to extract kinetic parameters of oxide growth and metal dissolution through it. From the experimental results, it can be concluded that ODA addition reduces the corrosion rate of both fresh and pre-oxidized carbon steel in secondary circuit significantly by slowing down both interfacial reactions and transport through the oxide layer.

Summary Nuclear power is attractive for helping carbon reduction and energy restructuring, but the low utilization of nuclear resource and technical problems on reprocessing are key issues for further nuclear development. Subcritical system might be a potential way to solve these problems, which include fusion-driven subcritical system and accelerator-driven subcritical system. Especially the new subcritical system conception spent fuel burner could be used for waste transmutation, nuclear fuel breeding, and energy production. In this contribution, 2 nuclear fuel cycle scenarios with subcritical system for spent fuel burning and pressurized water reactor were assessed with the performances of sustainability of resources, environmental impacts, and economy. The results showed that the advanced fuel cycle based on subcritical system is an attractive option for the long-term development of nuclear power in China.

The Helically coiled tube Once-Through Steam Generator (H-OTSG) is a key piece of equipment for compact small reactors. The present study developed and verified a thermal–hydraulic design and performance analysis computer code for a countercurrent H-OTSG installed in a small pressurized water reactor. The H-OTSG is represented by one characteristic tube in the model. The secondary side of the H-OTSG is divided into single-phase liquid region, nucleate boiling region, postdryout region, and single-phase vapor region. Different heat transfer correlations and pressure drop correlations are reviewed and applied. To benchmark the developed physical models and the computer code, H-OTSGs developed in Marine Reactor X and System-integrated Modular Advanced ReacTor are simulated by the code, and the results are compared with the design data. The overall characteristics of heat transfer area, temperature distributions, and pressure drops calculated by the code showed general agreement with the published data. The thermal–hydraulic characteristics of a typical countercurrent H-OTSG are analyzed. It is demonstrated that the code can be utilized for design and performance analysis of an H-OTSG.

The paper focuses on the presentation and comparison of basic nodal and expanded multi-nodal models of the Pressurized Water Reactor (PWR) core, which includes neutron kinetics, heat transfer between fuel and coolant, and internal and external reactivity feedback processes. In the expanded multi-nodal model, the authors introduce a novel approach to the implementation of thermal power distribution phenomena into the multi-node model of reactor core. This implementation has the form of thermal power distribution coefficients which approximate the thermal power generation profile in the reactor. It is assumed in the model that the thermal power distribution is proportional to the axial distribution of neutron flux in the un-rodded and rodded reactor core regions, as a result of control rod bank movements. In the paper, the authors propose a methodology to calculate those power distribution coefficients, which bases on numerical solutions of the transformed diffusion equations for the un-rodded and rodded reactor regions, respectively. Introducing power distribution coefficients into the expanded multi-nodal model allows to achieve advanced capabilities that can be efficiently used in design and synthesis of more advanced and complex control algorithms for PWR reactor core, for instance in the field of reactor temperature distribution control.

Structural integrity analysis of nuclear pipe containing defects should be maintained in the nuclear power plants. Deterministic approach has been often used for integrity evaluation even though there are many uncertainties to hinder a reasonable evaluation. Based on EPRI engineering estimation method, the formulas are derived to calculate the shape factor F′ and the full plastic solution coefficient h 1 ′ of primary coolant pipe with cracks in this paper. Considering the uncertainties in assessment parameters, the probabilistic method is introduced to establish probabilistic failure assessment curve (PFAC) according to the probabilistic model. Usually, probabilistic approach has strong dependences on data which is sensitive to the probabilistic model. In this respect, non-probabilistic method that only needs the upper and lower limits of parameters is as a promising alternative. Non-probabilistic failure assessment curve (NPFAC) is also provided based on interval method. The comparison between PFAC and NPFAC indicates that NPFAC is an appropriate methodology to conduct safety assessment with defects when data information is inadequate. A numerical example is given to show the application of probabilistic failure assessment and non-probabilistic failure assessment, the results suggest that it is effective and feasible to use non-probabilistic method for assessment.

A parametric optimization framework for the steam cycle of a typical pressurized water reactor (PWR) nuclear power plants (NPP) (i.e. Daya Bay nuclear plant) is proposed. The approach determines the optimal operation parameters of the steam cycle to maximize the power output with primary loop remaining unchanged. Thermodynamic model of the steam cycle system has been established to determine the power output under different operating conditions, and the availability and accuracy of the model are investigated. In addition, an improved genetic-simplex algorithm (IGSA) is put forward by using modified search strategies and integrating genetic algorithm (GA) with simplex algorithm (SA). Performance comparison between the improved algorithm and the original ones is performed by solving the optimization test problems. The improved algorithm enables the handling of nonlinear constrained optimization problem because of dramatically enhanced search capability. Based on the model, the parametric optimization for the steam cycle of Daya Bay PWR nuclear plant is carried out using the IGSA, GA and SA, respectively. In the optimized schemes it has been found that the IGSA can yield increase in power output as much as 23.8 MW by an average of 236 iterations over current practice, and the corresponding thermal efficiency increases from 33.87% to 34.69%. In comparison, even more iterations have been executed, the maximum power output increments gained by the GA and SA are 21.4 MW and 18.7 MW, respectively. The results show that the new algorithm IGSA is superior to original ones. It also demonstrates that the proposed approach can effectively increase the efficiency of the PWR nuclear power plant without burning any additional fuel or replacing the devices of the system.

A systematic uncertainty analysis strategy based on stochastic sampling method is developed for best estimate analysis of the rod ejection accident in pressurized water reactor (PWR). Self-developed SAMP module and DAKOTA code are used to randomly sample the multigroup nuclear data, important neutronics and thermal-hydraulic (T-H) input parameters respectively. The 3D coupled reactor core model for Almaraz PWR plant is established using RELAP5/PARCS code system and the uncertainties of the core power, fuel enthalpy and local power distribution are obtained. The results indicate that the uncertainties of nuclear data and input parameters can result in large uncertainty on the results which should be carefully considered.

A pressurized water reactor (PWR) is most suitable for load-following operation because of its self-regulating nature. However, rapid power maneuvering at an enhanced rate is heavily constrained by the overall power coefficient of reactivity which depends on both the reactor neutronics and thermal hydraulics. In this paper, a 100 MWe PWR module with an integrated steam generator, is modeled using the nonlinear equations representing both the neutronics and thermal hydraulics. Such a PWR is typically an integrated structure with a steam generator included in the same casing and can either be connected to the grid or connected with other such modules using a multimodular approach to constitute a Nuclear Steam Supply System (NSSS) driving a common turbine and an alternator. In this paper, Nonlinear Dynamic Inversion (NDI) technique is used to design a controller capable of controlling such a reactor for load-following with frequent changes in demand, in both the modes of operation, namely, reactor-follow-turbine and turbine-follow-reactor. With the assumption that the parameters of the reactor vary within intervals as the power varies, an interval approach coupled with NDI ensures that the controller satisfies the operating constraints. The methodology is established for both the modes with credible real-time Hardware-in-Loop (HiL) simulations.

The oxide film formed on alloy 690 in simulated pressurized water reactor primary water is decorated with NiFe 2 O 4 particles which remain epitaxial with matrix. The penetrative inner oxides on both surface and crack wall are composed of Cr-rich (Fe,Cr,Ni) 3 O 4 and Cr 2 O 3 which were formed by solid state reactions of substrate with inwards diffusing oxygen and have orientation relationships with the substrate. Cr 2 O 3 forms first along the widely spaced planes of substrate. Compact Cr 2 O 3 layer cannot develop as there is no long-range outward diffusion of Cr in substrate with low defect density at low temperature.