This paper presents a results of brittle fracture (BF) assessment of WWER-1000 reactor pressure vessel (RPV) subjected to pressurized thermal shook (PTS). Work is focused on the two factors affected on the BF safety margin. The first one is the definition of the heat transfer conditions from outer surface of RPV, which is commonly not considered due to the assumption of conservatism. The second one is the warm pre-stressing (WPS) of RPV metal during PTS. The procedure for the formation of thermal boundary conditions (BC) for the RPV (i.e. lower nozzle forging, cylindrical part and elliptical bottom), reactor supports, as well as for the reactor thermal insulation, is developed. This procedure takes into account the heat transfer due to radiation in the GA-301 room and the presence of forced supply of “cold” air by the ventilation system. A brief history of WPS implementation in Ukraine is presented. Also, the comparative review of the WPS experience in PWR countries and WPS approaches is made. The results of fracture mechanics calculations for the three most severe PTS scenarios of the Unit 3 RPV of South-Ukrainian NPP are presented. Wherein, both type of results is demonstrated: with and without consideration of the RPV outer surface heat transfer due to the air-cooling ventilation system. It is shown that the RPV air cooling consideration: leads to decreasing of the stress intensity factor; provides to satisfaction of the cladding integrity criterion according to VERLIFE-2008 methodology; can justify lifetime of the RPV nozzle region in terms of not to exceed the upper shelf of fracture toughness 200 MPa·m 0.5 . At the same time, consideration of the RPV air cooling alongside with the WPS approach, depending on WPS approach considered, can lead as to the decreasing the conservatism of RPV BF assessment, as to the increasing it. Also, concluded that Ukrainian WPS approach needs to be refined more carefully with consideration of experimental nature of pre-stressing or harmonized with some of the modern ones.

Following the tragic events at the Fukushima Daiichi power plant in 2011, priority was given to increasing the accident tolerance of fuel systems for the current fleet of nuclear reactors. These enhanced Accident Tolerant Fuel (ATF) concepts include a wide variety of fuel and cladding materials, both as variants of the current Zircaloy-UO 2 system and also as novel fuel and cladding concepts. In addition to testing at steady-state, prototypic, conditions within a nuclear reactor, performance of these ATF concepts in off-normal and transient conditions must be evaluated. The Transient Reactor Test (TREAT) facility at Idaho National Laboratory’s (INLs) Materials and Fuels Complex (MFC) was restarted in the Fall of 2017 and is well-suited to serve this purpose. September of 2018 marked the first fueled specimen to be tested in TREAT since its restart; testing of fuel specimens has been ongoing since then. Initial fuel tests focused on the traditional Zircaloy-UO 2 fuel system in order to gain a more thorough understanding of operating characteristics of both the test vehicle system and also the interactions between the reactor and the experiment itself. These tests also served to commission new test vehicles using the well-characterized Zircaloy-UO 2 system. The Separate Effects Test Holder, SETH Capsule, is a modular capsule designed such that it can support a wide variety of specimen geometries ranging from prototypic pressurized water reactor (PWR) fuel samples, heat sink based experiments, and more. The capsule itself is an additively manufactured titanium capsule, within which the experimental specimen is loaded. The SETH Phase I series of tests included five individual SETH capsules, each with a single fuel rodlet and instrumentation to measure temperature during irradiation in TREAT. Each fuel rodlet is representative of a fuel rod in a PWR, with UO 2 in Zircaloy-4 cladding. In August of 2019, TREAT irradiated the first ATF candidate fuel, U 3 Si 2 . This marked the first transient test of an ATF concept and is part of a larger campaign that will irradiate a total of four capsules containing ATF concepts. This test campaign, SETH Phase II, built upon the previous SETH Phase I campaign with a nearly identical design except for the fuel rodlet itself. Two of the four SETH capsules contained U 3 Si 2 fuel within Zircaloy-4 cladding, and the other two capsules contained U 3 Si 2 fuel within SiC cladding. This paper reviews the design, fabrication, and assembly efforts resulting in the four qualified SETH capsules for TREAT irradiation of these ATF concepts.

Inner surface of most of primary circuit components in PWR/BWR/WWER type reactors is covered by austenitic cladding that serves primarily as anticorrosion protection. This is also supported by the requirements for stress analysis of the vessel by most of the codes — austenitic cladding is not taken into account in the calculation of vessel wall thickness and on allowance of stress intensities for operating conditions. Its effect is taken, in some codes like for WWER components, in fatigue calculation and also for evaluation of vessel resistance against fast fracture during pressurized thermal shock (PTS) events.

As part of nuclear power plant ageing management, the increased probability of a need of repair welding must be taken into account along with the increase of plant lifetime. An essential prerequisite for successful and safe repair welding is that the applied welding procedures have been properly validated and qualified prior to their use. For instance, if no post-weld heat treatment can be performed and the desired tempering effect has to be based on temper-bead technique, a user needs to scan among several available repair welding procedures. A decision has to be made which of the procedures provides the maximum desired tempering effect with the case in question. This research is a part of a larger experimental effort developing repair welding techniques, and is a part of the Finnish Nuclear Power Plant Safety Research Programme SAFIR2022. The currently studied experimental repair welding case is a low-alloy steel mock-up with an austenitic cladding. Repair welding is assumed to represent a ‘worst-case’ scenario where a postulated linear crack-like defect exists beneath the cladding and might extend across the interface into the reactor pressure vessel steel side. This postulated defect will be removed by machining, and the thereby machined groove will be filled by repair welding using a nickel-base super alloy 52M filler metal by cold metal transfer-gas metal arc welding with a robotic arm. In this paper, different repair welding techniques and alternatives are shortly surveyed based on existing literature. Overall, published documentation was sparse. While only few studies were considered relevant in terms of established links to actual repair cases of under-cladding defects in reactor pressure vessels, others were mainly for modelling and simulation purposes without e.g. cladding groove preparation or the use of irradiation-embrittled material. Most of these procedures were based on the use of nickel-base alloy filler metal in the combination with temper-bead welding technique, with the aim at omitting both preheating and post-weld heat treatment. The main challenge in the repair weld design is to optimise all relevant welding parameters, including the thermal efficiency of temper-bead welding, in order to obtain a sound, defect-free weld with controlled reactor pressure vessel steel heat affected zone maximum hardness. In the simulations presented in the paper, the goal was to compute the resulting deformations, strains and stresses induced by the repair process and make a-priori estimates of the effectiveness of different repair techniques based on the numerical predictions. The numerical analyses allow the comparison of the procedures and enable selecting the one with most efficient combination of weld thermal cycles in terms of tempering and normalisation effects. The prediction of prevailing residual stresses is also important when further application of the component is considered. The paper is followed by Part II, in which the topics of experimental evaluation and material characterization of the repair weld are presented.

Aging management of the existing fleet of nuclear power plants is becoming an increasingly important topic, especially as many units are approaching their design lifetimes or are entering long-term operation. As these plants continue to age, there is an increased probability for the need of repairs due to extended exposure to a harsh environment. It is paramount that qualified and validated solutions are readily available. A repair method for a postulated through cladding crack into the low alloy steel of a nuclear power plant’s reactor pressure vessel has been investigated in this study. This paper is part of larger study that evaluates the current possibilities of such repair welds. The present paper documents the weld-trials and method selection. A parallel paper describes numerical simulations and optimization of weld parameters. The presented weld-trial represents a case where a postulated crack has been excavated and repaired using a nickel base Alloy 52M filler metal by gas metal arc welding-cold metal transfer with a robotic arm. A SA235 structural steel has been used as a base material in this weld-trial. No pre-heating or post-weld heat treatment will be applied, as it would be nearly impossible to apply these treatments in a reactor pressure vessel repair situation. While Alloy 52M presents good material properties, in terms of resistance to environmentally assisted degradation mechanisms, such as primary water stress corrosion cracking, it is notoriously difficult to weld. Some difficulties and challenges during welding include a sluggish weld puddle, formation of titanium and/or aluminium oxides and its susceptibility to lack of fusion defects and weld metal cracking, such as ductility dip cracking and solidification cracking. Moreover, gas metal arc welding-cold metal transfer is not traditionally used in the nuclear industry. Nonetheless, it presents some interesting advantages, specifically concerning heat input requirements and automation possibilities, as compared to traditional welding methods. The mechanical properties, in terms of indentation hardness, and microstructure of a weld-trial sample have been evaluated in this study. The fusion boundary and heat affected zone were the main areas of focus when evaluating the mechanical and microstructural properties. Detailed microstructural characterization using electron backscatter diffraction and nanoindentation were performed across the weld interface. Based on these results, the gas metal arc welding cold metal transfer is seen as a potential high-quality weld method for reactor pressure vessel repair cases.

The clad selection criteria proposed in this work supports a design approach provided in [PVP2020-21469] for cladded component for high temperature nuclear service. The proposed design method guards the clad material against the creep-fatigue failure and ratcheting strain accumulation in elevated temperature nuclear service without the requirement of long-term material properties. However, it limits the type of clad materials that can be used with the existing Class A materials qualified for ASME Section III, Division 5 rules. The analysis approach and design rules allow the use of two types of clad materials — soft clads that creep much faster than the base material and hard clads that creep much slower and have higher yield stress than the base material. This work proposes selection criteria for such soft and hard clad materials to use with a Class A metallic alloy — austenitic steel 316H. The criteria are developed based on the effect of relative elastic modulus and creep rate on the long term stress redistribution between the 316H base and the clad material. The proposed clad selection criteria are applicable up to a design temperature of 750°C and for 1% to 10% thick cladding. The selection criteria are evaluated on two materials — nickel and a molybdenum based alloy TZM — categorizing them as soft or hard clad for 316H base material.

The inner surfaces of ferritic steel pressure vessels are often coated with a weld-deposited layer of austenitic cladding for the purpose of corrosion resistance. UK practice of assessing the defect tolerance by postulating defects in components leads to the requirement for a fracture assessment that includes cladding. Section XI requires that the cladding be considered in the fracture analysis, including cladding-induced stresses in the K I evaluation; these stresses largely arise due to the different coefficients of thermal expansion of the cladding and base material, and can also be known as cladding residual stresses. It is thought that in some assessments of clad pressure vessels, the cladding residual stress is either not included, or assumed to be of a significantly reduced magnitude — due to it decreasing at typical operating temperatures and/or being relaxed by prior overloads such as a hydrostatic test. This paper describes finite element analysis to determine the effect of cladding residual stress, comparing J values for loading with and without cladding stresses. Comparisons are also made with J-estimations using the R6 fracture assessment procedure. It is argued that cladding-induced stresses would not contribute to ductile failure as they would be reduced by prior loading, with remaining residual stresses being overcome under large plasticity before failure, due to their secondary nature. Further, it may be misleading to include them in an R6 assessment when using simplified approaches, as their contribution can be exaggerated through pessimistic calculations of the V factor.

Different thermal analysis models were developed to simulate the dry cask simulator (DCS). The DCS is an experiment designed to simulate dry storage of a single boiling water reactor fuel assembly under a variety of heat loads and internal pressures. The DCS was set up and tested in both a vertical and horizontal configuration to determine cladding temperatures in vertical and horizontal dry cask storage systems. The models included a detailed STAR-CCM+ model with the fuel assembly geometry explicitly modeled, a porous STAR-CCM+ model with the fuel assembly geometry modeled as a porous media region with calculated effective properties, and a COBRA-SFS model. COBRA-SFS is a thermal-hydraulic code developed for steady-state and transient analysis of multi assembly spent-fuel storage and transportation systems. STAR-CCM+ is a commercial computational fluid dynamics (CFD) code. Both a detailed and porous STAR-CCM+ model were developed to look at the effective thermal conductivity (k eff ) approach to modeling a fuel assembly. A k eff fuel model is typically modeled in CFD thermal analyses due to its significantly lower computational costs. The models were run for a combination of low and high canister pressures (100 kPa and 800 kPa) and low and high internal heat loads (0.5 kW and 5 kW). Results from all three models were compared against experimental data taken from the DCS for the peak cladding temperature (PCT) and inlet air mass flow.

In this work, a two-dimensional (2D) geometrically-accurate model of the TN-32 cask is generated in ANSYS/Fluent to investigate the effect of backfill gases and their pressures on the peak cladding temperature (PCT). This model is similar to the cask being used in high-burnup (HBU) spent fuel data project lead by the Electric Power Research Institute (EPRI). Helium, nitrogen, argon, and water vapor fill gases are investigated at pressures ranging from atmospheric (∼10 5 Pa) to 100 Pa. Steady-state computational fluid dynamics (CFD) simulations that include the effect of gas rarefaction (temperature-jump) at the gas-solid interfaces are conducted. The PCT as a function of heat generation rate and pressure is reported as well as the heat generation rate that brings the cladding temperature to the radial hydride formation limit. The results show that there are competing effects between the temperature-jump and the thermal conductivity of the gas to increase the fuel rods’ temperature. The low pressures increased the PCT, with the increase being most significant for the helium backfill.

SA508 low-alloy steel for a reactor vessel was exposed to primary water in a pressurized water reactor (PWR) plant because the cladding layer of type 309 stainless steel for a reactor pressure vessel (RPV) was peeled off owing to an accident in which the thermal sleeve was detached. The advantage of electrochemical deposition (ECD) Ni plating techniques is that the RPV can be repaired without significant thermal effects, and Ni has sound corrosion resistance that can withstand exposure in primary water. The corrosion rate of the damaged part was assessed, and its trend was analyzed. The essential variables of Ni plating for repair of the damaged part were derived. These conditions are applicable variables for the repair plating device and have been carefully adjusted using the repair plating device. The process for establishing ASME technical standards called code case N-840 is described. The process for developing Ni-plating devices and the electrodeposition procedure specifications are described as well.

For structural integrity assessment of the reactor pressure vessel (RPV) in JEAC 4206-2016, it is required that the fracture toughness (K Jc ) be higher than the stress intensity factor at the crack tip of a postulated under-clad crack (UCC) near the inner surface of RPV steel under the pressurized thermal shock event. Previous analytical studies showed a low constraint effect at the crack tip of an UCC, compared with that of a normal surface crack. Such a low constraint effect may increase the apparent K Jc . In this study, we performed three-point bending (3PB) fracture toughness tests and finite element analysis (FEA) for RPV steel containing an UCC or a surface crack to quantitatively investigate the effect of cladding on the K Jc . The FEAs considering the anisotropic property of the cladding successfully reproduced the load vs. load-line displacement curves obtained from the tests. We found that the apparent K Jc for the UCC was considerably higher than that for the surface crack. FEA also showed that the constraint effect for the 3PB test specimen with the UCC was lower than that for the specimen with the surface crack owing to the cladding. Thus, a low constraint effect from an UCC may increase the apparent K Jc .

General Electric (GE) is working with the US Department of Energy (DOE) to develop advanced technology fuel (ATF) for light water reactors (LWR) that will have enhanced tolerance to failure under severe accident conditions. The development of materials for the current fuel is aimed at Generation III LWR but findings may be extended to future Generation IV reactors. One of the concepts pursued by GE is to use iron-chromium-aluminum (FeCrAl) or IronClad for the cladding due to its outstanding resistance to reaction with air and steam at temperatures higher than 1000°C. Ferritic FeCrAl alloys have been used for almost nine decades in the industry, but never in nuclear applications, therefore its fabrication and mechanical aspects for nuclear use needs to be evaluated. Results show that billets of FeCrAl can be produced via traditional melting and using powder metallurgy, and these billets can later be processed to high strength full length cladding tubes having less than half a millimeter wall thickness. The tubes can be joined to the caps via several welding processes.

Dissimilar Metal Welds (DMWs) are used in many industries to increase the service life of components. Typically a corrosion resistant material is bonded to the internal surface of a second material, this is often referred to as cladding. This type of composite component provides the desired mechanical properties and corrosion resistance without excessive cost. Structural integrity assessments can be used to estimate the remaining life of critical components but these methods are complicated and results are dependent on factors such as: material properties, defect size, defect location and the direction of defect propagation. The cladding process is generally conducted via welding which results in yield strength residual stresses. The interaction between primary and secondary stresses on crack driving force (CDF) are not well understood, particularly at the interface between multiple materials where the microstructure is not homogeneous. These unknown factors contribute to overly conservative structural integrity assessments. This work investigates the fracture toughness of defects orientated parallel the interface of a dissimilar metal clad weld using experimental and finite element (FE) methods. A nuclear grade austenitic clad ferritic steel was investigated. A fracture toughness test was conducted on a CT specimen and the influence of the materials boundary on crack propagation was examined using digital image correlation (DIC). DIC of the fracture toughness test shows that high amounts of strain exist both in the region immediately surrounding the crack tip and in the ferritic material at the material boundary. Post-test analysis of the specimen showed two defects initiated at the root of the notch on the outer surface of the specimen. One longer notch was found internally. Two-dimensional plane stress and plane strain FE analyses were conducted and compared to the experimental results.

Cladding structural components with a corrosion resistant material may greatly extend the design life of molten salt reactor concepts. A complete design methodology for such cladded, high temperature nuclear components will require addressing many issues: fabrication, corrosion resistance, metallurgical interaction, and the mechanical interaction of the clad and base materials under load. This work focuses on the final issue: the mechanical interaction of the base and clad under creep-fatigue conditions. Depending on the relative mechanical properties of the two materials the clad may substantially influence the long-term cyclic response of the structural system or its effect might be negligible. To quantify the effect of different clad material properties we develop an efficient method for simulating pressurized cladded components in the limiting case where the section of interest is far from structural discontinuities. Using this method we evaluate the mechanics of the clad/base system and identify different regimes of mechanical response. The focus is on situations relevant to high temperature nuclear components: thermal-cyclic Bree-type problems and similar axisymmetric structures. The insights gained from these structural studies will form the basis for developing design rules for high-temperature, nuclear, cladded components.

An evaluation of the fracture toughness of the heat-affected zone (HAZ), which is located under the weld overlay cladding of a reactor pressure vessel (RPV), was performed. Considering inhomogeneous microstructures of the HAZ, 0.4T-C(T) specimens were manufactured from the cladding strips locations, and Mini-C(T) specimens were fabricated from the distanced location as well as under the cladding. The reference temperature ( T o ) of specimens that were aligned with the middle section of a cladding strip (HAZ MCS ) was ∼12°C higher than that of specimens that were aligned with cladding strips at the overlap (HAZ OCS ). T o values of partial area in the HAZ were obtained using Mini-C(T) specimen. The T o values obtained near the side of the cladding were ∼13°C higher than those away from the cladding. T o values of HAZ for both 0.4T-C(T) and Mini-C(T) specimens were significantly lower than that of the base metal at a quarter thickness by 40°C–60°C. Compared to the literature data that indicated fracture toughness at the surface without overlay cladding and base metal of a quarter thickness in a pressure vessel plate, this study concluded that the welding thermal history showed no significant effect on the fracture toughness of the inner surface of RPV steel.

As one of the most promising Generation IV nuclear reactors, thorium molten salt reactor (TMSR) possesses inherent safety, simplified fuel cycle and high power generation efficiency. However, the structure material of TMSR must encounter challenges of high-temperature environment and serious molten fluoride salts corrosion. Although the UNS N10003 alloy has excellent performance in TMSR, the high cost of this alloy reduces the economy of TMSR. Although the austenitic stainless steel possesses excellent high temperature strength and lower cost, it exhibits poor corrosion resistance in molten fluoride salts. Therefore, the purpose of this work is using Ni cladding on austenitic stainless steel by gas tungsten arc welding (GTAW) for corrosion protection and investigating the effects of thermal ageing on microstructure and hardness of cladding. The elements distribution and segregation, the precipitates behavior and the microstructure evolution of interface and cladding layer under as-welded and thermal ageing conditions were studied by a series of characterization methods respectively. The hardness evolution of claddings under as-welded and thermal ageing conditions were tested and discussed.

Nuclear reactor pressure vessels must be able to withstand thermal shock due to emergency cooling during a loss of coolant accident. Demonstrating structural integrity during thermal shock is difficult due to the complex interaction between thermal stress, residual stress, and stress caused by internal pressure. Finite element and analytic approaches exist to calculate the combined stress, but validation is limited. This study describes an experiment which aims to measure stress in a slice of clad reactor pressure vessel during thermal shock using time-resolved synchrotron X-ray diffraction. A test rig was designed to subject specimens to thermal shock, whilst simultaneously enabling synchrotron X-ray diffraction measurements of strain. The specimens were extracted from a block of SA508 Grade 4N reactor pressure vessel steel clad with Alloy 82 nickel-base alloy. Surface cracks were machined in the cladding. Electric heaters heat the specimens to 350°C and then the surface of the cladding is quenched in a bath of cold water, representing thermal shock. Six specimens were subjected to thermal shock on beamline I12 at Diamond Light Source, the UK’s national synchrotron X-ray facility. Time-resolved strain was measured during thermal shock at a single point close to the crack tip at a sample rate of 30 Hz. Hence, stress intensity factor vs time was calculated assuming K-controlled near-tip stress fields. This work describes the experimental method and presents some key results from a preliminary analysis of the data.

During vacuum drying of used nuclear fuel (UNF) canisters, helium pressure is reduced to as low as 67 Pa to promote evaporation and removal of remaining water after draining process. At such low pressure, and considering the dimensions of the system, helium is mildly rarefied, which induces a thermal-resistance temperature-jump at gas–solid interfaces that contributes to the increase of cladding temperature. It is important to maintain the temperature of the cladding below roughly 400 °C to avoid radial hydride formation, which may cause cladding embrittlement during transportation and long-term storage. Direct Simulation Monte Carlo (DSMC) method is an accurate method to predict heat transfer and temperature under rarefied condition. However, it is not convenient for complex geometry like a UNF canister. Computational Fluid Dynamics (CFD) simulations are more convenient to apply but their accuracy for rarefied condition are not well established. This work seeks to validate the use of CFD simulations to model heat transfer through rarefied gas in simple two-dimensional geometry by comparing the results to the more accurate DSMC method. The geometry consists of a circular fuel rod centered inside a square cross-section enclosure filled with rarefied helium. The validated CFD model will be used later to accurately estimate the temperature of an UNF canister subjected to vacuum drying condition.

In this work, a geometrically-accurate two-dimensional (2D) computational fluid dynamic (CFD) model of a used nuclear fuel cask, that can contain up to 32 pressurized water reactor (PWR) used nuclear fuel (UNF) assemblies, is constructed. This model is similar to the TN-32 cask employed in the ongoing high-burnup (HBU) Spent Fuel Data Project lead by the Electric Power Research Institute (EPRI). This model is used to predict the peak cladding temperature under vacuum drying conditions. Due to the symmetry of the cask, only one-eighth of the cross-section is modeled. Steady-state simulations that include the temperature-jump boundary conditions at the gas-solid interfaces are performed for different heat generation rates in the fuel regions and a range of dry helium pressures, from ∼10 5 to 100 Pa. These simulations include conduction within solid-gas regions and surface-to-surface radiation across all gas regions. The peak cladding temperatures are reported for various heat generation rates and rarefaction conditions, along with the maximum allowable heat generation that brings the cladding temperatures to the radial hydride formation limit. The results showed that the decrease of helium pressure significantly increased the temperature of the cladding material compared to the atmospheric pressure condition.

To perform the integrity assessment of a reactor pressure vessel (RPV) related to Pressurized Thermal Shock (PTS), we model the RPV using the 3D finite element method (FEM). Accurate prediction of temperature and stress fields is determined by using 2-Phase computational fluid dynamics (CFD) simulation in combination with an appropriate finite element discretization of the RPV wall. The cladding and the ferritic low alloy steel are considered as two separated layers, which can be intersected by superficial cracks. The calculation of the stress intensity factor (SIF) in mode I is based on the linear fracture mechanics theory (LEFM) and hypothetical cracks are located in different locations to consider the most critical cases. In the present study, the sub-modeling technique is implemented to refine the mesh required by the fracture analysis in the region of interest. Three types of cracks are considered: axial, circumferential and inclined. The performed integrity assessment uses the master curve approach. The stress intensity factor in the deepest point of a surface crack was compared with the material’s fracture toughness. In previous studies the integrity of the RPV subjected to medium and small break Loss-of-Coolant Accident (MBLOCA and SBLOCA, respectively) has been evaluated, therefore the concern in this contribution is the large break of Loss-Of-Coolant Accident (LBLOCA). The combination of 3D FEM with CFD simulations allows us to study the influence of the dynamic cooling plume on the stress intensity in more detail than with the conventional one dimensional method.