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A printed circuit steam generator (PCSG) is being considered as the component for pressurized water reactor (PWR) type small modular reactor (SMR) that can further reduce the physical size of the system. Since a steam generator in many PWR-type SMR generates superheated steam, it is expected that dryout front oscillation can potentially cause thermal fatigue failure due to cyclic thermal stresses induced by the transition in boiling regimes between convective evaporation and film boiling. To investigate the fatigue issue of a PCSG, a reference PCSG is designed in this study first using an in-house PCSG design tool. For the stress analysis, a finite element method analysis model is developed to obtain the temperature and stress fields of the designed PCSG. Fatigue estimation is performed based on ASME Boiler and pressure vessel code to identify the major parameters influencing the fatigue life time originating from the dryout front oscillation. As a result of this study, the limit on the temperature difference between the hot side and cold side fluids is obtained. Moreover, it is found that the heat transfer coefficient of convective evaporation and film boiling regimes play an essential role in the fatigue life cycle as well as the temperature difference.

This study discusses the effect of using 232Th instead of 238U on the neutronic characteristics and the main operating parameters of the pressurized water reactor (PWR). MCNPX version 2.7 was used to compare the neutronic characteristics of UO2 with (Th, 235U)O2 and (Th, 233U) O2. Firstly, the infinity multiplication factor (Kinf), thermal neutron flux, and power distribution have been studied for the investigated fuel types. Secondly, the effect of Gd2O3 and Er2O3 on the Kinf and on the radial thermal neutron flux and thermal power has been investigated to distinguish which of them is more suitable than the other in reactivity management. Thirdly, to illustrate the effectiveness of 232Th in decreasing the inventory of both the actinides and non-actinides, the concentration of plutonium (Pu) isotopes and minor actinides (MAs) has been simulated with the fuel burnup. Besides, due to their large thermal neutron absorption cross-section, the concentrations of 135Xe, 149Sm, and 151Sm with the fuel burnup have been investigated. Finally, the main safety parameters such as the reactivity worth of the control rods (ρCR), the effective delayed neutron fraction βeff, and the Doppler reactivity coefficient (DRC) were calculated to determine to which extent these fuel types achieve the acceptable limits.

Multiple failures of main steam isolation valve (MSIV) stem components have occurred at nuclear power plants. Stems are typically composed of ASM SA564 Grade 630 precipitation hardened steel, commonly known as 17-4 PH steel, in the H1100 condition. Several past valve stem failures have been attributed to thermal embrittlement, a condition under which the stem base material exhibits degraded mechanical properties due to microstructural changes induced by long-term exposure to a temperature range of approximately 260–316 °C (500-600 °F). Failure mechanisms of embrittled MSIV stems include sudden fracture due to loads from test actuations during plant outages as well as in-service fatigue crack growth (FCG) caused by vibrations from the steam flow. Three MSIV stems were removed from service after nine years in a pressurized water reactor (PWR) and were metallurgically and mechanically evaluated to determine the severity and extent of thermal embrittlement present. Findings from these evaluations were used to develop a condition-based projection of risk over time of the embrittlement-induced failure of the stems. It was found that the stem material removed from service did not exhibit significant metallurgical or mechanical evidence of embrittlement and that the probability of failure of the material in its current aged condition was small.

Simulation of Computational Fluid Dynamic is applied to present the thermal performance of water-based Yttrium oxide nanofluid in subchannel of pressurized water reactor (PWR) system. Thermal hydraulic aspect such as pressure drop and heat transfer are estimated in typical conditions of pressurized water with flow rates ranged (20×103≤Re≤80×103) using fresh water (0 %vol.) and different volume fraction of water-Yttrium oxide nanofluid (2 and 4% vol.) as coolant fluid. Results were obtained and compared with correlations of single-phase pressure drop and convective heat transfer for the case of fully developed turbulent flow. The addition of Yttrium oxide nanoparticles to the coolant fluid in pressurized water reactor led to increase in convective heat transfer coefficient and pressure drop. Increasing the nanoparticle volume fraction of (2 and 4% vol.) causing an increase in the average Nu by 3.46% and 7.61%, respectively. The CFD model established in ANSYS software was validated by comparing the pressure drop of CFD results with Blasius correlation and Nu with Ma¨ıga et al. correlation and gave a good agreement.

This paper describes the methods and results of analytical TEM examinations and autoclave testing of two highly-irradiated flux thimble tube materials harvested from a commercial pressurized water reactor. The materials are cold-worked 316L, and accumulated 65 dpa and 100 dpa of radiation dose. To set the baseline for a broader study, the materials were examined in the as-irradiated condition and tested as O-ring specimens at relatively high constant loads in simulated PWR water conditions. Tests were also conducted with elevated hydrogen. For a given load, more rapid cracking was associated with higher radiation dose, and with the elevated hydrogen.

Exposure and slow strain rate tensile tests were conducted in a simulated pressurized water reactor (PWR) primary water to investigate the oxidation resistance and stress corrosion cracking (SCC) susceptibility of 308L and 309L stainless steel (SS) cladding layers. A double-layer structure oxide layer grown on 308L SS and 309L SS contained the Cr-enriched nanocrystalline internal layer and the Fe-enriched spinel oxide in the external layer. Ni-enrichment at the matrix/oxide boundary was observed. The internal oxide film on 309L SS was thicker and had a lower Cr content than that on 308L SS. Preferential dissolution of inclusions led to pits on 308L SS and 309L SS surfaces during the exposure tests. More inclusions in 309L would decrease its SCC resistance due to the pits’ ability to act as the SCC initiation site. 308L SS had a lower susceptibility to SCC than 309L SS in PWR primary water. Lower ferrite content and higher strength/hardness reduced the oxidation and SCC resistance of 309L SS cladding. The effect of ferrite on oxidation and SCC of the SS claddings is discussed.

Radioactive corrosion product materials collected from the control rod drive mechanism (CRDM) housing in a pressurized water reactor (PWR, HANBIT-1 KNPP) were analyzed using an electron probe micro...

This work is an experimental study into the hydrodynamics of the coolant flow in the in-vessel pressure duct of a pressurized water reactor. In the article, an experimental testbench and a nuclear reactor model under investigation are described, the measurement methods are set forth, and the operating variables at which the study was conducted and the obtained results are provided. In the experiments, the mixing of the loop coolant flows inside the model of the nuclear reactor downcomer was simulated. The study was conducted on the high-pressure aerodynamic testbench of Alekseev State Technical University in Nizhny Novgorod. The scale model of the nuclear reactor had the structural components characteristic of loop-type reactor units, such as the annular downcomer and the bottom pressure vessel. The experiments were conducted at Reynolds numbers within the 20 000–50 000 range measured in the annular gap of the downcomer of the model. The axial velocity field at the inlet to the reactor core simulator was investigated using a pneumometric probe. The temperature field was recorded in the experiment by the impurity diffusion method, i.e., by introducing of a contrast tracer into one of the loops of the model. The degree of mixing the flows was estimated by the admixture (tracer) concentration at the inlet to the core simulator. Propane was used as the contrast admixture. The study has yielded the spatial distribution of the tracer in the coolant flow in the annular downcomer and in the bottom pressure vessel. The data on the distribution of the contrast admixture are presented in the form of charts. The swirling of the coolant flow in the in-vessel pressure duct has been analyzed. It has been shown that the mixing intensity in the bottom pressure vessel is affected by the central vortex with the central axis. The parameters of mixing the admixture in the model of the in-vessel pressure duct have been estimated.

Inconel 690TT has an excellent high-temperature corrosion resistance and has been widely applied in steam generators of pressurized water reactor (PWR) nuclear power plants. In this paper, Inconel ...

This work presents the neutronic analysis of fuel design for a long-life core in a pressurized water reactor (PWR). In order to achieve a high burnup, a high enrichment U-235 is traditionally considered without special constraints against proliferation. To counter the excess reactivity, Erbium was selected as a burnable poison due to its good depletion performance. Calculations based on a standard fuel model were carried out for the PWR type core using SRAC code system. A parametric study was performed to quantify the neutronically achievable burnup at a number of enrichment levels and for a numerous geometries covering a wide design space of lattice pitch. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are also investigated. It was found that it is possible to achieve sufficient criticality up to 100 GWd/tHM burnup without compromising the safety parameters.

For the structural integrity evaluation of pressurized water reactor (PWR) steam generator (SG) tubes subjected to transient hydraulic loading, determination of the tube-to-tube gap velocity and static pressure distributions along the tubes is prerequisite. This paper addresses both computational fluid dynamics (CFD) and analytical approaches for predicting the tube-to-tube gap velocity and static pressure distributions during blowdown following a feedwater line break (FWLB) accident at a PWR SG. First of all, a comparative study on CFD calculations of the transient velocity and pressure distributions in the SG secondary sides for two different models having 30 or no tubes is performed. The result shows that the velocities of sub-cooled water flowing between any adjacent two tubes of a tubed SG model during blowdown can be roughly estimated by applying the specified SG secondary side porosity to those of the no-tubed SG model. Secondly, simplified analytical approximate solutions for the steady two-dimensional SG secondary flow velocity and pressure distributions under a given discharge flowrate are derived using a line sink model. The simplified analytical solutions are validated by comparing them to the CFD calculations.

A Large-Break Loss of Coolant Accident (LB-LOCA) in a generic, 1000 MWe and 3-loop Chinese Pressurized Water Reactor (PWR) with the In-Vessel corium Retention (IVR) strategy was simulated by using ASTEC V2.1.1. One of the criteria of assessing the validity of IVR is that the decay heat of molten pool could be removed successfully, which means that the local heat flux will not exceed the Critical Heat Flux (CHF) of outer nucleate boiling. ASTEC determines the process of core degradation, the corium relocation and the transient heat flux to indicate whether the vessel can maintain its integrity under the LB-LOCA. From the core degradation process, for this LB-LOCA scenario, the corium with fission products slumps into the lower plenum via the downcomer, which means the sideward relocation. After the initial relocation, the mass of compositions of molten pool in lower head are almost steady except the mass of steel. The augmentation of the steel in the lower head is consistent with the melting process of the lower support plate, the fluid distribution structures and the vessel inner wall. At the azimuth from 78° to 87°, the transient heat flux is much larger compared with the other locations. The transient heat flux reaches its maximum value which is about 1.1 MW/m2 at the azimuth of 84°. And the maximal ratio of the transient local heat flux to CHF is about 0.8. For the residual thickness of the lower head, the minimum is 37.35 mm at the azimuth from 78° to 87°. The results of thermal analysis conclude that the vessel can maintain its integrity during LB-LOCA scenario because the transient heat flux from the vessel wall to the external coolant do not exceed the CHF of outer nucleate boiling.

The study is aimed at performing calculated substantiation of the thermal reliability of a pressurized water reactor (PWR) with spherical fuel elements, commonly known as a fixed-bed nuclear reactor, and determining the most advantageous design of the spherical packing from the view point of thermal hydraulics. Spherical fuel elements have a number of advantages over cylindrical fuel rods; in particular, they feature better retention of fission products, enhanced nuclear safety (due to a high melting temperature of ceramic materials), more intense heat transfer due to increased coolant flow turbulence, and a reduced influence of thermal cyclic loads on fuel elements. To confirm these statements, a thermohydraulic calculation of the KLT-40S type reactor with a modified intra-channel filling of fuel assemblies (FAs) consisting of spherical fuel elements was carried out. To determine the optimal spherical filling, two types of spherical packing were calculated, with three different diameters of spherical fuel elements. In the course of the calculation, solutions to the following issues were proposed: how to take into account the channels flow area variability, how to calculate the Reynolds number for channels of a given shape, what formulas should be used to determine the Nusselt number, and how to determine the hydraulic resistance in the channels. As a result of the calculation, data on the following main thermohydraulic characteristics have been obtained: surface heat flux density, heat transfer coefficient, maximum fuel temperature, and hydraulic losses. These results were compared with the results of calculations for cylindrical fuel rods. The obtained results demonstrate the advantage of spherical fuel elements over cylindrical fuel rods in a number of basic parameters, which gives prospects for further study of the use of spherical fuel elements in reactors of this type. The obtained study results can be applied in designing reactor plants of low and medium capacity, as well as in modernizing the existing reactor plants.

This paper presents a numerical prediction of the transient hydraulic loads acting on the tubes and external supports of a pressurized water reactor (PWR) steam generator (SG) during blowdown following a sudden feedwater line break (FWLB). A simplified SG model was used to easily demonstrate the prediction. The blowdown discharge flow was treated as a flashing flow to realistically simulate the transient flow fields inside the SG and the connected broken feedwater pipe. The effects of the SG initial pressure or the broken feedwater pipe length on the intensities or magnitudes of transient hydraulic loads were investigated. Then predictions of the decompression pressure wave-induced impulsive pressure differential loads on SG tubes and the transient blowdown loads on SG external supports were demonstrated and the general aspects of transient responses of such transient hydraulic loads to the FWLB were discussed.

Grain boundary (GB) oxidation of proton-irradiated 304 nuclear grade stainless steel in primary water of pressurized water reactor was investigated. The investigation was conducted by studying microstructure of the oxide and oxide precursor formed at GB on an "atom-by-atom" basis by a combination of atom-probe tomography and transmission electron microscope. The results revealed that increasing irradiation dose promoted the GB oxidation, in correspondence with a different oxide and oxide precursor formed at the GB. Correlation of the oxide and oxide precursor with the GB oxidation behavior has been discussed in detail.

The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized water reactor is analyzed by using Computational Fluid Dynamics (CFD) method in this study. The geometry model of the full-scaled reactor vessel is built, which includes the cold and hot legs, downcomer, lower plenum, core, upper plenum, top plenum, and is verified with some parameters in DCD. Under normal condition, it is found that the flow skirt, core plate holes and outlet pipe cause pressure loss. The maximum and minimum flow coefficient is 1.028 and 0.961 respectively, and the standard deviation is 0.019. Compared with other reactor type, it shows relatively uniform of the flow distribution at the core inlet. The coolant mixing coefficient is investigated with adding additional variables, showing that mass transfer of coolant occurs near the interface. The coolant mainly distributes in the 90° area of the corresponding core inlet, and mixes at the interface with the coolant from the adjacent cold leg. 0.1% of corresponding coolant is still distributed at the inlet of the outer-ring components, indicating wide range of mixing coefficient distribution.

Reliable prediction of nuclear fuel rod behavior is of prime concern for assessing proper design parameters as well as operating conditions and sustainability aspects in a nuclear reactor. In this study, a numerical analysis has been conducted to investigate the thermomechanical behavior of solid and annular fuel rods individually, considering a pressurized water reactor (PWR). The transient heat conduction equation has been employed for the heat-generating uranium dioxide pellet along with the surrounding zircaloy cladding and solved numerically using the finite element method. Loss of coolant conditions has been simulated by reducing the coolant heat transfer coefficient and compared with the normal operating conditions. In all the cases considered, the annular fuel has shown superior properties in terms of minimized fuel surface heat flux, lower but relatively uniform radial temperature, and thermal strain distribution.

A small leak occurring on the surface of a fuel rod due to damage exposes UO2 to a steam atmosphere. During this time, fission gas trapped inside the fuel rod leaks out, and the gas leakage can be increased due to UO2 oxidation. Numerous studies have focused on the steam oxidation and its thermodynamic calculation in UO2. However, the thermodynamic calculation of the UO2 oxidation in a pressurized water reactor (PWR) environment has not been studied extensively. Moreover, the kinetics of the oxidation of UO2 pellet also has not been investigated. Therefore, in this study, the thermodynamics of UO2 oxidation under steam injection due to a damaged fuel rod in a PWR environment is studied. In addition, the diminishing radius of the UO2 pellet with time in the PWR environment was calculated through an experiment simulating the initial time of steam injection at the puncture.

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.

The crevice stress corrosion cracking (SCC) susceptibility of austenitic stainless steels was evaluated in simulated pressurized water reactor (PWR) environments. To simulate the abnormal condition in temporary clamping devices on leaking small bore pipes, crevice bent beam (CBB) tests were performed in the oxygenated as well as hydrogenated conditions. No SCC cracks were found for SS316 in both conditions. SS304 also showed good resistance in the hydrogenated condition. However, all SS304 specimens showed SCC cracks in the oxygenated condition, indicating poor crevice SCC resistance. It was found that residual ferrites were selectively dissolved because of the galvanic corrosion coupled with the neighbouring austenite phase, resulting in SCC initiation in SS304. Crack morphologies were mostly transgranular assisted by the damaged δ-ferrite and deformation-induced slip bands.

This paper is focused on overcoming reactor shutdown malfunction and diagnostics due to poor cooling and water drop level within a pressurized water reactor. So the temperature coupling analysis be...

Corrosion behavior and mechanism of irradiated 304 nuclear grade stainless steel were studied in simulated pressurized water reactor primary water. The microstructure of the oxide formed on the steel irradiated to different doses over an exposure period range of 25–1500 h was analyzed and compared. It was found that the general and intergranular corrosion rates of the steel were increased with irradiation dose, in correspondence with an evolution of the general oxide and the oxide formed at the grain boundary. Correlation of the oxide evolution with the corrosion kinetics and mechanism has been discussed in detail.

The structural integrity of welded joints in the reactor pressure vessel (RPV) is directly related to the safety of nuclear power plants. The RPV is made from SA508-III steel in a pressurized water reactor. In this study, we investigated the effects of temperature on the tensile and fracture toughness properties of Chinese SA508-III welded joint in different sampling areas in order to provide reference data for structural integrity assessments of RPVs. The specimens used in tensile and fracture toughness tests were fabricated from the base metal (BM), weld metal (WM), and the heat-affected zone (HAZ) in the welded joint. The representative testing temperatures included the ambient temperature (20 °C), upper shelf temperature (100 °C), and service temperature (320 °C). The results showed that temperature greatly affected the fracture toughness (JIC) values for the SA508-III welded joint. The JIC values for BM and HAZ both decreased remarkably from 20 °C to 320 °C. The fracture morphologies showed that the BM and HAZ in the welded joint exhibited fully ductile fracture at 20 °C, whereas partial cleavage fracture was mixed in ductile fracture mode at 100 °C and 320 °C. The WM exhibited the ductile and cleavage fracture mixed mode at various temperatures, and the JIC values showed slight changes.

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.

High fidelity nuclear reactor fuel assembly simulation using CFD method is an effective way for the structure design and optimization. The validated models and user practice guidelines play critical roles in achieving reliable results in CFD simulations. In this paper, the international benchmark MATiS-H is studied carefully and the best user practice guideline is achieved for the rod bundles simulation. Then a 5 × 5 rod bundles model in the advanced pressurized water reactor (PWR) is established and the detailed three-dimensional thermal-hydraulic characteristics are investigated. The influence of spacer grids and mixing vanes on the flow and hear transfer in rod bundles is revealed. As the coolant flows through the spacer grids and mixing vanes in the rod bundles, the drastic lateral flow would be induced and the pressure drop increases significantly. In addition, the heat transfer is enhanced remarkably due to the strong mixing effects. The calculation results could provide meaningful guidelines for the design of advanced PWR fuel assembly.

Passive containment cooling system (PCCS) is an important passive safety facility in the large advanced pressurized water reactor. Using the physical laws, such as gravity and buoyancy, the water film/air countercurrent flow is formed in the external annular channel to keep inside temperature and pressure below the maximum design values. Due to the large curvature radius of the annular channel, one of the short arc segments is taken out, as a rectangular channel, to analyze the main water film evaporation heat transfer characteristics. Two numerical methods are used to predict the water film evaporative mass flow rate during the heat transfer process in the large-scale rectangular channel with asymmetric heating when the water film temperature is not saturated. At the same time, these numerical simulation results are validated by the experiment which is set up to study water film/air countercurrent flow heat transfer on a vertical back heating plate with 5 m in length and 1.2 m in width. It is shown that the maximum deviation between numerical simulation and experiment is 30%. In addition, the influences on these parameters, such as heat flux, evaporative mass flow rate, and water film thickness, are evaluated under the different tilted angles of the rectangular channel and horizontal plane, water/air inlet flow rates, water/air inlet temperatures, heating surface temperatures, and air inlet relative humidities. All these results can provide a good guidance for the design of PCCS in the future.

Small modular reactors (SMRs), an emerging nuclear power plant technology, are suitable for large grids as well as remote load centers and offer load following and frequency response capabilities. While the SMRs have expectedly higher response rates, detailed dynamic models including reactor dynamics are necessary for power system dynamic studies. This paper presents a dynamic model of an integral pressurized water reactor (iPWR)-type SMR, modeled in Siemens PTI PSS/E, to assess the contribution of the reactor to the power system dynamics. The proposed SMR model mimics the heat generation process and subsequent heat transfer process with the inclusion of the reactor core based on point kinetics, primary coolant based on natural circulation, and a simplified three lump representation of the steam generator. Controllers are designed to operate the turbine valve and reactor control rod in closed loops. The SMR model is integrated with the modified turbine-governor system and a power system study is conducted. Results show the power system and internal reactor responses when subjected to electrical demand variations of 20 ${\%}$ rated electrical output (REO) with a valve rate limit of $\pm \text{80}{\%}$ REO/min.

Nuclear steam generators (NSGs) are key components of nuclear power plants (NPPs) since their reliability affects the overall plant performance. Corrosion control of steam generators under operation and during plant outages is attained by proper design, adequate material selection and control of water chemistry in the entire secondary cooling circuit. This article reviews the types of steam generators, the materials used for their construction (with a special focus on tubing alloys), the corrosion processes that may occur under operation and the strategies to preserve steam generators and other components of the secondary cooling system from corrosion during plant outages.

In the lower chamber of pressurized water reactor (PWR), the flow distribution device is the core module to distribute coolant into the core. It has complex structure and numerous design parameters...

The vertically inverted U-tube steam generator (UTSG) is widely used in the pressurized water reactor (PWR). The reverse flow behavior generally exists in some U-tubes of a steam generator (SG) under both single- and two-phase natural circulations (NCs). The behavior increases the flow resistance in the primary loop and reduces the heat transfer in the SG. As a consequence, the NC ability as well as the inherent safety of nuclear reactors is faced with severe challenges. The theoretical models for calculating single- and two-phase flow pressure drops in U-tubes are developed and validated in this paper. The two-phase reverse flow characteristics in two types of SGs are investigated base on the theoretical models, and the effects of the U-tube height, bending radius, inlet steam quality and primary side pressure on the behavior are analyzed. The conclusions may provide some promising references for SG optimization to reduce the disadvantageous behavior. It is also of significance to improve the NC ability and ensure the PWR safety during some accidents. Keywords: Two-phase natural circulation, Reverse flow, UTSG, U-tubes

The pressurized water reactor (PWR) employs UO2 pellets as nuclear fuel, which are packed in zirconium alloy tubes called nuclear fuel cladding. In the manufacture of the nuclear fuel, machining scraps are generated which are not easily discarded as scraps because of its high cost. These zirconium nuclear alloys are very costly and are not produced in Brazil. In this work, novel methods to recycle Zircaloy scraps using vacuum induction melting and electron beam furnaces were used to obtain ingots. The cast ingots were subjected to thermal treatments and then chemically analyzed, followed by microstructural characterization, mechanical properties evaluation, and X-ray diffraction. The results indicated the feasibility of the processes for obtaining alloys for application in the nuclear area, chemical industry or materials for biological applications such as dental prostheses.

The article presents the results from experimental investigations into local coolant flow fluid dynamics that were carried out on the fragmental model of a VVER-type pressurized water reactor’s mixed core consisting of two types of fuel assemblies (FAs): TVSA-T (one segment) and TVSA-T.mod.2 (two segments). The coolant flow processes in the fuel rod bundle were simulated on an aerodynamic test bench. The pressure field in the flow was measured using a five-channel pneumometric probe. The obtained pressure field in the flow was recalculated for the coolant velocity vector according to the dependences derived during calibration. For drawing up a detailed flow motion pattern, a characteristic model cross section area was separated, which included the space between the assemblies and four fuel rod rows in each TVSA-type fuel assembly. The spatial distribution of coolant flow velocity projections was analyzed, as a result of which it became possible to reveal the regularities associated with the streamlining of spacer-, mixing-, and combined-spacer grids in TVSA assemblies by coolant; to determine coolant cross flows resulting from streamlining of hydraulically nonidentical grids and determine their localization in the experimental model longitudinal and cross sections; and to reveal the effect of accumulating flow hydrodynamic disturbances in the model longitudinal and cross sections resulting from the staggered arrangement of hydraulically nonidentical grids. The results obtained from investigations of inter-assembly interaction of coolant between the neighboring TVSA-T and TVSA-T.mod.2 assemblies have been adopted for practical use at AO Afrikantov OKBM in estimating the thermal reliability of VVER-type reactor cores and have been included in the database for verification of computation fluid dynamics computer programs (CFD codes) and detailed cell-wise numerical analysis of the core of VVER reactors.

A linearized mathematical model for the pressurized water reactor (PWR) core dynamics is presented based on conservation of energy and mass balance. The core and coolant system are treated as a lumped parameter while considering the delayed neutron effect. The developed model is controlled by negative feedback reactivity from the coefficients of reactivity in the fuel and reactor coolant. The model state variables are linearized first order differential equations with a steady state initial value and put in state space using Matlab/Simulink. The reactor core dynamic response is investigated through transients introduced to the core model inlets represented in external reactivity insertion eg. (0.001 Δk/k), 10% step decrease in core flow rate and 10 oF step decrease in core inlet coolant temperature. The simulation results demonstrated the role of feedback reactivity form coolant temperature and Doppler Effect in stabilizing the core power and affecting the steady state values of core variables.

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.

Stress corrosion cracking (SCC) of an SA508-309 L/308 L–316 L dissimilar metal weld joint in primary pressurized water reactor environment was investigated by the interrupted slow strain rate tension tests following a microstructure characterization. The 308 L weld metal shows a higher content of δ ferrite than the 309 L weld metal. In addition, no obvious Cr-depletion but carbides precipitation at γ/δ phase boundaries was observed in both 308 L and 309 L weld metals. The slow strain rate tension tests showed that the SCC susceptibility of the base and weld metals of the dissimilar metal weld joint follows the order of SA508

In a previous study, delayed hydride cracking (DHC) assessment of pressurized water reactor (PWR) spent fuel during dry storage using the threshold stress intensity factor (KIH) was performed. However, there were a few limitations in the analysis of the cladding properties, such as oxide thickness and mechanical properties. In this study, those models were modified to include test data for irradiated materials, and the cladding creep model was introduced to improve the reliability of the DHC assessment. In this study, DHC susceptibility of PWR spent fuel during dry storage depending on the axial elevation was evaluated with the improved assessment methodology. In addition, the sensitivity of affecting parameters such as fuel burnup, hydride thickness, and crack aspect ratio are presented. Keywords: DHC, PWR spent fuel, dry storage, Sensitivity

After several years in the reactor core, irradiated nuclear fuel is handled and subsequently stored for a few years under water next to the core, to achieve thermal cooling and decay of very short-lived radionuclides. Thereafter, it might be sent to dry-cask interim storage before final disposal in a deep geological repository. Here, the spent nuclear fuel (SNF) is subject to a series of physicochemical phenomena which are of concern for the integrity of the nuclear fuel cladding. After moving the SNF from wet to dry storage, the temperature increases, then slowly decreases, leading the hydrogen in solid solution in the cladding to precipitate radially with consequent hydride growth and cladding embrittlement (Kim, 2020). Another phenomenon affecting the physical properties of the cladding during interim dry storage is the irradiation damage produced in the inner surface of the cladding by the alpha decay of the actinides present at the periphery of the pellet, particularly when the burnup at discharge is high. SNF pellets with high average burnup present larger fuel volumes at the end of their useful life due to accumulation of insoluble solid fission products and noble gases, which leads to disappearance of the as-fabricated pellet–clad gap. Further swelling is expected as a consequence of actinide decay and the accumulation of helium. This leads to larger cladding hoop stress and larger alpha decay damage. The present work first investigates the variation in diameter caused by pellet swelling in an irradiated Zircaloy-4 cladding after chemical digestion of the uranium oxide (UOx) pellet. Second, the irradiation damage produced during the 30 years elapsed since the end of irradiation in terms of displacements per atom (dpa) is studied by means of the FLUKA Monte Carlo code. The irradiation damage produced by the decay of actinides in the inner surface of the cladding extends for less than 3 % in depth. The considered cladded UOx pellet was extracted from a pressurized water reactor (PWR) fuel rod consisting of five segments, with an average burnup at discharge of 50.4 GWd (tHM)−1.

The safe-end of a steam generator (SG) nozzle dissimilar metal weld (DMW) for pressurized water reactors (PWRs) is the weakest point of failure which is crucial for the safe operation of a nuclear power station. Related to materials micro-structures, a uniaxial stress–strain relationship is the basic input parameter for nuclear power plant design, safety evaluation, and life management. In this paper, the micro-structure and tensile mechanical properties of a DMW of a European pressurized water reactor (EPR) were studied. Vickers hardness tester, optical microscope, and electron back scatter diffraction were used to analyze the micro-structure of the DMW joint. In addition, the residual strain of the DMW joint base material, heat-affected zone, weld metal, and fusion boundary region were studied. Based on digital image correlation (DIC) technology, tensile mechanical properties of the DMW joint were obtained. The results show that an accurate tensile stress–strain relationship of dissimilar metal welded joints can be obtained by using the DIC technique, the weld is the relatively weak link, and the residual strain is concentrated in the heat-affected zone. This study provides valuable engineering information regarding nuclear power plant design, in-service performance testing, and structural analysis and evaluation of welds containing defects.

A 3-dimensional (3D) fuel performance analysis program, able to simulate normal operating conditions and accident conditions for PWR fuel behaviors, was developed based on the Multiphysics Object-Oriented Simulation Environment (MOOSE) finite-element framework. By taking fission products swelling, densification and expansion of pellet, thermal and irradiation creep, gap heat transfer, fission gas release, and cladding crack propagation into consideration, detailed fuel behaviors have been simulated in a multiphysics coupling way. Local defects in fuel pellet caused during manufacturing and filling processes known as the missing pellet surface (MPS) can cause abnormal stress distribution of the cladding and it could even lead to cladding failure. Taking Stress Corrosion Cracking (SCC) phenomenon into consideration, a simulation of PWR fuel rodlet that consists of a pellet with an MPS defect and an intact pellet was conducted. The fuel rod has experienced with sorts of events, including normal operating conditions and a high-power ramp event. The simulation results indicated that: 1) The MPS defect affects the temperature and displacement distribution in the vicinity of the MPS defect. When the pellets are in contact with the cladding, the inner surface of the cladding presents a large tensile hoop stress, which accelerates the crack propagation. 2) During the ramp event, the crack propagation rate was higher than that under normal condition and crack length expanded by about 0.1 µm.

Nanofluid has drawn significant attention in enhancing the cooling performance and safety aspects of nuclear reactor technology. Herein, a numerical analysis has been accomplished to explore the thermal transport in subchannel geometry of a typical pressurized water reactor (PWR) replete with Al2O3–TiO2/water hybrid nanofluid. Constant heat flux is maintained on the nuclear fuel outer surface, while nanoparticle concentration and Reynolds number are varied within the subchannel. The computational fluid dynamics (CFD) model has been solved in ANSYS FLUENT adopting the steady-state k-e turbulence model. The results suggest a considerable enhancement in heat transfer coefficient, Nusselt number, and pressure drop upon increasing hybrid nanoparticle concentration in the base fluid.

A method for automated generation program for nuclear fuel reloading patterns in Pressurized Water Reactor (PWR) has been developed. This newly-developed method consists of six different steps to minimize the maximum F ΔH value, and maximize the reactor cycle length. Step 1 is initial fuel placement that is expected to produce the longest cycle length possible with the selected Fuel Assemblies (FAs) for the current cycle. Step 2 is aiming to decrease the F ΔH value of the FA with the maximum F ΔH . Step 3 aims to increase the F ΔH value of the old FA with the lowest F ΔH . Step 4 is rotating FA with the lowest F ΔH value to increase its F ΔH value, and rotating several old FAs in the neighboring FA with the maximum F ΔH value to decrease the maximum F ΔH value. Step 5 is aiming to increase the F ΔH value of FA with the lowest F ΔH value. The last step or step 6, will try to move FAs that have high k ∞ in the periphery zone, inward to increase the cycle length of the reactor. These steps are translated into code in the Python programming language to enable automatic execution in a computer. A 3D nuclear reactor core neutronic code, COCO, is used for the calculation of F ΔH value and reactor cycle length. Every nuclear power plant designer company will have their F ΔH peaking factor safety limit in accordance with their DNB experiments and calculations, and the F ΔH value safety limit used in this research is 1.46. A PWR loading pattern model is used to test this method. During the test, all the steps in this method are successfully executed in a total of 25 iterations plus one initialization calculation and produced acceptable results. The results of this method are all of the loading patterns found in all steps which have the maximum F ΔH value below the defined criterion values. In the mentioned PWR loading pattern model, four optimized loading patterns are found using this method, all of which can be selected in the PWR refueling loading pattern design.

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.

The neutron and photon flux rates are important parameters for safe reactor operation, refueling and decommissioning, scientific applications and radiation protection. For the Integrated small pressurized water reactor, the advanced reactor core analysis code CORAL, the source calculation code ORIGEN-II and the Monte Carlo code SuperMC are used to establish the reactor flux rate calculation model under normal operation and shutdown refueling condition. The results show that (1) In the normal operation of the reactor, the neutron flux rate is attenuated by 10 orders of magnitude from the outermost component to the inner surface of the pressure vessel, and the shielding effect of the coolant on neutrons is more significant. The neutron flux of the inner surface of the pressure vessel in 40 years is 3.723 × 1014 ncm2; the neutron flux in 60 years is 5.585 × 1014 ncm2. The photon flux rate is reduced by 10 orders of magnitude from the periphery of the core to the outer surface of the pressure vessel. High-quality density materials have better photon shielding effects. (2) In the case of reactor shutdown and refueling, the neutron flux rate is much smaller than the photon flux rate. On the outer surface of the pressure vessel, the maximum neutron and photon dose rates are 7.74 × 10−10 mSv · h−1 and 6.97 × 10−5 mSv · h−1, respectively, which belong to the supervision area. When the cover is opened, the radiation dose value of the workplace at the top of the reactor is less than 0.0025 mSv · h−1, which can ensure the radiation safety of the operation.

The properties of the oxide films formed on solution-annealed and cold-worked 316L stainless steel (SS) specimens with and without charged hydrogen in deaerated pressurized water reactor primary water at 300 °C were investigated. The outer oxide layers of all specimens were composed of magnetite (Fe3O4) and NiFe2O4. Charged hydrogen resulted in larger outer iron-bearing oxide particles forming due to hydrogen-enhanced outward diffusion of iron cations. Prior cold-work accelerates the oxidation was observed. Charged hydrogen led to local cracks in the oxide film and enhanced the penetration oxidation beneath the metal/oxide interface. The Cr-rich inner oxide layer grown on the prior cold-worked specimen with charged hydrogen was thicker than that on the cold-work specimen or the hydrogen-charged specimen, revealing the combined effects of charged hydrogen and prior cold-work on the acceleration of the oxidation process. The working mechanism of the solid-state hydrogen effect on the oxide film was discussed.

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI utilizes reference fatigue crack growth rate (FCGR) curves for flaw evaluations. Code Case N-809 describes a reference curve currently used in flaw evaluations of austenitic stainless steels exposed to a pressurized water reactor (PWR) environment in accordance with ASME, Section XI. Recently, an extensive database of Type 304, Type 304L, and Type 304/304L dual-certified stainless steels and the corresponding weld metal in PWR environments was assessed and an updated FCGR model generated. This database includes previously unreported FCGR data in 100°C PWR environments at an R-ratio of 0.7 and a rise time of either 51 sec or 510 sec. The results of this lower temperature testing are reported here and do not support the non-Arrhenius temperature relation in Code Case N-809, which predicts an increase in FCGRs with decreasing temperature below a temperature of 150°C. The updated model more accurately describes the FCGR behavior in the near-threshold, low ΔK regime. Additionally, the updated model eliminates the non-Arrhenius temperature dependence of the Code Case N-809 reference curves for temperatures below 150°C and replaces it with a single Arrhenius temperature dependence between 100°C and 338°C. Similar to the Code Case N-809 reference curve, this model does not describe the severely retarded FCGR behavior that has been observed to occur for austenitic stainless steel under certain conditions, nor does it attempt to predict the conditions under which severe retardation is likely to occur.

Alloy 82/182 Dissimilar Metal (DM) butt welds have been the subject of significant work due to their susceptibility to primary water stress corrosion cracking (PWSCC). The current rules for inspection of DM butt welds are contained in ASME Code Case N-770 (latest revision). These rules are based on crack growth rates that were originally published in MRP-115 in 2004. Since that time significant work has been performed to better understand the crack growth behavior of Alloy 82/182 materials in pressurized water reactors (PWRs) and updated crack growth rates have recently been published in MRP-420 (Rev. 1). These updated crack growth rates show a significant difference at lower temperatures which line up with the cold leg temperatures for PWRs (approx. 540–560°F). Additionally, the US Nuclear Regulatory Commission (NRC) working along with the Electric Power Research Institute (EPRI) recently released the xLPR computer code which is a probabilistic fracture mechanics code for evaluating DM butt welds. This paper evaluated various cold leg DM butt welds for PWRs using the xLPR probabilistic computer code to determine if a change in the frequency between inspections for unmitigated DM butt welds (Item B-1) from their current requirements in ASME Code Case N-770 is warranted. Failure probabilities were determined for the current technical basis using MRP-115 crack growth rates (with the current inspection duration) and compared to the probabilities using the new crack growth rates in MRP-420, Rev. 1 (with increased inspection durations). The evaluation results demonstrate that the duration between volumetric examinations for small and medium diameter DM butt welds from Code Case N-770, Inspection Item B-1, can be safely increased to once per 10-year ISI interval from the current requirement of once every 7 years when the specific requirements described in this paper are met.