Search

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.

Meta-heuristic algorithms have found their place in optimization problems. Henry gas solubility optimization (HGSO) is one of the newest population-based algorithms. This algorithm is inspired by Henry's law of physics. To evaluate the performance of a new algorithm, it must be used in various problems. On the other hand, the optimization of the proportional–integral–derivative (PID) gains for load-following of a nuclear power plant (NPP) is a good challenge to assess the performance of HGSO. Accordingly, the power control of a pressurized water reactor (PWR) is targeted, based on the point kinetics model with six groups of delayed-neutron precursors. In any optimization problem based on meta-heuristic algorithms, an efficient objective function is required. Therefore, the integral of the time-weighted square error (ITSE) performance index is utilized as the objective (cost) function of HGSO, which is constrained by a stability criterion in steady-state operations. A Lyapunov approach guarantees this stability. The results show that this method provides superior results compared to an empirically tuned PID controller with the least error. It also achieves good accuracy compared to an established GA-tuned PID controller.

A two-phase natural circulation test using SMART integral test loop (SMART–ITL) was conducted to explore thermo-hydraulic phenomena of two-phase natural circulation in the SMART reactor. Specifically, the test examined the natural circulation in the primary loop under a stepwise coolant inventory loss while keeping the core power constant at 5% of the scaled full power. Based on the test results, three flow regimes were observed: single-phase natural circulation (SPNC), two-phase natural circulation (TPNC), and boiler–condenser natural circulation (BCNC). The flow rate remained steady in the SPNC, slightly increased in the TPNC, and dropped abruptly and maintained in the BCNC. Using a natural circulation flow map, the natural circulation characteristic in the SMART–ITL was compared with those in pressurized water reactor simulators. In the SMART–ITL, a BCNC regime appeared instead of siphon condensation and reflux condensation regimes because of the use of once-through steam generators.

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.

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

During the long-term cooling (LTC) phase of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR), water is supplied from the containment sump to the reactor coolant system (RCS) by the flooded sump water to the Reactor Vessel (RV) through the broken pipes. As part of the technical efforts for resolving GSI-191 [( Reid and Crytzer, May. 2007) 1 , consideration is needed for the consequences of debris penetrating the sump screen and propagating downstream into the RV. Injection of debris (fiberglass) into the RV during the LTC recirculation phase needs special attention to assure that reactor core cooling is maintained. The point of concern is the potential for debris to adversely affect the reactor core flow paths or heat transfer [ 2 ]. However, all the experiments for proving the coolability of RV have been done with the assumption of the most of debris would be transferred to the RV and the bottom nozzle of the FAs. The purpose of the tests is to quantify the amount of the debris that would be accumulated at the lower plenum and the debris that passes through the FAs since non-conservatism of other researches assumptions that have been used in the past experimental or analytical programs.

The pressure control in the pressurized water reactor (PWR) primary loop is key to secure the safe operation. The pressurizer (PZR) unit is responsible for attaining this task. Thus, the PZR unit’s...

In Korea, the pressurized water reactor (PWR) is the most common nuclear power plant. Twenty-four PWRs are operating commercially or are under construction. According to the public requirements of the reactor pressure vessel (RPV) surveillance program, all PWR plants have surveillance capsules. One type of PWR, the OPR1000 nuclear power plant has a design life of 40 years. Before the plant began operation, six surveillance capsules were installed at the OPR1000 RPV inner wall in the downcomer region to obtain more accelerated embrittlement characteristics of the RPV material. However, the lead factor defined as the ratio of the fast (E > 1.0 MeV) neutron flux at the surveillance capsule to the fast neutron flux at the RPV peak fluence location had been estimated at about 1.3 for OPR1000 plants. Therefore, obtaining high-dose irradiation embrittlement data for life extension is difficult. Supplemental surveillance capsules were fabricated and installed at the Westinghouse plant, which has a relatively high lead factor to overcome the small lead factors of the OPR1000. This paper discusses the fabrication and installation of supplemental surveillance capsules and the expected withdrawal schedule based on neutron transport calculations and dosimetry evaluations. As a result, 60 years of RPV embrittlement data for OPR1000 are expected to be obtained if the supplemental capsule is irradiated at the Westinghouse capsule holder for 7 years approximately.

A detailed geometrical model of a Krško reactor core was developed using a Monte Carlo neutron transport code MCNP. The main goal of developing an MCNP core model is for it to be used in future research focused on ex-core calculations. A script called McCord was developed to generate MCNP input for an arbitrary fuel cycle configuration from the diffusion based core design package CORD-2, taking advantage of already available material and temperature data obtained in the nuclear core design process. The core model was used to calculate 3D power density profile inside the core. The applicability of the calculated power density distributions was tested by comparison to the CORD-2 calculations, which is regularly used for the nuclear core design calculation verification of the Krško core. For the hot zero power and hot full power states differences between MCNP and CORD-2 in the radial power density profile were

Measured decay heat data of light water reactor (LWR) spent nuclear fuel (SNF) assemblies are adopted to train machine learning (ML) models. The measured data is available for fuel assemblies irradiated in commercial reactors operated in the United States and Sweden. The data comes from calorimetric measurements of discharged pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies. 91 and 171 measurements of PWR and BWR assembly decay heat data are used, respectively. Due to the small size of the measurement dataset, we propose: (i) to use the method of multiple runs (ii) to generate and use synthetic data, as large dataset which has similar statistical characteristics as the original dataset. Three ML models are developed based on Gaussian process (GP), support vector machines (SVM) and neural networks (NN), with four inputs including the fuel assembly averaged enrichment, assembly averaged burnup, initial heavy metal mass, and cooling time after discharge. The outcomes of this work are (i) development of ML models which predict LWR fuel assembly decay heat from the four inputs (ii) generation and application of synthetic data which improves the performance of the ML models (iii) uncertainty analysis of the ML models and their predictions.

The main steam line break accident is an essential initiating event of the pressurized water reactor. In present work, the fuzzy set theory and the signal-based fault detection method has been used to detect the occurrence and diagnosis of the location and break area for the small scale MSLB. The models are validated by the AP1000 accident simulator based on MAAP5. From the test results it can be seen that the proposed approach has a rapid and proper response on accident detection and location diagnosis. The method proposed to evaluate the break area shows good performances for small scale MSLB with the relative deviation within ±3%.

This research serves to advance the development of engineering computational fluid dynamics (CFD) computing efficiency for the analysis of pressurized water reactor (PWR) core using rod-type fuel assemblies with mixing vanes (one kind of typical PWR core). In this research, a CFD scheme based on the reconstruction of the initial fine flow field (RIF CFD scheme) is proposed and analyzed. The RIF scheme is based on the quantitative regulation of flow velocities in the rod-type PWR core and the principle that the CFD computing efficiency can be improved greatly by a perfect initialization. In this paper, it is discovered that the RIF scheme can significantly improve the computing efficiency of the CFD computation for the rod-type PWR core. Furthermore, the RIF scheme also can reduce the computing resources needed for effective data storage of the large fluid domain in a rod-type PWR core. Moreover, a flow-ranking RIF CFD scheme is also designed based on the ranking of the flow rate, which enhances the utilization of the flow field with a closed flow rate to reconstruct the fine flow field. The flow-ranking RIF CFD scheme also proved to be very effective in improving the CFD efficiency for the rod-type PWR core.

A Nuclear Power Plant (NPP) is a complex dynamic system-of-systems with highly nonlinear behaviors. In order to control the plant operation under both normal and abnormal conditions, the different systems in NPPs (e.g., the reactor core components, primary and secondary coolant systems) are usually monitored continuously, resulting in very large amounts of data. This situation makes it possible to integrate relevant qualitative and quantitative knowledge with artificial intelligence techniques to provide faster and more accurate behavior predictions, leading to more rapid decisions, based on actual NPP operation data. Data-driven models (DDM) rely on artificial intelligence to learn autonomously based on patterns in data, and they represent alternatives to physics-based models that typically require significant computational resources and might not fully represent the actual operation conditions of an NPP. In this study, a feed-forward backpropagation artificial neural network (ANN) model was trained to simulate the interaction between the reactor core and the primary and secondary coolant systems in a pressurized water reactor. The transients used for model training included perturbations in reactivity, steam valve coefficient, reactor core inlet temperature, and steam generator inlet temperature. Uncertainties of the plant physical parameters and operating conditions were also incorporated in these transients. Eight training functions were adopted during the training stage to develop the most efficient network. The developed ANN model predictions were subsequently tested successfully considering different new transients. Overall, through prompt prediction of NPP behavior under different transients, the study aims at demonstrating the potential of artificial intelligence to empower rapid emergency response planning and risk mitigation strategies.

The digital acoustic model of a nuclear reactor (NRDAM) is represented as a self-oscillatory system belonging to a special class of nonlinear dissipative systems that can generate sustained oscillations whose parameters do not depend on the initial conditions and are only governed by the properties of the system itself. It has been found that a pressurized water reactor with coolant flowing in a turbulent mode is an open system of high complexity with a large number of components with links between them being probabilistic rather than predetermined in nature. The coolant loop components featuring negative dissipation (negative friction) are revealed. It is shown that chaotic turbulent pulsations and vortices are self-organized in these components into ordered wave oscillations, the frequency of which is determined according to the Thomson (Kelvin) formula. An electronic generator of self-oscillations with a transformer feedback used in radio engineering circuits has similar properties. A nozzle is an acoustic analog of a transformer. A negative resistance contained in nonlinear dynamic systems like a nozzle or a natural circulation loop results in that chaotic turbulent disturbances become self-organized, and self-oscillations are generated in the form of acoustic standing waves (ASW). Based on theoretical and experimental data, the certainty of the ability of a reactor together with the pipelines connected to it to simultaneously generate several ASWs—a property that has not been known previously—is confirmed. By using the NRDAM in designing and operation of nuclear power plants (NPPs), it becomes possible to reveal the sources of ASWs arising in the coolant, their occurrence conditions, and frequency. The use of the NRDAM is also necessary for determining the effect that the coolant circuit equipment geometrical parameters and layout have on the interaction of neutronic, thermal-hydraulic, and vibroacoustic processes. By applying the NRDAM, it becomes possible to optimize the engineering and design solutions in developing new-generation NPPs by eliminating the conditions causing the occurrence of undesirable self-oscillations of coolant and vibroacoustic resonances resulting from the coincidence of the ASW frequencies with the vibration frequencies of nuclear fuel and equipment in normal and emergency operation modes and also under the conditions of shock impacts and seismic loads.

Here we present a newly developed two-step oxidative-reductive in-service chemical decontamination process suitable for high temperature and high pressure circuit (HTHPC) decontamination in a pressurized water reactor (PWR). Various process parameters on the corrosion depth and decontamination factor were investigated by employing different kinds of detergents. Under optimal conditions, the average decontamination factor is about 60.0 and the maximum corrosion depth is 0.20 um. Actual decontamination by applying this process to HTHPC, the dose field in most locations decreased by more than 89% and it’s proved that the decontamination process is completely suitable for the decontamination of PWRs in service.

Nuclear emergency preparedness and response is an essential part to ensure the safety of nuclear power plant (NPP). Key support technologies of nuclear emergency decision-making usually consist of accident diagnosis, source term estimation, accident consequence assessment, and protective action recommendation. Source term estimation is almost the most difficult part among them. For example, bad communication, incomplete information, as well as complicated accident scenario make it hard to determine the reactor status and estimate the source term timely in the Fukushima accident. Subsequently, it leads to the hard decision on how to take appropriate emergency response actions. Hence, this paper aims to develop a method for rapid source term estimation to support nuclear emergency decision making in pressurized water reactor NPP. The method aims to make our knowledge on NPP provide better support nuclear emergency. Firstly, this paper studies how to build a Bayesian network model for the NPP based on professional knowledge and engineering knowledge. This paper presents a method transforming the PRA model (event trees and fault trees) into a corresponding Bayesian network model. To solve the problem that some physical phenomena which are modeled as pivotal events in level 2 PRA, cannot find sensors associated directly with their occurrence, a weighted assignment approach based on expert assessment is proposed in this paper. Secondly, the monitoring data of NPP are provided to the Bayesian network model, the real-time status of pivotal events and initiating events can be determined based on the junction tree algorithm. Thirdly, since PRA knowledge can link the accident sequences to the possible release categories, the proposed method is capable to find the most likely release category for the candidate accidents scenarios, namely the source term. The probabilities of possible accident sequences and the source term are calculated. Finally, the prototype software is checked against several sets of accident scenario data which are generated by the simulator of AP1000-NPP, including large loss of coolant accident, loss of main feedwater, main steam line break, and steam generator tube rupture. The results show that the proposed method for rapid source term estimation under nuclear emergency decision making is promising.

The lack of plant-side energy storage analysis to support nuclear power plants (NPP), has setup this research endeavor to understand the characteristics and role of specific storage technologies and the integration to an NPP. The paper provides a qualitative review of a wide range of configurations for integrating the energy storage system (ESS) to an operating NPP with pressurized water reactor (PWR). The role of ESS technologies most suitable for large-scale storage are evaluated, including thermal energy storage, compressed gas energy storage, and liquid air energy storage. The methods of integration to the NPP steam cycle are introduced and categorized as electrical, mechanical, and thermal, with a review on developments in the integration of ESS with an operating PWR. By adopting simplified off-design modeling for the steam turbines and heat exchangers, the results show the performance of the PWR steam cycle changes with respect to steam bypass rate for thermal and mechanical storage integration options. Analysis of the integrated system characteristics of proposed concepts for three different ESS suggests that certain storage technologies could support steady operation of an NPP. After having reviewed what have been accomplished through the years, the research team presents a list of possible future works.

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.

This paper presents the verification and validation (V&V) of a calculation module for isotope inventory prediction to control the back-end cycle of spent nuclear fuel (SNF). The calculation method presented herein was implemented in a two-step code system of a lattice code STREAM and a nodal diffusion code RAST-K. STREAM generates a cross section and provides the number density information using branch/history depletion branch calculations, whereas RAST-K supplies the power history and three history indices (boron concentration, moderator temperature, and fuel temperature). As its primary feature, this method can directly consider three-dimensional core simulation conditions using history indices of the operating conditions. Therefore, this method reduces the computation time by avoiding a recalculation of the fuel depletion. The module for isotope inventory calculates the number densities using the Lagrange interpolation method and power history correction factors, which are applied to correct the effects of the decay and fission products generated at different power levels. To assess the reliability of the developed code system for back-end cycle analysis, validation study was performed with 58 measured samples of pressurized water reactor (PWR) SNF, and code-to-code comparison was conducted with STREAM-SNF, HELIOS-1.6 and SCALE 5.1. The V&V results presented that the developed code system can provide reasonable results with comparable confidence intervals. As a result, this paper successfully demonstrates that the isotope inventory prediction code system can be used for spent nuclear fuel analysis.

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 presents an illustration of Nagra’s AMAC (Advanced Methodology for Activation Characterization) approach to the South Korean pressurized water reactor Kori-1 decommissioning. The results achieved are supported by the existing experience from the detailed AMAC applications to Swiss NPPs and are used not only for a demonstration of the applicability of AMAC to South Korean NPPs, but also for a first approximation of the activated waste volumes to be expected from Kori-1. A packaging concept based on the above activation characterization is also presented, using the AMAC algorithmic optimization software ALGOPACK leading to the minimum number of waste containers needed given the selected packaging constraints. Nagra’s AMAC enables effective planning before and during NPP decommissioning, including recommendations for cutting profiles for diverse reactor components and building structures. Finally, it is expected to lead to significant cost savings by reducing the number of expensive waste containers, by optimizing a potential melting strategy for metallic waste as well as by significantly limiting the number of radiological measurements. All information about Kori-1 used for the purpose of this study was collected from publicly available sources.

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 ...

The safe and secure designs of any nuclear power plant together with its cost-effective operation without accidents are leading the future of nuclear energy. As a result, the Reliability, Availability, Maintainability, and Safety analysis of NPP systems is the main concern for the nuclear industry. But the ability to assure that the safety-related system, structure, and components could meet the safety functions in different events to prevent the reactor core damage requires new reliability analysis methods and techniques. The Fault Tree Analysis (FTA) is one of the most widely used logic and probabilistic techniques in system reliability assessment nowadays. The Dynamic fault tree technique extends the conventional static fault tree (SFT) by considering the time requirements to model and evaluate the nuclear power plant safety systems. Thus this paper focuses on developing a new Dynamic Fault Tree for the Auxiliary Feed-water System (AFWS) in a pressurized water reactor. The proposed dynamic model achieves a more realistic and accurate representation of the AFWS safety analysis by illustrating the complex failure mechanisms including interrelated dependencies and Common Cause Failure (CCF). A Simulation tool is used to simulate the proposed dynamic fault tree model of the AFWS for the quantitative analysis. The more realistic results are useful to establish reliability cantered maintenance program in which the maintenance requirements are determined based on the achievement of system reliability goals in the most cost-effective manner.

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.

A comparative computational fluid dynamics (CFD) study was conducted on the three different types of pressurized water reactor (PWR) upper plenum, named TYPE 1 (support columns (SCs) and control rod guide tubes (CRGTs) with two large windows), TYPE 2 (SCs and CRGTs without windows), and TYPE 3 (two parallel perforated barrel shells and CRGTs). First, three types of upper plenum geometry information were collected, simplified, and adopted into the BORA facility, which is a 1/5 scale system of the four-loop PWR reactor. Then, the geometry, including the upper half core, upper plenum region, and hot legs, was built using the Salome platform. After that, an unsteady calculation to simulate the reactor balance operation at hot full power scenario was performed. Finally, the differences of flowrate distribution at the core outlet and temperature distribution and transverse velocity inside the hot legs with different upper plenum internals were compared. The results suggest that TYPE 1 upper plenum internals cause the largest flowrate difference at the core outlet while TYPE 3 leads to the most even distributed flowrate. The distribution and evolution pattern of the tangential velocity inside hot legs is highly dependent on the upper plenum internals. Two counter-rotating swirls exist inside the TYPE 1 hot leg and only one swirl revolving around the hog leg axis exist inside the TYPE 2 hot leg. For TYPE 3, two swirls like that of TYPE 1 rotating around the hot leg axis significantly increase the temperature homogenization speed. This research provides meaningful guidelines for the future optimization and design of advanced PWR upper plenum internal structures.

The small PWR has been paid more and more attention due to its diversity of application and flexibility in the site selection. However, the large core power density, the small containment space and the rapid accident progress characteristics make it difficult to control the containment pressure like the traditional PWR during the LOCA. The pressure suppression system has been used by the BWR since the early design, which is a suitable technique that can be applied to the small PWR. Since the configuration and operating conditions are different from the BWR, the pressure suppression system should be redesigned for the small PWR. Conducting the experiments on the scale down test facility is a good choice to reproduce the prototypical phenomena in the test facility, which is both economical and reasonable. A systematic scaling method referring to the H2TS method was proposed to determine the geometrical and thermohydraulic parameters of the pressure suppression containment response test facility for the small PWR conceptual design. The containment and the pressure suppression system related thermohydraulic phenomena were analyzed with top-down and bottom-up scaling methods. A set of the scaling criteria were obtained, through which the main parameters of the test facility can be determined.

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.

The BEAVRS benchmark problem describes detailed information such as the geometry and material specifications to construct a neutronic calculation model of a commercial Pressurized Water Reactor (PWR) core. The benchmark was modified several times; the main purpose of the current study is to investigate the effect of the modifications in the BEAVRS benchmark on sensitivities and uncertainties in cross section that in turn affect the multiplication factors and reactivity coefficients. The sensitivity coefficients are calculated for keff due to perturbation in cross-sections using the KSEN card of MCNP6 code for the two cores, where all control rods are in and out. Since MCNP6 cannot calculate uncertainties, the NJOY2021 was used to compute the relative covariance matrices of the cross-sections data library, thereafter, a python script was developed to calculate the uncertainties of keff as well as sensitivities and uncertainties of reactivity due to the control rods insertion. The results indicate that the uncertainties in reactivities caused by control rods insertion for the BV2 model is greater than the BV1 model, due to the effect of using higher coolant density in the nozzle and support plate structures in addition to the effect of the burnable absorber designs in the BV2 model.

Traditional light water reactor simulations are usually either high fidelity, requiring hundreds of node-hours, or low fidelity, requiring only seconds to run on a common workstation. In current re...

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.

In this paper, we validate the decay heat calculation capability via a two-step method to analyze spent nuclear fuel (SNF) discharged from pressurized water reactors (PWRs). The calculation method is implemented with a lattice code STREAM and a nodal diffusion code RAST-K. One of the features of this method is the direct consideration of three-dimensional (3D) core simulation conditions with the advantage of a short simulation time. Other features include the prediction of the isotope inventory by Lagrange non-linear interpolation and the use of power history correction factors. The validation is performed with 58 decay heat measurements of 48 fuel assemblies (FAs) discharged from five PWRs operated in Sweden and the United States. These realistic benchmarks cover the discharge burnup range up to 51 GWd/MTU, 23.2 years of cooling time, and spanning an initial uranium enrichment range of 2.100–4.005 wt percent. The SNF analysis capability of STREAM is also employed in the code-to-code comparison. Compared to the measurements, the validation results of the FA calculation with RAST-K are within ± 4%, and the pin-wise results are within ± 4.3%. This paper successfully demonstrates that the developed decay heat calculation method can perform SNF back-end cycle analyses.

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.

The article focused on the development of information technology for the optimization of control over complex dynamic systems at the stage of their design that should realize possibilities of modeling of linear and nonlinear dynamic systems, the analysis and synthesis of such systems, their optimization on various quality criteria. The purpose of this article is to develop the structure and elements of information technology to optimize the control of complex dynamic systems, including automated control systems. The general structure and functional model of information technology of the analysis and synthesis of control systems includes modeling of dynamic systems in the form of systems of differential equations and transfer functions, integration of systems of differential equations, calculation of system quality criteria, methods for solving various optimization tasks on solutions of optimization problems of dynamic systems is offered.

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.

In nuclear power plants the design extension conditions are more complex and severe than those postulated as design basis accidents, therefore, they must be taken into account in the safety analyses. In this study, many hypothetical investigated transients are applied on KONVOI pressurized water reactor during a 6 inches (182 cm2) cold leg small break loss-of-coolant-accident to revise the effects of all safety systems ways through their availability/non-availability on the thermal hydraulic behaviour of the reactor. The investigated transients are represented through three cases of small break loss-of-coolant-accident as, Case-1, without scram and all of the safety systems are a failure, Case-2, the normal scram actuation with a failure of all safety systems (non-availability), and finally Case-3, with normal actuation scram sequence and normal sequential actuation of all safety systems (availability). These three investigated transient cases are simulated by creating a model using analysis of thermal-hydraulics of leaks and transient code. In all transient cases, all types of reactivity feed- backs, boron, moderator density, moderator temperature, and fuel temperature are considered. The steady-state results are nearly in agreement with the plant parameters available in previous literature. The results show the importance of the reactivity feedback effects in loss-of-coolant-accident on the fallouts power as they are considered the key parameters for controlling the clad and fuel temperatures to maintain them below their melting point. Moreover, the calculated results in all cases show that the thermal-hydraulic parameters are in acceptable ranges and encounter the safety criterion during loss-of-coolant-accident design extension conditions accidents processes. Furthermore, the results show that the core uncover and fuel heat up do not occur in KONVOI pressurized water reactor the design extension conditions simulations, as all safety systems provide adequate core cooling by sufficient water inventory into the core to cover it.

Many experiments have been conducted on accidents and transients of pressurized water reactor (PWR) employing the rig of safety assessment/large-scale test facility (ROSA/LSTF). Recent research activities concerned with the OECD/NEA international joint projects included experimental investigation via the ROSA and ROSA-2 Projects, and counterpart testing with thermal-hydraulic integral test facilities under collaboration of the PKL-2, PKL-3, ATLAS, and ATLAS-2 Projects. Major results of the related integral effect tests (IETs) with the LSTF were reviewed to experimentally identify thermal-hydraulic phenomena involved, regarding the PWR accident sequences in accordance with the new regulatory requirements for the Japanese light-water nuclear power plants. Future separate effect test using the LSTF is planned to simulate loss of emergency core cooling system (ECCS) recirculation functions in a large-break loss-of-coolant accident (LOCA). Key results of the recent IETs utilizing the LSTF and future plans were presented relevant to multiple steam generator tube rupture accident with recovery operation, small-break LOCA with accident management measure on core exit temperature reliability, and small-break LOCA with thermal stratification under cold water injection from ECCS into cold legs. Also, main outcomes of the LSTF IETs were indicated for wide spectrum LOCA with core uncovery and anticipated transient without scram following small-break LOCA under totally failed high-pressure injection system.

In case of a postulated loss-of-coolant accident (LOCA) in a light water reactor, core degradation might occur if emergency measures fail. During this process a large amount of hydrogen might be generated by the oxidation of zirconium in the fuel rod cladding with steam. If the hydrogen leaks into the containment of a pressurized water reactor or the reactor building of a boiling water reactor and if the design concept of passive autocatalytic recombining is not adequate or fails due to adverse atmospheric conditions, it might form a combustible mixture with air. In case of ignition, pressure peaks can occur that are relevant for the integrity of the containment or the reactor building and safety systems. In the past 20 years 45 hydrogen deflagration tests (HD tests) have been performed in the THAI/THAI+ test facility with a free volume of 60–80 m3 which is operated by Becker Technologies GmbH in Germany. The lumped parameter code COCOSYS V3.0 (Containment Code System) which is part of the software package AC2 2019 developed by Gesellschaft fur Anlagen-und Reaktorsicherheit (GRS) gGmbH offers the models FRONT and COMB for the simulation of hydrogen deflagrations in the containment or reactor building of a nuclear power plant (NPP). COCOSYS simulations of select THAI+ HD tests are shown, and model development potential is deduced.

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.

In nuclear power plants, steam generators are important parts. This paper introduces a U-tube steam generator (UTSG) model implemented completely using MATLAB environment. The UTSG consists of four regions: upward and downward primary regions, upward and downward metal tube regions, and secondary regions, which contain heat transfer region, steam separation region, and subcooled water region. Governing equations are derived by applying energy and mass conservation equations in all regions. Accurate functions that describe the relationships between thermodynamic properties of the saturated steam are introduced instead of interpolation method that is widely used. Steady state and one transient case are presented as well.

Qinshan nuclear power plant is the first large Chinese-designed nuclear power station based on pressurized water reactor, and the second generation main stream active safety injection system is adopted for Qinshan nuclear power plant. In this paper, a novel passive safety injection system (PSIS) has been proposed for ocean-based Qinshan Phase One nuclear power plant to replace the original active one. The PSIS contains high-pressure, medium-pressure, and lower-pressure safety injection systems, and a two-stage automatic depressurization system. To evaluate the system performance, small-break LOCA has been investigated using RELAP5. Various break sizes and locations including 2-inch, 10-inch cold leg break, and double-ended direct vessel injection line break were analyzed. Key safety parameters such as safe injection mass flow rates, coolant level of the core, system pressure, and fuel cladding temperature were monitored during the accident process. The results illustrate that the performance of the safety injection system can guarantee the effective core cooling and submerged under different LOCA even with only half of the safety injection system, which can fulfill the single failure criteria. The thermal-hydraulic analysis for the Qinshan passive safety injection system is significant to master the related technologies for Chinese engineer and develop the Chinese-designed third-generation nuclear power plants, and the PSIS can guarantee the reactor submerged under LOCA even plus the station block out accident.

Reactors are widely used in the nuclear power plant due to the rapid demand for electricity by reducing the greenhouse effect. However, the effectiveness of the nuclear reactor depends on an adequate safety system. Hence, temperature and heat transfer are two critical parameters for any reactor in operation for which intelligent temperature control with an integrated safety system is essential. Therefore, the present study has emphasized the development of a simulated small-scale water-based reactor with intelligent control and safety system and examined through the analysis of thermal-hydraulic parameters. Radial heat transfer of an electric rod used as fuel in the primary circuit has been analyzed by taking sensor reading in various positions of the core. The developed system is self-controlled with all possible active and passive safety systems. Consecutively, the prototype has also been designed including manual adjustment to ensure a fail-safe environment. The system is capable to operate at temperatures between 80°C and 120°C, although the design can withstand up to 200°C. The data of the experiment are taken under the pressure of 200 kPa at 120°C temperature. Results show that heat output of 2116.09 kJ has been obtained from the system against heat input of 2514.80 kJ, which gives an efficiency around 16% of the developed system.

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.