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The paper describes the characteristics of small modular reactors (SMR), their differences from existing nuclear power plants and the features of their use in modern hybrid electric power systems in combination with renewable energy sources and Nuclear Renewable Hybrid Energy System (NRHES) electricity storage, their advantages and problems of use in the future power system of Ukraine. Prospects for cooperation in this area with companies from the United States of America were considered. Small modular SMR reactors are suitable for electrical systems of various capacities, their modular system allows flexibility and speed of construction, reduces capital investment needs and eases financing requirements. The smaller size and variety of reactors can also mean that they can be built in locations not traditionally suitable for large nuclear power plants and, importantly, near energy-intensive industries or remote communities, i.e. as elements of distributed generation. They can ensure a reliable supply of not only electricity, but also heat. SMRs can also be deployed at decommissioned coal-fired power plants. Taking advantage of existing infrastructure, including switchgear and coal-fired turbines, could reduce SMR construction costs and avoid the need to add new transmission lines from these facilities. Problems with the implementation of such networks are revealed. Attention was drawn to the increased complexity of the system in connection with the use of various sources of generation and processes of distribution and consumption of electricity. The main technical characteristics, features of the reactor design, the station building, safety systems and control systems are given. The study concludes that SMRs have significant advantages for use in modern networks, due to their modular design and modular deployment, to meet a variety of power switching and scaling requirements., В работе приведена характеристика малых модульных реакторов (SMR), их отличий от существующих АЭС и особенностей применения в современных гибридных электроэнергетических системах в комбинации с возобновляемыми источниками энергии и накопителями электроэнергии Nuclear Renewable Hybrid Energy System (NRHES), их преимуществами и проблемами использования в будущем энерго. Украины. Рассмотрены перспективы сотрудничества в этой области с компаниями Соединенных Штатов Америки. Малые модульные реакторы SMR подходят для электрических систем разных мощностей, их модульная система дает гибкость и быстроту в построении, уменьшают потребности в капиталовложениях и облегчают требования к финансированию. Меньший размер и разнообразие реакторов также может означать, что их можно строить в местах, традиционно не пригодных для крупных АЭС, и, что немаловажно, вблизи энергоемких производств или отдаленных населенных пунктов, то есть как элементы распределенной генерации. Они могут обеспечить надежную поставку не только электроэнергии, но и тепла. SMR также могут быть развернуты на угольных ТЭС, выходящих из эксплуатации. Использование преимуществ существующей инфраструктуры, включая электрораспределительные установки и турбины угольных электростанций, могло бы снизить затраты на строительство SMR и избежать необходимости добавления новых линий электропередачи с этих объектов. Раскрыты проблемы с внедрением таких сетей. Обратите внимание на повышенную сложность системы в связи с применением различных источников генерации и процессов распределения и потребления электроэнергии. Приведены основные технические характеристики, особенности конструкции реактора, здания станции, систем безопасности и систем управления. По результатам исследования сделан вывод, что SMR имеют значительные преимущества для использования в современных сетях благодаря своей модульной конструкции и модульному развертыванию, чтобы удовлетворить разнообразные требования к изменению и масштабированию мощности., В роботі приведена характеристика малих модульних реакторів (SMR), їх відмінностей від існуючих АЕС і особливостей застосування в сучасних гібридних електроенергетичних системах в комбінації з відновлюваними джерелами енергії та накопичувачами електроенергії Nuclear Renewable Hybrid Energy System (NRHES), їх переваги та проблеми використання в майбутній енергосистемі України. Розглянуті перспективи співробітництва в цій сфері з компаніями Сполучених Штатів Америки. Малі модульні реактори SMR підходять для електричних систем різних потужностей, їх модульна система дає гнучкість і швидкість у побудові, зменшують потреби в капіталовкладеннях і полегшують вимоги до фінансування. Менший розмір і різноманітність реакторів також може означати, що їх можна будувати в місцях, традиційно не придатних для великих АЕС, і, що важливо, поблизу енергоємних виробництв або віддалених населених пунктів, тобто як елементи розподіленої генерації. Вони можуть забезпечити надійне постачання не тільки електроенергії, а й і тепла. SMR також можуть бути розгорнуті на вугільних ТЕС, що виходять з експлуатації. Використання переваг існуючої інфраструктури, включаючи електророзподільні установки та турбіни вугільних електростанцій, могло б зменшити витрати на будівництво SMR та уникнути необхідності додавання нових ліній електропередачі з цих об’єктів. Розкрито проблеми з впровадженням таких мереж. Звернуто увагу на підвищену складність системи в зв’язку з застосуванням різних джерел генерації та процесів розподілення та споживання електроенергії. Наведено основні технічні характеристики, особливостей конструкції реактора, будівлі станції, систем безпеки та систем керування. За результатами дослідження зроблено висновок, що SMR мають значні переваги для використання в сучасних мережах, завдяки своїй модульній конструкції та модульному розгортанню, щоб задовольнити різноманітні вимоги до зміни та масштабування потужності.

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.

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 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.

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.

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.

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.

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.

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.

Due to excess reactivity, fresh nuclear fuel often contains burnable poisons. This research looks at six different burnable poisons and their impacts on reactivity, material attractiveness, and waste management. An MCNP simulation of a PWR fuel pin was performed with a fuel burnup of 60 GWd/MTHM to determine when each burnable poison fuel type would decrease below a k∞ of 1. For determining the plutonium material attractiveness in each burnable poison fuel type, the plutonium isotopic content of the used fuel was evaluated using Bathke's Figure of Merit formula. For the waste management analysis, the thermal output of each burnable poison fuel type was determined through ORIGEN decay simulations at 100 and 300 years after being discharged from the core. The performance of all six burnable poisons varied over the three criteria considered and no single burnable poison performed best in all three considerations.

The phenomenon identification and ranking table (PIRT) is an important basis in the nuclear power plant (NPP) thermal-hydraulic analysis. This study focuses on the importance ranking of the input parameters when lacking the PIRT, and the target scenario is the small break loss of coolant accident (SBLOCA) in a pressurized water reactor (PWR) CPR1000. A total of 54 input parameters which might have influence on the figure of merit (FOM) were identified, and the sensitivity measure of each input on the FOM was calculated through an optimized moment-independent global sensitivity analysis method. The importance ranking orders of the parameters were transformed into the Savage scores, and the parameters were categorized based on the Savage scores. A parameter importance ranking table for the SBLOCA scenario of the CPR1000 reactor was obtained, and the influences of some important parameters at different break sizes and different accident stages were analyzed.

Based on the current high-level radioactive waste management basic plan and the analysis results of spent nuclear fuel characteristics, such as dimensions and decay heat, an improved geological disposal concept for spent nuclear fuel from domestic nuclear power plants was proposed in this study. To this end, disposal container concepts for spent nuclear fuel from two types of reactors, pressurized water reactor (PWR) and Canada deuterium uranium (CANDU), considering the dimensions and interim storage method, were derived. In addition, considering the cooling time of the spent nuclear fuel at the time of disposal, according to the current basic plan-based scenarios, the amount of decay heat capacity for a disposal container was determined. Furthermore, improved disposal concepts for each disposal container were proposed, and analyses were conducted to determine whether the design requirements for the temperature limit were satisfied. Then, the disposal efficiencies of these disposal concepts were compared with those of the existing disposal concepts. The results indicated that the disposal area was reduced by approximately 20%, and the disposal density was increased by more than 20%.

The Korea Atomic Energy Research Institute (KAERI) has developed geological repository systems for the disposal of high-level wastes and spent nuclear fuels (SNFs) in South Korea. The purpose of the most recently developed system, the improved KAERI Reference Disposal System Plus (KRS+), is to dispose of all SNFs in Korea with improved disposal area efficiency. In this paper, a system-level safety assessment model for the KRS+ is presented with long-term assessment results. A system-level model is used to evaluate the overall performance of the disposal system rather than simulating a single component. Because a repository site in Korea has yet to be selected, a conceptual model is used to describe the proposed disposal system. Some uncertain parameters are incorporated into the model for the future site selection process. These parameters include options for a fractured pathway in a geosphere, parameters for radionuclide migration, and repository design dimensions. Two types of SNF, PULS7 from a pressurized water reactor and Canada Deuterium Uranium from a heavy water reactor, were selected as a reference inventory considering the future cumulative stock of SNFs in Korea. The highest peak radiological dose to a representative public was estimated to be 8.19×10-4 mSv‧yr-1, primarily from 129I. The proposed KRS+ design is expected to have a high safety margin that is on the order of two times lower than the dose limit criterion of 0.1 mSv‧yr-1.

This paper proposes the application of a pattern recognition–based technique to enhance the process of control rod position identification in pressurized water reactors (PWRs). The proposed techniq...

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.

We have developed a practical-scale dry disassembling process to dismantle PWR (Pressurized Water Reactor) spent nuclear fuel assembly in the order of several tens of kilograms of heavy metal/batch to supply rod-cuts (cladding tube and UO2 pellets) for mechanical decladding process. Dry head-end disassembling process has advantages over the wet head-end process because of the lower risk of proliferation and treatment of spent fuel with relatively high heat and radioactivity. This study describes the main design considerations for the disassembling process of the spent nuclear fuel assembly during the dry head-end process. The down-ender, dismantling, extraction, and cutting technologies are analyzed and models have been designed for testing. The purpose of dry head-end disassembly process is to test the main device performance and to obtain scale-up data for practical-scale disassembling. With this in mind, design considerations were analyzed based on remoteness, and basic verification tests were performed. However, the authors used simulated fuel, instead of the actual spent fuel, owing to a lack of joint determination. In addition, in the present study, we did not consider the heat generated from minor actinides or the radioactivity of the fission product; these aspects will be considered in a future study. During the basic test performed in this study, a simulated assembly was completely disassembled using new methods, such as dismantling, extraction, and cutting processes. The practical-scale dry disassembling technology can be tested using scale-up data for reuse of the spent fuel.

The water level control system in the pressurizer is carried out for safety purposes when the PWR reactor is operating. Control is done by maintaining the water level in the pressurizer at a certain level. The height of the water level in the pressurizer will change according to the conditions of the nuclear reactor. This research aimed to obtain the best safety control system using the fractional order PID (FOPID) control system and prove that using the control system will provide better results than other existing system applications. In this research, the optimum gain control of FOPID is obtained with K p equals 416.4, K i equals 416.4, K d equals 104.1, constitutes integral equals 0.15, and derivative order equals 1.9, which results in 324 times faster settling time and maximum overshoot 1.5 times lower when compared to conventional PID controllers combined with CMAC. DOI: http://dx.doi.org/10.17977/um024v5i22020p071