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The aim of this paper is to present the isotopic changes in nuclear fuel during the first reactor cycle of the Korean Advanced Power Reactor 1400 (APR1400). The neutron transport and burnup calculations were performed using the Monte Carlo continuous energy burnup code—MCB. The three-dimensional numerical model consisting of the reactor pressure vessel with core internals was developed using available geometrical and material data as well as the reactor's operating conditions. The reactor core was divided into 11 axial and 22 radial burnup zones in order to recreate the spatial distribution of the fuel burnup. The isotopic changes due to the nuclear transmutations and decays were calculated in each burnup zone until the desired average burnup of 17.571 GWd/tHMint was reached. The calculations include changes in the boric acid concentration at defined time steps and the burnout of the gadolinia burnable absorber embedded in the nuclear fuel. This study shows the spatial distribution of minor and major actinides at the end of the reactor cycle as well as the depletion of uranium, the build-up of plutonium, and the formation of neptunium, americium, and curium during the reactor's operation. [ABSTRACT FROM AUTHOR]

A new method for the generation of axial power distributions using artificial neural network (ANN) technique and real-time measured planar radial peaking factor, Fxy (hereinafter 'Live Fxy') method for digital core protection system of pressurized water reactors is presented. ANN with Simulated Annealing (SA) technique to find global optimum solution was used to calculate core average axial power distribution. In Live Fxy method, axial node-wise Fxys were calculated by multiplying pin/box factors to the measured assembly powers inferred from in-core detector signals without calculating detailed 3D power distributions. The hot pin power distribution for DNBR and LPD was then obtained by multiplying the core average axial power distribution and Fxys at axial nodes. The validation of the method was performed for various core conditions (e.g. core power levels, control rod positions, etc.) of Korean OPR1000 power plant. The result showed a decrease in RMS errors by 2.2% ~ 7.0% (3.6% on average), and the minimum thermal margins for DNBR and LPD were increased by 6.4% and 15.6%, respectively. The application of the method would improve the accuracy of axial power distribution and thermal margin, contributing to the operability and safety of digital core protection system. [ABSTRACT FROM AUTHOR]

In order to achieve a full mechanistic understanding of stress corrosion cracking (SCC), the key operating mechanisms need to be identified but also quantified. In this study, we summarize and rationalize key findings from the last 15 years of high-resolution characterization of SCC in our group. A comprehensive characterization of a set of austenitic alloys with different Ni content and constant Cr level, tested under simulated pressurized water reactor (PWR) primary water conditions at various temperatures, has revealed evidence for at least two operating mechanisms: one diffusion-related and the other deformation-related. For their relevance to the nuclear industry, two additional alloys with increased Cr content were also studied (A800 and A690). Key precursors for SCC initiation and propagation are identified and their effect on alloy degradation discussed. A list of key materials' properties that ensure low SCC susceptibility is proposed. [ABSTRACT FROM AUTHOR]

This study introduces a transient analysis module developed for RAST-V and validates it using the Kalinin-3 benchmark problem. For the benchmark analysis, RAST-V standalone and STREAM/RAST-V calculations were performed. STREAM supplies the few-group constants and RAST-V conducts a 3D core simulation utilizing few-group cross-sectional data. To improve accuracy, the main solver was developed based on the advanced semi-analytic nodal method. To evaluate the computational capability of the transient analysis module in RAST-V, Kalinin-3 benchmark is employed. Kalinin-3 represents a coolant transient benchmark that offers experimental data during the deactivation of the Main Circulation Pumps. Consequently, the transient calculations reflected the changes in the reactor flow rate. Benchmark comprising steady-state and transient calculations. During the steady state, the STREAM/RAST-V combination demonstrated a 30 ppm root mean square difference from 0 to 128.50 EFPD. For the transient calculations, STREAM/RAST-V showed power differences within ±7 % over a range of 0–300 s. Axial offset differences were within ±3 %, and the RMS difference in radial power ranged within 2.596 % at both 0 and 300 s. Overall, this study effectively demonstrated the newly developed transient solver in RAST-V and validated it using the Kalinin-3 benchmark problem.

AbstractThis research presents the development of a one-dimensional analytical model to investigate the impact of pressure variations in the primary loop on natural circulation (NC). The model takes into account a sinusoidal input heat distribution and derives equations for the parameters of NC. The model covers a broad spectrum of NC patterns, spanning from fully single-phase to fully two-phase flow. The research demonstrates a smooth and continuous transition between various kinds of NC. Moreover, the research demonstrates that NC is capable of efficiently dissipating the decay heat generated inside the core of a pressurized water reactor, encompassing a range from 100% to 60% of the total inventory present within the primary loop. The findings of this study are compared to prior research outcomes and demonstrate a reasonable level of consistency. Additionally, comparisons are made with uniform input power distribution to demonstrate that there are no significant differences in the NC parameters when using sinusoidal heat input. [ABSTRACT FROM AUTHOR]

The fatigue life of Type 316 stainless steel in the pressurized water reactor (PWR) water environment and its dependency on that in air were investigated. Specimens from two heats of the material were prepared for fatigue tests, and additionally, previous test findings obtained using another heat were applied. The fatigue test results obtained for various strain ranges and strain rates revealed that the fatigue life in the water environment Nf(env) did not depend on that in air Nf(air) when the strain rate was 0.4%/s. The variation in the ratio of Nf(env) to Nf(air) was larger when Nf(air) obtained for each heat of the material was used than that when a general Nf(air) was used. The changes in the penalty factor Fen with applied strain range and strain rate could be predicted by assuming Nf(env) depended not on strain rate but on rise time. Highlights: Fatigue life in PWR water environment did not depend on that in air.Fen should be determined using a common life in air regardless of material heat.Rise time better to represented test speed to predict the environmental effect.Changes in Fen with strain range and rate could be predicted using rise time. [ABSTRACT FROM AUTHOR]

A major subject for evaluating the corrosive conditions in the pressurized water reactor (PWR) primary coolant is to determine the optimal hydrogen concentration for mitigating primary water stress corrosion cracking (PWSCC) without adverse effects on major structural materials. An analytical method combining water radiolysis and electrochemical corrosion potential (ECP) analyses was proposed for evaluating the corrosive conditions in PWR primary coolant. These procedures originally developed for boiling water reactors (BWRs) were successfully extended to PWRs with different water chemistry parameters, including pH, temperature, and radiation qualities, after minor changes in the original procedures and major input parameters, such as the inclusion of the effects of alpha radiolysis and Li+ (Na+) and H+ effects for the anodic polarization curve. This study discusses the results of water radiolysis analysis for PWR primary coolant, and the characteristic behavior of hydrogen peroxide (H2O2) as a function of hydrogen inlet concentrations in PWR primary coolant conditions (higher pH and α-ray irradiation). A possible reaction scheme involving eaq– was proposed for H2O2 suppression under alkaline conditions. The corrosive conditions were discussed in the following publication, using ECP as the major index for PWR corrosive conditions. [ABSTRACT FROM AUTHOR]

The aim of this paper is to present the isotopic changes in nuclear fuel during the first reactor cycle of the Korean Advanced Power Reactor 1400 (APR1400). The neutron transport and burnup calculations were performed using the Monte Carlo continuous energy burnup code—MCB. The three-dimensional numerical model consisting of the reactor pressure vessel with core internals was developed using available geometrical and material data as well as the reactor’s operating conditions. The reactor core was divided into 11 axial and 22 radial burnup zones in order to recreate the spatial distribution of the fuel burnup. The isotopic changes due to the nuclear transmutations and decays were calculated in each burnup zone until the desired average burnup of 17.571 GWd/tHMint was reached. The calculations include changes in the boric acid concentration at defined time steps and the burnout of the gadolinia burnable absorber embedded in the nuclear fuel. This study shows the spatial distribution of minor and major actinides at the end of the reactor cycle as well as the depletion of uranium, the build-up of plutonium, and the formation of neptunium, americium, and curium during the reactor’s operation.

Actinide minor is a reactor waste with high toxicity and a long half-life. Minor actinides can be reduced by reusing them as fuel mixtures in reactors. This research uses PWR reactors with the primary fuel UN-PuN or Uranium Plutonium Nitride with a burning time of 5 years. The fuel consists of enriched Uranium, reactor-grade Plutonium from LWR waste, and minor actinides including Neptunium-237, Americium-241, and Curium-244. The purpose of this study was to find a design that is effective in reducing minor actinide waste. There are six designs or cases used in the addition of minor actinides. Each case has six minor actinide pins in each assembly. The addition of minor actinides is arranged in heterogeneous cores. The analysis was carried out by observing the values of k-eff, excess reactivity, and mass of minor actinides obtained from simulations using OpenMC code 0.13.2 and the ENDF/B-VIII library. The homogeneous core obtained an excess reactivity of 9.7 % with a percentage of plutonium of 8 %. The results of the homogeneous core are used as a reference for preparing a heterogeneous core. The heterogeneous core obtained an excess reactivity of 9.9 % with a percentage of plutonium F1: 5.5 %, F2: 8 %, and F3: 10.5 %. Np-237 can be reduced by 53 kg, and Am-241 can be reduced by 61 kg with minor actinide pins in case 1. Cm-244 can be reduced by 363 kilograms with minor actinide pins in case 6. Excess reactivity in the addition of Np-237 and Am-241 decreased to 5.3 %, while the accumulation of Cm-244 increased to 12.1 %. [ABSTRACT FROM AUTHOR]

In the field of nuclear reactor design, precise calculation of reactor dimensions is crucial for operational efficiency and safety. This study focuses on the critical dimension prediction for a spherical nuclear thermal reactor, employing a computational approach to address this challenge. Utilizing the two-group neutron diffusion equation, the research aims to establish a mathematical model for neutron distribution within the reactor's geometry. A key aspect of this model involves the prediction of neutrons' spatial distribution, essential for understanding the reactor's behavior under operational conditions. The methodology adopted in this investigation involves using a MATLABbased program specifically developed for solving the two-group diffusion equation in spherical reactor geometry. This approach facilitates the determination of exact dimensions and optimal fuel mass for the reactor. The study's findings indicate a critical core radius of 21.7 cm, with a water mass of 40.5 kg and a U235 fuel mass of 1.12 kg. Additionally, the ratio of fast flux to slow flux was approximately 1.5. These results not only align closely with prior research in this domain but also enhance the understanding of spherical reactor design. Crucially, the study's outcomes demonstrate a high degree of compatibility with existing literature, thereby reinforcing the validity of the computational model used. This research contributes significantly to the nuclear engineering field by providing a robust method for determining the critical dimensions and mass of fuel required for the efficient and safe operation of spherical nuclear thermal reactors. Which still regarded nuclear energy as an essential partner in friendly energy production in the world to reduce CO2 emissions. The implications of these findings are substantial, offering a pathway to optimized reactor design and a deeper understanding of neutron behavior in complex geometrical configurations. [ABSTRACT FROM AUTHOR]

The CRUD on the fuel cladding under the pressurized water reactor (PWR) operating condition causes several issues. The CRUD can act as thermal resistance and increases the local cladding temperature which accelerate the corrosion process. The hideout of boron inside the CRUD results in axial offset anomaly and reduces the plant's shutdown margin. Recently, there are efforts to revise the acceptance criteria of emergency core cooling systems (ECCS), and additionally require the modeling of the thermal resistance effect of the CRUD during the performance analysis. There is an urgent need for the evaluation of the effect of the CRUD deposition on the cladding heat transfer under PWR operating conditions, but the experimental database is very limited. The experimental facility called DISNY was designed and constructed to analyze the CRUD-related multi-physical phenomena, and the performance analysis of the constructed DISNY facility was conducted. The thermal-hydraulic and water chemistry conditions to simulate the CRUD growth under PWR operating conditions were established. The design characteristics and feasibility of the DISNY facility were validated by the MARS-KS code analysis and separate performance tests. In the current study, detailed design features, design validation results, and future utilization plans of the proposed DISNY facility are presented.

The load control of a nuclear reactor is important due to the nonlinear nature of its dynamics and the dependence of some parameters on the output power. the Proportional-Integral-Derivative controller (PID) is commonly regarded as an easy choice for reliable control. In this research, the relative neutron density in the point kinetics model of a Pressurized Water Reactor (PWR) is controlled by an optimized PID with the meta-heuristic Differential Evolution (DE) algorithm. The Integral of Time-Absolute Error (ITAE) performance index has been used for optimization with this algorithm. The simulation results show that the optimized control system with the DE algorithm has the appropriate efficiency and accuracy in response to power demand.

This research article explores the potential of optical fibers as sensors, highlighting their ability to measure various parameters such as temperature, pressure, stress, and radiation dose. The study focuses on investigating the material compatibility of optical fibers in challenging sensing environments like Gen II/II+ and advance nuclear reactors, as well as concentrated solar power (CSP) plants. Material compatibility tests were conducted to determine the feasibility of using fluorine and germanium optical fiber sensors in these environments. The study found that raw fibers were corrosion-resistant to lead bismuth eutectic at 600 °C, regardless of the coating. In molten salt environments, raw fibers were incompatible with FLiNaK but showed corrosion resistance to MgCl₂-NaCl-KCl. However, the survivability of raw fiber optics improved with a gold coating in FLiNaK. Raw fiber optics were found to be incompatible in high-temperature steam at 1200 °C and in a pressurized water reactor (PWR) at 300 °C.

This study proposes an effective repair technology using arc additive manufacturing for pressurized water reactors (PWRs) in nuclear power plants (NPP) aimed at avoiding complete replacements and post-weld-heat treatments (PWHTs) of component parts while ensuring safety and reliability. Effective repair technology is defined as economic and process efficiency, because of maintenance costs and radiation exposure, and it is critical in related industries. The technology is designed to relieve the hardness and martensite fraction of the welding heat affected zone (HAZ) of low alloy steels (SA508) in PWRs penetration/nozzles by heat source generated in the WAAM process, thus ensuring structural integrity. In the first layer of wire arc additive manufacturing processes, 89.6% of the martensite phase was formed in the HAZ of SA508, which was significantly reduced to 45.7% due to repetitive thermal behaviors at the third layer. The resulting process dramatically reduced hardness from 450Hv in the initial layer additive manufacturing to 320Hv in 3 layers without additional heat treatments. Moreover, the study quantitatively investigated the martensite starting temperature (Ms) and bainite starting temperature (Bs) and analyzed the microstructure and mechanical behavior of the developed process using thermodynamic calculation (CALPHAD), finite element method (FEM) simulation, and microstructure quantitative analysis by electron backs-catter diffraction (EBSD). The proposed technologies and their quantitative analysis results can be a substantial alternative to the repair technology for penetration/nozzles in nuclear primary water cooling reactor applications, complying with ASME Sec.IX Qw-462.12 and ISO 15614-1.

In thiswork, a 24-month two-batch fuel management strategy for the APR1400 using LEU + has been investigated, where enrichments of 5.9 and 5.2 w/o are utilized in lieu of the conventional 4–5 w/o UO2 fuel. In addition, an Accident Tolerant Fuel (ATF) clad based on the swaging technology is applied to APR1400 fuel assemblies. In this special ATF clad design, both outer and inner SS316 layers protect the conventional zircaloy clad. Erbia (Er2O3) is introduced as a burnable absorber with two-fold goals to lower the critical boron concentration in the long-cycle LEU + loaded core as well as to handle the LEU + fuel in the existing front-end fuel facilities without renewing the license. Two types of fuel assemblies with different loading of gadolinia (Gd2O3) are considered to control both the reactivity and the core radial power distribution. The erbia burnable absorber is uniformly admixed with UO2 in all fuel pins except for the gadolinia-bearing ones. In this study, two core designs were devised with different erbia loading, and core performance and safety parameters were evaluated for each case in comparison with a core design without any burnable absorbers. The core analysis was done using the two-step method. First, cross-sections are generated by the SERPENT 2 Monte Carlo code, and the 3-D neutronic analysis is performed with an in-house multi-physics nodal code KANT.

This research article explores the potential of optical fibers as sensors, highlighting their ability to measure various parameters such as temperature, pressure, stress, and radiation dose. The study focuses on investigating the material compatibility of optical fibers in challenging sensing environments like Gen II/II+ and advance nuclear reactors, as well as concentrated solar power (CSP) plants. Material compatibility tests were conducted to determine the feasibility of using fluorine and germanium optical fiber sensors in these environments. The study found that raw fibers were corrosion-resistant to lead bismuth eutectic at 600 °C, regardless of the coating. In molten salt environments, raw fibers were incompatible with FLiNaK but showed corrosion resistance to MgCl₂-NaCl-KCl. However, the survivability of raw fiber optics improved with a gold coating in FLiNaK. Raw fiber optics were found to be incompatible in high-temperature steam at 1200 °C and in a pressurized water reactor (PWR) at 300 °C. [ABSTRACT FROM AUTHOR]

The investigation of thermal-hydraulic parameters during steam-line break (SLB) accidents is performed by applying the personal computer transient analyzer (PCTRAN) simulator model of the VVER-1200 nuclear power plant (NPP). Five cases, namely, 0.005 m2 (Case-1), 0.01 m2 break (Case-2), 0.02 m2 break (Case-3), 0.04 m2 (Case-4), and 0.08 m2 (case-5) of SLB accident inside containment with the concurrent loss of AC power have been simulated. There was no variation in the timing of the trip of the reactor coolant pumps, the main feedwater pumps, or the turbine in any of the five SLB accidents. However, the reactor scram's onset time varies slightly between the five scenarios. Pressure and temperature in the reactor coolant system (RCS) quickly reached a peak following the start of the SLB accident, fell shortly after the reactor scram, and eventually stabilized in all cases. In comparison to the larger breaks in the SLB accident, the smaller breaks result in a higher RCS temperature and pressure. After the SLB accident, the pressurizer's liquid level rises and then quickly drops in all cases. The break mass flow rate from the steam line rapidly increases until the occurrence of the reactor scram and then decreases to a stabilized value. Steam generator A has a faster rate of heat removal rate than steam generator B, and its pressure and liquid level decrease more quickly than those of steam generator B. The thermal power of the reactor, peak cladding temperature, and fuel temperature showed a rapid drop after the initiation of the SLB accident. There was no increase in these parameters from the initial state of the simulation. The radiation in the air of the reactor building and steam line was very low during the simulation period. Therefore, there was no violation of the safety aspects of the SLB accident of the PCTRAN simulation of the VVER-1200 NPP model.

Current plans and efforts of reactor operators and vendors to include extended-enrichment (EE) fuel and accident-tolerant fuel (ATF) in current reactor fleets motivate the study of these changes in reactor physics analysis. This work uses the U.S. Nuclear Regulatory Commission's core simulator PARCS to do the core calculation and the SCALE Polaris lattice physics code to generate the homogenized, few-group constants. In this work, both pressurized water reactor and boiling water reactor (BWR) colorset models are used to verify the proposed approach. The accuracy presented in the colorset models verified the capability of the PARCS/Polaris procedures for the transition core analysis in light water reactors. For the whole-core calculation, the ATF and EE-ATF transition core models were incorporated, in addition to the nominal core model. The BWR model was chosen to represent the entire core calculation due to its challenging design. The core parameters studied are the core power distribution, power peaking factor, Doppler temperature coefficients, and control rod worth at cold zero power and hot full power. When the core parameters of the transition cores are compared with those of the nominal core in PARCS, the results suggest that there is no drastic change in the core parameters for the implementation of ATF and EE fuels. [ABSTRACT FROM AUTHOR]

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58Co and 60Co are the primary radionuclides and contributors to collective dose of PWRs, which may exist in ionic, particle and colloidal state in the primary loop. The pore diameter of the most advanced filter used in chemical and volume control system (CVCS) in domestic PWRs is 0.1 μm, the removal efficiency of Co colloids with the diameter of less than 0.1 μm is very low. Activated corrosion products such as Co colloids may deposit on the surface of the primary loop and auxillary systems during shutdown, contributing to collective dose. The chemical conditions of primary loop during the oxidation operation process greatly affect the removal efficiency of radionuclides such as Co colloids. To achieve the best dissolution and removal effect of Co colloids, the dissolution behavior of Co colloids need to be studied. In this paper, simulated Co colloids were first synthesized in the laboratory. The average particle size of simulated Co colloids was characterized by TEM to be 40-50 nm. The composition of simulated Co colloids was characterized by TEM, SEM and EDS as CoO and Co. Then the effects of temperature (60-80 ℃), Li concentration (pH, 0.1-3 mg/kg), H2O2 concentration (0.20 mg/kg) and reaction time (0-6 h) on the dissolution behavior of Co colloids were investigated on simulated chemical condition of the oxidation operation process. The results show that the dissolution of Co colloids is quite fast, reaching reaction equilibrium in 0.5-1 h. The dissolution behavior of Co colloids is affected by temperature, Li concentration and H2O2 concentration. The dissolution of Co colloids is promoted by addition of H2O2 due to the weak acidity and strong oxidizing power of H2O2. However, too much H2O2 is unfavourable for the dissolution of Co colloids, which may be due to the possible formation of passive film on Co colloids. Low Li concentration or low pH is favourable for the dissolution of Co colloids. The oxidizing power of H2O2 is stronger at higher temperature promoting the dissolution of Co colloids. Yet high temperature is not favorable for the dissolution of Co colloids, as the dissolution of Co colloids is an exothermal process. Therefore a moderate temperature is the best choice for the dissolution of Co colloids. To summarize, the best chemical condition for the dissolution and removal of Co colloids during the oxidation operation process is at 70 ℃ with as low Li concentration as possible and about 10 mg/kg H2O2. The results in this paper provide important references for the improvement of the oxidation operation process of PWRs.

Abstract Striving to peak carbon emissions and achieve carbon neutrality (known as the "Dual-Carbon" goal) is an inevitable requirement for elevating the environmental resource constraints and realizing harmonious coexistence between the mankind and the earth. In the energy system, nuclear energy offers various advantages, such as high energy density, low carbon emission, strong environmental adaptability and large potential for energy co-generation and co-supply. It is one of the supporting energy sources for the transformation and upgradation of the energy system to a clean, efficient and low-carbon way. In this paper, the opportunities and challenges for innovation-driven nuclear energy development in the fields of electricity generation, hydrogen production, heat supply and seawater desalination under the goal of "Dual-Carbon" are discussed and analyzed. Besides, the relevant research on improving the safety and economy of the pressurized water reactor (PWR) and sodium-cooled fast reactor (SFR) conducted by the Nuclear THermal–hydraulic research Lab (NuTHeL) of Xi’an Jiaotong University is briefly introduced.

This paper presents the development of a nodal diffusion code, RAST-V, and its verification and validation for VVER (vodo–vodyanoi energetichesky reactor) analysis. A VVER analytic solver has been implemented in an in-house nodal diffusion code, RAST-K. The new RAST-K version, RAST-V, uses the triangle-based polynomial expansion nodal method. The RAST-K code provides stand-alone and two-step computation modes for steady-state and transient calculations. An in-house lattice code (STREAM) with updated features for VVER analysis is also utilized in the two-step method for cross-section generation. To assess the calculation capability of the formulated analysis module, various verification and validation studies have been performed with Rostov-II, and X2 multicycles, Novovoronezh-4, and the Atomic Energy Research benchmarks. In comparing the multicycle operation, rod worth, and integrated temperature coefficients, RAST-V is found to agree with measurements with high accuracy which RMS differences of each cycle are within ±47 ppm in multicycle operations, and ±81 pcm of the rod worth of the X2 reactor. Transient calculations were also performed considering two different rod ejection scenarios. The accuracy of RAST-V was observed to be comparable to that of conventional nodal diffusion codes (DYN3D, BIPR8, and PARCS).

Computational fluid dynamics (CFD) is adopted to calculate and analyze the performance and separation mechanism of a steam-water separation equipment for a pressurized water reactor (PWR) steam generator. The steam-water two-phase flow is simulated by the Euler two-fluid model, and several representative water droplet diameters are selected among 50 to 400 μm. The influence mechanism of water droplet diameter on the performance is investigated by analyzing the flow parameters such as phase volume fraction, velocity, and turbulent kinetic energy. The results show that the integrated calculation and analysis of the separation equipment can more truly reflect the flow state between the separators, improving the reliability of the performance prediction and mechanism analysis. With the increase in water droplet diameter, the separation efficiency of the separation equipment and each separator gradually increases, the outlet wetness gradually decreases, and the pressure loss first decreases and then stabilizes. When 400 μm is taken as the characteristic value of the actual droplet diameter distribution at the inlet of the separation equipment, the performance prediction is more accurate, the pressure loss of each separator is relatively close, and the pressure loss of the primary separator is less affected by the droplet diameter and smaller than that of the swirl-vane primary separator, which is conducive to achieving a higher circulation ratio (CR). Re-entrainment occurs at the perforations of the primary separator and the outlet of the secondary separator, and the corresponding structure is suggested to be optimized to further improve separation efficiency of the separation equipment. [ABSTRACT FROM AUTHOR]

Reduction in ambient pressure within the containment of a light water reactor in the event of a hypothetical Loss of Coolant Accident (LOCA) is crucial for containment integrity. This article introduces the distinct concept of a Passive Containment Cooling System (PCCS). In the proposed PCCS concept, the containment is divided into two compartments with appropriate volumes. An external heat exchanger is connected between the two compartments of the containment for heat transfer and pressure reduction in the case of LOCA. This concept has unique characteristics and provides many advantages over previous designs. A RELAP5 model was developed to perform thermal-hydraulic analysis of the proposed PCCS concept. The performance of the PCCS after its deployment on a small reactor (998.6 MWth) has been assessed under various conditions following LOCA. The results indicate that partitioning of the containment creates a differential pressure that acts as a driving force for flow through the PCCS heat exchanger. Consequently, a decrease in long-term containment pressure is observed. In the case of equal volume fractions in the two containment compartments, the PCCS reduces long-term pressure in the containment by 16.5%. Sensitivity analysis was conducted in which the flow area of the PCCS inlet line, the heat transfer area, and the size of the heat exchanger tank were varied. The PCCS heat transfer capacity increases as these parameters are increased. However, the former is less sensitive than the latter two. The performance of the PCCS heat exchanger under steady-state conditions was validated using RELAP5 simulations, empirical correlations, and analytical models. The evaluation of heat exchanger parameters by various methods (presented in Appendix 1) was found in close agreement. [ABSTRACT FROM AUTHOR]

Gas–liquid two-phase countercurrent flow limitation (CCFL) phenomena widely exist in nuclear power plants. In particular, the gas–liquid countercurrent flow limitation phenomena in a pressurized water reactor (PWR) during a loss-of-coolant accident (LOCA) or a small-break loss-of-coolant accident (SBLOCA) play an important role in nuclear reactor safety research. Over several decades, a series of experimental investigations and numerical studies have been carried out to study the CCFL phenomena in a PWR. For the experimental investigations, numerous experiments have been conducted, and different CCFL mechanisms and CCFL characteristics have been obtained in various test facilities simulating different scenarios in a PWR. The CCFL phenomena are affected by many factors, such as geometrical characteristics, liquid flow rates, and fluid properties. For the numerical studies, more and more numerical models were presented and applied to the calculations of two-phase countercurrent flow over the past several decades. It is considered that the computational fluid dynamics (CFD) tools can simulate most of the two-phase flow configurations encountered in nuclear power plants. In this paper, the experimental investigations and the numerical studies on two-phase countercurrent flow limitation in a PWR are comprehensively reviewed. This review provides a further understanding of CCFL in a PWR and gives directions regarding future studies. It is found that relatively fewer investigations using steam–water under high system pressures are performed due to the limitation of the test facilities and test conditions. There are a number of numerical studies on countercurrent two-phase flow in a PWR hot leg geometry, but the simulations in other flow channels were relatively rare. In addition, almost all of the numerical simulations do not include heat and mass transfer. Thus, it is necessary to investigate the effects of heat and mass transfer experimentally and numerically. Furthermore, it is of significance to perform numerical simulations for countercurrent two-phase flow with a fine computational grid and suitable models to predict the formation of small waves and the details in two-phase flow. [ABSTRACT FROM AUTHOR]

In recent years, interest in approximate computing has been increasing significantly in many disciplines in the context of saving energy and computation cost by trading off on the quality of numerical simulation. The hardware acceleration based on the low-precision floating-point arithmetic is anticipated by the upcoming generation of microprocessors and code compilers and has already proven to be beneficial for weather and climate modelling and neural network training. The present work illustrates the application of low-precision arithmetic for the nuclear reactor core uncertainty analysis. We studied the performance of an elementary transient reactor core model for the arbitrary precision of the floating-point multiplication in a direct linear system solver. Using this model, we calculated the reactor core transients initiated by the control rod ejection taking into account the uncertainty of the model input parameters. Then, we evaluated the round-off errors of the model outputs for different precision levels. The comparison of the round-off errors and the model uncertainty showed the model could be run using a 15-bit floating-point number precision with an acceptable degradation of the result's accuracy. This precision corresponds to a gain of about 6× in the bit complexity of the linear system solution algorithm, which can be actualized in terms of reduced energy costs on low-precision hardware. [ABSTRACT FROM AUTHOR]

Arabidopsis HOS15/PWR/HDA9 repressor complex, which is similar to the TBL1/NcoR1/HDAC complex in animals, plays a well-known role in epigenetic regulation. PWR and HDA9 have been reported to interact with each other and modulate the flowering time by repressing AGL19 expression, whereas HOS15 and HDA9, together with the photoperiodic evening complex, regulate flowering time through repression of GI transcription. However, the role of the HOS15/PWR/HDA9 core repressor complex as a functional unit in the regulation of flowering time is yet to be explored. In this study, we reported that the loss-of-function hos15-2/pwr/hda9 triple mutant accumulates higher transcript levels of AGL19 and exhibits an early flowering phenotype similar to those of hos15, pwr, and hda9 single mutants. Interestingly, the accumulation of HOS15 in the nucleus was drastically reduced in pwr and hda9 mutants. As a result, HOS15 could not perform its role in histone deacetylation or interaction with H3 in the nucleus. Furthermore, HOS15 is also associated with the same region of the AGL19 promoter known for PWR-HDA9 binding. The acetylation level of the AGL19 promoter was increased in the hos15-2 mutant, similar to the pwr and hda9 mutants. Therefore, our findings reveal that the HOS15/PWR/ HDA9 repressor complex deacetylates the promoter region of AGL19, thereby negatively regulating AGL19 transcription, which leads to early flowering in Arabidopsis. [ABSTRACT FROM AUTHOR]

There has been a lot of interest in generating electricity using nuclear energy recently. This interest is due to the features of such a source of energy. The main part of the nuclear energy system is the reactor core, especially the most widely used Pressurized Water Reactor (PWR). This reactor is the hottest part of the nuclear system; security risks and economic possibilities must be considered. Controlling this reactor can increase the security and efficiency of nuclear power systems. This study presents a dynamic model of the (PWR), including the reactor's core, the plenums of the upper and lower, and the connecting piping between the reactor core and steam generator. In addition, an adaptive neuro-fuzzy (ANFIS) self-tuning PID Controller for the nuclear core reactor is presented. This adaptive controller is used to enhance the performance characteristics of PWR by supporting the profile of the reactor power, the coolant fuel, and hot leg temperatures. The suggested proposed ANFIS self-tuning controller is estimated through a comparison with the conventional PID, neural network, and fuzzy selftuning controllers. The results showed that the proposed controller is best over traditional PID, neural network, and fuzzy self-tuning controllers. All simulations are throughout by using MATLAB/SIMULINK. [ABSTRACT FROM AUTHOR]