Advanced Pressurized Water Reactor Research Articles

Experimental investigation and prediction of CHF in wire-wrapped rod bundles

The wire-wrapped tight lattice fuel assembly is one of the options for small modular reactors. To clarify the characteristics of critical heat transfer for departure from nucleate boiling, experiments were conducted to investigate the critical heat flux in a wire-wrapped 19-rod bundle. The critical heat flux coincident with associated thermal–hydraulic parameters was obtained under typical pressurized water reactor conditions. Validation of data accuracy was performed to provide confidence in the experiment results. It was found that the critical heat flux in the wire-wrapped rod bundle is sensitive to the change in steam quality at some steam quality regions. Besides, the critical heat flux in wire-wrapped rod bundles has a similar reflection under the effects of local pressure, local mass flux, and steam quality. However, the current correlations tend to overestimate the critical heat flux for a wire-wrapped rod bundle. Hence, further development of KfK-3 correlations was proposed for better prediction of the critical heat flux in wire-wrapped bundles. This work is helpful for rod bundle designs of advanced pressurized water reactors.

Techno-economic feasibility study of coupling low-temperature evaporation desalination plant with advanced pressurized water reactor

The increasing demand for freshwater necessitates sustainable desalination solutions, and nuclear power plants offer a promising avenue by utilizing their low-grade waste heat. This study assesses a techno-economic feasibility of coupling a 5 MWth low-temperature evaporation plant with a UAE-based Advanced Pressurized Water Reactor (APR1400). The system addresses freshwater demand, aligning with sustainability goals through low-grade heat utilization. The investigation explores three extraction points for low-grade heat steam, with temperatures ranging from 80 °C to 130 °C. Thermodynamic evaluations using DE-TOP illustrate power requirements and losses, considering variations in maximum brine temperature from 50 °C to 65 °C. Economic analysis using DEEP estimates water production costs. Findings reveal negligible variances in power plant parameters and a minimal reduction in overall efficiency (<0.5 %). The power loss ratio increases proportionally (10 % to 18.6 %) with higher-temperature heat extraction, while the total power requirements for the desalination plant rises by around 30 %, with a negligible power output reduction ratios (0.03 % to 0.07 %). A consistent linear correlation between water production rate and maximum brine temperature doubles water production (∼900 to 1800 m3/day). Applying multiple extraction points across low-grade heat sources demonstrates scalability, reaching three times that of single-point extraction, with marginal increases in power requirements and losses, while maintaining the power reduction ratio below 0.15 %. Economic feasibility indicates competitive water production costs, ranging from 1.53 to 0.87 $/m3 for desalination capacities between 900 and 5400 m3/day. This study provides valuable insights into sustainable water production at the nexus of nuclear energy and desalination, with implications for necessary policy intervention.

Development of scaling approach based on experimental and CFD data for thermal stratification and mixing induced by steam injection through spargers

Advanced Pressurized Water Reactors (APWRs) and Boiling Water Reactors (BWRs) employ a suppression pool as a heat sink to prevent containment overpressure. Steam can be discharged into the pool through multi-hole spargers or blowdown pipes in both normal and accident conditions. Direct Contact Condensation (DCC) creates sources of momentum and heat. The competition between these two sources determines the development of thermal stratification or mixing of the pool. Thermal stratification is of safety concern as it reduces the cooling capability compared to a completely mixed pool condition. In this work we develop a scaling approach to prediction of the thermal stratification in a water pool induced by steam injection through spargers. Experimental data obtained from large-scale pool tests conducted in the PPOOLEX and PANDA facilities, as well as simulation results obtained using validated codes are used to develop the scaling. Two injection orientations, namely radial injection through multi-hole Sparger Head (SH) and vertical injection through Load Reduction Ring (LRR), are considered. We show that the erosion rate of the cold layer can be estimated using the Richardson number. In this work, scaling laws are proposed to estimate both the (i) transient erosion velocity and (ii) the stable position of the thermocline. These scaling laws are then implemented into a 1D model to simulate the thermal behavior of the pool during steam injection through the sparger.

Thermal-Hydraulic comparative analysis of G-III + BWR and PWR passive safety features under Loss of coolant accident

As one of symbolic design features of Generation-III and III+ (G-III and G-III + ) nuclear reactors, the passive safety system has been implemented on majority of G-III and III + PWRs and BWRs. The passive safety systems in those reactors are designed differently yet adhere to identical fundamental principles. It would be interesting to examine the similarities and differences between the passive safety systems of G-III + PWR and BWR. System code simulations for Small Break and Large Break Loss of Coolant Accident for the Economic Simplified Boiling Water Reactor (ESBWR) and China’s advanced PWR had been conducted, respectively. The results showed that the passive safety systems of ESBWR and China’s Advanced PWR were comparable in design, and when dealing with the similar type of SBLOCA and LBLOCA, the passive safety components with the same functionality could all contribute to the core's cooling with stability and reliability, and the injection flow rate values were nearly equivalent. Separating the injection lines of ESBWR's safety systems guarantees that the action sequence could well remain stable. This design could prevent the injection flow rate and action sequence from being impacted by the interaction of safety injection systems. This work could serve as a reference for the design of subsequent new reactors, thereby contributing to the design of reactors with enhanced economics and safety.

Parameter study for critical heat flux under external reactor vessel cooling condition based on a mechanism model

External reactor vessel cooling has been adopted an important method to implement In Vessel Retention severe accident mitigation strategy for many advanced passive pressurized water reactors. Critical heat flux is one of the important criteria to evaluate the success or failure of this strategy. The uncertainty of engineering factors in practical design and operation conditions may make influence on critical heat flux more or less. In this paper, a parameter study for critical heat flux was carried out based on a new critical heat flux mechanism model. This mechanism model is developed based on the mechanism of liquid film dryout, and the predicted results of this model are in good agreement with the experimental results. Five important parameters, namely, coolant subcooling degree, geometry of lower head, system pressure, flow rate and heating power shape are analyzed based on this mechanism model. The results of parameter study show that the decrease of inlet coolant temperature and lower head radius are beneficial to the increase of critical heat flux. And the critical heat flux increases with the increase of system pressure. While the change of flow rate and heating power shape have little effect on critical heat flux. Under the condition of same circumference, the critical heat flux of spherical lower head is higher than that of elliptical lower head in the middle and low inclination angle region, while the trend is opposite in high inclination angle region, but not significant.

Analysis of Optimum Accident Tolerant Fuel and Cladding Behavior in Advanced Pressurized Water Reactor

Analysis of Optimum Accident Tolerant Fuel and Cladding Behavior in Advanced Pressurized Water Reactor

Economic evaluation of hydrogen production in second and third units of Bushehr nuclear power plant regarding future need of nuclear fission technology

Today, the most important challenge for hydrogen production is the use of suitable energy source for sustainable production. The dual use of nuclear power plants allows them to expand into other markets by eliminating greenhouse gas emissions from processes such as hydrogen production. The significance of this matter especially in countries with ongoing new nuclear power plants is high and shows the future need for nuclear fission technology. In this paper, an economic assessment of hydrogen production in the second and third units of the Bushehr nuclear power plant (BNPP) with the Hydrogen Economy Evaluation Program (HEEP) is carried out and a comparison of CO2 emission in nuclear and non-nuclear methods of hydrogen production is presented. Most nuclear reactors, including Advanced Pressurized WaterReactor (APWR), High Temperature Gas Reactor (HTGR), and Water-Water EnergeticReactor (WWER), can be used as a source of thermal and electrical energy for the process of hydrogen production. In this research, hydrogen production processes based on the Conventional Electrolysis (CE), High Temperature Steam Electrolysis (HTSE), and Sulfur-Iodine (SI) Process using different storage and transportation options are compared to choose the cost-effective method for coupling with the WWER1000 reactor. Hydrogen production, thermal energy, and electricity costs have been calculated and explained in detail. The results showed that the cost-effective option for hydrogen production is using the High Temperature Steam Electrolysis (HTSE) method, Compressed Gas (CG) storage, and pipeline transportation, by 2.88 $/kg. Also, applying this method to produce hydrogen, the nuclear power plant thermal efficiency, and the hydrogen plant efficiency are kept in the best possible conditions by 51.98% and 58.21%, respectively.

Cladding of nuclear fuel with iron based alloy: Penalty and solution

Replacing zirconium-based alloys fuel cladding by that of iron in nuclear fuel of Light Water Reactors (LWRs) needs increasing the fuel enrichment to achieve the same core cycle length. The penalty of using iron based alloys fuel cladding in LWRs can be overcome by direct recycling of nuclear spent fuel in CANDU reactors. The problems associated with using zirconium-based alloys fuel cladding in LWRs is not significant in the case of CANDU reactors. This is because CANDU reactors discharge the spent fuel with low burnup which is in the withstand tolerance of zirconium-based alloys and the moderator acts as an effective heat sink in case of severe accident.The case studied in this research is direct recycling of the spent fuel discharged from the European Advanced Pressurized Water Reactor (EU-APWR) in CANDU-6 reactor. The calculations using MCNPX computational code showed that replacing fuel clad from zircaloy-4 to stainless steel-348 requires increasing the fuel enrichment from 4.95 %w U-235 to 6.8 %w U-235 with an increase of 37%. After direct spent fuel recycling in CANDU-6, the total burnup of the fuels cladded with zircaloy-4 and with stainless steel-348 are about 70 and 96 GWd/t, respectively with an increase of about 37% in the case of the fuel cladded with stainless steel which compensates the increase of the enrichment.

Experimental analysis of boron dilution transient flow mixing characteristics based on planar laser induced fluorescence measurement

In Reactor Pressure Vessels (RPV), flow mixing between high and low concentration coolants will cause uneven distribution of boric acid. Severe cases will lead to boron dilution transient and re-criticality safe operation accident. Therefore, accurate measurement of coolant flow mixing characteristics and boron concentration distribution in RPV and study of the mechanism affecting the mixing process are of great significance to improve reactor operation safety. Based on the structural characteristics of large advanced pressurized water reactor pressure vessel, a large-scale visual experimental device was built by 1:6 scale modular design. Optical path correction scheme is used to solve the problem of optical path distortion existing in PLIF technology, and improved ratio calibration method is used to reduce the experimental error of concentration field distribution measurement. Flow mixing process and boron concentration field distribution in pressure vessel under single loop conditions were obtained based on plane laser induced fluorescence technique (PLIF). The influence of Reynolds number on boron diffusion behavior was studied. The experimental results show that during single loop operation, the localized eddy current formed in the left area of the lower head will result in a high concentration retention zone and a V-shaped boric acid concentration distribution in the core. With the increase of the ratio of cold tube flow to injection flow, the time when the high concentration coolant reaches the core hardly changes with the boric acid solution flow rate. The mixing diffusion behavior of boron concentration is the best when the ratio of injection to Reynolds number in the cold tube is at a certain value.

Critical review of nuclear power plant carbon emissions

Nuclear power plays a crucial role in achieving the target of carbon neutrality to build a sustainable society. However, it is not “carbon-free” when considering its entire life cycle. Therefore, accurate accounting and monitoring of its generated carbon emissions are required to avoid miscalculations of nuclear energy as a clean energy source. In this study, the life-cycle carbon emissions of nuclear power plants (NPPs) with different reactor types are reviewed. In addition to the characteristic differences among different reactors, disparities in the review results originate from the varying emissions at the respective stages of the nuclear fuel cycle, technology choices at each stage and accounting methods and boundaries. The carbon emissions resulting from NPP construction and operation are underestimated due to the limited data and methods, which creates uncertainty in the evaluation of NPP carbon emissions. An integrated framework for carbon emissions accounting considering the construction and operation of NPPs (CACO-NPP) is proposed. This integrated framework aims to improve the accounting accuracy for carbon emissions originating from NPPs. An emerging Generation III NPP with the latest technology, HPR1000 (an advanced pressurized water reactor), was adopted as a case study. The results show that the total emissions resulting from vegetation loss, equipment manufacturing and labor input during construction and operation are 1232.91 Gg CO2 with a carbon intensity of 1.31 g CO2/kWh, indicating the notable mitigation capability of Generation III NPPs. By combining the maturity of HPR1000 technology with successive design improvements, the carbon emissions of such reactor types could be further reduced. This development is very important for realizing China’s carbon neutrality target.

A comparative economic study of nuclear hydrogen production, storage, and transportation

The cost of hydrogen production from environmentally friendly energy resources is a primary barrier to fully realizing a hydrogen economy. Therefore, a detailed analysis of hydrogen production costs from various sustainable routes is required. This paper presents a comparative economic study of hydrogen production using several nuclear reactors integrated with electrolysis hydrogen production methods. The Hydrogen Economic Evaluation Program (HEEP) software is used for comparing hydrogen production using different technologies. The HEEP tool was developed by International Atomic Energy Agency (IAEA) and offers comprehensive data for the different nuclear power plants, hydrogen production processes, and various storage and transportation options. The current study utilizes HEEP capabilities to evaluate the cost of nuclear hydrogen production and compare it with non-nuclear sources. Furthermore, it considers the cost of hydrogen storage and transportation options. The reactors considered are Advanced Pressurized Water Reactor (APWR) and High Temperature Gas-cooled Reactor (HTGR) to power low- and high-temperature electrolyzers. In addition, the costs of hydrogen produced using electricity from grid, solar, and wind are evaluated and compared with that of nuclear plants. It is found that the cost of hydrogen production and storage using Compressed Gas (CG) from an APWR plant is 8.2 $/kg for a plant capacity of 360 MWe; the cost is reduced to 6.06 $/kg when the capacity increased to 1117 MWe. It is also found that HTGR integrated with HTSE can produce hydrogen at 3.51 $/kg, which is slightly lower than the case of using the grid electricity, which costs 3.55 $/kg.

Parametric neutronic analysis of different cladding options for ThO2 pellet of advanced dual-cooled annular PWR assembly

This study presents a parametric neutronic evaluation of different candidate claddings for a purposed candidate fuel (Th-233U-235U)O2 in an advanced dual-cooled annular PWR assembly. The investigated claddings include Zircaloy-2/4, zirconium carbide (ZrC), silicon carbide (SiC), ferritic Fe-20Cr-5Al (FeCrAl) and APMT alloys, and austenitic 310 (SS-310), compared to the by design cladding for the AP1000 assembly, Zirlo™. Neutronic calculations were performed using the deterministic code DRAGON, and the parameters included infinite multiplication factor, criticality periods, neutron flux spectra, radial power distributions, rim effects, and reactivity feedback coefficients.The results show that the candidate fuel with Zirlo™ achieves a higher criticality period than UO2 fuel in both solid and annular configurations. The candidate fuel with SS-310 cladding exhibits a higher criticality period penalty of about 12.60% compared to Zirlo™, APMT, and FeCrAl claddings show a criticality period penalty of about 10.94% and 8.79%, respectively. The candidate fuel with ceramic claddings, SiC and ZrC, can achieve slightly higher criticality periods compared to Zirlo™ of about 1.11% and 0.24%, respectively, due to their low thermal neutron absorption cross-sections. Regarding the power profile, the annular (Th-233U-235U)O2 fuel exhibits a flatter behavior compared to UO2 fuel. However, the claddings with lower capture cross-sections exhibit slightly increased fission power through the fuel pellet. The fuel temperature coefficient for candidate fuel exhibits more negative values as the capture cross-section of the cladding increases, while the molybdenum and zirconium-containing claddings exhibit more favorable moderator temperature coefficients.

Neutronic Characterization for a Pressurized Water Reactor Spent Fuel Assembly

The transfer of nuclear spent fuel from the reactor storage pool to dry storage or for reprocessing or final disposition requires information about its isotopic composition, decay heat, and other thermomechanical properties. The spent nuclear fuel assembly of a typical advanced pressurized water reactor, AP-1000, was characterized using the Monte Carlo MCNPX code and SCALE/ORIGEN code. The simulation of operational history started from the operation of the first fresh core for an average fuel assembly with certain physical isotopic parameters until 25 GWd/tonne U discharge burnup. The analysis considered the calculations of the radionuclide inventories, activity, neutron emission spectrum, gamma-ray emission spectrum, and decay power after 700 effective full power days and for post different time ranges until a 1 million–year cooling period. The comparison of some results of the two codes showed small differences due to the consideration of the continuous-energy variation for neutrons in the MCNPX code and the discrete energy assumption in the SCALE/ORIGEN code.

Uncertainty and sensitivity analysis of iodine release in severe accidents of advanced pressurized water reactors based on the Latin Hypercube method and Grey Correlation Coefficients

Based on an HPR1000 pressurized water reactor (PWR), a large break loss of coolant accident (LBLOCA) model was created in ASTEC software to simulate several accident scenarios. By varying the initial conditions, the release of iodine was evaluated to analyse the influencing factors. It was shown that the break size contributes to the release rate of iodine, while in-vessel retention (IVR) and the containment spray system (CSS) play significant roles in inhibiting the release of iodine into the environment, and more than 90% of the iodine is distributed in the containment when a break occurs in a hot leg or surge pipeline. With the use of the Latin hypercube method, grey correlation coefficients were calculated to quantitatively estimate the input parameters. The results showed that the main factors affecting iodine release vary in the different phases, and the temperature of the cooling water in external vessel reactor cooling (EVRC) displays the highest correlation coefficients, with values greater than 0.85. Overall, IVR is an important measure to prevent iodine release into containment and environmental settings.

Numerical simulation of air discharged in subcooled water pool

Turbulent jet discharges in subcooled water pools are essential for safety systems in nuclear power plants, specifically in the pressure suppression pool of boiling water reactors and In-containment Refueling Water Storage Tank of advanced pressurized water reactors. The gas and liquid flow in these systems is investigated using multiphase flow analysis. This field has been extensively examined using a combination of experiments, theoretical models, and Computational Fluid Dynamics (CFD) simulations. ANSYS CFX offers two approaches to model multiphase flow behavior. The non-homogeneous Eulerian-Eulerian Model has been used in this work; it computes global information and is more convenient to study interpenetrated fluids. This study utilized the Large Eddy Simulation Model as the turbulence model, as it is better suited for non-stationary and buoyant flows. The CFD results of this study were validated with experimental data and theoretical results previously obtained. The figures of merit dimensionless penetration length and the dimensionless buoyancy length show good agreement with the experimental measurements. Correlations for these variables were obtained as a function of dimensionless numbers to give generality using only initial boundary conditions. CFD numerical model developed in this research has the capability to simulate the behavior of non-condensable gases discharged in water.

Neutronic investigation of enriched gadolinium, natural gadolinium and Zirconium-erbium as burnable poisons in the advanced PWR assemblies by MCNPX code

In Pressurized Water Reactors, burnable absorbers are implemented for controlling the excess reactivity and avoiding the high power peaking factor. A feasibility study was carried out for investigating the effect of enriched gadolinium (157Gd2O3), natural gadolinium (152 - 160Gd2O3) and Zirconium-Erbium (ZrEr2) as burnable absorbers on the safety parameters of advanced PWR assembly. Three poisoned assemblies are modeled and simulated by the well-known MCNPX code. To hold a fair comparison between these assemblies, the author adjusted the concentration of fissile materials and burnable absorbers in each assembly to provide approximately the same initial reactivity. The MCNPX simulation showed that the enriched gadolinium model achieved a number of advantages over Zirconium-Erbium as: 1) prolonging the fuel cycle length by 1.04%, 2) increasing the negativity of moderator temperature coefficient and fuel temperature coefficient by 10.3% and 9.3% at the BOL and this contributes to core safety, 3) minimizing residual reactivity penalty at the higher burnup steps and 4) more hardening of neutron spectrum at the BOL. On the other hand, the Zirconium – Erbium layer presented a number of benefits over enriched gadolinium as 1) reducing the peaking factor by 1.3%, 2.6% and 2% at 20, 30 and 50 GWd/ton, 2) increasing the reactivity worth by 1.6% at the BOL, 3) increasing the power flattening by 7.6% at the EOL, and this is considered an enormous economic benefit. Therefore, the modeling of enriched gadolinium and zirconium erbium in this study is suggested for improving the neutronic and safety parameters of PWR cores.

Depressurization strategy of automatic depressurization system for advanced pressurized water reactor

China’s advanced pressurized water reactor is simulated using a system analysis code. A strategy for operating the reactor’s automatic depressurization system (ADS) when it is inadvertently opened is investigated. Results show that under four typical ADS pipeline failure conditions, the ADS can effectively release the primary loop pressure and maintain the safety of reactor core. The redundancy of the two sets of independent ADS depressurization pipelines is designed to be 2 × 100%. The step-by-step opening of pipelines ADS-1–ADS-3 can effectively improve the depressurization and cooling rates of the primary loop. Pipeline ADS-2 is effective for accelerating depressurization, and pipeline ADS-3 can be regarded as the redundancy of ADS-2. The step-by-step opening of the pipelines increases the inlet pressure and widens the inlet pressure curve of the sprayer, enabling the sprayer to maintain a long and stable critical jet state.

Improvement and validation of aerosol models for natural deposition mechanism in reactor containment

Nuclear safety is the lifeline for the development and application of nuclear energy. In severe accidents of pressurized water reactor (PWR), aerosols, as the main carrier of fission products, are suspended in the containment vessel, posing a potential threat of radioactive contamination caused by leakage into the environment. The gas-phase aerosols suspended in the containment will settle onto the wall or sump water through the natural deposition mechanism, thereby reducing atmospheric radioactivity. Aiming at the low accuracy of the aerosol model in the ISAA code, this paper improves the natural deposition model of aerosol in the containment. The aerosol dynamic shape factor was introduced to correct the natural deposition rate of non-spherical aerosols. Moreover, the gravity, Brownian diffusion, thermophoresis and diffusiophoresis deposition models were improved. In addition, ABCOVE, AHMED and LACE experiments were selected to validate and evaluate the improved ISAA code. According to the calculation results, the improved model can more accurately simulate the peak aerosol mass and respond to the influence of the containment pressure and temperature on the natural deposition rate of aerosols. At the same time, it can significantly improve the calculation accuracy of the residual mass of aerosols in the containment. The performance of improved ISAA can meet the requirements for analyzing the natural deposition behavior of aerosol in containment of advanced PWRs in severe accident. In the future, further optimization will be made to address the problems found in the current aerosol model.

Simulation of jets induced by steam injection through multi-hole sparger using effective heat and momentum models

Direct contact condensation (DCC) of steam in the pressure suppression pool (PSP) is used to control containment pressure in Boiling Water Reactors (BWRs) and Advanced Pressurized Water Reactors (APWRs). The competition between momentum and heat sources induced by steam injection through multi-hole spargers and blowdown pipes determines whether the pool is thermally stratified or mixed. Development of thermal stratification affects capacity of the PSP to condense steam. To enable computationally efficient modeling of the PSP transients, the Effective Heat Source (EHS) and Effective Momentum Source (EMS) models have been proposed previously. The EHS/EMS models enable simulation of the large scale pool behavior without explicit modeling of the steam water interface and DCC phenomena. One of the problems for an optimal implementation of the EHS/EMS models is the definition of the boundary conditions that impose distribution of the momentum and heat sources. In this work, EHS/EMS models are implemented using “Unit Cell” approach in ANSYS Fluent to provide detailed numerical analysis of the individual turbulent jets induced by steam injection through the sparger holes. The model is validated against data from Particle Image Velocimetry (PIV) and temperature measurements for a range of steam injection conditions in the PANDA HP5 tests. Good agreement between the test data and simulations suggests that the proposed model can provide sufficiently accurate prediction of both local and large scale phenomena induced by steam injection into the pool.

Numerical Investigation on the Thermal-Hydraulic Characteristics near a Corroded Lower Head of the Reactor Vessel During IVR-ERVC

Third-generation advanced pressurized water reactors adopt the external reactor vessel cooling (ERVC) strategy to ensure the pressure vessel is not at risk of melt-through in severe accidents, thereby completely controlling radioactive materials in the pile. However, due to its long-term service, vessel aging, steel corrosion, and oxidation may lead to deformation at different locations on its outer surface, forming various shapes of sawtooth structures, thus affecting the heat transfer behavior of the high-temperature walls during ERVC. In this paper, the Fluent code, coupled with boiling heat and mass transfer equations based on user-defined functions (UDFs) was used to simulate the thermal-hydraulic processes on the lower head with three typical sawtooth structures. The distribution of the stagnation zone for vapor buildup was the main focus. By varying the heat flux installed on the lower head and the inlet velocity of the flow channel, the onset time of critical boiling and the development of the location of critical heat flux over time were further investigated. It was found that a long period of unstable bubble generation and detachment before the onset of critical boiling may occur on the lower head. These findings can provide technical support for the safety design of advanced nuclear reactors.