PWR Reactor Research Articles

Analytic network process (ANP) based decision support tool for nuclear power plant location and reactor type selection

The nuclear energy industry has seen a resurgence due to interest in sustainable energy. Regulatory agencies are evaluating innovative technology and facility designs for the first time in a long time. The nuclear energy business needs a workable instrument for decision-making in light of the shift in public and regulatory opinion. This study has two main motivations. The first is to determine the set of criteria to be used in determining suitable locations for a nuclear power plant (NPP) to be established in Turkey and to determine suitable regions with Analytic Network Process (ANP) based analyzes in the light of these criteria. The second is to select the most suitable nuclear reactor type for the specified location. Plant efficiency and lifetime vary depending on the reactor type used. For this reason, in the study, both the location of the nuclear power plant and the selection of the appropriate reactor type for the plant were evaluated together. In this context, a criterion set consisting of 3 main criteria and 20 sub-criteria to be used in NPP installation location and reactor selection was determined. In the study, site selection was made by focusing on economic, social and technical factors, and reactor selection was made by focusing on technical factors. Giresun, Trabzon, Rize and Artvin provinces located in the Eastern Black Sea region were evaluated for NPP installation. For reactor selection, we focused on reactor types according to their coolants and in this context, PWR, PHWR, GCR and BWR reactors were examined. NPP location and reactor selection in the study; It was realized as Giresun (PWR) > Rize (PWR) > Trabzon (BWR) > Artvin (BWR).

Convective Heat Transfer in PWR, BWR, CANDU, SMR, and MSR Nuclear Reactors—A Review

Convective Heat Transfer in PWR, BWR, CANDU, SMR, and MSR Nuclear Reactors—A Review

Fractional neutron point kinetics model for reactivity transients of the NuScale and comparison with the classical kinetics approach

In this work, the behavior of the neutron and natural circulation parameters for Reactivity Transients in the PWR reactor called NuScale are analyzed through the novel system equations. It model includes non-integer and ordinary order differentials coupling the neutron point kinetic with the thermos-hydraulic core of the reactor. Transient processes of increase and decrease of the feed flow and the coolant inlet temperature are analyzed to identify the feedback of the thermohydraulic effects with the neutron phenomena in a non-integer order scheme. In the results, it is observed that for short times the power and reactivity have the same behavior with the classical model and the non-integer order model for all values of the anomalous diffusion coefficient simulated. However, for long times of simulation, there is more reactivity and less power, in contrast to the classical model, when reducing the value of the anomalous diffusion coefficient these counter-intuitive effects are increased. Additionally, a delay is observed between the change in reactivity and power. In conclusion, it is observed that sub-diffusive processes are relevant when there are slow changes in reactivity, which can be analyzed through the Fractional Order Neutron Density Equation.

Analysis of variation minor actinide pin configurations Np-237, AM-241, and Cm-244 in UN-PuN fueled pressurized water reactor

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 %.

Effect of Protactinium-231 (Pa-231) Addition on Effective Multiplication Factor Value of ThN-U233N Fueled PWR Reactor

This study aims to determine the optimum percentage of addition of protactinium-231 in a ThN-U233N fueled PWR reactor.In ThN-U233N, uranium-233 is used as the fissile material. Pa-231 as burnable poisson is added to ThN-U233N fuel to reduce the keff value. In this research conducted benchmarking keff values ​​of OpenMC and SRAC codes, homogeneous and heterogeneous core configurations, and the addition of Pa-231 in ThN-U233N fuel. Neutronic analysis was performed using the OpenMC code. In the homogeneous core configuration, the enrichment percentage of U-233 is varied by 1%-15%. Meanwhile, the heterogeneous core configuration uses 3 variations of ring geometry and fuel. The results of the benchmarking of the OpenMC and SRAC code keff values ​​show a maximum error of 1,906%. The homogeneous core configuration produces the optimum value for U-233 enrichment of 3%. Meanwhile, the optimum value of the heterogeneous core configuration is achieved in ring geometry F1 = F2 = F3 = 3 rings with enrichment of U-233 at F1 = 0,5%, F2 = 3%, F3 = 5,5% and addition of Pa-231 of 1.5%. The optimum calculation results has an excess reactivity value 8.37%.

KIT reactor safety research for LWRs: Research lines, numerical tools, and prospects

This paper describes the main research lines of the Karlsruhe Institute of Technology (KIT) relevant to Light Water Reactors (LWR) that permit both the safety evaluation of different reactor designs and the prediction of the radiological source term including its impact in the case of hypothetical severe accidents. The numerical tools for core analysis, nuclear power plant analysis of design basis and severe accidents, and the estimation of the radioactive dispersions of fission products after a core meltdown accident are described and selected results are discussed. The analytical investigations at KIT are complemented by experimental investigations at two facilities, namely the COSMOS-H/-L and the QUENCH facilities devoted to safety-relevant Thermal-Hydraulic (TH) phenomena and in-vessel severe accident early-phase for different kinds of fuel rods including Accident Tolerant Fuel (ATF). These data are crucial for the validation of the numerical tools used for safety demonstration. The paper is complemented by a brief description of the in-house tools under development for improved core/plant analysis based on the multi-physics/-scale coupling methodologies, where the in-house tools are combined with external system TH, Monte Carlo, and thermo-mechanic code to increase the prediction accuracy of the core/plant behaviour under stationary and accidental conditions. Selected results are presented and discussed. It can be stated that the long-lasting safety research activities at KIT devoted to LWR led to the development of a computation route combining external with in-house codes that is very much appropriate for safety evaluations and risk assessment; it includes unique multi-physics/-scale coupled tools for improved core and plant analysis of different reactor designs, e.g. PWR, VVER, BWR, SMR, and research reactors.

Impact of the resonance elastic scattering on reactor-core physics calculation results based on the deterministic method

NECP-Atlas and similar nuclear data processing codes have developed the function of incorporating the resonance elastic scattering effect into the multi-group cross sections and scattering matrices. However, the hyper-fine group method is commonly used in deterministic-based lattice physics codes and whole-core neutronics codes for resonance self-shielding calculations. In this method, the neutron slowing-down equation is established based on the conventional asymptotic scattering model where the neutron up-scattering caused by the resonance elastic scattering is ignored. Consequently, the deterministic codes cannot directly use the multi-group data produced by the nuclear data processing codes. In this paper, a correction factor method is proposed to correct the multi-group cross sections and scattering matrices obtained by the hyper-fine group calculations to consider the resonance elastic scattering. The proposed method has been implemented in the lattice physics code LOCUST. The few-group constants produced by the LOCUST code are provided to reactor-core analysis code SPARK to simulate a CNP-1000 type PWR reactor operated in China. Some core parameters measured during zero-power physics tests at 600 K and operations are analyzed to show the effect of the resonance elastic scattering on the core parameters. The numerical results show that the boron concentration reduces by 10 ∼ 14 ppm at 900 K during operations, which is closer to the measured value and improves the accuracy of reactor-core physics calculation.

Анализ международного опыта решения проблемы образования радиоизотопов сурьмы в первом контуре реакторных установок

The article summarizes and presents an analysis of international experience in solving the problem of formation of radioisotopes of antimony in the primary circuit of reactor facilities of various types. The experience of twelve countries operating power units with VVER, PWR, BWR, and PHWR reactors in identifying sources of stable antimony isotopes in the primary circuit and implementing measures to eliminate them is reviewed. The recommendations of international organizations and requirements of normative documents of individual countries established to limit the formation of radioisotopes of antimony in the primary circuit and their implementation in new designs of reactor facilities of foreign design have been analyzed. A list of the main possible sources of antimony in the primary circuit is determined.

Coolant Hydrodynamics in the PWR Reactor Core with TVS-Kvadrat of Various Designs

. The paper presents the results of experimental studies of the hydrodynamics of the coolant in the core of the PWR reactor with fuel assemblies “TVS-Kvadrat” of various designs. Experimental studies of hydrodynamics consisted in studying the velocity fields and the process of cross-flow of coolant between adjacent fuel assemblies “TVS-Kvadrat” of various designs in homogeneous and mixed cores on large-scale experimental models, including fragments of two adjacent fuel assemblies “TVS-Kvadrat” and a gap between them. The test models differ in the number of installed grids, this is due to the fact that in a homogeneous core the hydraulic load along the height of the fuel assemblies is identical, and in a mixed core the hydraulic load is uneven due to the different number of installed grids. The hydrodynamics of the coolant has been studied on an aerodynamic open-loop based on the theory of hydrodynamic similarity, using individually calibrated pneumometric sensors. Based on the data obtained, the fields of local axial and transverse velocities in various sections along the length of the studied section of the test models have been constructed. The fields of local dimensionless transverse and axial flow velocities in the fuel bundles and the gap between the fuel assemblies “TVS-Kvadrat” of the PWR reactor core have been studied. Peculiarities of the process of transverse flow between adjacent “TVS-Kvadrat” of a homogeneous and the mixed core of the PWR reactor have been revealed. The results of experimental studies have been transferred to the developer of “TVS-Kvadrat”, that is JSC “Experimental Design Bureau of Mechanical Engineering named after I. I. Afrikantov”, for further use in justifying the design and operating modes of a nuclear power plant.

Solusi Persamaan Difusi Neutron Pada PWR (Pressurized Water Reactor) Berbentuk Heksagonal dengan Bahan Bakar Uranium Daur Ulang

The Research on solution of the neutron diffusion equation with a PWR reactor using recycled uranium fuel at 1⁄6 section of the reactor core with a hexagonal IGT-6 geometry. The purpose of this research is to determine the distribution of the neutron flux in the PWR of recycled uranium fuel. The solution is done by computational simulation using the Dev-C++ programming. The parameters used in this study determine the specifications of the reactor core, determine the volume fraction, determine the atomic density, calculate the macroscopic cross-section with the PIJ module, calculate the neutron diffusion equation, calculate ϕ (x,y) using the Gauss Seidel method. The results obtained in this study are the neutron diffusion equation without a source obtaining the highest relative neutron flux value in group 1 of 4,5729×〖10〗^(-2), with a fission source obtaining the highest relative neutron flux value in group 3 of 7,3327×〖10〗^(-4), with fission and scattering sources obtaining the highest relative neutron flux value found in group 2 of 1,5157×〖10〗^(-3), and 3,200 MW of power is added to the source fission, the value of the neutron flux does not change. This is because the addition of power does not affect the value of the neutron flux.

Bibliometric Review and Technical Summary of PWR Small Modular Reactors

Given the real problem of climate change and the consequent need to reduce greenhouse gas emissions, nuclear energy can be considered a real technological option in the absence of a technology that can supply the population with sufficient electrical energy in a sustainable way. The research and development advances of new nuclear reactor designs are evident in the case of Small Modular Reactor (SMR) technology. Two goals are clear in such a technology: (i) reducing onsite construction costs and time; (ii) promoting economic efficiency through the mass production of nuclear reactors. In view of the interest generated by this technology, a technical summary of PWR-SMR design is included in this paper, utilizing the scientific method based on: (i) a bibliometric review of scientific publications related to SMRs; and (ii) a technical analysis of aspects of the different PWR-SMR designs. The results obtained in the present research show a growing interest on the part of the scientific community in research on PWR reactors integrated within SMR designs. Few countries are doing research in this area; however, those that are doing so have achieved greater technological advances in their designs, resulting in greater technological variety and a higher degree of maturity.

A system of automated power control during disturbances that occur inside a PWR nuclear reactor

In order to increase the service life of NPPs is proposed to change the power of the nuclear reactor depending on the needs of consumers. However, power maneuvering increases the risk of an emergency situation, since it is very difficult to maintain a steady state of the reactor when moving from one power level to another. For this, it is necessary to adapt the methods of operational control when changing the power of a nuclear reactor. An approximation model is proposed, with the help of which it is possible to improve the system of automated power control in case of internal disturbances of a nuclear reactor. The change in reactivity that occurs as a result of a change in the concentration of xenon when maneuvering the power of a nuclear power plant was attributed to internal disturbances. An approximation model of the processes was built using the Padé approximation.

Structural integrity assessment of a closure head of a PWR reactor pressure vessel under pressurized thermal shock

The pressurized thermal shock induced during the accidental conditions is the potential threat to the structural integrity of reactor pressure vessel. The pressurized thermal shock puts the reactor pressure vessel under detrimental thermo-mechanical stresses, and a small crack on the surface, particularly at highly stressed locations, may initiate, propagate, become critical, and eventually fail the component. Therefore, critical components are needed to be assessed. For structural integrity analysis of class-I components under pressurized thermal shock, a methodology is proposed in this research article. As a case study, the structural integrity assessment of the closure head (class-I component) of the RPV under pressurized thermal shock is presented here.

Fracture mechanics analysis of a closure head of a PWR reactor pressure vessel the effect of bolt pre-load and thermal stresses

The effect of the bolt pre-load and thermal stresses in addition to pressure stresses on the structural integrity of a closure head of an RPV is presented here. The trends of the stress intensity factors (SIFs) of the surface cracks have been investigated under the combined effects of the loading conditions. A contradictory behavior of the surface cracks at the inside and outside locations of the head is observed due to thermal stresses, while the bolt pre-load decreases the integrity of the head at both locations. The maximum SIFs of the analyzed cracks are also found well below the material’s fracture toughness.

Detection of Numerical Power Shift Anomalies in Burnup Modeling of a PWR Reactor

This paper presents the detection and interpretation of numerical power density anomalies in the full-core PWR reactor model, developed at the level of RPV. The study shows that in the Monte Carlo neutron transport coupled with burnup modeling, power density anomalies may occur and should be carefully investigated. The power density anomalies originate from the initial core asymmetry due to different designs of top and bottom reflectors and non-uniform initial axial neutron flux distribution. The power density anomalies increase with increasing fuel burnup due to spatial changes in 135Xe concentrations and following power density redistribution along the reactor core. A method for anomaly detection for the semi-symmetric reactor core is proposed.

A Raman spectroscopic and ab initio investigation of aqueous boron speciation under alkaline hydrothermal conditions: evidence for the structure and thermodynamic stability of the diborate ion.

Raman spectra of aqueous sodium borate solutions, with and without excess NaOH, NaCl, and LiCl, have been obtained from perpendicular and parallel polarization measurements acquired using a custom-built sapphire flow cell over the temperature range 25 to 300 °C at 20 MPa. The solvent-corrected reduced isotropic spectra include a large well-defined band at 865 cm-1 which overlaps with the boric acid B(OH)3 band at 879 cm-1, and becomes increasingly intense at elevated temperatures. This band does not correspond to the spectrum of any other previously reported aqueous polyborate ions, all of which have symmetric stretching bands at frequencies below that of borate, [B(OH)4]-, at 745 cm-1. Based on the classic high-temperature potentiometric titration study by R. E. Mesmer, C. F. Baes and F. H. Sweeton, Acidity measurements at elevated temperatures. VI. Boric acid equilibriums, Inorg. Chem., 1972, 11, 537-543, the new band was postulated to arise from a diborate ion, [B2(OH)7]- or [B2O(OH)5]-. Ab initio density functional theory (DFT), together with chemical modelling studies, suggest that it is most likely [B2(OH)7]-. Thermodynamic formation quotients derived from the peak areas showed variations with ionic strength as well as charge-balance discrepancies, which suggest one or more unidentified minor equilibrium species may also be present. The most likely candidate is the divalent diborate species [B2O2(OH)4]2- which is also predicted to have a band near 865 cm-1 and is postulated to be present as a sodium ion pair. These are the first quantitative Raman spectra ever reported for borate-rich solutions under such conditions and provide the first spectroscopic evidence of a diborate species at PWR reactor coolant temperatures.

Fracture mechanics analysis of a closure head of a PWR reactor pressure vessel LEFM-based SIF

In this study, fracture mechanics analysis of a closure head of a PWR reactor pressure vessel has been performed under accidental conditions. For the purpose, a wide range of elliptical surface cracks starting from minute cracks to worst-case cracks was postulated at the highest stress concentration locations of the head. The analyzed accidental conditions are small break loss of coolant accident (SB-LOCA) and Rancho-Seco transient (RST). The maximum linear elastic fracture mechanics-based stress intensity factors of the worst-case crack ‘a = 0.25 th; a = 0.33 c’ under SB-LOCA and RST are 104.37 MPa√m and 137.12 MPa√m respectively, which are less than the fracture toughness value of the material i.e. 155.6 MPa√m. Hence, the accidental conditions do not threaten the structural integrity of the proposed closure head.

An Analysis of Water Characteristics of Shiddhirganj 210MW Steam Power Plant, Sirajganj 225MW Combined Cycle Power Plants and Rooppur 2400MW Nuclear Power Plant in Bangladesh: A Comparative Study of Power Plant Grade Water

Water quality has long been an important part of the operation of nuclear and thermal power plants. Water is used as a working and cooling fluid in power plants. The quality of source water to be used in the power plants after treatment should conform to the prescribed values of Physico-Chemical properties like pH, EC, TDS, Alkalinity, Hardness, presence of Chloride content, Silica, and Heavy metals as recommended by technical guidelines. To analyze the water characteristics, water samples were collected from various sources. Two river water samples and a groundwater sample (marked as PP1, PP2, and NPP site) were tested to determine the water properties. The Physico-chemical properties of the water samples were determined using various analytical tests. Heavy metals were also investigated using the Atomic Absorption Spectrophotometric (AAS) process. The Physico-chemical properties of investigated samples were so good as to the recommended values of the World Health Organization (WHO), United States of Public Health (USPH), and Bangladesh Environmental Department. But the values were very far from the recommended values for thermal power plants operation. So, it needs to treat to use in the boilers. Gravitation, Carbon filtration, Ion exchange method, and Reverse Osmosis (RO) are good ways to treat the water before use in power plants. The Padma River water is better than the other two samples when the Physico-chemical characteristics of the sample waters were analyzed to the prescribed levels for the boiler water of a thermal power plant and the secondary water circuit of a third-generation nuclear PWR reactor. The aim of this work is to explore the difference between the source water quality parameters values and those of the recommended values of technical guidelines. To minimize the water treatment cost, a new site for the installation of a power plant project in Bangladesh can be advocated depending on the availability of good quality water.

On the use of time-dependent success criteria within risk-informed analyses. Application to LONF-ATWS sequences in PWR reactors

The classical Probabilistic Safety Analysis (PSA) does not include any time dependence explicitly. However, the success criteria (SC) could evolve during the cycle for some initiating events. In that sense, there is a type of sequence in which this time-dependency is quite important, the family of Anticipated Transient without Scram (ATWS) sequences in Pressurized Water Reactors. Therefore, a new risk-informed approach is proposed in this paper, which makes it possible to obtain the time-dependent SC evolution of the safety functions affected by the Moderator Temperature Coefficient (MTC) value. Then, the evolution of the ATWS conditional core damage probability (CCDP) could be obtained using a PSA model. To quantify the CCDP, the average values of the time-dependent failure probabilities must be computed. Finally, the comparison between the CCDP obtained through the application of the classical PSA approach and the new one makes it possible to quantify the impact of time-dependence on the SC of the headers that this new risk-informed ATWS approach can provide.

New approach to find the equilibrium cycle in PWR reactors using genetic algorithms

This work presents a methodology applied for performing in-core fuel management PWR (ICFM), using genetic algorithms (GAs). A system was developed in which the GAs interact directly with a reactor physics code created for this purpose. The nodal code, called NOD2ABC, is responsible for providing the neutron characteristics of the reactor that one wants to carry out the ICFM and sending them to the GAs to optimize them, based on some criteria. This code was developed in 2-D and uses the Nodal Expansion Method (NEM). The reference reactor of the Central Nuclear Almaraz II, made available by the International Atomic Energy Agency (IAEA), was chosen and used in this work with the objective of qualifying and validating all the methodology used in the ICFM. The Assemblies of Almaraz II were modeled using the SCALE 6.0 system, developed by the Reactor and Nuclear Systems Division (RNSD), of the Oak Ridge National Laboratory (ORNL), with the purpose of generating the homogenized multigroup cross sections applied in the implemented nodal code. A new ICFM approach, using GAs, was structured to find a possible equilibrium cycle of the modeled reactor. The equilibrium cycle found reached all the physical conditions provided, lasts 240 days with the reactor operating at full power and has an average discharge burnup of 28.93 MWd/kgHM.