Reactor Primary Circuit Research Articles

Streamline rotodynamic pump model for two-phase flow simulation

Abstract This study is concerned with the simulation of water-vapor two-phase flows in pressurized water reactors (PWR) primary pumps. The correct modeling of such flows is essential to predict the consequences of a Loss Of Coolant Accident (LOCA), an accident scenario involving a rupture in the reactor primary circuit. During such accidents, pressure loss leads to important coolant vaporization in the circuits, pumps and other system components. A review of turbomachinery streamline models, used to represent pumps in system codes, is presented. Particular attention is paid to the model used in the CATHARE-3 system code, as its development is the motivation for the present work. Numerical simulations were then performed using CATHARE-3 streamline pump model, currently validated for single-phase liquid flows. These simulations included test cases for pure liquid and vapor conditions, as well as one two-phase flow case, with inlet void fraction of approximately 10 %. For pure liquid water, the agreement is good for discharge between 75% and 150% of nominal discharge. Results are satisfying up to 300% of nominal discharge for the single-phase vapor case. The two-phase results are close to experimental values around 75 and 130 % of nominal discharge. Further work is required in order to consolidate the physical analysis and to improve and validate CATHARE-3 streamline pump model for two-phase flow conditions.

Reduced-order method for nuclear reactor primary circuit calculation

Reduced-order method for nuclear reactor primary circuit calculation

Evaluation of the unavailability of the primary circuit of Triga SSR reactor, importance factors and risk criteria for its components

Abstract Probabilistic Safety Assessment is a tool for improving a nuclear installation and its safety in all its function phases. PSA provides a methodical way of identifying sequences resulting from a wide range of initiating events concerning an accident, and includes systematic and realistic determination of accident frequencies and their consequences. This paper analyzes the unavailability of the TRIGA SSR reactor primary circuit, obtaining the system unavailability value, identifying main contributors to system failure, and calculating risk importance factors. Two cases were considered: one without considering Common Cause Failure (CCF) in the developed fault tree, and the second considering CCF on pumps and heat exchangers lines. The aim was to assess their influence on the primary system’s unavailability. Two importance factors, Fussell–Vesely and Risk Achievement Worth, were employed to identify components with a strong impact on the risk (probability of system failure). These factors aid in enhancing maintenance activities and testing. The analysis utilized the EDFT code, part of the PSAMAN code package developed in SCN-Pitesti.

Probabilistic safety analysis of the large primary circuit leakages accident scenarios in the VVER-reactor

Probabilistic safety analysis of the loss of coolant accidents in the VVER-type reactor plant has been performed taking into account the internal initiating events of the reactor primary circuit large leakages. Swedish program code RiskSpectrum PSA was used for probabilistic safety analysis. Logical probabilistic models of the accident scenarios of the large primary circuit leakages in the VVER-type reactor were developed taking into account different operation modes of the power plant unit. Critical paths and probabilities of the accident scenarios occurrence of the large leakages were identified. The critical path of development of the reactor primary circuit large leakages accident with large leakages of 140–346 mm diameters is the accident sequence with a leak in the pipeline to the reactor pressurizer vessel. It has been established that the greatest contribution to the failure of safety functions during this initiating event is made by the failures due to common causes of the supporting systems (cooling and ventilation systems), critical consumers cooling circuit, emergency injection system. The critical path of the accident with large leakages of 346–850 mm diameters is the accident sequence with the rupture of any of the four primary circuit loops. The greatest contribution to the failure of safety functions during this initiating event is made by the failures due to common causes of the emergency injection system elements. Based on the accident analysis, recommendations for increase of performance reliability of safety functions during the large leakages accidents under all operation states of the nuclear power plant unit were given. In order to increase reliability of safety systems, it is necessary to eliminate failures due to common causes of equipment, increase the reliability of operation of supporting systems, change the maintenance and checking equipment procedures.

Corrosion of Chromium Coating Fabricated on Zircaloy-4 Substrate.

In the nuclear industry, coated cladding is a topical problem and it is chosen as the near-term and most promising ATF (Accident-Tolerant Fuel) cladding concept. The main objective of this concept is to enhance the accident tolerance of nuclear power plants and accordingly, the performance of cladding is expected to be improved. This work assesses the corrosion performance of a Zircalloy-4 alloy coated with a thin chromium coating by MS (magnetron sputtering), tested under a CANDU (CANada Deuterium Uranium) reactor primary circuit simulated condition (LiOH solution, 10 MPa, 310 °C, pH = 10.5). The anticorrosive performance is evaluated by a gravimetric analysis, a metallographic analysis, X-ray diffraction, electronic microscopy, and electrochemical methods. A four times less gain mass was noticed compared to uncoated Zircaloy-4, indicating a smaller corrosion rate. The SEM micrographs illustrate that the coatings are still adherent, and they are keeping the initial morphological characteristics during the autoclaving process. A SEM cross-section analysis shows values of the thickness of the coatings between 0.8 and 1.46 µm. By XRD, the presence of Cr2O3 oxide is identified. Electrochemical testing confirms good stability and good corrosion performance of Cr coating over time under autoclave conditions.

Stress Corrosion Cracking of High-Strength Ni-Base Alloys in Pressurized Water Reactor Primary Water Containing KOH vs. LiOH

The stress corrosion cracking (SCC) behavior of high-strength Ni-base Alloys X-750 and 718 has been investigated to assess the effect of replacing LiOH with KOH as the pH moderator in pressurized water reactor (PWR) primary circuit coolant. The SCC initiation behavior of X-750 was evaluated by multitensile specimen constant load test in either LiOH- or KOH-containing PWR primary water. The SCC growth behavior of X-750 and Alloy 718 was evaluated on compact tension specimens with on-the-fly changes between LiOH- and KOH-containing water in three different PWR primary water chemistries. Direct current potential drop technique was used for in situ monitoring of crack initiation time and crack extension in SCC initiation and propagation tests, respectively. The results suggest no significant effect of KOH vs. LiOH on the SCC initiation or propagation of the tested materials.

Transient three-dimensional numerical simulation on asymmetrical flow and thermal characteristics of CEFR sodium pool under one secondary pump shaft stuck accident

The China Experimental Fast Reactor (CEFR) primary circuit adopts a pool-type structure. The sticking of the main circulating pump shaft in the secondary circuit is an important design-based accident for CEFR, and it is necessary to investigate the three-dimensional thermal-hydraulic phenomena in the sodium pool under the accident. However, conducting relevant experiments directly is almost impossible due to the large dimensions and intricate flow paths within the reactor. In the present work, the “modular modeling and integrated coupling calculation” method is used to build the detailed model and calculate the primary circuit of CEFR in a 1:1 scale. The three-dimensional thermal-hydraulic phenomena inside the sodium pool under the accident condition of one secondary circuit pump shaft stuck are simulated. According to the calculation results, the backflow phenomenon of the fault loop pressure pipe and the intermediate heat exchanger (IHX) begins to appear around 50 s. An obvious thermal stratification phenomenon occurs in the hot pool at the beginning of the accident. As the accident progresses, the temperature difference between the hot and cold pools decreases gradually, and the thermal stratification of the hot pool gradually diminishes. The temperature difference between the inlet and outlet of the same IHX decreases from a maximum of 170 °Cto 2 °C. During the course of the accident, the peak temperatures calculated for the fuel assembly cladding and fuel pellet are 541.78 °C and 1040.17 °C, respectively, which is far below the designed temperature limit of 800 °C and 2800 °C, respectively. Moreover, comparative analysis is conducted on the simulation results of pump shaft stuck in the primary and secondary circuits. The results suggest that when the shaft of one secondary circuit pump is stuck, the average temperature of the sodium at the outlet of the core is higher and the temperature difference between the inlet of different loop pumps is greater, which places more substantial demands on the reactor component structure. It provides significant numerical reference for the safety design and assessment of CEFR.

The Role of Determinism in the Prediction of Corrosion Damage

This paper explores the roles of empiricism and determinism in science and concludes that the intellectual exercise that we call “science” is best described as the transition from empiricism (i.e., observation) to determinism, which is the philosophy that the future can be predicted from the past based on the natural laws that are condensations of all previous scientific knowledge. This transition (i.e., “science”) is accomplished by formulating theories to explain the observations and models that are based on those theories to predict new phenomena. Thus, models are the computational arms of theories, and all models must possess a theoretical basis, but not all theories need to predict. The structure of a deterministic model is reviewed, and it is emphasized that all models must contain an input, a model engine, and an output, together with a feedback loop that permits the continual updating of the model parameters and a means of assessing predictions against new observations. This latter feature facilitates the application of the “scientific method” of cyclical prediction/assessment that continues until the model can no longer account for new observations. At that point, the model (and possibly the theory, too) has been “falsified” and must be discarded and a new theory/model constructed. In this regard, it is important to stress that no amount of successful prediction can prove a theory/model to be “correct”, because theories and models are merely the figments of our imagination as developed through imperfect senses and imperfect intellect and, hence, are invariably wrong at some level of detail. Contrariwise, a single failure of a model to predict an observation invalidates (“falsifies”) the theory/model. The impediment to model building is complexity and its impact on model building is discussed. Thus, we employ instruments such as microscopes and telescopes to extend our senses to examining smaller and larger objects, respectively, just as we now employ computers to extend our intellects as reflected in our computational prowess. The process of model building is illustrated with reference to the deterministic Coupled Environment Fracture Model (CEFM) that has proven to be highly successful in predicting crack growth rate in metals and alloys in contact with high-temperature aqueous environments of the type that exist in water-cooled nuclear power reactor primary coolant circuits.

Hydrogen Sensor to Monitor the Conditions in the Primary Circuit of a Nuclear Reactor

AbstractA hydrogen concentration measurement device is developed for nuclear reactor primary circuits water. Despite the necessity to monitor hydrogen, only one producer offers a really selective hydrogen sensor on the world market because of the complexity of the device preparation when long‐term signal stability is required. Therefore, the Mass Transfer Laboratory at UCT Prague shares the experience gained during the sensor development. The results of the amperometric hydrogen sensor laboratory tests are presented, which demonstrate how to improve the sensor durability. An operational test in Dukovany nuclear power plant confirmed a significant improvement of the latest sensor version and showed that a simple principle could be realized in the form of a reliable device for industrial measurements.

IEA-R1 Renewed Primary System Pump B1-B Nozzles Stress Analysis

The present report is a summary of the structural analysis of the pump nozzles applying the finite element method by using the Ansys computer program. The IEA-R1 RR is an open pool-type moderated and cooled by light water using beryllium/graphite as a reflector. The reactor can reach up to 5MW of thermal power cooled by the primary and secondary systems. The primary coolant system consists of a piping arrangement, a decay tank, two pumps, and two heat exchangers. The primary pump B1-B presented some failures requiring refurbishment by a new one. The pump used in the IEA-R1 must meet the requirements inherent to the nuclear installation, in addition to the operational requirements for rotating equipment, such as flow and pressure, and structural integrity of the body and nozzles. The supplier specified the type of pump suitable for the System. The pump furnished granted mechanical allowable loads for the nozzles that were lower than the loads imposed by the piping on the nozzles. To enable the installation of the pump in the primary circuit, new support was inserted in the piping system next to the pump minimizing efforts and deformations. A piping stress analysis was carried out to obtain the new efforts imposed on the nozzles. For validation of the motor pump set, a verification of the nozzles was done compared with API 610 standard loads, and the allowable loads of the provider. Finally, a structural analysis of the pump nozzles with the new loads was developed using the finite element method. The calculated stresses meet the limits prescribed by the ASME code; therefore, the new B1-B Pump is approved for operation at the IEA-R1 Nuclear Research Reactor primary circuit.

Thermohydraulic studies of alkali liquid metal coolants for justification of nuclear power facilities

The paper presents and discusses the results of experimental and computational studies obtained by the authors on hydrodynamics and heat exchange in fuel assemblies of the alkali liquid metal cooled fast reactor cores, and experimental data on hydrodynamics of flow paths in the heat exchanger and reactor header systems. Investigation results are presented on in-tank coolant circulation obtained using a well-developed theory of approximation simulation of the nonisothermic coolant velocity and temperature fields in the fast neutron reactor primary circuit and demonstrating stable stratification and thermal fluctuations in the coolant. Results are presented from experimental and computational simulation of the alkali liquid metal boiling process based on fuel assembly models during an emergency situation caused by an operational occurrence involving simultaneous loss of power for all reactor coolant pumps and the reactor scram rod failure. Objectives are formulated for further studies, achieving which is essential for the evolution of the liquid metal technology, as dictated by the need for the improved safety, environmental friendliness, reliability and longer service life of nuclear power facilities currently in operation and in the process of development.

Study on pressure surge characteristics of reactor coolant pump shaft stuck accident condition based on wavelet analysis

In order to study the internal transient pressure surge law of the reactor coolant pump (RCP) during the shaft stuck transition. Based on HPR1000 reactor primary circuit closed system (RPCCS), the steam generator tube bundle area and the reactor pressure vessel core area are simplified. The hydraulic characteristics of the RCP can be matched with the resistance characteristics of the system pipeline. The transient numerical calculation of the RCP shaft stuck transition process is carried out by establishing a three-dimensional simplified model of RPCCS. The pressure surge characteristics of each monitoring point in the impeller, guide vane and volute passage under different shaft stuck mass flow periods (TQi) are obtained. The collected transient pressure surge data is analyzed by wavelet transform method (WTM). The results show that: in the stable operation state, the wavelet intensity value in the high frequency band in each flow passage change periodically and alternately. During the process of shaft stuck transition, the demarcation point (TN0) is taken when the impeller speed just drops to 0 r/min. The pressure of each monitoring point in the flow passage reaches the extreme value at this point, and then gradually tends to the reference pressure of 15.5 MPa. The regularity of high-band wavelet response in each flow passage of the RCP is gradually destroyed. Taking TN0 as the demarcation point, the low-band wavelet response firstly increases and then decreases. And the low-band wavelet response in the guide vane passage is generally higher than that in other flow passages at the same time. In this process, there is significantly different in the vortex structures on both sides of the impeller blade. The high vorticity magnitude regions appear in the impeller passage on the side of the blade working surface and in the guide vane passage. In this paper, by revealing the time–frequency characteristics of pressure surge in the RCP under shaft stuck accident (SSA), which provides a certain support for the dynamic safety evaluation of the reactor hydrodynamic system.

Sorption of 137Cs and 60Co on Titanium Oxide Films in Light Water Reactor Primary Circuit Environment.

This paper discusses the processes of the long-lived 137Cs and 60Co immobilization on titanium surfaces in simulated light water reactor primary circuit environments. This study is prompted by numerous problems in both the maintenance of equipment during reactor operation and the dismantling of the reactor after the completion of the operation, which is associated with contamination of working surfaces with long-lived radionuclides. The composition of the oxide films formed on the surface of commercial titanium alloy ПT-3B has been studied with specimens prepared in autoclave test conditions and surface samples from the pipeline sections to which the primary coolant was applied. These films on the coolant pipeline surface consist of a titanium dioxide layer tightly adhered to the pipeline metal surface and weakly fixed deposits—crystallites comprised of titanium oxides and other corrosion products (oxides and hydrated oxides of iron, nickel, chromium etc.). The radionuclide composition of the samples was studied by gamma-spectrometry. It is shown that the mechanism of titanium-surface contamination with 137Cs is by physisorption, contamination level increases upon the presence of dispersed particles. For 60Co, both sorption and deposition onto surfaces are observed.

Some fundamental understandings of Zn-injection water chemistry on material corrosion in pressurized water reactor primary circuit

The optimization of water chemistry in pressurized water reactor (PWR) primary circuit is one of the most effective ways to achieve both safety and economy for operating PWR nuclear power plants (NPPs). Special attention has been paid to fundamental research and engineering application of Zn-injection water chemistry (ZWC) in PWR primary circuit in recent years in China due to the rapid development of nuclear power. The present paper mainly reports the status of PWR NPPs in China and some fundamental understandings of ZWC on material corrosion. Effects of temperature (T), pH at T value, Zn concentration, Zn-injection sequence and lasting time on electrochemical corrosion behavior, oxide film characteristic and stress corrosion cracking (SCC) susceptibility of nuclear-grade austenitic alloys were investigated. A modified point defect model was used to discuss the effects of ZWC on the oxide films in high-temperature pressurized water. Some water chemistry parameters in PWR primary coolant system are proposed for mitigating corrosion and SCC of nuclear-grade austenitic alloys.

The Multifunctional Nuclear Magnetic Flowmeter for Control to the Consumption and Condition of Coolant in Nuclear Reactors

The necessity of coolant flow consumption measurement accuracy increase in the nuclear reactor primary circuit has been substantiated. Additionally, the need to control the coolant condition in the current flow inside the pipeline is shown. Nowadays, the real-time coolant’s condition control function is not implemented at stationary nuclear power plants or mobile nuclear power plants used in moving objects. It is shown that a coolant consumption measurement error decreases and its condition data availability increases the heat transfer efficiency and the electrical energy generation (without the nuclear reactor and steam generator design change). Problems arising during coolant consumption control using various flowmeters models in the nuclear reactor primary circuit are considered. It has been found that nuclear magnetic flowmeters can solve these problems. New difficulties are noted as emerging when using pulsed nuclear magnetic flowmeters designs developed for measuring hydrocarbons, water, biological compounds consumption, and condition control. A new nuclear magnetic flowmeter design has been developed using a modulation technique for nuclear magnetic resonance signal recording. Methods for measuring the coolant flow’s longitudinal T1 and transverse T2 relaxation times are presented. Investigations of coolant flow parameters (consumption and relaxation times) inside the pipeline have been carried out. It is found that the measurement error for these parameters does not exceed 1%. The prospects of using the developed nuclear magnetic flowmeter-relaxometer design in the nuclear reactor first circuit are shown.

Core physical conceptual design of small reactor for unmanned underwater vehicle

In order to improve the endurance of unmanned underwater vehicle (UUV) and meet the needs of long-term underwater mission, the conceptual design of the 3.68 MWt long-life small reactor JCZER-20 is carried out in an UUV application scenario with a propulsive power of 850 kW. According to the design content, the concept of active zone partition design is put forward, and a new circulating scheme of coolant in the reactor primary circuit is given. The layout of control body in the reactor and the design of control system are carried out. The control rod structure of JCZER-20 adopts special design to deal with special conditions such as shaking during operation. The reactivity distribution scheme is given, and the reactor physical characteristics are analyzed for the design scheme. This scheme can lengthen coolant channels, effectively improve the total temperature rise of the coolant, but not increase the extra heat load. MCNP-ORIGEN2 simulation results show that the JCZER-20 reactor can operate at full power for more than 30 years without refueling. The reactor core reactivity can be effectively controlled by the control system of reactor. The core design meets preset design requirements.

Thermodynamic Analysis and Crystallographic Properties of MFe2O4, MCr2O4 and MAl2O4 (M = Fe, Ni, Zn) Formed on Structural Materials in Pressurized Water Reactor Primary Circuit under Zinc and Zinc-aluminum Water Chemistry.

Zinc injection technology (zinc water chemistry, ZWC) was widely applied in pressurized water reactor (PWR) primary circuits to reduce radiation buildup and improve corrosion resistance of structural materials. The simultaneous injection of zinc-aluminium (ZAWC) is a novel implement created to replace part of Zn2+ by Al3+. It was reported ZAWC can improve further corrosion resistance of carbon steels and stainless steels. However, ZAWC sometimes showed even negative effects on Nickel-alloys. In this study, mechanism of formation of oxide film on metals was investigated. The reactions of Fe2+ Ni2+ in oxide films replaced by Zn2+, or Fe3+ replaced by Al3+ in ZAWC were analysed. The thermodynamic data and solubility of mixed oxides (ZnFe2O4, ZnCr2O4, and ZnAl2O4), the products of replace reactions, were calculated. According to the Gibbs free energy difference between products and reactants, values of the formation reaction of ZnFe2O4, ZnCr2O4, and ZnAl2O4 are extremely negative. Solubility of ZnAl2O4 is the lowest among mixed oxide products, which implies the oxide film composites of ZnAl2O4 may show a lower corrosion rate. In addition, the preferential formation of NiAl2O4 on Ni-based-alloy, under ZAWC, was discussed based on crystallographic properties of spinel, which was considered as the cause of negative effects of ZAWC on corrosion resistance of Nickel-alloys. This research provides an analytical basis for the study of thermodynamic stability of oxide films under different chemical chemistry and a theoretical basis for improving corrosion resistance of different metals and optimizing the chemical conditions of PWR primary circuit.

Study on Hydraulic Characteristics of Reactor Coolant Pump Shutdown Transition Process Based on Primary Circuit Closed System

According to the structural characteristics of a steam generator and a reactor pressure vessel, the reactor primary circuit system of HPR1000 is simplified to make the reactor coolant pump (RCP) characteristics match the resistance characteristics of the system pipeline, so a simplified model of reactor primary circuit closed system is formed. On this basis, the RCP shutdown accident in a single circuit of the reactor is numerically simulated. Results show that the pressure in the pipeline system has changed greatly compared. The flow rate drops rapidly and decreases to 1.7% of stable operation at 10s. The torque drops sharply to −51.89% of stable operation, and then rises slowly to zero. The pressure at the inlet and outlet of the impeller changes dramatically which leads to the change of blade load and the velocity in the impeller passage decreases gradually. The pressure and velocity in the guide vane continue to decrease.

Three-dimensional numerical research on natural circulation characteristics of decay heat removal system in pool-type fast reactor CEFR

The pool-type sodium-cooled fast reactor decay heat removal system (DHRS) is an important engineered safety system to ensure the establishment of natural circulation in the case of station blackout (SBO). It is of great significance to study characteristics of DHRS for evaluating its passive residual heat removal capacity. Due to numerous core subassemblies and small scale of gaps between subassemblies, it is difficult to use the same scale grids to describe the three-dimensional flow paths in sodium pool with full core. Therefore few researches, either experiments or numerical simulation, have been performed on such complicated phenomena so far. In order to identify the natural circulation flow paths of fast reactor, an integrated three-dimensional simulation model of China Experimental Fast Reactor (CEFR) primary circuit system with full core was established, including 712 subassemblies and small gaps between subassemblies. FLUENT was used to carry out three-dimensional numerical simulation. Through the transient simulation of 5000 s, the main flow paths including the intra-subassembly flow path, the inter-wrapper flow path and the Main Vessel Cooling System (MVCS) backflow path were identified. At 5000 s, the intra-subassembly flow rate, the inter-wrapper flow rate, and the MVCS backflow rate were 3.07 kg/s, 1.53 kg/s, and −1.94 kg/s respectively. Thesymmetrical thermo-hydraulic characteristics of two loops were obtained to analyze thermal stratification in sodium pool. The results provided a key three-dimensional numerical reference for the safety of CEFR.

Development of a methodological approach for the computational investigation of the coolant flow in the process of the sodium cooled reactor cooldown

A methodological approach has been developed for the computational investigation of the thermal-hydraulic processes taking place in a sodium cooled fast neutron reactor based on a Russian computational fluid dynamics code, FlowVision. The approach takes into account the integral layout of the reactor primary circuit equipment and the peculiarities of heat exchange in the liquid metal coolant, and makes it possible to model, using well-defined simplifications, the heat and mass exchange in the process of the coolant flowing through the reactor core, and the reactor heat-exchange equipment. Specifically, the methodological approach can be used for justification of safety during the reactor cooldown, as well as for other computational studies which require simulation of the integral reactor core and heat-exchange equipment. The paper presents a brief overview of the methodological approaches developed earlier to study the liquid metal cooled reactor cooldown processes. General principles of these approaches, as well as their advantages and drawbacks have been identified. A three-dimensional computational model of an advanced reactor has been developed, including one heat-exchange loop (a fourth part of the reactor). It has been demonstrated that the FlowVision gap model can be applied to model the space between the reactor core fuel assemblies (interwrapper space), and a porous skeleton model can be used to model the reactor’s heat-exchange equipment. It has been shown that the developed methodological approach is applicable to solving problems of the coolant flow in different operating modes of liquid metal cooled reactor facilities.