Reactor Primary Coolant Research Articles

Corrosion of 316L stainless steel produced by laser powder bed fusion and powder metallurgy in pressurized water reactor primary coolant

Corrosion mechanism of 316 L stainless steel produced by laser powder bed fusion-hot isostatic pressing (LPBF-HIP) and powder metallurgy-hot isostatic pressing (PM-HIP) is studied with in-situ electrochemical impedance measurements coupled to detailed oxide film characterization. Quantitative analysis of impedance spectra using the Mixed-Conduction Model and estimation of local kinetic and transport parameters by interpretation of in-depth elemental composition profiles indicated lower corrosion and oxidation rates of LPBF-HIP and PM-HIP materials in comparison to conventional wrought 316 L. This owes to a higher fraction of low-angle grain boundaries, smaller grain size, the presence of nano-sized oxide particles and elevated Cr and Ni contents.

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.

Oxidation of Alloy 690 in Simulated Pressurized Water Reactor Primary Coolant – Experiments and Modeling

During normal operation of the primary loop of a water cooled-water moderated energy reactor (WWER), the associated water chemistry evolves from an initially high-boron high-potassium (beginning-of-cycle, 1200 ppm B and 8 ppm K) to significantly lower concentrations (ca. 400 ppm B and 3.5 ppm K) at the end of the respective campaign. In addition to that, the concentration of Li produced by transmutation of B reaches ca. 0.7 – 0.9 ppm. To the best of our knowledge, the effect of such water chemistry changes on pressurized water reactor (PWR) internals such as Alloy 690 have not been studied yet. Such a study appears relevant also within the frames of a discussion for a possible transition from B-Li to B-K water chemistry in PWRs.In the present paper, the effect of evolution of primary water chemistry during operation on the corrosion rate and conduction mechanism of oxide films on Alloy 690 is studied by in-situ impedance spectroscopy at 280 °C / 8 MPa during 1-week exposures in an autoclave connected to a re-circulation loop. Impedance measurements were performed in three-electrode mode featuring a Pt sheet counter electrode and a Pd electrode negatively polarized with a current of -10 µA vs. another Pt to approximate the reversible hydrogen electrode (RHE). Spectra were recorded at the corrosion potential in a frequency range 10 kHz – 0.1 mHz using an ac amplitude of 50 mV (rms). At the end of exposure, the samples were polarized in the range of potentials -0.1 to 0.5 V vs. RHE to evaluate the stability of the passive oxide. Separate samples of Alloy 690 were simultaneously exposed in the autoclave and subsequently analyzed by Glow-Discharge Optical Emission Spectroscopy (GDOES) in order to estimate the thickness and the in-depth composition of the formed oxides.Impedance data were quantitatively interpreted using the Mixed-Conduction Model for Oxide Films (MCM) to estimate the rates of metal oxidation at the alloy/oxide interface, oxide dissolution and restructuring at the film/coolant interface and ion transport in the protective layer. The effect of water chemistry evolution on corrosion rate and conduction mechanism in the oxide on Alloy 690 in primary coolant is discussed on the basis of the obtained parameters.Acknowledgments: This study is funded by the European Union-NextGenerationEU, through the National Recovery and Resilience Plan of the Republic of Bulgaria, project № BG-RRP-2.004-0002 (BiOrgaMCT).

Irradiation of ultrasonic sensors and adhesive couplants for application in light water reactor primary loop piping and components

The Electric Power Research Institute (EPRI) Nuclear Sector and US Department of Energy Light Water Reactor Sustainability Program are committed to engaging in research and development endeavors to address materials aging issues specific to long term operation of light water power reactors. To this effect, EPRI launched an industry initiative to develop nondestructive evaluation systems for online monitoring of existing cracks in light water reactor primary coolant loop piping and components. One of the goals of this initiative is to develop a sensor system (or systems) that can determine nondestructively if cracks are growing or arrested and, in the case of the former, to characterize their growth rates. A missing component of this initiative is an experimental assessment of how sensors and adhesive couplants will perform in service when exposed to chronic energetic neutron radiation, particularly at the primary coolant loop hot and cold leg dissimilar metal welds, which join the primary loop piping to the reactor pressure vessel and reside in the vicinity of the reactor core. The objective of this experimental study was to determine how ultrasonic transducers and adhesive couplants perform when exposed to irradiation in a test reactor to simulate and accelerate in-service exposure. To achieve this objective, the signal stability of piezoelectric transducers and performance of adhesive couplants as a function of accumulated fast neutron fluence were characterized by collecting ultrasonic data in-situ during irradiation. Of particular interest were the ultrasonic signal quality and time decay of the amplitude of acoustic reflections as a function of fast neutron fluence. The results of the study showed that, of the 8 transducer/substrate sample assemblies tested, only 3 generated usable ultrasonic signals through the conclusion of the irradiation campaign. It was found that high temperature epoxy tends to ultrasonically couple the sensors to the substrates better than three types of refractory ceramic cements studied, as is supported by post irradiation examination. The results obtained through this experimental study will be utilized in the achievement of the overall goal of development of a sensor system to perform online monitoring of primary loop components.This manuscript has been authored in part by UT-Battelle, LLC, under contract DE AC05-00OR22725 with the US Department of Energy (DOE). The publisher acknowledges the US government license to provide public access under the DOE Public Access Plan (https://energy.gov/downloads/doe-public-access-plan [energy.gov]).

Optimization of dissolved hydrogen concentration for mitigating corrosive conditions of pressurized water reactor primary coolant under irradiation (1) evaluation of water radiolysis

ABSTRACT 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.

Study on the basic of spallation target for accelerator driven system

A study on the basic of spallation target for Accelerator Driven System (ADS) has been done. ADS is a subcritical assembly combined with a high-power proton accelerator. In the center of the subcritical assembly, there is a spallation target for producing primary neutrons that will be multiplied by the subcritical core surrounding it. The neutrons are produced via spallation reaction which occurs when the spallation target is bombarded by high-energy protons. The spallation target has to be designed so that the number of neutrons escaping from the target per incident proton is maximum and the power density deposited on it is high (~1 MW/litter). The spallation target can be made of a solid or liquid material such as Pb, Hg, U, W, Ta, or Pb-Bi which are the most preferred materials for spallation targets. But, today Pb-Bi is the reference material for ADS applications. The spallation target has a large thermal load due to the conversion of about 10 MW proton into a neutron. Therefore, the target needs a cooling system; there are two options that can be considered i.e., cooling by means of the subcritical reactor primary system coolant and cooling by means of an independent loop.

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.

Long-Term Oxidation of Zirconium Alloy in Simulated Nuclear Reactor Primary Coolant—Experiments and Modeling

Oxidation of Zr-1%Nb fuel cladding alloy in simulated primary coolant of a pressurized water nuclear reactor is followed by in-situ electrochemical impedance spectroscopy. In-depth composition and thickness of the oxide are estimated by ex-situ analytical techniques. A kinetic model of the oxidation process featuring interfacial reactions of metal oxidation and water reduction, as well as electron and ion transport through the oxide governed by diffusion-migration, is parameterized by quantitative comparison to impedance data. The effects of compressive stress on diffusion and ionic space charge on migration of ionic point defects are introduced to rationalize the dependence of transport parameters on thickness (or oxidation time). The influence of ex-situ and in-situ hydrogen charging on kinetic and transport parameters is also studied.

파울링된 피복관 표면의 가압 경수로 운전환경에서의 미포화 비등 열전달 성능평가를 위한 실험장치 구축 연구

The CRUD (Chalk River Unidentified Deposit or Corrosion Related Unidentified Deposit) is a fouling deposit of metal or metal oxide on the nuclear fuel cladding under PWR (Pressurized Water Reactor) operating conditions and causing the several operational and safety-related problems. The major concerns related to the CRUD are AOA (Axial Offset Anomaly) caused by the accumulation of the boron species inside the porous CRUD from the primary reactor coolant and CILC (CRUD Induced Localized Corrosion) caused by locally increased cladding surface temperature due to the additional thermal resistance of the CRUD layer. The growth of CRUD and the CRUD-related problems are known to be governed by the sub-cooled nucleate boiling phenomena inside the upper part of the reactor core, especially on the hot fuel assembly and hot fuel pin. In the current study, the experimental facility called DISNY (crud DeposItion Simulator for Nuclear energY) was proposed as a testbed to simulate the CRUD growth and to investigate the sub-cooled nucleate boiling heat transfer performance of the CRUD deposited cladding surface under prototypical PWR operating conditions.

Optimization of dissolved hydrogen concentration for mitigating corrosive conditions of pressurised water reactor primary coolant under irradiation (2) evaluation of electrochemical corrosion potential

ABSTRACT One of the major subjects for evaluating the corrosive conditions in the PWR primary coolant was to determine the optimal hydrogen concentration for mitigating PWSCC without any adverse effects on major structural materials. As suitable procedures for evaluating the corrosive conditions in PWR primary coolant, a couple of procedures, i.e. water radiolysis and ECP analyses, were proposed. The previous article showed the radiolysis calculation in the PWR primary coolant, which was followed by an ECP study here. The ECP analysis based on a couple of a mixed potential model and an oxide layer growth model was developed originally for BWR conditions, which was extended to PWR conditions with adding Li+ (Na+) and H+ effects on the anodic polarization curves. As a result of comparison of the calculated results with INCA in-pile-loop experiment data as well as other experimental data, it was confirmed that the ECPs calculated with the coupled analyses agreed with the measured within ±100 mV discrepancies.

Thermal Fatigue of Pressurized Water Reactor Coolant System Loop Drain Lines Due to Outflow Activities

Abstract Thermal cycling induced fatigue is widely recognized as one of the major contributors to the damage of nuclear plant piping systems, especially at locations where turbulent mixing of flows with different temperature occurs. Thermal fatigue caused by swirl penetration interaction with normally stagnant water layers has been identified as a mechanism that can lead to cracking in dead-ended branch lines attached to pressurized water reactor (PWR) primary coolant system. Electric Power Research Institute (EPRI) has developed screening methods, derived from extensive testing and analysis, to determine which lines are potentially affected as well as evaluation methods to perform evaluations of this thermal fatigue mechanism for the U.S. PWR plants. However, recent industry operating experience (OE) indicates that the model used to predict thermal fatigue due to swirl penetration is not fully understood. There are limitations with the EPRI generic evaluation. In addition, cumulative effects from other thermal transients, especially those resulted from outflow activities, may also contribute to the failure of reactor coolant system (RCS) branch lines. In this paper, we report direct OE from one of our PWR units where thermal fatigue cracking is observed at the RCS loop drain line close to the welded region of the elbow. A conservative analytical approach that takes into account the influence of thermal stratification, in accordance with ASME section III class 1 piping stress method, is also proposed to evaluate the severity of fatigue damage to the RCS drain line, as a result of various transients, particularly the transients from outflow activities. Finally, recommendations are made for future operation and inspection based on results of the evaluation.

An innovation idea for SBLOCA mitigation strategy: Self-coagulation system in comparison with human hemostatic mechanism

The loss of coolant accident (LOCA), as one of the design basis accidents (DBAs), is a hypothetical accident that is usually considered in the design of nuclear power plant. LOCA is caused by small/large breaks in the reactor primary coolant and pressure boundary and may result in a loss of reactor coolant at a rate in excess of the reactor makeup system capability. Currently, the typical mitigation measures for LOCA are through the vessel high/low pressure coolant injection (HPCI/LPCI), core spray system, or reactor core isolation cooling (RCIC) system, to protect reactor from fuel melting and core degradation. However, is it possible to directly repair the small breaks passively? Inspired by the hemostasis and coagulation mechanism of human blood vessels, the authors are considering to develop a technology which would allow the reactor self-coagulation during the small break loss of coolant accident (SBLOCA). With the assistance of this innovative technology, the speed of losing coolant through reactor primary loop breaks could be significantly reduced or even the loss is stopped. With more and longer existing coolant in the core, the possibility of core degradation will also be significantly reduced. This innovative technology needs to be designed feasible and simple, which can be directly used into the existing nuclear power plants to improve the inherent safety of the cooling system, benefit the economy and safety of the nuclear power plant. In this paper, the authors compare the reactor primary loop of nuclear power plant with the blood circulation system of human beings from comprehensive and multi-angle perspectives, discuss the similarity, feasibility, and economy of nuclear power plant self-coagulation system based on the existing hemostasis and coagulation mechanism of human blood vessels. If possible, this technology will play a revolutionary positive role in alleviating DBAs related to SBLOCA.

CFD study on onset of liquid entrainment through ADS-4 branch line in AP1000

The consequences of liquid entrainment through the fourth stage Automatic Depressurization System (ADS-4) of Advanced Passive nuclear power reactor (AP1000) are decreased reactor core coolant inventory and increased resistance to quick depressurization of the reactor primary coolant system. Therefore, the accurate knowledge of the Onset of Liquid Entrainment (OLE) is critical to nuclear reactor safety (NRS) assessment. In this study, three-dimensional computational fluid dynamics (CFD) simulations of OLE were carried out using a two-fluid Eulerian model of ANSYS Fluent CFD code along with the renormalization group (RNG) κ-ε turbulence model for each phase. In OLE phenomena, the interfacial drag has a significant effect but available choices of interfacial drag coefficient in Fluent were not suitable due to the presence of droplet clusters in entrainment. Therefore, a modified interfacial drag coefficient accounting for droplet clusters was used. The transient simulations were performed with this modified CFD model and validated against ADS-4 Depressurization and Entrainment TEst Loop (ADETEL) data. The comparison between CFD calculations and experimental data shows a good agreement. Furthermore, the effects of the liquid mass flow rates and single inlet on OLE were also investigated with the validated CFD model. To improve the understanding of OLE phenomena, the liquid volume fraction and gas velocity field distributions were also studied.

Oxidation and carburization behaviour of two type 316H stainless steel casts in simulated AGR gas environment at 550 and 600 °C

• Duplex oxide and carburized layers form on steels exposed to AGR reactor coolant • Cracking in the oxide layer increases the oxidation and carburization rates • Carbon locates in the austenite lattice and predominately M 23 C 6 carbides • Diffusion of substitutional solutes is the rate-limiting mechanism of carburization • Small precipitates lead to a reduced carburization rate at 550ºC The time-dependent oxidation and carburization behaviour of two type 316H austenitic stainless steels with varying Mn content and differing average austenite grain size of 142 µm in ‘Low Mn’ (0.98 wt.% Mn) and 81 µm in ‘High Mn’ (1.52 wt.% Mn), were assessed at 550 °C and 600 °C in the presence of simulated reactor primary gas coolant containing 500 vppm H 2 O, 100 vppm H 2 , 300 vppm CH 4 and 1 vol.% CO, balanced with CO 2 . Rupture and spallation of the initial protective chromia layer occurs in the ‘High Mn’ steel after 2000 h at 550 °C, and leads to the formation of a magnetite/spinel oxide layer that reaches 75% surface coverage only after 8000 h. In contrast, ‘Low Mn’ steel reaches ~ 85% coverage after only 1000 h at the same temperature. The development of an inner carburized layer occurs gradually in both steels once the duplex oxide layer is forming. The differences in steel behaviour are reduced at 600 °C, where both oxidation and carburization are significantly accelerated. Only 0.02 wt.% carbon remains in solid solution in the austenite lattice in the carburized layer, the excess carbon atoms precipitating out in the form of Cr-rich M 23 C 6 particles, present both inter- and intragranularly. The experimental values of the activation energy for carburization suggest that diffusion of substitutional solutes as the rate-limiting mechanism of the process. Previously observed carburization depths of ≥ 200 µm in ex-service austenitic stainless steels, may be achieved by pre-conditioning 316H steel in simulated reactor gas coolant for ≥ 6000 h at 600 °C.

Influence of hydrazine decomposition kinetics on mass transfer of corrosion products in propulsion reactor primary coolant

Influence of hydrazine decomposition kinetics on mass transfer of corrosion products in propulsion reactor primary coolant

Prognostic systems representation in a function-based Bayesian model during engineering design

Prognostics and Health Management (PHM) systems are usually only considered and set up in the late stage of design or even during the system’s lifetime, after the major design decision have been made. However, considering the PHM system’s impact on the system failure probabilities can benefit the system design early on and subsequently reduce costs. The identification of failure paths in the early phases of engineering design can guide the designer toward a safer, more reliable and cost-efficient design. Several functional failure modeling methods have been developed recently. One of their advantages is to allow for risk assessment in the early stages of the design. Risk and reliability functional failure analysis methods currently developed do not explicitly model the PHM equipment used to identify and prevent potential system failures. This paper proposes a framework to optimize prognostic systems selection and positioning during the early stages of a complex system design. A Bayesian network, incorporating the PHM systems, is used to analyze the functional model and failure propagation. The algorithm developed within the proposed framework returns the optimized placement of PHM hardware in the complex system, allowing the designer to evaluate the need for system improvement. A design tool was developed to automatically apply the proposed method. A generic pressurized water nuclear reactor primary coolant loop system is used to present a case study illustrating the proposed framework. The results obtained for this particular case study demonstrate the promise of the method introduced in this paper. The case study notably exhibits how the proposed framework can be used to support engineering design teams in making better informed decisions early in the design phase.

Degradation of impact toughness in cast stainless steels during long-term thermal aging

Cast austenitic stainless steels (CASSs) have been extensively used for the large components of nuclear reactor primary coolant systems. Since the cast steels inevitably contain degradable metastable phases and replacement of the large coolant system components is impractical, the thermal embrittlement of CASS components has been a serious concern in the extended-term operation of nuclear power plants. This study aimed to systematically measure and analyze the effect of long-term thermal aging on the Charpy impact toughness to provide a comprehensive understanding of thermal degradation behavior and a practical aging model to predict the degree of thermal degradation in the cast stainless steels. The materials tested in the research include eight CASS alloys (two CF3s, one CF3M, three CF8s, and two CF8Ms) and two reference wrought materials (304L and 316L), in which the nominal δ-ferrite content ranges from ~2% to 33%. These stainless steels have been thermally aged at two light water reactor (LWR) temperatures (290 and 330 °C) and at two accelerated-aging temperatures (360 and 400 °C) for up to 30,000 h; these include both under-aged and over-aged conditions relative to the extended service lifetime (80 years). Charpy impact testing was performed for aged and non-aged specimens, and the impact (absorbed) energy parameters were correlated with a new aging parameter (A). Both the reduction of impact fracture toughness and the shift of ductile-brittle transition temperature were strongly dependent on the δ-ferrite content and degree of thermal aging. A linear relationship was found between the increasing rate of the index transition temperature T41J and aging parameter A; base on which an empirical model was proposed for prediction of the transition temperature as a simple function of the aging parameter (A) and δ-ferrite content (F). Finally, the critical aging parameter for embrittlement (AC) was evaluated and compared with the existing δ-ferrite content criteria.

Operator Action–Induced Two-Phase Flow Condition Resulting in Performance Degradation of Interfacing Passive System

This study investigates the degradation of the heat transfer performance of a closed-circuit intermediate natural circulation heat transport loop used as a passive safety system in a nuclear power plant (NPP). The degradation arises from the strong thermal-hydraulic (TH) coupling of the loop operating characteristics, saturation temperature and pressure, and natural circulation flow rate, which determine the heat rejection rate to the TH boundary conditions imposed on the hot side of the loop by the transitory state of the primary reactor coolant system (RCS) of the NPP. Several operator actions related to a feed-and-bleed emergency operating procedure (F&B) are postulated, and system TH code simulations are performed to demonstrate how the F&B can induce two-phase flow conditions in the RCS. Natural circulation two-phase flow regimes in the RCS hot leg can significantly reduce the heat transfer to the circulating working fluid of the interfacing heat transport loop over long periods, sometimes lasting over 24 h, of passive system mission time. A transient performance indicator for the passive system mission is introduced for use in the passive reliability assessment and quantitative comparison of transient simulations. The need to consider human factors in the design and operation of NPP passive safety systems is stressed.

Integral PWR-Type Small Modular Reactor Developmental Status, Design Characteristics and Passive Features: A Review

In recent years, the trend in small modular reactor (SMR) technology development has been towards the water-cooled integral pressurized water reactor (iPWR) type. The innovative and unique characteristics of iPWR-type SMRs provide an enhanced safety margin, and thus offer the potential to expand the use of safe, clean, and reliable nuclear energy to a broad range of energy applications. Currently in the world, there are about eleven (11) iPWR-type SMRs concepts and designs that are in various phases of development: under construction, licensed or in the licensing review process, the development phase, and conceptual design phase. Lack of national and/or internatonal comparative framework for safety in SMR design, as well as the proprietary nature of designs introduces non-uniformity and uncertainties in regulatory review. That said, the major primary reactor coolant system components, such as the steam generator (SG), pressurizer (PRZ), and control rod drive mechanism (CRDM) are integrated within the reactor pressure vessel (RPV) to inherently eliminate or minimize potential accident initiators, such as LB-loss of coolant accidents (LOCAs). This paper presents the design status, innovative features and characteristics of iPWR-type SMRs. We delineate the common technology trends, and highlight the key features of each design. These reactor concepts exploit natural physical laws such as gravity to achieve the safety functions with high level of margin and reliability. In fact, many SMR designs employ passive safety systems (PSS) to meet the evolving stringent regulatory requirements, and the extended consideration for severe accidents. A generic classification of PSS is provided. We constrain our discussion to the decay heat removal system, safety injection system, reactor depressurization system, and containment system. A review and comparative assessment of these passive features in each iPWR-type SMR design is considered, and we underline how it maybe more advantageous to employ passive systems in SMRs in contrast to conventional reactor designs.

Investigating on multi-RANS models coupling simulation for flow mixing experiments

Prediction and analysis of coolant flow and mixing problems in the reactor pressure vessel (RPV) are some of the most critical studies to ensure system safety and balance of nuclear power plant. The flow mixing problem in the reactor primary system with complicated interior structure is a large domain three-dimensional problem. In this paper, we investigated the performance of utilizing multiple RANS models (MRANS) coupling method in the computational fluid dynamics (CFD) code ANSYS FLUENT to study flow mixing. Based on the domain decomposition technique, the 2-equation and 7-equation RANS models were coupled and applied to simulate the flow mixing phenomenon in the Rossendorf Coolant Mixing Model (ROCOM) test facility. First, a CFD model of ROCOM was developed, and the explicit domain decomposition coupling simulation was achieved with a user-defined Python control program. To ensure the best CFD model was utilized, three different turbulence models and varying resolution grids were used for turbulence model feasibility analysis and grid sensitivity testing. Then the applicability of MRANS coupling method was evaluated by comparing the results from the MRANS simulations with those obtained from flow mixing experiments. In addition, single and MRANS model simulation results were compared. Results show good agreements between the MRANS coupling simulations and the experimental data. Compared with overall simulations by single RANS models, the MRANS coupling method also have superior performance in accuracy and efficiency for the CFD simulation study of mixing phenomena in the reactor primary coolant system.