Integral Pressurized Water Reactor Research Articles

Application of the passive turbocharger to PRHRS in integral PWR type SMR with an sCO2 power cycle

One of the potential design concepts for small modular reactor(SMR) is coupling an integral pressurized water reactor(IPWR) with a supercritical CO2(sCO2) power cycle to reduce the plant footprint. However, the passive safety systems of this SMR concept have not been well established compared to steam power cycle based SMRs. In this study, a passive turbocharger utilizing the unique characteristics of sCO2 is proposed as a way to improve the passive safety of sCO2 power cycle. This is the concept of installing a turbine-compressor set at the connection of passive residual heat removal system(PRHRS) to effectively remove the high level of residual heat in the early stages of an accident rather than relying only on natural circulation cooling. Based on the findings, it can be concluded that proposed sCO2 turbocharger concept has the potential to enhance performance of PRHRS and reduce the height of natural circulation loop leading to smaller volume.

Implementing Component Degradations into a Modelica Model of an iPWR System to Develop Health Monitoring Techniques

The economic operation of small modular reactors will partly rely on managing and reducing inspection and maintenance activities while supporting new operational paradigms like load-following. Turbine control valves throttle the steam from the steam generator into the steam turbine while maintaining the pressure within the steam generator at a constant set point. Degradation of these components could impact the ability to manage electrical power production. Utilizing the Idaho National Laboratory Hybrid repository and the Oak Ridge National Laboratory TRANSFORM library developed for multiphysics simulations in Dymola/Modelica, an integral pressurized water reactor system was modeled based on the available specifications of the NuScale power module. The effects of various component degradation modes have been implemented into the model in order to simulate faulted plant data during both steady-state and load-following operations. The fault modes resemble different physical fault modes that may occur at an operating nuclear power plant; a leaking turbine control valve and a valve actuator failure due to loss of hydraulic pressure have been implemented. A neural network autoencoder is employed in conjunction with statistical analysis, namely, simple signal thresholding (SST) or sequential probability ratio testing (SPRT), to identify the presence of a fault. Fuzzy logic is additionally employed in a novel and promising manner to classify the state of the system based on the cumulative sum of the neural network residuals. SST and SPRT are both successfully validated using healthy data and proved capable of identifying both fault types; fuzzy logic identified the false positives and classified the faulted data correctly.

Analysis of transient characteristics and design improvement of the passive residual heat removal system of NHR-200-II

NHR-200-II is a small integrated pressurized water reactor with 200 MW core thermal power. The core heat is transferred to two independent intermediate circuits via fourteen in-vessel primary heat exchangers (PHE), and the heat in the intermediate circuits is transferred to feedwater by two steam generators (SG) in the two intermediate circuits respectively. A passive residual heat removal (PRHR) branch is connected to each intermediate circuit to remove core decay heat under postulated accidents. During normal operation, PRHR branches are isolated by valves while SG branches in intermediate circuits are open. The valves in PRHR branches will be opened and the isolation valves of SG branches will be closed during decay heat removal scenarios. The decay heat removal capacity of NHR-200-II PRHRS could be seriously deteriorated once the isolation valves for SG branches fail to close, which was confirmed in a scaled integral test loop previously. Current understanding of PRHRS’s thermal-hydraulic characteristics with possible isolation failure in SG branches is limited. In this paper, the NHR-200-II PRHRS is modeled with RELAP5 considering the case of success and fail to isolate SG branches. A series of numerical simulations are carried out to study the impact of various parameters, such as the initial temperature, the size of the intermediate circuits’ header, and the initial flow direction in the intermediate circuits. Oscillatory flow is found when SG branches fail to be isolated under certain parameters combinations. An improved PRHRS design is purposed to eliminate possible flow oscillations, and the purposed improved design are tested by numerical simulations.

CFD Analysis of Coolant Flow Characteristics in Reactor Pressure Vessel: A Comprehensive Review

This comprehensive review explores the application of Computational Fluid Dynamics (CFD) in analyzing coolant flow within Reactor Pressure Vessels (RPVs) of nuclear power plants. By synthesizing existing research, methodologies, and advancements specific to RPVs, the paper offers in-depth insights into critical aspects such as boundary conditions, turbulence modeling, heat transfer mechanisms, and validation techniques. Examining a range of studies encompassing various reactor types from Pressurized Water Reactors (PWRs) to Integral Pressurized Water Reactors (IPWRs), the review underscores CFD's pivotal role in enhancing safety, efficiency, and performance optimization in nuclear reactors. Through systematic exploration, this study underscores the critical importance of precise modeling in facilitating safety assessments, operational optimization, and design enhancements across various reactor systems. Accurate modeling serves as a cornerstone for informed decision-making processes aimed at maximizing reactor performance while ensuring the highest standards of safety and reliability. The paper navigates through challenges such as computational limitations and turbulence modeling intricacies, while also discussing emerging trends like the porous media method aimed at improving computational efficiency. By offering a comprehensive understanding of thermal-hydraulic behavior in nuclear reactors, the review underscores CFD's contribution to enhancing safety and reliability in nuclear power generation. Overall, this review underscores the indispensable role of CFD in advancing our understanding of nuclear reactor dynamics, thereby contributing significantly to the overarching goals of improved safety and reliability in nuclear power generation.

Analysis of heat-loss mechanisms with various gases associated with the surface emissivity of a metal containment vessel in a water-cooled small modular reactor

In various small modular reactor (SMR) designs currently under development, the conventional concrete containment building has been replaced by a metal containment vessel (MCV). In these systems, the gap between the MCV and the reactor pressure vessel is filled with gas or vacuumed weakly, effectively suppressing conduction and convection heat transfer. However, thermal radiation remains the major mode of heat transfer during normal operation. The objective of this study was to investigate the heat-transfer mechanisms in integral pressurized water reactor (IPWR)-type SMRs under various gas-filled conditions using computational fluid dynamics. The use of thermal radiation shielding (TRS) with a much lower emissivity material than the MCV surface was also evaluated. The results showed that thermal radiation was always the dominant contributor to heat loss (48–97%), while the conjugated effects of the gas candidates on natural convection and thermal radiation varied depending on their thermal and radiative properties, including absorption coefficient. The TRS showed an excellent insulation performance, with a reduction in the total heat loss of 56–70% under the relatively low temperatures of the IPWR system, except for carbon dioxide (13%). Consequently, TRS can be utilized to enhance the thermal efficiency of SMR designs by suppressing the heat loss through the MCV.

NuScale-like SMR Model Development and Applied Safety Analyses with the Code Chain Serpent-DYN3D-ATHLET

NuScale is an integral pressurized water reactor (iPWR) operated with light water driven by natural circulation through two helical coil steam generators (HCSGs). This work reports the safety analyses of a boron dilution and steam line break accidental sequences in a developed plant computational model based on the specifications from the Final Safety Analysis Report (FSAR). Multi-physics calculations are performed with the code chain Serpent-DYN3D-ATHLET. A state-of-the-art multi-dimensional vessel topology is developed with ATHLET for the accurate representation of flow and temperature fields, as well as spatial core power distribution. The static calculation results show agreement with the reference values from the FSAR. The boron dilution sequence shows a homogeneous core power excursion by the boron feedback, and the reactor is tripped upon “high pressurizer pressure” signal. During the steam line break sequence, the affected HCSG depressurizes rapidly and the reactor is tripped upon “low main steam pressure” signal. None of the transients violate safety margins. The adopted 3-D vessel modelling approach and applied multi-physics calculations are able to capture both transient physics within the reactor domain, and conclude that symmetric arrangement of the HCSG tubes enhance coolant mixing and prevent a heterogeneous core power excursion.

Risk identification of integral pressurized water reactor (IPWR) cooling system using a combination HAZOP, FMEA, and FTA methods

AbstractIntegral Pressurized Water Reactor (IPWR) is gaining attention for use in Indonesia due to lower initial capital investment, sitting flexibility, and high level of inherent safety. IPWR relies heavily on passive cooling systems that have many advantages but lacks operational experience because no one operates commercially. Therefore, it is necessary to carry out a risk analysis related to passive cooling system failures that may occur during operation. Even if conventional risk analysis methods are applied singularly, the results still have shortcomings. This study aims to integrate the use of the Hazard and Operability Analysis (HAZOP), Fault Tree Analysis (FTA), and Failure Mode and Effect Analysis (FMEA) methods to identify risks in the IPWR primary cooling system during normal operation. HAZOP was used to identify hazards, FTA analyzed the root causes of the critical problems, and the basic event's FMEA was developed for risk ranking and mitigation measures. This combined methodology makes risk assessment more systematic and comprehensive because it manages entity‐wide risk. The results offer valuable insights for stakeholders involved in decision‐making regarding the development of nuclear power plants (NPPs) in Indonesia.

Numerical simulation and optimization of In-Vessel primary heat exchangers of NHR200-II

NHR200-II is a multipurpose small integral pressurized water reactor. The primary heat exchangers of NHR200-II are in-vessel single-phase heat exchangers. The primary heat exchangers adopted a tube-in-tube bundle design to improve compactness; however, a significant weakness of the tube-in-tube bundle is the nonuniform thermal expansion between the inner and outer tubes owing to the inhomogeneous tube-temperature distribution. To minimize the thermal stress, the inner and outer tube diameters were optimized using numerical simulations. The nonuniform mass flow rates and temperature fields in each channel and tube were calculated using the optimal tube sizes, and the axially averaged temperatures of the inner and outer tubes were also calculated. The present prediction of nonuniform temperature fields can be used as the input to estimate the magnitude of thermal stress in the assessment of the structural integrity of the primary heat exchangers of NHR200-II.

A non-invasive measurement approach of boric acid concentration for anticipating reactivity excursion in light water integrated small modular reactors

This article presents a viability study about a non-invasive and a higher accurate experimental approach to measure boric acid concentration in water under a vast combination of pressure and temperature conditions. This novel approach has been developed to be used in a test facility simulating integral Pressurized Water Reactor - Small Modular Reactors (iPWR-SMRs), that is under development and will be operated by the Brazilian Nuclear Energy National Commission. Besides, the passive acting of the main systems is one of the most important requirements that characterize Generations III + and IV nuclear reactors. The integration of such systems in a unique hermetic enclosure has brought new technological challenges to the nuclear engineering. Among such challenges is the need for adequate and efficient instrumentation for assessing operating conditions and phenomena present in iPWR-SMRs. The boric acid solution has been used in conventional PWRs as an important option to core reactivity control, demanding the need to characterize boric acid concentrations in reactor vessels any time, justifying the interest of this research. However, in an iPWR-SMR, water flow (with boron) between the pressurizer and the nuclear core occurs under passive convection regime and is invariably very slow and small. This situation may require a non-invasive measurement approach. The objective of this work is to analyze the viability of using ultrasonic technique to characterize the concentration of boric acid in a liquid medium non-invasively under experimental conditions of a specific test facility. Ultrasonic tests were performed relating different boric acid concentrations with the path time of the ultrasonic wave. Factors able to mask the characterization of these concentrations by ultrasonic technique were also evaluated. The results showed that this technique allows the characterization of boric acid concentrations in liquid medium under very simple conditions, requiring further studies for complex conditions like those found in nuclear reactors.

Proliferation resistance of light water reactor versus pebble bed reactor

Intrinsic proliferation resistance (PR) against the weaponization of U and Pu in used nuclear fuel discharged from a 500 MWt pebble bed reactor (PBR) and an integral pressurized water reactor (iPWR) was assessed. Four inherent characteristics (spontaneous fission neutron emission, heat load, radiation exposure, and Rossi-α) of U and Pu in the used fuel were the basis for the PR evaluation. These four characteristics for a significant quantity of U (75 kg of 235U) and Pu (8 kg) metallic sphere made from used fuel were quantified by performing reactor neutronics simulations. The intrinsic PR values for PBR used fuel were found to be lower compared to the iPWR for all the cases analyzed due to the higher quantity and better quality of nuclear material. Online refueling with an effective safeguards approach and a better choice of PBR fuel composition are needed to increase the PR of its used fuel.

Small modular reactors: An assessment of workforce requirements and operating costs

In the past seven years, global warming discourse has revolved around restricting temperature increases to under 2 °C above pre-industrial levels by utilizing such low-carbon energy sources as nuclear, solar, and wind energy. However, the use of nuclear energy for electricity generation via conventional nuclear power plants (NPPs) has met stiff challenges due to high capital costs, thereby constraining newcomer nations. A solution contending high capital costs for NPPs is the introduction of small modular reactors (SMRs) that has garnered attention over the past ten years because of their lower initial capital investment, scalability, and ease of deployment for less developed electricity grid and small areas of low electricity demand. In this study, the workforce requirements and salaries component for operating workers of 300 MWe integral Pressurized Water Reactors (iPWR) SMR is modeled and assessed using an upgraded International Atomic Energy Agency (IAEA) Nuclear Power Human Resources (NPHR) modeling tool from 2018 to 2055. The results show that at a discount rate of 10% an average annual salary payment, the net present cost (NPC) for the operating workers is approximately 431 million USD with a standard deviation of 8%. Moreover, the results further show that at the end of themodeling period and considering 43 function areas in a typical NPP, the top three functions to be recruited and contributing to the total domestic workers for a typical newcomer nuclear nation are Maintenance/Construction (18.17%), Operations (9.02%), and Management (8.22%) workers. The results enable the estimation of human resource requirements to operate an SMR over its life cycle, which in turn enables informed budgeting for human resources.

Computational model for thermohydraulic analysis of an integral pressurized water reactor with mixed oxide fuel (Th, Pu)O2

The use of advanced generation III+ and IV nuclear reactors, and their applications, has become important, seen as a means capable of contributing to the global transition to more sustainable, affordable and reliable energy systems. This technology, which could be integrated into future carbon-free electric power generation systems with high proportions of different renewable energy sources, includes Small Modular Reactors (SMR). There are about 100 different proposed projects of Generation III+ and IV, of which about 50 are SMR concepts, in various stages of development and of different types of technologies. Other important issues for achieving the long-term sustainability of nuclear energy are the proper use of its fuel sources and the improvement of nuclear waste management. Therefore, fuels based on a mixture of oxides have been used successfully in several countries. In addition, the incorporation of thorium-based fuel is a current challenge for the new designs of advanced reactors. The present paper focuses on the analysis of a small modular integral pressurized water reactor (iPWR) with Thorium-Uranium Oxide (Th-U MOX) mixtures. A thermohydraulic model is developed using the Ansys CFX program, which allows the calculation of the temperature distribution in the section where the highest power is produced within the SMR IPWR core (critical section). The temperature distributions in the fuel, clad and coolant were calculated with the objective of verifying that they were within the safety limits.

Aerosol retention within submerged Containment-Type iPWRs: Experiments and modeling

Integral pressurized water reactor (iPWR) designs employing submerged, high-pressure steel containment vessels rely on water in the outer operating pool as the primary heat sink during an accident. This drives very strong steam condensation, and along with it, strong diffusiophoretic deposition rates for fission product aerosols. In this study, a loss-of-containment isolation accident is proposed and simulated with MELCOR, where the strong condensation/diffusiophoresis is the last barrier preventing direct radionuclide releases to the environment. This is supported by experiments that measured the thermal stratification, steam condensation rates, and aerosol deposition rates that would be experienced in a prototypic, iPWR-like geometry. The development of the steam layer, whose thickness was proportional to steam injection rates, impacted the aerosol deposition by limiting the area over which condensation-driven diffusiophoresis could occur. Diffusiophoresis was still confirmed to be the primary mechanism for aerosol deposition, with observed rates that were in line with established theory.

Analysis of depressurization ability for IRIS containment during SBLOCAs using RELAP5 and CONTEMPT

As an integral and modular pressurized water reactor (PWR), the international reactor innovative and secure (IRIS) eliminates the threat of large break LOCAs (LBLOCAs) by integrating all major reactor coolant system components into the reactor vessel and provides compact steel containment with a suppression pool. Unlike large reactor containment, IRIS containment has modest volume and compact internal structure. The analysis of traditional dry containment cannot be applied to the IRIS reactor. Despite the disappearance of LBLOCAs (exceeding 4 in), small steel containment still faces the threat of high–frequency small break LOCAs (SBLOCAs) (e.g., the largest possible 4 in chemical and volume control system (CVCS) line break and the lowest elevation 2 in direct vessel injection (DVI) line break). To identify the impact of these characteristics on the integrity of IRIS reactor containment, a safety assessment framework that combines deterministic and probabilistic methods is proposed for the pressure response of containment under SBLOCAs of IRIS reactor. First in the modeling phase, primary and secondary loop models have been built considering the design of IRIS, and accident consequences under typical SBLOCAs were evaluated using RELAP5/MODE4. A simplified model of containment was presented herein based on the primary structure of IRIS containment. Second in the input parameters identification and uncertainty quantification stage, the responses of containment were simulated using CONTEMPT–LT. Thus, the main factors affecting the depressurization capacity of the containment were identified, and the reasonable probability density distributions were determined. Third, in the uncertainty propagation stage, Monte Carlo method was used to map the uncertainty of input parameters to the output variable. Fourth in the sensitivity analysis and uncertainty analysis phase, the contribution of each input parameter was measured using sensitivity analysis, and their interactions were investigated. The range of deviations of the containment pressure response was measured using uncertainty analysis, and finally, the calculated results were combined with the containment fragility curve to evaluate containment failure probability.

Radiological impact and emergency zones for small iPWR with different approaches for source term calculation

Assessment of the potential radiological impact and emergency zones for small Integral Pressurized Water Reactors (iPWR) is very important for their employment in the near term. In this work, an attempt is made to provide an analysis and estimation of the envelope of the potential impact from a severe accident in an iPWR. The source term is calculated from the solution of the lumped aerosol concentration equation in the containment, by using different approaches for estimating the various parameters. The first approach comprises a straightforward use of the methodology and parameters used for large reactors differing only in the lower power level of an iPWR. To account for the anticipated enhanced aerosol retention in the containment and the slower core damage progression in case of iPWRs, two additional approaches are then used. The values of the deposition velocities required are specified using relevant data from available detailed computational and experimental studies. Conservative, yet representative, accidental conditions in the containment are specified, based on insights offered by available analyses with severe accident codes available in the open literature and disclosed regulatory safety assessments. The radiological impact is assessed using a Gaussian plume model, whereas the exposure is assessed against the latest reference levels and dose criteria. Fetus exposure was found to determine the level of the radiological impact during the plume phase. The radioactive releases and the associated radiological impact may differ decisively depending on the approach used to estimate the source term. Consideration of the enhanced safety attributes of iPWRs (namely, enhanced radioactivity retention in the containment and slower accident progression) leads to a reduction of more than one order of magnitude, from about 75TBq to 4TBq and 6TBq to 0.3TBq, in terms of I131 and Cs137 releases to the atmosphere, respectively. Correspondingly, emergency distances are estimated to span from about 4 km - a level commensurate with the generic distances suggested in international standards for low power nuclear reactors - to a distance as low as 600 m if the favorable characteristics of iPWRs are considered, thus implying that no significant measures are required off-site.

Impact of core power density on economics of a small integral PWR

Increasing the core power density of nuclear reactors allows higher power generation from the same volume, reducing the cost of the plant. The lower power rating and higher degree of inherent safety for small modular reactors motivate exploring the highest achievable power density to improve their economics and overcome lack of economy of scale. Given the urgency to deploy carbon free energy sources in the near term, the current nuclear fuel supply must be adopted. Assuming standard 17x17 pressurized water reactor (PWR) fuel assemblies with enrichment below 5%, a parametric study on core power density of an integral PWR (iPWR) is performed. The reference core design is rated at 60 kW/L and composed of 77 reduced-height fuel assemblies. The reactivity control during normal operation is assumed to utilize soluble boron and fuel burnable poisons. The commercial STUDSVIK code package is used to design cores ranging from 20 kW/L to 170 kW/L. This range results in respective output powers between 50 and 420 MWe for an assumed 32% plant thermal efficiency. Core safety parameters, such as peaking factors, Minimum Departure from Nucleate Boiling Ratio and maximum fuel temperatures are evaluated for all designs. Simplified economic evaluation of lifetime-levelized unit costs for the fuel cycle, operation and maintenance and capital expenditures was performed as a function of power density. The relatively low core discharge burnup (32 MWD/kgU) leads to high lifetime-levelized fuel cycle unit cost for the reference core. However, increase in power density allows significant reduction in operation, maintenance and capital unit costs. Doubling power density resulted in 80% reduction in total levelized cost of electricity. The unique features of a forced circulation iPWR technology combined with compact steam generator technology could allow achieving high power densities (greater than 125 kW/L). Specific examples on relationship of cost and safety margins are also discussed.

Determination of Emergency Planning Zones distances and scaling-based comparison criteria for downsized Nuclear Power Plants

Nuclear Small Modular Reactors (SMR) present innovative features in design and safety compared to existing Nuclear Power Plants (NPP). Some of these features may turn into a reduction in radioactive releases –both in magnitude and frequency.It seems reasonable to think that smaller plants, i.e. featuring lower power levels, should present lower radioactive releases and offsite consequences. However, how small is small enough cannot be addressed through direct comparison. Furthermore, since multiple nuclear sites to compensate for the same amount of power provided by a single, large nuclear site, might be needed, such direct comparison is ill-conceived.Scaling-based criteria to facilitate comparison of radioactive releases between plants of different size –i.e. power– are proposed. Such criteria set limits to radioactive releases and offsite radioactive risk for SMRs so that offsite consequences of replacing large NPPs by a wider set of SMR sites are not exceeded when assessed globally, i.e. in terms of power generation.The first, deterministic-oriented criterion looks at the radioactive source strength of the SMR based on the maximum size of the SMR Emergency Planning Zone (EPZ) in order not to exceed the EPZ of another NPP when normalizing the EPZ to the power generating unit. The second, risk-informed criterion, looks at the sum of the radioactive harm caused by any event considered under probabilistic analysis application weighted by its frequency, i.e. standard definition of risk.Aside from these two scaling-based criteria for offsite nuclear hazard comparison between nuclear sites of different size, this paper presents two different methods for the determination of EPZ distances. The first method is straightforward based on plant-specific data for dose-consequence calculation, whereas the second, inverse method, builds on extrapolating the EPZ distance of a reference, large NPP down to the SMR.The scaling-based criteria and the methods for the determination of the EPZ distance are put into practice in an application exercise for an integral Pressurized Water Reactor (PWR) SMR.

Evaluating the Diffusion Approximation Capability on the Integral Pressurized Water Reactor (IPWR) Core Calculation

Diffusion approximation is an important approximation used to model a nuclear reactor core with a quite good accuracy and less computational cost. This approximation has been used widely around the globe for various kinds of nuclear reactors. This diffusion approximation is applied in a two-step method, a method combining a high fidelity transport code and low fidelity diffusion code. Meanwhile, innovations in the nuclear core model continue to make the nuclear reactor core safer, more robust, and smaller. The trend of creating smaller and more modular reactor core is emerging in the last ten years. These innovations will affect the core modeling system. Consequently, for smaller reactors, it is important to evaluate the capability of diffusion approximation if one wants to use a computationally cheaper method to model the reactor core. In this paper, neutron diffusion calculation for 160 MWth integral pressurized water reactor (IPWR) core was conducted using the PARCS nodal diffusion code employing the few-group spatially homogenized cross-sections generated by the Serpent Monte Carlo code. Due to its capability to model any reactor geometry in the high-resolution calculation, the results from Serpent were also used as a reference. Two important parameters are compared between PARCS and Serpent: effective neutron multiplication factor and core power distribution. For the full IPWR core model, a discrepancy of 564 pcm between PARCS and Serpent k eff was observed, while the radial power distribution had a maximum error of 4.71 %. It can be said, to some extent, that the diffusion approximation can be applied to IPWR core analysis. However, further improvement is indeed required if one wants more accurate results with low computational costs.

Mixed-oxide fuel strategies in an integral pressurized water reactor

The integral pressurized water reactor (iPWR) will play an important role in nuclear energy development considering its outstanding safety features. To achieve high proliferation resistance, extended fuel cycles have been proposed. Other important issues to achieve nuclear energy long-term sustainability are the proper use of its fuel sources and the improvement of nuclear waste management. Therefore, fuels based on oxides mixture have been successfully used in several countries. Also, Thorium-based fuel incorporation is a current challenge for the new advanced reactor designs. This paper compared different fuel compositions using recycled Plutonium and Uranium in a conceptual reactor of the type iPWR. The use of Thorium as fertile material is presented as well. The presented study considers a standard fuel cycle with slightly enriched Uranium (below 5%) as reference. One-third of the conceptual iPWR’s core is loaded with MOX fuel assemblies and the other two-thirds with conventional UO2. These proposed fuel cycles are compared with the standard full-UO2 cycle, regarding the behavior of main safety parameters, fuel depletion, and waste production. The studied fuel compositions reached a fuel cycle length of 4 years, approximately. The results for the core parameters show that using MOX and Thorium-based fuels in iPWR fulfills the safety constraints.

The simulation research on the natural circulation operation characteristic of FNPP in rolling and inclined condition

Abstract Recently, the FNPP (Floating Nuclear Power Plant) has got more and more attention and rapid development due to very wide prospect application on remote areas or islands. In general, the IPWR (Integral Pressurized Water Reactor) is adopted to meet the requirements of the limited space, the nuclear safety and the maneuverability in marine. The IPWR could depend on natural circulation operation to remove the residual heat of core under accident or low load operation condition. Because the driving head is low, the natural circulation flow is likely to be influenced by rolling and inclined condition. To clarify the natural circulation flow characteristics of the core in FNPP rolling motion and inclined condition, based on the modified THEATRe code by adding the ocean motion module and spatial coordinate convert module, the main thermal-hydraulic parameters variation in rolling and inclined condition were obtained. The effect of inclined angle, rolling amplitude and period on the natural circulation flow were discussed. The natural circulation flow in the core fluctuates periodically with rolling motion. And the inclination and rolling will also cause the degree of steam superheat of OTSG secondary side fluctuate, which could impact on the stable operation of secondary side system.