Analysis Of Pressurized Water Reactor Research Articles

CFD modelling of flashing flows for nuclear safety analysis: possibilities and challenges

Abstract Because of its relevance for the safety analysis of pressurized water reactors (PWR), many research activities on flashing flows in pipes and nozzles arose from the mid of last century. Most of them have been focused on the critical mass flow rate and transient pressure or temperature fluctuations by means of experiments and system codes. Since the beginning of this century, owing to the increase in computer speed and capacity, computational fluid dynamics (CFD) is being used more and more in the investigation of flashing flows, which has the advantage of providing three-dimensional insights in the internal flow structure as well as its evolution. This work presents an overview of relevant flashing scenarios in the nuclear safety analysis, and focuses on the discussion about possibilities and challenges of using CFD modelling. It is shown that a two-fluid model with the thermal phase-change model is superior to a mixture model with pressure phase-change, relaxation and equilibrium models, respectively, in terms of interfacial mass transfer, however, efforts are still required to improve the interphase heat-transfer model. Furthermore, since flashing is accompanied with high void fraction and broad bubble size ranges, a poly-disperse two-fluid model is recommended, but the effect of phase change on bubble coalescence and breakup needs further research. In addition, during flashing the flow pattern may change from single phase to bubbly flow, churn flow, annular flow, and even mist flow. The rapid change of interfacial topology as well as its influence on the applicability of closure models remains a challenge.

Safety analysis of pressurized water reactors with annular fuel during different transients

The dual-cooled annular fuel has two channels, with a shorter heat transfer path and a larger heat transfer area than solid fuel, which can effectively improve the safety of the reactor. It has excellent potential to improve the economy of the pressurized water reactor (PWR) and is considered to have a good development prospect. In this paper, the safety of the reactor with annular fuel assemblies under different transient conditions is analyzed using the system code NUSOL-SYS and the subchannel code NACAF, with the CPR1000 as the reference power plant. The analysis mainly includes the large break loss of coolant accident (LBLOCA), main steam line break accident (MSLB), loss of flow accident (LOFA), and rod ejection accident (REA) under nominal power and higher power density. The flow blockage accident and the combined accidents of LBLOCA and the flow blockage accident are also evaluated. The results show that during LBLOCA, the peak cladding temperature (PCT) of the core is only 833 K, much lower than that of the solid fuel. In addition, the departure from nucleate boiling ratio (DNBR) of annular fuel is larger during MSLB and LOFA, which means a more negligible risk of fuel damage. In REA, the average temperature of annular fuel is below 900 K, and the average enthalpy is below 200 kJ/kg, which is much lower than that of solid fuel. The results at 150 % power show that the performance of the core with annular fuel is still better than that of solid fuel at nominal power, except that the PCT during LBLOCA is slightly higher than solid fuel at nominal power. During the flow blockage accident, the results with different blockage areas show that a small proportion of blockage of the inner channel does not have a significant effect on the cladding temperature and that there is a risk of rupture when the inner channel is completely blocked. Finally, the results of the assumed combined accident show that the blockage of the inner channel before scram will significantly increase the PCT of the core and the time to complete the reflood during LBLOCA.

Development of supporting platform for the fine flow characteristics of reactor core

This study presents the Supporting platform for reactor fine flow characteristics calculation and analysis (Cilian platform), a user-friendly tool that supports the analysis and optimization of pressurized water reactor (PWR) cores with mixing vanes using computational fluid dynamics (CFD) computing. The Cilian platform allows for easy creation and optimization of PWR's main CFD calculation schemes and autonomously manages CFD calculation and analysis of PWR cores, reducing the need for human and computational resources. The platform's key features enable efficient simulation, rapid solution design, automatic calculation of core scheme options, and streamlined data extraction and processing techniques. The Cilian platform's capability to call external CFD software reduces the development time and cost while improving the accuracy and reliability of the results. In conclusion, the Cilian platform exemplifies an innovative solution for efficient computational fluid dynamics analysis of pressurized water reactor (PWR) cores. It holds great promise for driving advancements in nuclear power technology, enhancing the safety, efficiency, and cost-effectiveness of nuclear reactors. The platform adopts a modular design methodology, enabling the swift and accurate computation and analysis of diverse flow regions within core components. This design approach facilitates the seamless integration of multiple computational modules across various reactor types, providing a high degree of flexibility and reusability.

Improvement of the subcooled boiling model using a new net vapor generation correlation inferred from artificial neural networks to predict the void fraction profiles in the vertical channel

In the one-dimensional thermal-hydraulic (TH) codes, a subcooled boiling model to predict the void fraction profiles in a vertical channel consists of wall heat flux partitioning, the vapor condensation rate, the bubbly-to-slug flow transition criterion, and drift-flux models. Model performance has been investigated in detail, and necessary refinements have been incorporated into the Safety and Performance Analysis Code (SPACE) developed by the Korean nuclear industry for the safety analysis of pressurized water reactors (PWRs). The necessary refinements to models related to pumping factor, net vapor generation (NVG), vapor condensation, and drift-flux velocity were investigated in this study. In particular, a new NVG empirical correlation was also developed using artificial neural network (ANN) techniques. Simulations of a series of subcooled flow boiling experiments at pressures ranging from 1 to 149.9 bar were performed with the refined SPACE code, and reasonable agreement with the experimental data for the void fraction in the vertical channel was obtained. From the root-mean-square (RMS) error analysis for the predicted void fraction in the subcooled boiling region, the results with the refined SPACE code produce the best predictions for the entire pressure range compared to those using the original SPACE and RELAP5 codes.

Development of a Critical Heat Flux Correlation based on a Mechanistic Model under Subcooled Flow Boiling Conditions

In this study, a mechanistic model-based correlation for the critical heat flux is proposed to meet the practical requirements for the design and safety analysis of pressurized water reactors. The correlation is based on a liquid sublayer dryout model for a single tube, which includes physical parameters, such as sublayer thickness and vapor blanket velocity, expressed as an explicit nondimensional equation and modified using a look-up table for critical heat flux to ensure a wide applicable range in coolant conditions. The correlation predicts critical heat flux for the subcooled region under the pressure range of 0.1–18 MPa, mass velocity range of 500–8000 kg/m2s, and thermal equilibrium quality ranging from −0.5 to −0.05 with high accuracy of the departure from nucleate boiling ratio limit based on a 95% probability and a 95% confidence of 1.20.

Validation of the thermal-hydraulic thunder code and application for PWR core in transient conditions

The THUNDER computer code allows performing thermal-hydraulic core analysis of pressurized water reactors aiming at the design and safety analysis of power and naval reactors in steady state and transient conditions. It uses the subchannel and control volume methods to perform a tridimensional analysis of the nuclear reactor core. The conservation equations are solved simultaneously for each control volume based on geometric and operation conditions. The Integral Momentum Method is used for solving the conservation equations decoupling the momentum equation from the mass and energy conservation equations. The drift-flux model is adopted to treat the two-phase flow condition. The analysis is performed in two-stages: first, a macroscopic analysis is conducted with subchannel equivalent to a fuel assembly (bundle), and second, with a subchannel equivalent to an individual coolant channel. Only the hottest fuel assembly is analyzed with detail. To validate the THUNDER code we used the experimental data from the NUPEC PWR benchmark (subchannel and bundle tests), in which we verify the accuracy of the coolant temperature at steady-state in 4 different void fraction conditions and the accuracy of the void fraction at transients initiated by four different conditions: core power increase, reduction of inlet mass flow rate, increase of inlet coolant temperature and depressurization of the inlet coolant. At steady-state conditions, the mean relative discrepancy for the coolant temperature was 0.09. At transient conditions, the void fraction mean absolute discrepancy for power increase was 0.045, for mass flow reduction was 0.079, for inlet coolant depressurization was 0.032 and for temperature increase was 0.092. The code was applied for the Brazilian 1350 MW nuclear power plant ANGRA 2 for evaluating the thermal-hydraulic safety parameters at steady-state, as well as the operational transients of loss of one or several primary coolant pumps. The THUNDER code results showed that the safety limits were not exceeded in concordance to the Final Safety Analysis Report of the plant. These results indicate that the THUNDER code is sufficiently accurate to perform thermal-hydraulic analyses for PWR power plants at steady-state and transient conditions.

A new pin-resolved ultra-fine-group method based on global–local resonance treatment framework

In order to accurately consider intra-pin effects such as the spatial self-shielding effect and non-uniform temperature effect, several pin-resolved resonance self-shielding methods have been developed recently. The basic strategy of these pin-resolved resonance self-shielding methods is to divide resonance calculation into the whole domain calculation (global) and the 1-D fuel pin calculation (local), which we call the global–local resonance treatment framework. The spatial self-shielding effect, non-uniform temperature effect and resonance interference can be considered accurately by the 1-D fuel pin calculation with the ultra-fine-group method. Due to the simplified geometry, the computation time is short. In addition, the 1-D fuel pin calculation can be performed in parallel. This paper reviews pin-resolved resonance self-shielding methods that have been proposed and proposes a new pin-resolved ultra-fine-group method. These methods are implemented in the numerical nuclear reactor neutronics code SHARK developed by Nuclear Power Institute of China. Numerical results show good accuracy in the pressurized water reactor analysis.

An Artificial-Intelligence and Machine-Learning-Based Methodology to Conduct Seemingly Strain-Controlled Fatigue Test in a Pressurized-Water-Reactor-Test-Loop-Autoclave, While Not Controlling the Strain

Abstract In general, low cycle fatigue analysis of pressurized water reactor (PWR) components, requires strain-controlled fatigue test data such as using strain versus life (ε–N) curves. Conducting strain-controlled fatigue tests under in-air conditions is not an issue. However, controlling strain in a PWR-test-loop-autoclave is a challenge, since an extensometer cannot be placed in a narrow autoclave (typically used in a high-temperature-pressure PWR-test-loop). This is due to lack of space inside an autoclave that houses the test specimen. In addition, installing a contact-type extensometer in the path of a high-pressure flow can be a challenge. These difficulties of using an extensometer inside an autoclave led us to use an outside-autoclave displacement sensor which measures the displacement of pull-rod-specimen assembly. However, in our study (based on in-air fatigue test data), we found that a pull-rod-controlled based fatigue test can lead to substantial cyclic hardening/softening resulting in substantially different cyclic strain amplitudes and their rates compared to the desired cyclic strain amplitudes and its rates. In this paper, we propose an Artificial-Intelligence and Machine-Learning based technique such as using k-means clustering technique to improve the pull-rod-control based fatigue test method, such that the gage-area strain amplitude and rates can reasonably be achieved. In support of this, we present the fatigue test results for both 316 SS base and 81/182 dissimilar-metal-weld specimens.

Design and analysis of RIF scheme to improve the CFD efficiency of rod-type PWR core

This research serves to advance the development of engineering computational fluid dynamics (CFD) computing efficiency for the analysis of pressurized water reactor (PWR) core using rod-type fuel assemblies with mixing vanes (one kind of typical PWR core). In this research, a CFD scheme based on the reconstruction of the initial fine flow field (RIF CFD scheme) is proposed and analyzed. The RIF scheme is based on the quantitative regulation of flow velocities in the rod-type PWR core and the principle that the CFD computing efficiency can be improved greatly by a perfect initialization. In this paper, it is discovered that the RIF scheme can significantly improve the computing efficiency of the CFD computation for the rod-type PWR core. Furthermore, the RIF scheme also can reduce the computing resources needed for effective data storage of the large fluid domain in a rod-type PWR core. Moreover, a flow-ranking RIF CFD scheme is also designed based on the ranking of the flow rate, which enhances the utilization of the flow field with a closed flow rate to reconstruct the fine flow field. The flow-ranking RIF CFD scheme also proved to be very effective in improving the CFD efficiency for the rod-type PWR core.

Korean Development of Advanced Thermal-Hydraulic Codes for Water Reactors and HTGRs: Space and Gamma

The Korea nuclear industry has been developing the thermal-hydraulic system analysis Safety and Performance Analysis CodE (SPACE) and the GAs Multicomponent Mixture Analysis (GAMMA) code for safety analysis of pressurized water reactors (PWRs) and high-temperature gas-cooled reactors (HTGRs), respectively. SPACE will replace outdated vendor-supplied codes and will be used for the safety analysis of operating PWRs and for the design of an advanced PWR. SPACE consists of up-to-date physical models of two-phase flow dealing with multidimensional two-fluid, three-field flow. GAMMA consists of multidimensional governing equations consisting of the basic equations for continuity, momentum conservation, energy conservation of the gas mixture, and mass conservation of n species. GAMMA is based on a porous media model so that thermofluid and chemical reaction behaviors in a multicomponent mixture system and heat transfer within solid components, free and forced convection between a solid and a fluid, and radiative heat transfer between solid surfaces can be dealt with. GAMMA has a two-dimensional helium turbine model based on the throughflow calculation and a coupled neutronics-thermal-hydraulic model. Extensive code assessment has been performed for the verification and validation of SPACE and GAMMA.

DEVELOPMENT OF THE SPACE CODE FOR NUCLEAR POWER PLANTS

The Korean nuclear industry is developing a thermal-hydraulic analysis code for safety analysis of pressurized water reactors (PWRs). The new code is called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE). The SPACE code adopts advanced physical modeling of two-phase flows, mainly two-fluid three-field models which comprise gas, continuous liquid, and droplet fields and has the capability to simulate 3D effects by the use of structured and/or nonstructured meshes. The programming language for the SPACE code is C++ for object-oriented code architecture. The SPACE code will replace outdated vendor supplied codes and will be used for the safety analysis of operating PWRs and the design of advanced reactors. This paper describes the overall features of the SPACE code and shows the code assessment results for several conceptual and separate effect test problems.

Thermal-hydraulic modeling and transient analysis of pressurized water reactors

This paper examines a new dynamic moving boundary thermal-hydraulic fuel pin model (FUELPIN) for the transient analysis of a pressurized water reactor (PWR). FUELPIN is developed to accommodate the reactor core thermal-hydraulic model of the fuel pin and adjacent coolant flow channel, with detailed thermal conduction in fuel elements. Transient analyses using a known thermal-hydraulic analysis code, COBRA, and FUELPIN linked with a PWR system analysis code show that the thermal margin gains more by a transient MDNBR approach than the traditional quasi-steady methodology for a PWR. The studies of the nuclear reactor system show that moving boundary formulation is highly suitable for the transient thermal-hydraulic analysis of PWRs.

Two-Group, Flux-Coupled Xenon Oscillation Model with an Equation System of Axial Difference Parameters

An axial xenon oscillation model is developed for pressurized water reactor analysis. The model employs an equation system for axial difference parameters that is derived from xenon and iodine balance equations coupled with two-group, one-dimensional neutron diffusion equations. To treat nonlinear xenon-flux-coupled terms, the spatial distributions of xenon, iodine, and flux are expanded by the Fourier sine series. The equation with respect to the axial difference parameters can be analytically solved with the initial condition related to axial power difference, which can be measured in the reactor. The axial power difference variation during xenon oscillation is directly obtained, and it provides a prediction of xenon oscillation behavior. The accuracy of the model is verified by benchmark calculations with a three-dimensional reference core calculation code and measured data from a core startup test at Yonggwang Unit 3.

Validation of SCALE-4 for Burnup Credit Applications

In the past, criticality analysis of pressurized water reactor (PWR) fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. If credit is allowed for fuel burnup in the design of casks that are used in the transport of spent light water reactor fuel to a repository, the increase in payload can lead to a significant reduction in the cost of transport and a potential reduction in the risk to the public. A portion of the work has been performed at Oak Ridge National Laboratory (ORNL) in support of the U.S. Department of Energy (DOE) efforts to demonstrate a validation approach of criticality safety methods to be used in burnup credit cask design. The date, the SCALE code system developed at ORNL has been the primary computational tool used by DOE to investigate technical issues related to burnup credit. The SCALE code package is a well-established code system that has been widely used in away from reactor applications. Criticality safety analyses are performed via the criticality safety analysis sequences (CSAS) and spent-fuel characterization via the shielding analysis sequence (QSAS) and spent-fuel characterization via the shielding analysis sequence (SAS2H). The SCALE 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data has been used for all calculations. The American National Standards Institute/American Nuclear Society (ANSI/ANS)-8.1 criticality safety standard requires validation and benchmarking of the calculational methods used in evaluating criticality safety limits for applications outside reactors of correlation against critical experiments that are applicable. Numerous critical experiments for fresh PWR-type fuel in storage and transport configurations exist and can be used as part of a validation database. However, there are no critical experiments with burned PWR-type fuel in storage and transport configurations. As an alternative, commercial reactors offer an excellent source of measured critical configurations. Part of the work that has been performed to date to validate the SCALE-4 code system for burnup credit applications using measured critical configurations includes: 1. fresh fuel critical experiments having geometric and nuclear characteristics similar to PWR spent fuel in storage and transport configurations 2. commercial PWR hot-zero-power and hot-full-power reactor critical configurations. The ability to closely predict reactor critical conditions is important in the validation of a methodology for spent-fuel applications because input data are determined based on relatively little detail of reactor core operation. Such limited information is expected to be representative of data available when burnup credit calculations are being performed in the determination of optimum cask loadings. The results reported demonstrate the ability of the ORNL SCALE-4 methodology to predict a value of k eff very close to the known value of 1.0, both for fresh fuel criticals and for the more complex reactor criticals. Beyond these results, additional work in the determination of biases and uncertainties is necessary prior to use in burnup credit applications

Assessment of the Safety Function for the Anticipated Transient without Trip Mitigation System Actuation Circuitry at Maanshan Nuclear Power Station

Due to the potential threat of reactor coolant system (RCS) overpressurization, loss-of-normal-feed-water (LONF) transients without reactor trip have received special attention in the analysis of pressurized water reactor (PWR) anticipated transients without trip (ATWT). The U.S. Nuclear Regulatory Commission requires every PWR to be equipped with an ATWT mitigation system actuation circuitry (AMSAC) so that the turbine will be tripped and auxiliary feedwater (AFW) added when an LONF transient occurs. An AMSAC design will be installed in both units of the Maanshan Nuclear Power Station (MNPS) to deal with ATWTs under LONF transient conditions. A best-estimate transient analysis performed with the RETRAN-02/MOD3 code is used to assess the safety function of the actuation circuitry designed for MNPS. Analytical results show that the peak RCS pressure will not exceed the 22.16-MPa safety limit if the moderator temperature coefficient is sufficiently negative and the actuation circuitry functions normally. Effects of the moderator temperature coefficient, the Doppler coefficient, pressurizer power-operated relief valves, effective time of the AFW system, the steam dump system, and the automatic control rod system are discussed.

ATHLET Analysis of the Three Mile Island Unit 2 Accident

The thermal-hydraulic code A THLET used to analyze the Three Mile Island Unit 2 accident was developed at the Gesellschaft für Reaktorsicherheit in the Federal Republic of Germany for safety analysis of pressurized water reactors with U-tube steam generators.First calculations of phase 1 of the accident without simulating the secondary sides were preliminary and have shown the need for accurate heat flow rates in the steam generators. The available values resulted in an incorrect system pressure behavior.Second calculations with detailed simulation of the once-through steam generator provided interesting information like the influence of flow resistance in the aspirator on cold-leg temperature and the dependence of system behavior on the rate and position of auxiliary feedwater injection.

Analysis results from the Los Alamos 2D/3D program

Los Alamos National Laboratory is a participant in the 2D/3D program. Activities conducted at Los Alamos National Laboratory in support of 2D/3D program goals include analysis support of facility design, construction, and operation; provision of boundary and initial conditions for test facility operations based on analysis of pressurized water reactors; performance of pretest and posttest predictions and analyses; and use of experimental results to validate and assess the single- and multidimensional nonequilibrium features in the Transient Reactor Analysis Code (TRAC). During Fiscal Year 1986, Los Alamos conducted analytical assessment activities using data from the Cylindrical Core Test Facility and the Slab Core Test Facility. Los Alamos also continued to provide support analysis for the planning of Upper Plenum Test Facility experiments. Finally, Los Alamos either completed or is currently working on three areas of TRAC modeling improvement. In this paper, Los Alamos activities during Fiscal Year 1986 are summarized; several significant accomplishments are described in more detail to illustrate the work activities at Los Alamos.

Moderator Feedback Effects in Two-Dimensional Nodal Methods for Pressurized Water Reactor Analysis

A method was developed for incorporating moderator feedback effects in two-dimensional nodal codes used for pressurized water reactor (PWR) neutronic analysis. Equations for the assembly average quality and density are developed in terms of the assembly power calculated in two dimensions. The method is validated with a Westinghouse PWR using the Electric Power Research Institute code SIMULATE-E. Results show a several percent improvement is achieved in the two-dimensional power distribution prediction compared to methods without moderator feedback.