A Statistical Method for Benchmarking Nuclear Plant Models, Using ACAP

One of the requirements for validating nuclear reactor plant models is to benchmark the predicted results of selected transients against measured plant data or another qualified code. A major challenge with benchmarking is the criteria for validating a model are not always well defined and rely heavily on human judgment, thus introducing the possibility of human bias or inconsistent conclusions. The validation process can also be time consuming. A new method is presented to aid in the validation of nuclear reactor plant models, using the Automated Code Assessment Program (ACAP), which is a tool developed at Pennsylvania State University under contract by the U. S. Nuclear Regulatory Commission (NRC). The proposed method was developed specifically for real-time best-estimate nuclear operator training simulator transients. However, the tool can be easily customized for most applications (e.g., design models, steady state data). Four distinct statistical metrics and weightings were chosen, as deemed appropriate for transient nuclear operator training simulator applications. The metrics account for errors in magnitude and trend, and incorporate an experimental uncertainty. The four metrics are then combined into a single Figure of Merit (i.e., a statistical level of agreement between the predicted and benchmarking data sets). The use of ACAP for benchmarking is demonstrated by comparing experimental data from the Loss-of-Fluid-Test (LOFT) facility Large Break Loss-of-Coolant Experiment L2-5 with code predictions from a RELAP5-3D (Version 2.9.3+) model previously developed and published by Idaho National Laboratories. The proposed method is shown to have several advantages over conventional validation methods, in that it greatly reduces the possibility of human bias, generates reproducible results, can be easily automated to improve efficiency, and can be easily documented. After the initial validation, the tool can also be used to re-validate models after computer hardware changes, model changes, tool upgrades, and operating system upgrades.

Safety Enhancement Study of CAP1400 Spent Fuel Storage Racks

Safety enhancement is essential to spent fuel storage racks of nuclear power plant after Fukushima accident. Criticality safety, cooling safety and structural safety enhancement for CAP1400 (a larger advanced pressurized water reactor developed by China) spent fuel storage racks of are proposed and prominent results have been obtained. The criticality analysis results indicate that neutron absorber insert is efficient to increase criticality safety of storage rack. The CFD simulation and seismic analysis reveal that these storage racks can meet the cooling safety and structural integrity requirements. Both neutron absorber inserts and passive spent fuel pool cooling technology have good application prospect.

Investigation on Creep Mechanisms of Alloy 709

To improve efficiency, safety, and reliability of nuclear reactors, structural materials for Gen-IV reactors are being designed and developed. Alloy 709, a 20Cr-25Ni austenitic stainless steel, has superior mechanical properties to be a preferred candidate material for Sodium Fast Reactor structural application. Creep tensile tests were performed at temperatures of 700 °C, 725 °C and 750 °C and range of stresses from 100 MPa to 250 MPa. The apparent stress exponent and activation energy were found to be 10.3±0.4 and 368.6±14.7 kJ/mol. Linear extrapolation method was used to rationalize the higher stress exponent and activation energy relative to the mechanism in power law creep yielding to a true stress exponent of 7.1 ± 0.3 and a true activation energy of 277 ± 12.8 kJ/mol which is close to the lattice diffusion of iron in Fe-20Cr-25Ni. Hence, the lattice diffusion controlled dislocation climb process is believed to be the rate controlling creep deformation mechanism in this range of stresses and temperatures. The appropriate constitutive equation was developed based on the results; however, microstructural evaluations are under investigation to confirm the rate controlling mechanism. In addition, creep tests at higher temperatures and lower stresses are being conducted to extend the stress and strain-rate ranges to observe possible transition in creep mechanism.

ASME PTC 46 Performance Tests in Digital Plant Context

ASME PTC46 is nowadays a well known widely used Performance Test Code for Plant thermal performances. The Code — originally published in 1996, revised in 2015 — delivers all required guidance and methods for determination of the thermal performances in optimum precision conditions. Present paper considers going a step further in the performance test methodology in the modern context, ie specifically for Plants supported by Digital methods. Although a PTC46 test will deliver valuable information, this information — ie the test result — is valid solely at the point of time of the test on site, after which Plant Operator will not necessarily again have a very sharp vision of the level of Performances. Recent Digital tools and methods are a major change in this context and at present Operators of Digitalised Plants have access to a methodic systematic continuous measurement of the gaps between on-line performances and ideal “as should” status. Availability of such advanced functionalities is now opening the door to a radical change of the guarantees definitions, delivering to Plant developers better inputs without arbitrary assumptions as well as methodic and transparent measurement of the actual performance over significant periods of time, in opposition to the few hours of a PTC46 traditional test. Present paper addresses the possible definition of such new guarantees, as well as updates of the correction approach that are needed in order to make use of the Code principles in the modern Digital context. Pro and cons of both methods are discussed in detail , as well as analogy with other Power Plant areas.

Solution of Governing Equations for Six-Field System Code

Modeling of two phase flows in nuclear power plants is very important for design, licensing, and operator training and therefore must be performed accurately. As requirements have increased, the form and accuracy of the models and computer codes have improved along with them. Early formulations for the field equations include: single phase liquid with algebraic drift flux for the gas phase, modeled with mass, momentum and energy governing equations; and two separate fields, liquid phase and gas phase, typically modeled with six governing equations. These lump bubbles and droplets into the gas and liquid phases respectively and use flow regime maps based upon available experimental data. However, the experiments do not cover the entire spectrum of reactor conditions, so that transitions and extrapolations, which are inherently inaccurate, must be employed. Further, some reactor scenarios, such as boiling and condensation, can be more accurately resolved by modeling bubbles or droplets separately from the continuous fields. Introduction of an additional field, droplet or bubble, apart from the continuous liquid and gas fields, generally uses nine governing equations. Despite the successful development of the above-mentioned methods for modeling reactor coolant flow in modern software, such as RELAP, TRAC, TRACE, CATHARE and many others, there remain reactor scenarios that require greater resolution to model. This is particularly true of conditions during reflood, where emergency spray flows dominate the cooling profile within the core. Existing system codes use a lumped approach for two phase flows that groups the fields by their phase, thereby losing track of the physical interactions between the discrete fluid fields. The accuracy of these accident analysis system codes can be improved by characterizing the interactions between additional coolant fields. To capture the effect of the various field interactions, governing equations involving six-fields have been developed. The six fields are 1) continuous liquid, 2) continuous vapor, 3) large droplets, 4) small droplets, 5) large bubbles and 6) small bubbles. The additional fields and the related governing equations introduce additional variables and source terms that require new closure relationships and primary variables. This article presents the equations and variables and develops the discrete set of 18 equations that must be solved to model the system.

Experimental Study on Heat Transfer to Supercritical CO2 Flowing in Vertical Upward Tube at Medium Mass Flux

An experimental study was performed on heat transfer characteristics of supercritical pressure CO2 (SC-CO2) flowing at medium mass flux conditions in a vertically-upward tube of 16 mm inner diameter at the Heat Transfer and Flow test loop of Supercritical CO2 (HTF-SCO2) in Xi’an Jiaotong University. Experimental parameters included the pressure ranging from 7.5 to 10.5 MPa, the mass flux of 400–600 kg/m2s, and the heat flux of 20–100 kW/m2. Based on the experimental data, effects of mass flux, heat flux and operation pressure on heat transfer characteristics of SC-CO2 were thoroughly discussed. With the decrease of mass flux and increase of heat flux, heat transfer characteristics of SC-CO2 becomes worse and worse. The wall temperature rises to high levels with the occurrence of a wall temperature peak and the wall temperature peak also rises remarkably with the decrease in mass flux and increase in heat flux. Especially, effect of pressures on the heat transfer of SC-CO2 was found to be quite different from that previously reported in literature. When the heat flux is low (such as 30 kW/m2), the HTD was diminished with the increase in pressures, but when the heat flux is up to 50 kW/m2, the HTD is surprisingly intensified by the increase of pressure. The buoyancy effect was considered to explain this distinct influence of pressure on the heat transfer of SC-CO2 by employed a non-dimensional parameter Bu. With the increase of pressure, buoyancy effect was diminished owing to the decrease of density difference between fluids near the wall and the center. When heat flux was lower, the Bu was located between 5×10−6 and 10−4, where buoyancy effect impaired heat transfer, so the heat transfer coefficient increased by rising pressure. But when heat flux was larger, the Bu was above 10−4, where buoyancy effect began to enhance heat transfer, as a result, the heat transfer coefficient was reduced by weakened buoyancy effect because of the increase of pressure. (CSPE)

Improvement of the RELAP5-3D Model of Condensation in the Presence of Noncondensables

Condensation of steam on the primary side of a steam generator in a pressurized water reactor (PWR) is one means of removing decay heat during some accident scenarios, including small break loss of coolant accident (SBLOCA). However, when noncondensable gasses mix with steam, it impairs condensation. To correctly predict plant behavior, nuclear power plant (NPP) safety analysis codes such as RELAP5-3D must model the effect of condensation in the presence of noncondensables properly. A potential error in the RELAP5-3D code was reported in the condensation model in the presence of noncondensables by the University of Wisconsin[1]. The report indicated that the calculated condensation heat flux was under-predicted due to the modeling of the mass transfer in the gas mixture. The original documentation describing the implementation of the model was reviewed and compared with alternative formulations. An alternative that uses saturation vapor density at temperature of total pressure instead of the saturation vapor density at vapor partial pressure for calculating vapor mass flux was implemented. Comparison of the alternative method with the original against experimental data for several test cases showed improvement for most test cases.

Flow Testing and Analysis of the FSP-1 Experiment

The U.S. High Performance Research Reactor conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Size Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water channel velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A pressure differential versus flow rate curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported.

Investigation of Transient Flow and Heat Transfer for Passive Nuclear Reactor Direct Safety Injection

This paper describes a transient flow and heat transfer characteristics for a 1400MW passive pressurized-water reactor (PWR) direct vessel injection (DVI) system in different accident transient processes. The study components include reactor pressure vessel and a series reactor internal such as core barrel and radiation surveillance capsules, the flow channel include downcomer and lower plenum. Furthermore, the inject device is designed with special structures: first, a venturi type tube nozzle is connected to pressure vessel, second, a flow deflector is arranged in the downcomer which is facing the nozzle. This special structures will make the flow mixing and heat transfer very complicate and hard to predict. This study considers characteristics of the loops temperature and flow rate for both injection loop and reactor cold leg loop which are continuous change and long duration. Computational fluid dynamics (CFD) method is used in this study. Before this study, the physical model and numerical method are verified by an independent scaled model experiment. In this real reactor scale study, two typical accident transient processes are analyzed in this study, and temperature distribution on both reactor vessel and reactor internals are obtained. According to results analysis, the characteristics of heat distribution in downcomer were obtained: The injection fluid which is supposed to flow to core barrel is driven to the side of reactor vessel by the reflector. With the injection fluid flows in downcomer, the injection flow shape comes to a triangle. In addition, the transient results show that correlation degree of temperature distribution and injection flow character is gradually decreased with the increase of time history for passive injection. Overall, the exercise complements the activities in reactor safety analysis areas in understanding the origins of thermal load in reactor vessel, and being able to quantify them. Results of this study can be directly used by analyzing of reactor fatigue mechanics. (CSPE)