Fatih Aydogan

Development of Governing Equations Based on Six Fields for the RELAP Code

Abstract The RELAP5-3D code is used to analyze nuclear reactor systems during steady-state and transient operations. Reactor transients that result in significant two-phase flow conditions and phase change, such as reflood scenarios, loss-of-coolant accidents, and others, can tax the current capabilities of the code to model the flow fields. Current codes, such as RELAP5-3D, RELAP-7, and TRACE, have mass, momentum, and energy governing equations for only two fields (liquid and vapor). The representation of two-phase flow phenomena is improved by increasing the number of fields. Therefore, governing equations based on six fields (liquid, vapor, small bubble, large bubble, small droplet, and large droplet) are derived in this paper for implementation in RELAP5-3D.

Advanced small modular reactors

Abstract Small modular reactors (SMRs) are different than other reactors based on several reasons: (1) reduced power level; (2) reduced physical size and spatial foot prints; (3) increased modularity of reactor architecture; (4) increased safety margin; (5) increased security feature; (6) reduced financial risk of a reactor unit; and (7) increased flexibility of using the reactor unit for various energy needs. Among the new SMRs, US Department of Energy (DOE) has begun to support SMR activities in the US in the recent years by issuing solicitations, such as “Financial Assistance Funding Opportunity Announcement – Cost-Shared Development of Innovative Small Modular Reactor Designs.” US-DOE supports SMRs because of all these benefits. This chapter compares the advanced SMRs in different technical aspects.

Development of Conservative Form of RELAP5 Thermal Hydraulic Equations: Part I — Theory

The design and analysis of the thermal/hydraulic systems of nuclear power plants necessitates system codes that can be used in the analysis of steady state and transient conditions. RELAP5 is one of the most commonly used system codes in nuclear organizations. RELAP5 is based on a two-fluid, non-equilibrium, non-homogeneous, hydrodynamic model for the transient simulation of the two-phase system behavior. This model includes six governing equations to describe the mass, energy, and momentum of the two fluids. The “non-conservative” numerical approximation form (which is the current form of RELAP5 code versions) is obtained through the manipulation of selected derivative terms in the equations including the linearization of the product terms in the time derivatives of the equations. For non-conservative technique, the truncation errors introduced in the linearization process can produce mass and energy errors for some classes of transients during time advancements, either resulting in (a) automatic reduction of time steps used in the advancement of the equations and increased run times or (b) the growth of unacceptably large errors in the transient results. To eliminate these difficulties, a new, optional numerical approach has been introduced in RELAP/SCDAPSIM/MOD4.0. This new option uses a more consistent set of the “conservative” numerical approximation to solve non-linearized mass and energy governing equations. The RELAP/SCDAPSIM/MOD4.0 code, being developed as part of the international SCDAP (Severe Core Damage Analysis Package) Development and Training Program (SDTP), is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards. This paper provides an overview of the original RELAP5 numerical approximations and describes the new theoretical approach. Then the second article introduces the solution strategy of conservative approach and presents some examples of transient problems that have been run using this new approach.Copyright

Quantitative and Qualitative Comparison of Light Water and Advanced Small Modular Reactors (SMRs)

In the recent years, several Small Modular Reactors (SMRs) have been developed. These Nuclear Power Plants (NPPs) not only offer small power size (less than 300MWe), foot print, compact designs fabricated in factories to transport to the sites but also passive safety features. On one hand, some of the Light Water (LW) SMRs have been aggressively competing to win Department of Energy’s Funding Opportunity Announcements (FOA): NuScale, W-SMR, etc. These new LW-SMRs are mainly inspired by the early LW-SMRs (such as, Process Inherent Ultimate Safety (PIUS), International Reactor Innovative and Secure (IRIS), Small Innovative Reactor (SIR), etc). LW-SMRs employ significantly less number of components to decrease cost and increase simplicity. However, new physical challenges appeared with these changes. On the other hand, advanced SMRs (such as, PBMR, MHR Antares, Prism, 4S, Hyperion, etc.) are dazzled with their improved passive safety features. This paper compares most of the LW and Advanced SMRs in respect to reactors, nuclear fuel, containment, reactor coolant systems, re-fueling and emergency coolant systems quantitatively and qualitatively. The detailed comparisons in this paper show that one reactor is not the absolute winner in this comparison since each reactor is designed to meet different needs.Copyright

Comprehensive Analyses of Nuclear Safety System Codes

Many nuclear system codes have been developed for the main purpose of analyzing reactor performance of a nuclear power plant system during steady state and transient conditions. These codes generally include power plant component models for pumps, pipes, steam generators, pressurizers and other components. The parallel development of these nuclear system codes has been supported by government laboratories, universities, private entities and other organizations throughout the world. This has resulted not only in multiple codes, but multiple versions of the same code with different capabilities. The development paths of each code version have been driven by specific needs. The challenge for the user is to select a code that performs well for the desired analysis problem. Therefore, this work compares different aspects of various nuclear system codes. Firstly, it compares the governing equations for mass, momentum and energy in the evaluated system codes. Secondly, it compares all the codes’ closure models. Closure models are used in system codes to model thermal and mechanical non-equilibrium as well as the coupling of the phases. Thirdly, it compares the Separate Effect Tests (SET) and Integral Effect Tests (IET) employed for the verification and validation (V&V) during the development of each system code. These comparisons cover several thermal and hydraulic models, such as heat transfer coefficients for various flow regimes, two phase pressure correlations, two phase friction correlations, drag coefficients and interfacial models between the fields. Fourthly, major assumptions about the governing and closure equations in these codes are compared and discussed. Fifthly, numerical approach of every code is benchmarked with each other since numerical approach not only affects the speed of the system codes but also the accuracy of the results. Sixthly, the limitations of the codes are evaluated because these codes are challenged by analyzing not only existing nuclear power plants, but also next generation nuclear power plants. The nuclear industry is developing new, innovative reactor designs, such as Small Modular Reactors (SMRs), High-Temperature Gas-cooled Reactors (HTGRs) and others. Sub-types of these reactor designs utilize pebbles, prismatic graphite moderators, helical steam generators, innovative fuel types, and many other design features that may not be fully analyzed by current system codes. The results of this work serve as a guide for development of these system codes and indicate areas where models must be improved to adequately address issues with new reactor design and development activities.Copyright

Modeling Loss-of-Flow Accidents and Their Impact on Radiation Heat Transfer

Long-term high payload missions necessitate the need for nuclear space propulsion. The National Aeronautics and Space Administration (NASA) investigated several reactor designs from 1959 to 1973 in order to develop the Nuclear Engine for Rocket Vehicle Application (NERVA). Study of planned/unplanned transients on nuclear thermal rockets is important due to the need for long-term missions. In this work, a system model based on RELAP5 is developed to simulate loss-of-flow accidents on the Pewee I test reactor. This paper investigates the radiation heat transfer between the fuel elements and the structures around it. In addition, the impact on the core fuel element temperature and average core pressure was also investigated. The following expected results were achieved: (i) greater than normal fuel element temperatures, (ii) fuel element temperatures exceeding the uranium carbide melting point, and (iii) average core pressure less than normal. Results show that the radiation heat transfer rate between fuel elements and cold surfaces increases with decreasing flow rate through the reactor system. However, radiation heat transfer decreases when there is a complete LOFA. When there is a complete LOFA, the peripheral coolant channels of the fuel elements handle most of the radiation heat transfer. A safety system needs to be designed to counteract the decay heat resulting from a post-LOFA reactor scram.

Qualitative and Quantitative Evaluation of Coupling Approaches for Coupling of RELAP5 and LabVIEW

Abstract In the study and design of a nuclear power plant, extensive system modeling is necessary to determine how the reactor will perform in any given situation, not only in the normal performance of the reactor, but also in transients including unanticipated transients without scram and hypothetical accidents. One type of nuclear power plant under study is the hybrid energy system, which uses nuclear power to generate both electricity and heat for facilities. Obviously, the second steam cycle in the nuclear power plant requires several design updates and experiments. Unfortunately, the current versions of the Reactor Excursion and Leak Analysis Program (RELAP) do not allow online data streams from experimental facilities to the computational model of the secondary steam loop. Therefore, this study develops a coupling between RELAP5 and Laboratory Virtual Instrument Engineering Workbench (LabVIEW) to model primary and secondary coolant loops. In this way, the LabVIEW model can easily be connected to an experimental apparatus to provide an online data stream and the online transient behavior of an entire nuclear power plant system. This study shows two different coupling approaches and makes qualitative and quantitative comparisons between these approaches. This paper demonstrates the results of different couplings between the primary and secondary systems of a typical pressurized water reactor (PWR). The primary loop model is a four-loop PWR. The model has been executed with steady state and transients (in this case, a loss-of-coolant accident). The results of both coupling methods have been compared with the typical RELAP5 results.

Mathematical Models of Spacer Grids

The fuel rods in Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) cores are supported by spacer grids. Even though spacer grids add to the pressure loss in the reactor core, spacer grids have several benefits in Light Water Reactors (LWRs). Some of these benefits are: (i) increasing the turbulence at the bottom of the reactor core for better heat transfer in single phase region of the LWRs, (ii) improving the departure nucleate boiling ratio results for PWRs, and (iii) improving critical power ratio (CPR) values by increasing the thickness of film in annular flow regime in the top section of the reactor core of BWRs. Several mathematical models have been developed for single and two phase pressure loss across the grid spacer. Almost all of them significantly depend on Reynolds Number. Spacer designs have evolved (incorporating mixing vanes, springs, dimples, etc), resulting in the complexity of the analysis across the grid, all the models have been compared not only theoretically but also quantitatively. For the quantitative comparisons, this work compares the results of mathematical spacer models with experimental data of BWR Full Size Fine Mesh Bundle Tests (BFBT). The experimental data of BFBT provides very detailed experimental results for pressure drop by using several different boundary condition and detailed pressure drop measurements. Since one CT-scanner was used at the bundle exit and three X-ray densitometers were used for the chordal average void distribution at different elevations to generate the BFBT results, detailed two phase parameters have been measured in BFBT database. Two bundle types of BFBT, the current 8×8 type and the high burn-up 8×8 type, were simulated. Three combinations of radial and axial power shapes were tested: 1) beginning of cycle (BOC) radial power pattern/cosine axial power shape (the C2A pattern); 2) end of cycle (EOC) radial power pattern/cosine axial power shape (C2B pattern); and 3) beginning of cycle radial power pattern/inlet peaked axial power shape (C3 pattern) in BFBT. The pressure drop in BFBT database was measured in both single-phase flow and two-phase flow conditions that cover the normal operational behavior. BFBT database gives the three combinations of high burnup assemblies with different radial and axial power shapes, namely C2A, C2B and C3, which were utilized in the critical power measurements. There are two types of spacers in this program — ferrule type and grid type. Therefore, detailed experimental data of BFBT was used for analyzing mathematical models of spacer grid for various boundary conditions of BWR in this paper. It was observed and discussed that pressure drop values due to spacer models can be significantly different.© 2016 ASME

Survey of Coupling Schemes in Traditional Coupled Neutronics and Thermal-Hydraulics Codes

The design and the analysis of nuclear power plants (NPPs) require computational codes to predict the behavior of the NPP nuclear components and other systems (i.e., reactor core, primary coolant system, emergency core cooling system, etc.). Coupled calculations are essential to the conduct of deterministic safety assessments.Inasmuch as the physical phenomena that govern the performance of a nuclear reactor are always present simultaneously, ideally computational modeling of a nuclear reactor should include coupled codes that represent all of the active physical phenomena. Such multi-physics codes are under development at several institutions and are expected to become operational in the future. However, in the interim, integrated codes that incorporate modeling capabilities for two to three physical phenomena will remain useful. For example, in the conduct of safety analyses, of paramount importance are codes that couple neutronics and thermal-hydraulics, especially transient codes. Other code systems of importance to safety analyses are those that couple primary system thermal-hydraulics to fission product chemistry, neutronics to fuel performance, containment behavior and structural mechanics to thermal-hydraulics, etc. This paper surveys the methods used traditionally in the coupling of neutronic and thermal-hydraulics codes.The neutron kinetics codes are used for computing the space-time evolution of the neutron flux and, hence, of the power distribution. The thermal-hydraulics codes, which compute mass, momentum and energy transfers, model the coolant flow and the temperature distribution. These codes can be used to compute the neutronic behavior and the thermal-hydraulic states separately. However, the need to account with fidelity for the dynamic feedback between the two sets of properties (via temperature and density effects on the cross section inputs into the neutronics codes) and the requirement to model realistically the transient response of nuclear power plants and to assess the associated emergency systems and procedures imply the necessity of modeling the neutronic and thermal-hydraulics simultaneously within a coupled code system. The focus of this paper is a comparison of the methods by which the coupling between neutron kinetics and thermal-hydraulics treatments has been traditionally achieved in various code systems. As discussed in the last section, the modern approaches to multi-physics code development are beyond the scope of this paper.From the field of the most commonly used coupled neutron kinetic-thermal-hydraulics codes, this study selected for comparison the coupled codes RELAP5-3D (NESTLE), TRACE/PARCS, RELAP5/PARCS, ATHLET/DYN3D, RELAP5/SCDAPSIM/MOD4.0/NESTLE. The choice was inspired by how widespread the use of the codes is, but was limited by time availability. Thus, the selection of codes is not to be construed as exhaustive, nor is there any implication of priority about the methods used by the various codes.These codes were developed by a variety of institutions (universities, research centers, and laboratories) geographically located away from each other. Each of the research group that developed these coupled code systems used its own combination of initial codes as well as different methods and assumptions in the coupling process. For instance, all these neutron kinetics codes solve the few-groups neutron diffusion equations. However, the data they use may be based on different lattice physics codes. The neutronics solvers may use different methods, ranging from point kinetics method (in some versions of RELAP5) to nodal expansion methods (NEM), to semi-analytic nodal methods, to the analytic nodal method (ANM). Similarly, the thermal-hydraulics codes use several different approaches: different number of coolant fields, homogenous equilibrium model, separate flow model, different numbers of conservation equations, etc. Therefore, not only the physical models but also the assumptions of the coupled codes and coupling techniques vary significantly. This paper compares coupled codes qualitatively and quantitatively. The results of this study are being used both to guide the selection of appropriate coupled codes and to identify further developments into coupled codes.Copyright

Coupling of RELAP5-SCDAP MOD4.0 and Neutronic Codes

High-fidelity and accurate nuclear system codes play a key role in the design and analysis of complex nuclear power plants, which consist of multiple subsystems, such as the reactor core (and its fuel, burnable poisons, control elements, etc.), the reactor internal structures, the vessel, and the energy conversion subsystem and beyond to grid demand. Most commonly the interplay between these various subsystems is modeled using coupled codes, each of which represents one of the subsystems. And the most common direct coupling is that of thermal-hydraulics and neutronics codes.The subject of this paper is the coupling of codes that model not only thermal-hydraulics and neutronics, but also structural components damage. Furthermore, the neutronic component is not limited to the sole core solver. The coupled code system encompasses thermal-hydraulics, material performance of the fuel, neutronic solver, and neutronic data preparation. Thus, this paper presents a framework for coupling RELAP5/SCDAPSIM/MOD4.0 with a suite of neutron kinetics codes that includes NESTLE, DRAGON and a version of the ENDF library.The version of the RELAP5/SCDAPSIM/MOD4.0 code used in this work is one developed by Innovate System Software (ISS) as part of the international SCDAP Development and Training Program (SDTP) for best-estimate analysis to model reactor transients including severe accident phenomena. This RELAP5/SCDAPSIM/MOD4.0 code version is also capable of predicting nuclear fuel performance. It uses nodal power distributions to calculate mechanical and thermal parameters such as heat-up, oxidation and meltdown of fuel rods and control rods, the ballooning and rupture of fuel rod cladding, the release of fission products from fuel rods, and the disintegration of fuel rods into porous debris and molten material. On the neutronics side, this work uses the NESTLE and DRAGON codes. NESTLE is a multi-dimensional static and kinetic neutronic code developed at North Carolina State University. It solves up to four energy groups neutron diffusion equations utilizing the Nodal Expansion Method (NEM) in Cartesian or hexagonal geometry. The DRAGON code, developed at Ecole Polytechnique de Montreal, performs lattice physics calculations based on the neutron transport equation and is capable of using very fine energy group structures.In this work, we have developed a coupling approach to exchange data among the various modules. In the coupling process, the generated nuclear data (in fine multigroup energy structure) are collapsed down into two- or four-group energy structures for use in NESTLE. The neutron kinetics and thermal-hydraulics modules are coupled at each time step by using the cross-section data. The power distribution results of the neutronic calculations are transmitted to the thermal-hydraulics code. The spatial distribution of coolant density and the fuel-moderator temperature, which result from the thermal-hydraulic calculations, are transmitted back to the neutron kinetics codes and then the loop is closed using new neutronics results. Details of the actual data transfers will be described in the full length paper.Copyright