Development of the coupled 3D neutron kinetics/thermal-hydraulics code DYN3D-HTR for the simulation of transients in block-type HTGR

Kerntechnik ◽  
2011 ◽  
Vol 76 (3) ◽  
pp. 166-173
Author(s):  
U. Rohde ◽  
S. Baier ◽  
S. Duerigen ◽  
E. Fridman ◽  
S. Kliem ◽  
...  
Keyword(s):  
Block Type ◽  
Thermal Hydraulics ◽  
Neutron Kinetics

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

2015 ◽  
Author(s):  
Sabahattin Akbas ◽  
Victor Martinez-Quiroga ◽  
Fatih Aydogan ◽  
Abderrafi M. Ougouag ◽  
Chris Allison
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Primary System ◽  
Thermal Hydraulics ◽  
Neutron Kinetics ◽  
Coupled Codes ◽  
Physical Phenomena ◽  
Selection Of

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.

Coupling of RELAP5-SCDAP MOD4.0 and Neutronic Codes

2015 ◽  
Author(s):  
Victor Martinez-Quiroga ◽  
Sabahattin Akbas ◽  
Fatih Aydogan ◽  
Abderrafi M. Ougouag ◽  
Chris Allison
Keyword(s):  
Power Distribution ◽  
Power Plants ◽  
Neutron Transport ◽  
Severe Accident ◽  
Energy Structure ◽  
Thermal Hydraulics ◽  
Neutron Transport Equation ◽  
Time Step ◽  
Fuel Rods ◽  
Neutron Kinetics

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.

Initial version of an integrated thermal hydraulics and neutron kinetics 3D code X3D

2006 ◽  
Vol 236 (14-16) ◽  
pp. 1533-1546
Author(s):  
S. Jewer ◽  
A. Thompson ◽  
A. Hoeld ◽  
P.A. Beeley
Keyword(s):  
Thermal Hydraulics ◽  
Neutron Kinetics

The three-dimensional neutron kinetics coupled with thermal-hydraulics in RBMK accident analysis

2008 ◽  
Vol 238 (4) ◽  
pp. 1002-1025 ◽  
Author(s):  
F. D’Auria ◽  
S. Soloviev ◽  
V. Malofeev ◽  
K. Ivanov ◽  
C. Parisi
Keyword(s):  
Three Dimensional ◽  
Accident Analysis ◽  
Thermal Hydraulics ◽  
Neutron Kinetics

Analysis of Shutdown System Effectiveness in the Canadian Super Critical Water Reactor Using Coupled Thermal Hydraulics and Three-Dimensional Neutron Kinetics

2019 ◽  
Vol 5 (4) ◽  
Author(s):  
David William Hummel ◽  
David Raymond Novog
Keyword(s):  
Three Dimensional ◽  
Thermal Hydraulics ◽  
Flow Tube ◽  
Reactivity Effect ◽  
Neutron Kinetics ◽  
Power Pulse ◽  
Density Reduction ◽  
Positive Reactivity ◽  
Fuel Elements ◽  
Supercritical Water Cooled Reactor

Abstract The Canadian supercritical water-cooled reactor concept features a re-entrant fuel channel wherein coolant first travels down a center flow tube and then up around the fuel elements. Previous work demonstrated that in cases of sudden coolant flow reduction or reversal (such as that which would result from a large pipe break near the core inlet), the coolant density reduction around the fuel has a positive reactivity effect that results in a power excursion. Such a transient is inherently self-terminating since the inevitable density reduction in the center flow tube has a very large negative reactivity effect. Nevertheless, a brief power pulse would ensue. In this work, the possibility of mitigating the power pulse with a fast-acting shutdown system was explored. The shutdown system model, consisting of bottom-inserted neutron absorbing blades and realistic estimates of insertion rates and trip conditions, was added to a full-core coupled spatial neutron kinetics and thermal-hydraulics model. It was demonstrated that such a system can effectively mitigate both the peak magnitude of the power excursion and its duration.

Development of Neutronics and Thermal-Hydraulics Coupled Code for Accelerator Driven Subcritical Systems

2018 ◽  
Author(s):  
Run Luo ◽  
Pengfei Wang ◽  
Xinyu Wei ◽  
Shripad T. Revankar ◽  
Fuyu Zhao
Keyword(s):  
Steady State ◽  
Transient Analysis ◽  
Numerical Differentiation ◽  
Fuel Pellet ◽  
Spallation Neutron Source ◽  
Thermal Hydraulics ◽  
Neutron Kinetics ◽  
Reactor Axis ◽  
Differentiation Formulas ◽  
Subcritical System

A new neutronics and thermal-hydraulics coupled code named ARTAP is developed to analyze the steady-state and transient characteristics of accelerator driven subcritical system (ADS) in this paper. Monte Carlo simulations for spallation neutron source and deterministic calculations for the subcritical core are performed in the steady-state analysis module. The ADS core is divided into a number of nodes both along the reactor axis and the fuel pellet radius for the calculation of spatial temperature distributions. The power iteration method is adopted to solve the coupled neutronics and thermal-hydraulics problems. The transient analysis module consists of space-time neutron kinetics model and thermal-hydraulics dynamic model, which is calculated by using numerical differentiation formulas (NDFs) method. The new code is verified by comparing its predictions for both the steady-state and transient cases of the OECD/NEA ADS benchmark. Results of numerical simulations indicate that ARTAP is reliable and efficient to be applied for the ADS analysis.

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