Yagnik S.,EPRI |
Pytel M.,EPRI |
Lyon B.,ANATECH Corporation
LWR Fuel Performance Meeting, Top Fuel 2013 | Year: 2013
EPRI initiated a complete redesign of its fuel performance code FALCON in 2008. The resulting new code is much simpler to use, maintain, and develop. It is now simply renamed Falcon. This paper discusses Falcon's new programming architecture and its versatile input and output graphical user interfaces (GUIs). While the redesign fundamentally upgraded and modernized the legacy code, Falcon's inherent analytical strengths as a of a general purpose, best estimate, 2-D code with steady-state and transient analyses capabilities validated to high burn-ups were preserved. The redesign has enabled in Falcon at least two very desirable attributes for a fuel performance code. These attributes are believed to-date to be unique to Falcon. First, from the perspective of basic analysis needs, certain standard computations such as fuel centerline temperature, cladding corrosion, fuel rod growth, etc., can be readily performed using a built-in configuration schema by users of very basic skill set. Examples of such schema applications performed in a user-friendly and intuitive fashion are discussed in the paper. Second, the user is able to input many material properties and some behavioral models through dynamically library "plug-in " system, thus minimizing proprietary information restrictions and eliminating the need to interact with and understand the source code. The paper provides specific details of the advantages of redesign in the areas of basic and fuel duty related operational margin assessments and an example of user-supplied plug-in implementation for IFBA fuel.
Capps N.,University of Tennessee at Knoxville |
Montgomery R.,Pacific Northwest National Laboratory |
Sunderland D.,Pacific Northwest National Laboratory |
Sunderland D.,ANATECH Corporation |
And 2 more authors.
Nuclear Engineering and Design | Year: 2016
Missing pellet surface (MPS) defects are local geometric defects in nuclear fuel pellets that result from pellet mishandling or manufacturing. The presence of MPS defects can cause significant clad stress concentrations that can lead to through-wall cladding failure for certain combinations of fuel burnup, and reactor power level or power change. Consequently, the impact of MPS defects has limited the rate of power increase, or ramp rate, in both pressurized and boiling water reactors (PWRs and BWRs, respectively). Improved three-dimensional (3-D) fuel performance models of MPS defect geometry can provide better understanding of the probability for pellet clad mechanical interaction (PCMI), and correspondingly the available margin against cladding failure by stress corrosion cracking (SCC). The Consortium of Advanced Simulations of Light Water Reactors (CASL) has been developing the Bison-CASL fuel performance code to consider the inherently multi-physics and multi-dimensional mechanisms that control fuel behavior, including cladding stress concentrations resulting from MPS defects. This paper evaluates the cladding hoop stress distributions as a function of MPS defect geometry with discrete pellet radial cracks for a set of typical operating conditions in a PWR fuel rod. The results provide a first step toward a probabilistic approach to assess cladding failure during power maneuvers. This analysis provides insight into how varying pellet defect geometries affect the distribution of the cladding stress, as well as the temperature distributions within the fuel and clad; and are used to develop stress concentration factors for comparing 2-D and 3-D models.
Hursin M.,University of California at Berkeley |
Downar T.J.,University of Michigan |
Montgomery R.,ANATECH Corporation |
Montgomery R.,Pacific Northwest National Laboratory
Nuclear Engineering and Design | Year: 2013
When applied to reactivity initiated accidents (RIAs) analysis, codes such as DeCART can provide a detailed radial, azimuth, and axial power distribution within a fuel rod. The work reported here is aimed at quantifying the sensitivity of the cladding thermo-mechanical response, calculated by the fuel performance code FALCON to the more accurate and detailed neutronic solution provided by DeCART for full PWR core RIA analysis. As a basis of comparison, the neutronics analysis is also performed with the U.S. NRC PARCS code, which is representative of the methodology used by the industry. Based on the DeCART solutions, several fuel rods are chosen for analysis with FALCON according to several relevant criteria. For each of the selected fuel rods, a FALCON study is performed using the boundary conditions provided by the neutronic solvers to predict the cladding response in terms of Strain Energy Density (SED) to the power pulse during the transient. The results of the analysis led to the following conclusions:The largest impact on the cladding response can be attributed to the differences in the kinetic parameters in PARCS and DeCART.The modeling of fuel pin exposure in the current industry standard "two step" methodology can result in some significant discrepancies in terms of SED during RIA analysis.The effect of azimuthal power variation within a given fuel rod has a 10% impact on the SED and should be taken into consideration during RIA analysis, especially for high exposure fuel. © 2013 Elsevier B.V. All rights reserved.
Khvostov G.,Paul Scherrer Institute |
Lyon W.,ANATECH Corporation |
Zimmermann M.A.,Paul Scherrer Institute
Annals of Nuclear Energy | Year: 2013
A methodology for the analysis of cladding failures caused by Pellet-Cladding Interaction (PCI) that may result in the Stress Corrosion Cracking (SCC) during power ascension at a PWR reactor start-up is presented. The proposed approach is based on the capabilities of EPRI's FALCON MOD01 code - as developed by ANATECH Corp. - with the PSI in-house model GRSW-A for the micro-structural processes occurring in the fuel. The methodology allows for analysis of the impact of missing pellet surface (MPS) on the failure-related characteristics of the cladding, particularly the peak local hoop stress, along with the accounting for the transient gaseous fuel swelling and FGR. The application of the developed methodology to the ramp tests with PWR fuel samples from the SUPER-RAMP project, carried out in Studsvik (Sweden) in 1980s, is presented. This analysis has been conducted in the framework of the PSI participation in Fuel Modeling Programme FUMEX III, recently carried out by IAEA. As a result, the capability of the new methodology to differentiate between the power ramps with failure and without failure of the claddings of non-defect fuel rods is shown, and the appropriate failure thresholds for the selected criteria are determined. The results of calculation for the stress-concentration factors caused by MPS, as a function of the angular size of the MPS defect, are in good agreement with previous similar studies, specifically with the one undertaken by the principal FALCON MOD01 code developer - ANATECH Corp. Furthermore, the predicted effects of MPS defect size is compared with the effects related to the power ascension rate, with the power ramp level being kept the same. A reduced power ascension rate is determined, which is capable of 'neutralizing' the detrimental effect on the local stress concentration in the cladding caused by the MPS defect under consideration. © 2013 Elsevier Ltd. All rights reserved.
Rashid J.Y.R.,ANATECH Corporation |
Yagnik S.K.,EPRI |
Montgomery R.O.,Pacific Northwest National Laboratory
JOM | Year: 2011
Light water reactor fuel is a multicomponent system required to produce thermal energy through the fission process, efficiently transfer the thermal energy to the coolant system, and provide a barrier to fission product release by maintaining structural integrity. The operating conditions within a reactor induce complex multi-physics phenomena that occur over time scales ranging from less than a microsecond to years and act over distances ranging from inter-atomic spacing to meters. These conditions impose challenging and unique modeling, simulation, and verification data requirements in order to accurately determine the state of the fuel during its lifetime in the reactor. The capabilities and limitations of the current engineering-scale one-dimensional and two-dimensional fuel performance codes is discussed and the challenges of employing higher level fidelity atomistic modeling techniques such as molecular dynamics and phase-field simulations is presented. © 2011 TMS.