The ‘Fuel Rod Analysis ToolBox’: A general program for preparing the input of a fuel rod performance code

Laßmann, K.; Arndt, Sebastian; Laar, J. van de; Uffelen, P. Van

doi:10.1016/j.anucene.2015.03.012

Cited by 17 publications

(7 citation statements)

References 2 publications

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“…Thus the only input data unique to each validation case includes the rod geometry, power history and axial power profile, fast neutron flux and coolant boundary conditions. Generally, before usage as BISON input, experimental data histories were condensed using the Fuel Rod Analysis ToolBox program [40]. A comprehensive list of the standard models and assumptions used, and a more detailed description of each validation case, is available in the BISON validation report [41].…”

Section: Validation Standardsmentioning

confidence: 99%

Validating the BISON fuel performance code to integral LWR experiments

Williamson

Gamble

Perez

et al. 2016

Nuclear Engineering and Design

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BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that 1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, 2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and 3) comparison of rod diameter results indicates a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. Initial rod diameter comparisons are not satisfactory and have led to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.

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Section: Validation Standardsmentioning

confidence: 99%

Validating the BISON fuel performance code to integral LWR experiments

Williamson

Gamble

Perez

et al. 2016

Nuclear Engineering and Design

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“…In the first FINIX validation in Publication IV, data from OECD/NEA International Fuel Performance Experiments database [101] is used, specifically the Halden tests IFA-429 and IFA-432. In later validations, the data is directly obtained from Halden and condensed with the FuelRod Analysis Toolbox [102]. Data condensation is necessary, as Halden reports measurements typically at 15-minute intervals, and such an interval yields a very large dataset over an irradiation lasting multiple years.…”

Section: Integral Validationmentioning

confidence: 99%

Determination of tolerance limits for fuel assembly fission gas release summary statistics

Loukusa

Tulkki

2020

Nuclear Engineering and Design

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“…The Halden test IFA-677.1 [40,41,43] Figure 2.1. The Halden raw data were condensed using the Fuel Rod Analysis ToolBox developed by K. Lassmann [44]. The thermal boundary conditions at the cladding outer surface were determined using Bison's internal coolant channel model, with the coolant inlet temperature history determined from the Halden raw data.…”

Section: Description Of Halden Test Ifa-6771mentioning

confidence: 99%

M3MS-19IN0101201 Modeling Intra-granular Fission Gas Bubble Evolution in Uranium Dioxide during Transients

Pastore

2019

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The SciDAC milestone M3MS-19IN0101201 on Modeling Intra-granular Fission Gas Bubble Evolution in Uranium Dioxide during Transients was completed with the development, testing and implementation in the Bison code of an engineering model for intra-granular fission gas bubble evolution that is applicable to transient conditions. The model extends the previous work on modeling bubble evolution during normal operating conditions to include additional complexity that arises during transient situations, in particular, the so-called intra-granular bubble coarsening. This phenomenon refers to the appearance of a second population of coarsened bubbles with diameters of tens to hundreds of nm, which co-exist with nanometric bubbles present also during normal operating conditions and are responsible for a large portion of transient fuel swelling. The developed model considers growth of bubbles along dislocations, associated with the absorption of vacancies available in the dislocation core region, and favored by pipe diffusion of fission gas atoms. The model also utilizes the formulation for the gas atom resolution rate developed in the present SciDAC project through molecular dynamics simulations by other institutions. The present work therefore demonstrates an operational multiscale approach, based on collaboration within the project. The model is validated to an extensive experimental dataset of ramp tested uranium dioxide fuel samples. Additionally, we present new applications of the base model for bubble evolution during normal operating conditions to Bison engineering simulations of Cr 2 O 3-doped fuel rods irradiated in the Halden reactor. Simulation results in terms of integral fission gas release in the fuel rods are compared to the available experimental data, pointing out significant improvements in the engineering predictions compared to application of the empirical model that was available in Bison before the SciDAC developments. 2 Application of SciDAC model to Halden fuel rod simulations using Bison

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The ‘Fuel Rod Analysis ToolBox’: A general program for preparing the input of a fuel rod performance code

Cited by 17 publications

References 2 publications

Validating the BISON fuel performance code to integral LWR experiments

Validating the BISON fuel performance code to integral LWR experiments

Determination of tolerance limits for fuel assembly fission gas release summary statistics

M3MS-19IN0101201 Modeling Intra-granular Fission Gas Bubble Evolution in Uranium Dioxide during Transients

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