CTF User's Manual (V.4.2)

Salko, Robert; Avramova, Maria; Wysocki, Aaron; Hu, Jianwei; Toptan, Aysenur; Porter, Nathan; Blyth, Taylor S.; Dances, Christopher A.; Gómez, Ángel Ignacio Pérez; Jernigan, Caleb; Kelly, Joeseph

doi:10.2172/1737480

Cited by 6 publications

(3 citation statements)

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“…First, LHRs were selected as surrogates to characterize fuel rod temperature behaviors. Fuel rod temperature estimates calculated by VERA's thermal-hydraulics module, CTF [6], [7], are available in the HDF5 files. However, these results were expected to be less realistic than the LHRs calculated by VERA's neutronics module, MPACT [8], [9].…”

Section: Parameters Of Interestmentioning

confidence: 99%

Assessment of the Effect of Prototypic High-Burnup Operating Conditions of Fuel Fragmentation, Relocation, and Dispersal Susceptibility

Capps¹,

Hirschhorn²,

Greenquist³

et al. 2022

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The US nuclear energy industry is investigating strategies that further reduce the cost of energy production by using its existing fleet of nuclear generating stations. Most nuclear power plant operating costs are associated with purchasing fresh fuel assemblies or the efficiency of the reactor core design. Material costs are typically beyond the operator's control; however, the core design optimizations offer potential operational savings. The core design envelope available to operators is constrained by two primary regulatory criteria: an enrichment limit of 5% 235 U and a burnup limit of 62 GWD/tU. These constraints have resulted in renewed efforts by the nuclear industry to pursue extending the peak rodaverage burnup beyond 62 GWd/tU. This effort will likely require additional safety analyses beyond what is currently accepted by the US Nuclear Regulatory Commission. The purpose of this work is to demonstrate a best estimate plus uncertainty pin-by-pin high-burnup loss of coolant accident analysis technique to assess full-core high-burnup fuel fragmentation, relocation, and dispersal (FFRD) and identify approaches for minimizing or potentially mitigating FFRD through core design optimizations.* T indicates 18-24 month transition cycles + HBu indicates 24 month cycles with increased enrichment and burnup

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Section: Parameters Of Interestmentioning

confidence: 99%

Assessment of the Effect of Prototypic High-Burnup Operating Conditions of Fuel Fragmentation, Relocation, and Dispersal Susceptibility

Capps¹,

Hirschhorn²,

Greenquist³

et al. 2022

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“…COBRA-TF [62,63,66] was developed using FORTRAN 77 in 1980 by PNL (Pacific Northwest Laboratories, Washington, WA, USA), sponsored by the NRC (Nuclear Regulation Commission) and has been continuously updated. It is a LWR (square geometry) thermal hydraulics subchannel code developed for the purpose of studying general nuclear reactor behaviour and accident scenarios.…”

Section: Ctf Subchannel Codementioning

confidence: 99%

“…Second, a CTF description comprehending general features, updates, etc. [62,63] was undertaken to present the second code used in the coupling inner iterations within one outer iteration verification. Third, the specifications description covering the KAIST (Korean Advanced Institute of Science and Technology) benchmark [64] was performed to present the data used in the coupling inner iterations within one outer iteration verification.…”

Section: Introductionmentioning

confidence: 99%

DYN3D and CTF Coupling within a Multiscale and Multiphysics Software Development (Part I)

et al. 2021

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Understanding and optimizing the relation between nuclear reactor components or physical phenomena allows us to improve the economics and safety of nuclear reactors, deliver new nuclear reactor designs, and educate nuclear staff. Such relation in the case of the reactor core is described by coupled reactor physics as heat transfer depends on energy production while energy production depends on heat transfer with almost none of the available codes providing full coupled reactor physics at the fuel pin level. A Multiscale and Multiphysics nuclear software development between NURESIM and CASL for LWRs has been proposed for the UK. Improved coupled reactor physics at the fuel pin level can be simulated through coupling nodal codes such as DYN3D as well as subchannel codes such as CTF. In this journal article, the first part of the DYN3D and CTF coupling within the Multiscale and Multiphysics software development is presented to evaluate all inner iterations within one outer iteration to provide partially verified improved coupled reactor physics at the fuel pin level. Such verification has proven that the DYN3D and CTF coupling provides improved feedback distributions over the DYN3D coupling as crossflow and turbulent mixing are present in the former.

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