Analysis of the OECD/NRC BWR Turbine Trip Benchmark by the Coupled-Code System ATHLET-QUABOX/CUBBOX

Langenbuch, S.; Schmidt, K.-D.; Velkov, K.

doi:10.13182/nse04-a2457

Cited by 3 publications

(2 citation statements)

References 1 publication

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“…Some of the coupled codes that have been used in different facilities and the power stations under different circumstances are given below The coupled code DYN3D/ATHLET has been utilized for the transient analysis of BWR .The coupling code SKTECH/TRAC has been utilized for the analysis of control rod ejection transients by Asaka et al . .The coupled code NEM/TRAC‐PF has been used for MSLB transient analysis of PWR and BWR core transients .The coupled code QUABOX/ATHLET has been used for the analysis of OECD/NRC BWR turbine trip benchmark problem by Langenbuch .The coupled cope PANBOX/RELAP has been used for the analysis of MSLB in OECD/NEA PWR Sanchez‐Espinoza et al . Verification of this coupled technique was done by Jackson et al .…”

Section: Introductionmentioning

confidence: 99%

“…Verification of this coupled technique was done by Jackson et al . for the core transient analysis .ATHLETE code coupled with 3D core neutronic model DYN3D has been used for the analysis of BWR Turbine trip benchmark, and the validation of this code for PWR and BWR was done by Glaeser .Similarly, the code like ATHLET‐QUABOX/CUBBOX has been used by Langenbuch for the analysis of OECD/NRC BWR Turbine trip benchmark .RELAP code has been coupled with point kinetic codes for the analysis of OECD/NEA PWR MSLB by Sanchez‐Espanioza and Nigro et al . .CATHORE code has been coupled with CRONOS2‐FLICA4 for BWR turbine trip analysis .…”

Section: Introductionmentioning

confidence: 99%

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Neutronics and thermal hydraulic coupling analysis of integrated pressurized water reactor

Khan

Peng

Khan

2012

Int. J. Energy Res.

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In this paper, three-dimensional (3D) power distribution of newly designed small nuclear reactor core has been achieved by using neutron kinetic/thermal hydraulic (NK/TH) coupling. This is pressurized water reactor-based small nuclear reactor in which plate type fuel element has been used and the core of the reactor has hexagonal type geometry. This paper depicts the design of the reactor core by using coupling approach of neutronics(Neutron Kinetic) and thermal hydraulic studies. For this purpose, neutronic analysis has been obtained by using lattice physics code, i.e. HELIOS and neutron kinetic code, i.e. REMARK. HELIOS code gives the cross-section data which is being used as input to the REMARK code. At the same time, THEATRe code was used for the thermal hydraulic analysis of the reactor core. In the coupling process, some data (fuel temperature, moderator temperature, void fraction, etc.) from THEATRe code has been used in conjunction with HELIOS and REMARK codes. After finalizing the NK/TH coupling, 3D evaluation of the power distribution of the reactor core has been achieved and is included in the paper. The purpose of this paper is to evaluate the design and get the normal operational behavior of the reactor core by NK/TH coupling approach.

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Section: Introductionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Neutronics and thermal hydraulic coupling analysis of integrated pressurized water reactor

Khan

Peng

Khan

2012

Int. J. Energy Res.

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Application of the best-estimate plus uncertainty approach on a BWR ATWS transient using the NURESIM European code platform

Perin

Escalante

2017

Nuclear Engineering and Design

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Monte Carlo Neutronics and Thermal Hydraulics Analysis of Reactor Cores with Multilevel Grids

Bernnat

Mattes

Guilliard

et al. 2014

SNA + MC 2013 - Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo

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Power reactors are composed of assemblies with fuel pin lattices or other repeated structures with several grid levels, which can be modeled in detail by Monte Carlo neutronics codes such as MCNP6 using corresponding lattice options, even for large cores. Except for fresh cores at beginning of life, there is a varying material distribution due to burnup in the different fuel pins. Additionally, for power states the fuel and moderator temperatures and moderator densities vary according to the power distribution and cooling conditions. Therefore, a coupling of the neutronics code with a thermal hydraulics code is necessary. Depending on the level of detail of the analysis, a very large number of cells with different materials and temperatures must be regarded. The assignment of different material properties to all elements of a multilevel grid is very elaborate and may exceed program limits if the standard input procedure is used. Therefore, an internal assignment is used which overrides uniform input parameters. The temperature dependency of continuous energy cross sections, probability tables for the unresolved resonance region and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. The method is applied with MCNP6 and proven for several full core reactor models. For the coupling of MCNP6 with thermal hydraulics appropriate interfaces were developed for the GRS system code ATHLET for liquid coolant and the IKE thermal hydraulics code ATTICA-3D for gaseous coolant. Examples will be shown for different applications for PWRs with square and hexagonal lattices, fast reactors (SFR) with hexagonal lattices and HTRs with pebble bed and prismatic lattices.

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Analysis of the OECD/NRC BWR Turbine Trip Benchmark by the Coupled-Code System ATHLET-QUABOX/CUBBOX

Cited by 3 publications

References 1 publication

Neutronics and thermal hydraulic coupling analysis of integrated pressurized water reactor

Neutronics and thermal hydraulic coupling analysis of integrated pressurized water reactor

Application of the best-estimate plus uncertainty approach on a BWR ATWS transient using the NURESIM European code platform

Monte Carlo Neutronics and Thermal Hydraulics Analysis of Reactor Cores with Multilevel Grids

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