Validation of the thermal-hydraulic system code ATHLET based on selected pressure drop and void fraction BFBT tests

Marcello, V. Di; Escalante, Javier Jiménez; Espinoza, Víctor Sánchez

doi:10.1016/j.nucengdes.2015.04.003

Cited by 10 publications

(3 citation statements)

References 7 publications

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“…(3) Liquidvapor interphase mass and energy transfer model; (4) The drift-flux model providing a one-dimensional description of the velocity differences between liquid and vapor phases taking into consideration the void fraction across the flow channel; (5) The form pressure loss and the wall friction pressure loss determining the irreversible pressure loss in a flow channel. ATHLET has incorporated a large spectrum of models as illustrated in Figure 2A (Di Marcello et al, 2015;Wielenberg et al, 2019). With user's interfaces, the other independent modules like the GRS containment code COCOSYS and the Computational Fluid Dynamics (CFD) codes can be coupled.…”

Section: Externalsource Termsmentioning

confidence: 99%

“…The ATHLET has been successfully applied in the case of pre and post-test calculations of both large and small-scale experiments in the frame of International Standard Problems (ISPs), benchmarks and various international and national projects, e.g., the LSTF, PKL, and UPTF test facilities (Yousif et al, 2017;Hollands et al, 2019). The code's capabilities were investigated with the experimental data of test facilities named ATLAS and INKA (Di Marcello et al, 2015). Moreover, the code is validated is against the experimental data of facilities like MYHRRA, KASOLA, and TALL for the Accelerator-Driven Subcritical (ADS) systems and the future Generation IV nuclear applications (Hollands et al, 2019).…”

Section: Externalsource Termsmentioning

confidence: 99%

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Study on Typical Design Basis Conditions of HPR1000 With Nuclear Safety Analysis Code ATHLET

et al. 2020

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The third-generation nuclear power plant Hua-long Pressurized Reactor (HPR1000) is developed based on the experience of Chinese commercial Nuclear Power Plant (NPP) designs, construction, operation and maintenance. It improves the concept of defense in depth and strengthens severe accident prevention and mitigation strategies. The HPR1000 has implemented a number of active and passive innovative safety systems and accident management procedures for design basis conditions, e.g., the employment of Medium Pressure Rapid Cooldown (MCD) and Atmospheric Steam Dump System (ASDS) for the activation of Middle Head Safety Injection (MHSI), the application of Secondary Passive Residual Heat Removal System (SPRHR) for the residual heat removal. In the article, calculations are carried out for HPR1000 nuclear power plant with nuclear system safety analysis code ATHLET (Analysis of Thermal-Hydraulics of Leaks and Transient) 3.1 (Lerchl et al., 2016). By means of conservative deterministic safety analysis approach, transient analyses concerning selected typical design basis conditions, i.e., Large Break Loss-Of-Coolant Accident (LB-LOCA), Small Break Loss-Of-Coolant Accident (SB-LOCA), Steam Generator Tube Rupture accident (SGTR), and Feed water Line Break (FLB) are performed. The ATHLET results are also compared with the results performed by CGN-CNPTRI (China General Nuclear-China Nuclear Power Technology Research Institute) with their own code LOCUST with similar assumptions. The comparisons indicate that, although some discrepancies are detected, the trends of system responses predicted by the two codes are generally in agreement with each other for different accident scenarios. The results also demonstrate that the acceptance criteria for each accident can be met with significant safety margin. Thus, the effectiveness of safety system configuration and accident management procedures is guaranteed.

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Section: Externalsource Termsmentioning

confidence: 99%

Section: Externalsource Termsmentioning

confidence: 99%

Study on Typical Design Basis Conditions of HPR1000 With Nuclear Safety Analysis Code ATHLET

et al. 2020

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“…CFD and subchannel codes such as TRIO_U [35], SUB-CHANFLOW [36], FLICA4 [37], NEPTUNE [38], and TransAT [39] provide full thermal hydraulics at the fuel pin level and the boundary conditions used in other codes. System codes such as CATHARE (Code for Analysis of Thermal Hydraulics during an Accident of Reactor and Safety Evaluation) [40] and ATHLET (Analysis of Thermal Hydraulics of Leaks and Transients) [41] provide simplified thermal hydraulics at the nuclear power plant level and the boundary conditions used in other codes. Finally, fuel performance codes such as DRACCAR [42] and SCANAIR (Systems of Codes for Analysing Reactivity Initiated Accidents) [43] provide full thermo mechanics at the fuel pin level and the boundary conditions used in other codes.…”

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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Validation with the MARS and TASS/SMR codes based on experimental results of a pressurizer safety valve line break at the SMART-ITL facility

Chung

Jeon

Bae

et al. 2020

Annals of Nuclear Energy

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Validation of the thermal-hydraulic system code ATHLET based on selected pressure drop and void fraction BFBT tests

Cited by 10 publications

References 7 publications

Study on Typical Design Basis Conditions of HPR1000 With Nuclear Safety Analysis Code ATHLET

Study on Typical Design Basis Conditions of HPR1000 With Nuclear Safety Analysis Code ATHLET

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

Validation with the MARS and TASS/SMR codes based on experimental results of a pressurizer safety valve line break at the SMART-ITL facility

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