Computational studies on pressurized thermal shock in reactor pressure vessel

Chouhan, Rakesh; Kansal, Anuj Kumar; Maheshwari, N. K.; Sharma, Avaneesh

doi:10.1016/j.anucene.2020.107987

Cited by 5 publications

(2 citation statements)

References 10 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Fekete T. has reviewed the prospect of modern thermomechanics in structural integrity calculations of pressure vessels [8]. Chouhan R. et al have also presented computational studies on pressurized thermal shock in an RPV [9]. e integrity analyses of a reactor pressure vessel under PA PTS scenarios using the TRACE-XFEM approach have been performed by Mora D. F. et al [10].…”

Section: Introductionmentioning

confidence: 99%

Thermomechanical Analysis of a Reactor Pressure Vessel under Pressurized Thermal Shock Caused by Inadvertent Actuation of the Safety Injection System

Annor-Nyarko

Xia

Ayodeji

2022

Science and Technology of Nuclear Installations

View full text Add to dashboard Cite

The damage induced pressurized thermal shock (PTS) may pose to a reactor pressure vessel (RPV) is a critical safety requirement assessed as part of the ageing management programme of pressurized water reactors (PWRs). A number of researches have studied PTS initiated mainly by postulated accidents such as loss of coolant accidents (LOCAs). However, investigations on PTS-induced threat on RPV caused by inadvertent actuation of the safety injection, a frequent anticipated transient, have not been thoroughly studied. In this paper, a simplified multistep analysis method is applied to study the thermomechanical status of a two-loop PWR under PTS loads caused by inadvertent actuation of the safety injection system. A direct-coupling thermomechanical analysis is performed using a three-dimensional (3D) RPV finite element model. A 3D finite element submodel (consisting of the highiest stress concentration area in the RPV) and an assumed crack are then used to perform fracture mechanics analysis. Subsequently, the critical integrity parameter-stress intensity factor (SIF) is estimated based on FRANC3D-M-integral method coupled in the multistep simulation. The material fracture toughness of the vessel is computed based on the master curve method with experimental fracture toughness data. The results obtained from the direct coupling stress analysis in comparison with sequential coupling approach demonstrate the effectiveness of the proposed multistep method. Also, comparing SIF results obtained with that calculated based on the conventional virtual crack-closure technique (VCCT) and extended finite element method (XFEM) show good agreement. This study provides a useful basis for future studies on anticipated transient-induced crack propagation and remaining service life prediction of ageing reactor pressure vessels.

show abstract

Section: Introductionmentioning

confidence: 99%

Thermomechanical Analysis of a Reactor Pressure Vessel under Pressurized Thermal Shock Caused by Inadvertent Actuation of the Safety Injection System

Annor-Nyarko

Xia

Ayodeji

2022

Science and Technology of Nuclear Installations

View full text Add to dashboard Cite

show abstract

“…The PTS in pressurized nuclear reactors have been the subject of plentiful numerical studies employing various approaches, such as a coupling of multiple RANS models (standard k-, realizable k-and Reynolds-Stress models) (Hu et al 2020), a two-phase CFD model including the effect of free-surface condensation (Cremer et al 2019), a Direct Numerical Simulation (DNS) of 1 3 the flow and thermal fields with a discussion on the validity of the Boussinesq hypothesis (Shams et al 2019). Moreover, the mixing process in nuclear reactor downcomers have been recently experimentally and numerically analyzed using RANS turbulence models (Eltayeb et al 2021), or Large Eddy Simulation (LES) (Chouhan et al 2021) with sensitivity to the coolant injection rate (Li et al 2021). CFD is also used to evaluate the risk of crack initiation and propagation in a reactor pressure vessel (Uitslag-Doolaard et al 2020;Ruan and Morishita 2021).…”

Section: Introductionmentioning

confidence: 99%

Wall-Resolved Large Eddy Simulations of the Transient Turbulent Fluid Mixing in a Closed System Replicating a Pressurized Thermal Shock

Angeli

2021

Flow Turbulence Combust

View full text Add to dashboard Cite

The isothermal mixing of a heavy and a light liquid of different physical properties is numerically investigated by means of Large Eddy Simulations. The validation is based on experimental data held in a system reproducing various components of a pressurized water nuclear reactor, during a scenario of cold water injection at a low Atwood number of 0.05. The flow has two distinct stages: first a buoyancy-driven phase is characterized by a fluid front development in the cold leg and gives rise to Kelvin-Helmholtz whorls under the action of density changes. Then, the heavy liquid discharges into the downcomer filled with light liquid, which causes a turbulent mixing. These phenomena are analyzed through a single-phase approach where the density of the working fluid is either variable or modeled by the Boussinesq approximation. The influence of grid refinement is deeply examined, which shows that the mesh convergence is well achieved for the main flow quantities, unlike the low-magnitude spanwise components. Overall, the numerical solutions are found to reproduce the experimental measurements with a fair accuracy for both physical models used. These latter exhibit similar trends, due to the small density difference under consideration. The predictions in the downcomer appear to be more challenging owing to a strongest turbulence than in the cold leg, some flow features being not properly captured. However, the experimental data in the downcomer are found to be incomplete and somewhat dubious for a strict validation of the numerical simulations. Lastly, the flow distribution in the dowcomer is investigated, providing further insight on the mixing process.

show abstract

Verification & validation of CFD predictions regarding pressurized thermal shock (PTS) situations in ROCOM installation: Comparison with IAEA benchmark

Ayad¹,

Baghdad²,

Bouaichaoui³

et al. 2023

Nuclear Engineering and Design

View full text Add to dashboard Cite

Computational studies on pressurized thermal shock in reactor pressure vessel

Cited by 5 publications

References 10 publications

Thermomechanical Analysis of a Reactor Pressure Vessel under Pressurized Thermal Shock Caused by Inadvertent Actuation of the Safety Injection System

Thermomechanical Analysis of a Reactor Pressure Vessel under Pressurized Thermal Shock Caused by Inadvertent Actuation of the Safety Injection System

Wall-Resolved Large Eddy Simulations of the Transient Turbulent Fluid Mixing in a Closed System Replicating a Pressurized Thermal Shock

Verification & validation of CFD predictions regarding pressurized thermal shock (PTS) situations in ROCOM installation: Comparison with IAEA benchmark

Contact Info

Product

Resources

About