M. Annor-nyarko | Harbin Engineering university china (original) (raw)

Papers by M. Annor-nyarko

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

Science and Technology of Nuclear Installations, 2022

e damage induced pressurized thermal shock (PTS) may pose to a reactor pressure vessel (RPV) is a... more e 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 PTSinduced 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 twoloop 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. e material fracture toughness of the vessel is computed based on the master curve method with experimental fracture toughness data. e 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. is study provides a useful basis for future studies on anticipated transient-induced crack propagation and remaining service life prediction of ageing reactor pressure vessels.

Research paper thumbnail of Numerical Fracture Analysis of a Reactor Pressure Vessel based on Abaqus-FRANC3D Co-simulation Method

Procedia Structural Integrity, 2022

The reactor pressure vessel (RPV) is subjected to many cyclic loadings critical of which is the i... more The reactor pressure vessel (RPV) is subjected to many cyclic loadings critical of which is the integrity risk posed by pressurized thermal shock (PTS) loadings induced by one of the most frequent anticipated operational occurrences-inadvertent operation of the safety injection system (SIS). In this paper, the fracture mechanics analysis for an ageing pressurized water reactor subjected to PTS induced by inadvertent actuation of the SIS is performed using a proposed simplified Abaqus-FRANC3D co-simulation method. A three-dimensional (3-D) finite element half-symmetric model of a typical pressured water reactor RPV is used to perform thermal-mechanical stress coupling analysis. 3-D fracture mechanic submodel with an assumed surface crack is created for the computation of the transient integrity parameter-stress intensity factor (SIF), using M-integral approach coupled in the proposed co-simulation method. Subsequently ASME method is used to evaluate the vessel's material fracture toughness. Finally, the SIFs obtained with the simplified co-simulation method is compared with the conventional virtual crack-closure technique (VCCT), and the result show good agreement. This work serves as a useful reference for fast crack propagation and life prediction analysis of ageing PWR RPVs.

Research paper thumbnail of Integrity Evaluation of a Reactor Pressure Vessel Based on a Sequential Abaqus-FRANC3D Simulation Method

Science and Technology of Nuclear Installations, 2021

e safety-risk pressurized thermal shock (PTS) have on a reactor pressure vessel (RPV) is one of t... more e safety-risk pressurized thermal shock (PTS) have on a reactor pressure vessel (RPV) is one of the most important studies for the lifetime ageing management of a reactor. Several studies have investigated PTS induced by postulated accidents and other anticipated transients. However, there is no study that analyzes the effect of PTS induced by one of the most frequent anticipated operational occurrences-inadvertent operation of the safety injection system. In this paper, a sequential Abaqus-FRANC3D simulation method is proposed to study the integrity status of an ageing pressurized water reactor subjected to PTS induced by inadvertent actuation of the safety injection system. A sequential thermal-mechanical coupling analysis is first performed using a three-dimensional reactor pressure vessel finite element model (3D-FEM). A linear elastic fracture mechanics submodel with a postulated semielliptical surface crack is then created from the 3D-FEM. Subsequently, the submodel is used to evaluate the stress intensity factors based on the M-integral approach coupled within the proposed simulation method. Finally, the stress intensity factors (SIFs) obtained using the proposed method are compared with the conventional extended finite element method approach, and the result shows a good agreement. e maximal thermomechanical stress concentration was observed at the inlet nozzleinner wall intersection. In addition, e ASME fracture toughness of the reactor vessel's steel compared with SIFs show that the PTS event and crack configuration analysed may not pose a risk to the integrity of the RPV. is work serves as a critical reference for the ageing management and fatigue life prediction of reactor pressure vessels.

Research paper thumbnail of Control Volume Finite Difference Analysis of Thermal Stresses Induced in GHARR-1 Reactor Vessel by Coolant Heating

INTERNATIONAL JOURNAL OF SCIENTIFIC & TECHNOLOGY , 2014

Control volume finite difference analysis of thetransient temperature distributions and associate... more Control volume finite difference analysis of thetransient temperature distributions and associated induced thermal stresses in Ghana Research Reactor-1 (GHARR-1) Reactor vessel due to coolant heating has beenmodelledto investigatethe structural integrity of thevessel after 15 years of operation.Theinduced thermal stresses within the thickness of the cylindrical reactor vessel were also solved analytically using Bessel transforms. Computational flow chart translating into numericalalgorithms wasdeveloped and implemented inMATLAB to generate data for analysis and simulations. Results obtained indicated that both temperature and thermal stress distributions were below the limits imposed by the vessel material composition (melting point of 933 K and allowable yield stress of 480 MPa). The low level of induced thermal stresses indicated that the structural integrity of the reactor vessel has been maintained to forestall the incidence of crack propagation and other premature failure modes over the operational period.

Research paper thumbnail of Finite difference analysis of the transient temperature profile within GHARR-1 fuel element

Annals of Nuclear Energy, 2014

Mathematical model of the transient heat distribution within Ghana Research Reactor -1 (GHARR-1) ... more Mathematical model of the transient heat distribution within Ghana Research Reactor -1 (GHARR-1) fuel element and related shutdown heat generation rates have been developed.

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

Science and Technology of Nuclear Installations, 2022

e damage induced pressurized thermal shock (PTS) may pose to a reactor pressure vessel (RPV) is a... more e 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 PTSinduced 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 twoloop 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. e material fracture toughness of the vessel is computed based on the master curve method with experimental fracture toughness data. e 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. is study provides a useful basis for future studies on anticipated transient-induced crack propagation and remaining service life prediction of ageing reactor pressure vessels.

Research paper thumbnail of Numerical Fracture Analysis of a Reactor Pressure Vessel based on Abaqus-FRANC3D Co-simulation Method

Procedia Structural Integrity, 2022

The reactor pressure vessel (RPV) is subjected to many cyclic loadings critical of which is the i... more The reactor pressure vessel (RPV) is subjected to many cyclic loadings critical of which is the integrity risk posed by pressurized thermal shock (PTS) loadings induced by one of the most frequent anticipated operational occurrences-inadvertent operation of the safety injection system (SIS). In this paper, the fracture mechanics analysis for an ageing pressurized water reactor subjected to PTS induced by inadvertent actuation of the SIS is performed using a proposed simplified Abaqus-FRANC3D co-simulation method. A three-dimensional (3-D) finite element half-symmetric model of a typical pressured water reactor RPV is used to perform thermal-mechanical stress coupling analysis. 3-D fracture mechanic submodel with an assumed surface crack is created for the computation of the transient integrity parameter-stress intensity factor (SIF), using M-integral approach coupled in the proposed co-simulation method. Subsequently ASME method is used to evaluate the vessel's material fracture toughness. Finally, the SIFs obtained with the simplified co-simulation method is compared with the conventional virtual crack-closure technique (VCCT), and the result show good agreement. This work serves as a useful reference for fast crack propagation and life prediction analysis of ageing PWR RPVs.

Research paper thumbnail of Integrity Evaluation of a Reactor Pressure Vessel Based on a Sequential Abaqus-FRANC3D Simulation Method

Science and Technology of Nuclear Installations, 2021

e safety-risk pressurized thermal shock (PTS) have on a reactor pressure vessel (RPV) is one of t... more e safety-risk pressurized thermal shock (PTS) have on a reactor pressure vessel (RPV) is one of the most important studies for the lifetime ageing management of a reactor. Several studies have investigated PTS induced by postulated accidents and other anticipated transients. However, there is no study that analyzes the effect of PTS induced by one of the most frequent anticipated operational occurrences-inadvertent operation of the safety injection system. In this paper, a sequential Abaqus-FRANC3D simulation method is proposed to study the integrity status of an ageing pressurized water reactor subjected to PTS induced by inadvertent actuation of the safety injection system. A sequential thermal-mechanical coupling analysis is first performed using a three-dimensional reactor pressure vessel finite element model (3D-FEM). A linear elastic fracture mechanics submodel with a postulated semielliptical surface crack is then created from the 3D-FEM. Subsequently, the submodel is used to evaluate the stress intensity factors based on the M-integral approach coupled within the proposed simulation method. Finally, the stress intensity factors (SIFs) obtained using the proposed method are compared with the conventional extended finite element method approach, and the result shows a good agreement. e maximal thermomechanical stress concentration was observed at the inlet nozzleinner wall intersection. In addition, e ASME fracture toughness of the reactor vessel's steel compared with SIFs show that the PTS event and crack configuration analysed may not pose a risk to the integrity of the RPV. is work serves as a critical reference for the ageing management and fatigue life prediction of reactor pressure vessels.

Research paper thumbnail of Control Volume Finite Difference Analysis of Thermal Stresses Induced in GHARR-1 Reactor Vessel by Coolant Heating

INTERNATIONAL JOURNAL OF SCIENTIFIC & TECHNOLOGY , 2014

Control volume finite difference analysis of thetransient temperature distributions and associate... more Control volume finite difference analysis of thetransient temperature distributions and associated induced thermal stresses in Ghana Research Reactor-1 (GHARR-1) Reactor vessel due to coolant heating has beenmodelledto investigatethe structural integrity of thevessel after 15 years of operation.Theinduced thermal stresses within the thickness of the cylindrical reactor vessel were also solved analytically using Bessel transforms. Computational flow chart translating into numericalalgorithms wasdeveloped and implemented inMATLAB to generate data for analysis and simulations. Results obtained indicated that both temperature and thermal stress distributions were below the limits imposed by the vessel material composition (melting point of 933 K and allowable yield stress of 480 MPa). The low level of induced thermal stresses indicated that the structural integrity of the reactor vessel has been maintained to forestall the incidence of crack propagation and other premature failure modes over the operational period.

Research paper thumbnail of Finite difference analysis of the transient temperature profile within GHARR-1 fuel element

Annals of Nuclear Energy, 2014

Mathematical model of the transient heat distribution within Ghana Research Reactor -1 (GHARR-1) ... more Mathematical model of the transient heat distribution within Ghana Research Reactor -1 (GHARR-1) fuel element and related shutdown heat generation rates have been developed.