Pressurized Thermal Shock analysis of the reactor pressure vessel

Trampus, Péter

doi:10.1016/j.prostr.2018.12.204

Cited by 11 publications

(7 citation statements)

References 2 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The influence of cladding was almost negligible in this case. This result verifies our previous measurements on thermally treated cladded blocks [31].…”

Section: Yoke Asupporting

confidence: 92%

“…The influence of cladding was almost negligible in this case. This result verifies our previous measurements on thermally treated cladded blocks [31]. It should be mentioned, however, that the registered curve is rather noisy if measurement is performed through cladding.…”

Section: Yoke Asupporting

confidence: 90%

“…The reason is that plasticity and the elevated fracture toughness of cladding can provide additional strength to the pressure wall and this process can justify an extended reactor lifetime [30]. In addition, in thermal expansion coefficients, significant differences can be found with respect to pressure vessel base metals, which can cause a stress peak [31]. This is the so-called pressurized thermal shock and it is a potential risk of interfacial crack initiation and propagation.…”

Section: Introductionmentioning

confidence: 99%

See 2 more Smart Citations

Magnetic Investigation of Cladded Nuclear Reactor Blocks

et al. 2022

View full text Add to dashboard Cite

The wall, made of ferromagnetic steel, of a nuclear reactor pressure vessel is covered by an austenitic (very weakly ferromagnetic) cladding. In this work, we investigated how the base material and the cladding can be inspected separately from each other by nondestructive magnetic measurements. It was found that with the proper choice of the magnetizing yoke, these two different materials could be measured independently of each other. The effect of the yoke’s size was studied by the numerical simulation of magnetic flux, pumped into the material during magnetic measurements. Measurements were performed by two different sizes of yokes on pure base material, on base material under cladding and on cladding itself. Experiments verified the results of the simulation. Our results can help for the future practical application of magnetic methods in the regular inspection of nuclear power plants.

show abstract

“…The influence of cladding was almost negligible in this case. This result verifies our previous measurements on thermally treated cladded blocks [31].…”

Section: Yoke Asupporting

confidence: 92%

Section: Yoke Asupporting

confidence: 90%

Section: Introductionmentioning

confidence: 99%

See 1 more Smart Citation

Magnetic Investigation of Cladded Nuclear Reactor Blocks

et al. 2022

View full text Add to dashboard Cite

show abstract

“…e reactor pressure vessel (RPV), a critical component of a pressurized power reactor (PWR), is subjected to high temperature, high pressure, and fast neutrons emitted from the heat-generating core during its operations [1]. A key safety principle in the operations of an ageing PWR is to maintain the safety margins of the nonreplaceable RPV [2]. Furthermore, a critical-size defect in a vessel with a significant loss in ductility can grow rapidly under some reactor operating conditions leading to damage or, in severe cases, failure.…”

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

“…In addition, the re-licensing regime of ageing NPPs is largely premised on the integrity assessment of critical components such as the reactor pressure vessel (RPV). e structural integrity of the RPV is a key safety priority in the operation of ageing pressurized water reactor (PWR) NPPs since it technically determines the feasible lifetime of the reactor [2]. Furthermore, under some NPP transient conditions, a small defect of the size of the nondestructive testing limit in an ageing RPV may rapidly grow leading to damage or failure.…”

Section: Introductionmentioning

confidence: 99%

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

Annor-Nyarko

Xia

2021

Science and Technology of Nuclear Installations

View full text Add to dashboard Cite

The 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. The maximal thermomechanical stress concentration was observed at the inlet nozzle-inner wall intersection. In addition, The 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. This work serves as a critical reference for the ageing management and fatigue life prediction of reactor pressure vessels.

show abstract

Pressurized Thermal Shock analysis of the reactor pressure vessel

Cited by 11 publications

References 2 publications

Magnetic Investigation of Cladded Nuclear Reactor Blocks

Magnetic Investigation of Cladded Nuclear Reactor Blocks

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

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

Contact Info

Product

Resources

About