Reactor pressure vessel integrity assessment by probabilistic fracture mechanics – A plant specific analysis

Chen, Bo–Yi; Huang, Chun-Ming; Chou, Hsoung-Wei; Lin, Hsiu Mei; Liu, Ru-Feng; Weng, Tung-Li; Chang, Han-Jou

doi:10.1016/j.ijpvp.2013.10.011

Cited by 6 publications

(7 citation statements)

References 3 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…As the time passes, the functional capabilities of structures, systems and components (SSCs) deteriorate by ageing [10,11]. Ageing is a process in which the reliability of any equipment depreciates, and hence the likelihood of that equipment to fail increases [12].…”

Section: Introduction mentioning

confidence: 99%

Physical Ageing of The Research Reactor Core Structural Materials Due To Neutron Irradiation Exposure: A Review

Purba

2017

Jurnal Pengembangan Energi Nuklir

View full text Add to dashboard Cite

PHYSICAL AGEING OF THE RESEARCH REACTOR CORE STRUCTURAL MATERIALS DUE TO NEUTRON IRRADIATION EXPOSURE: A REVIEW.A research reactor (RR) is a nuclear reactor that has function to generate and utilize neutron flux and radiation ionization for research purposes and industrial applications. More than 60% of current operating RRs have been operated for 30 years or more. As the time passes, the functional capabilities of structures, systems and components (SSCs) of those RRs deteriorate by physical ageing, which can be caused by neutron irradiation exposure such as irradiation induced dislocation and microstructural changes. To extend the lifetime and/or to avoid unplanned outages, ageing on the safety related SSCs of RRs need to be properly managed. An ageing management is a strategy to engineer, operate, maintenance, and control SSC degradation within acceptable limits. The purpose of this study is to review physical ageing of the core structural materials of the RRs caused by neutron irradiation exposure. In order to achieve this objective, a wide range of literatures are reviewed. Comprehensive discussions on irradiation behaviors are limited only on reactor vessel and core support structure materials made from zirconium and beryllium as well as their alloys, which are widely used in RRs. It is found that the stability of the mechanical properties of zirconium and beryllium as well as their alloys was mostly affected by the neutron fluences and temperatures. ABSTRAK PENUAAN MATERIAL STRUKTUR TERAS REAKTOR RISET SECARA FISIK YANG DISEBABKAN OLEH PAPARAN RADIASI NEUTRON: SEBUAH KAJIAN.Reaktor riset adalah reaktor nuklir yang difungsikan untuk pembangkitan dan pemanfaatan fluks neutron dan ionisasi radiasi untuk tujuan penelitian dan aplikasi industri. Lebih dari 60% reaktor riset yang ada saat ini telah beroperasi selama 30 tahun atau lebih. Seiring dengan berjalannya waktu, kemampuan struktur, sistem dan komponen (SSK) tentunya mengalami penurunan yang dikenal dengan penuaan fisik. Penuaan fisik ini dapat disebabkan oleh karena adanya paparan radiasi neutron seperti irradiasi yang menyebabkan terjadinya dislokasi dan perubahan mikrostruktur material. Untuk memperpanjang umur reaktor dan untuk mengurangi terjadinya pemadaman yang tidak diinginkan maka perlu dilakukan manajemen penuaan SSK, khususnya yang terkait dengan keselamatan operasi. Tujuan dari penelitian ini adalah untuk mengkaji penuaan fisik yang terjadi pada material struktur teras reaktor riset yang disebabkan oleh paparan iradiasi neutron. Untuk mencapai tujuan ini maka dilakukan kajian pada berbagai macam pustaka yang terkait. Pembahasan yang komprehensif tentang sifat-sifat iradiasi dibatasi hanya pada bejana reaktor dan material struktur pendukung teras yang terbuat dari zirconium dan beryllium dan senyawanya yang umum digunakan oleh reaktor riset. Dari kajian yang telah dilakukan, ditemukan bahwa stabilitas sifat-sifat mekanik zirconium dan beryllium serta senyawanya sangat dipengaruhi oleh fluent dan temperatur neutron. Kata kunci: Reaktor riset, p...

show abstract

Section: Introduction mentioning

confidence: 99%

Physical Ageing of The Research Reactor Core Structural Materials Due To Neutron Irradiation Exposure: A Review

Purba

2017

Jurnal Pengembangan Energi Nuklir

View full text Add to dashboard Cite

show abstract

“…Probabilistic of fracture mechanic on hot piping was analyzed especially in aging PWR [5]. In the same way, the reactor safety were also analyzed with the probabilistic of fracture mechanics [6]. The other research on mechanical fracture analysis of pressure vessel was analyzed for the brittle steel SA-508 CL 3 in the pressure vessel of the reactor [7].…”

Section: Introductionmentioning

confidence: 99%

Evaluation on Mechanical Fracture of PWR Pressure Vessel and Modeling Based on Neural Network

Susmikanti

Himawan²,

Hafid³

et al. 2016

Tri Dasa Mega

View full text Add to dashboard Cite

EVALUATION ON MECHANICAL FRACTURE OF PWR PRESSURE VESSEL AND MODELING BASED ON NEURAL NETWORK. The important component of the PWR is a pressure vessel. The material resistance in the pressure vessel needs to be evaluated. One way of evaluation is by the mechanical fracture analysis. The modeling needs to know the phenomena of the analysis result in general. A number of researches have been completed on the calculation of mechanical fracture in the pressure vessel with an internal load. The mechanical fracture was modeled using a neural network approach. In relation to the material resistance of the pressure vessel, which is used in PWR AP1000, the material must be evaluated because of the effect of the load. The modeling is needed to predict the effect of the load. The aim of this study is to evaluate the material resistance through mechanical fracture analysis because of the influence load on the pressure vessel on PWR AP1000. The material, which was observed, is SA 508. This analysis consists of the calculation of stress intensity factor and J-integral with some load at the crack propagation position. The fracture mechanic was analyzed by finite element simulation. The result of Stress Intensity factor and J-Integral was compared with fracture toughness to know the durability of the material. The modeling of J-Integral and Stress Intensity Factor were obtained for some load based on neural network approach. Keywords: Material resistance, mechanical fracture, neural network, PWR, pressure vessel, crack propagation. ABSTRAK EVALUASI FRAKTUR MEKANIK PADA BEJANA TEKAN PWR DAN PEMODELAN BERBASIS NEURAL NETWORK. Komponen penting dari PWR adalah bejana tekan. Ketahanan bahan di bejana tekan perlu dievaluasi. Salah satu cara adalah dengan analisis fraktur mekanik. Pemodelan diperlukan untuk mengetahui fenomena hasil analisis pada umumnya. Terdapat penelitian untuk perhitungan fraktur mekanik dalam bejana tekan dengan beban internal. Penelitian lain adalah hasil dari fraktur mekanik dimodelkan menggunakan pendekatan jaringan syaraf. Sehubungan dengan ketahanan material dari bejana tekan yang digunakan dalam PWR AP1000, bahan harus dievaluasi karena efek dari beban. Pemodelan diperlukan untuk memprediksi pengaruh beban pada bahan dalam bejana tekan. Tujuan dari penelitian ini adalah untuk mengevaluasi ketahanan material melalui analisis fraktur mekanik karena pengaruh beban pada bejana tekan. Bahan yang diamati, adalah SA 508. Analisis ini terdiri dari perhitungan faktor intensitas tegangan dan J-integral dengan beberapa beban pada posisi perambatan retak. Fraktur mekanik dianalisis dengan metode elemen hingga. Hasil faktor intensitas tegangan dan J-Integral dibandingkan dengan ketangguhan patah untuk mengetahui daya tahan material. Pemodelan J-Integral dan faktor intensitas stres diperoleh untuk beberapa beban berdasarkan jaringan saraf. Kata kunci: Ketahanan bahan, teknik patahan, jaringan syaraf, PWR, bejana tekan, perambatan retak.

show abstract

“…In order to minimize the probability of radioactive releases into the environment, nuclear power plant confinement is designed with several barriers. One of the barrier is reactor pressure vessel [1][2]. The reactor pressure vessel is a pressure boundary in the pressurized water reactor (PWR) type, which serves as third barrier in the confinement of radioactive material resulting from chain reactions.…”

Section: Introductionmentioning

confidence: 99%

“…The reactor pressure vessel is a pressure boundary in the pressurized water reactor (PWR) type, which serves as third barrier in the confinement of radioactive material resulting from chain reactions. Therefore, the integrity of the reactor pressure vessel must be guarant either in a normal operation or accident conditions [1][2][3].…”

Section: Introductionmentioning

confidence: 99%

“…In practice, for analizing the material reliability or predicting the possibility of material failure, the initial conditions of the system are not known precisely due to uncertainties. To accomodate the uncertainties, a probabilistic method can be used, therefore the calculation of safety margin take into account the uncertainty of input that will affect the output value [1,2,4,7]. Uncertainty analysis includes physical variables of the loading conditions, such as internal pressure and material properties [7][8][9][10][11].…”

Section: Introductionmentioning

confidence: 99%

See 1 more Smart Citation

Fracture Mechanics Uncertainty Analysis in the Reliability Assessment of the Reactor Pressure Vessel: (2d) Subjected to Internal Pressure

Hartini¹,

Himawan²,

Susmikanti³

2016

Tri Dasa Mega

View full text Add to dashboard Cite

FRACTURE MECHANICS UNCERTAINTY ANALYSIS IN THE RELIABILITY ASSESSMENT OF THE REACTOR PRESSURE VESSEL: (2D) SUBJECTED TO INTERNAL PRESSURE.The reactor pressure vessel (RPV) is a pressure boundary in the PWR type reactor which serves to confine radioactive material during chain reaction process. The integrity of the RPV must be guaranteed either in a normal operation or accident conditions. In analyzing the integrity of RPV, especially related to the crack behavior which can introduce break to the reactor pressure vessel, a fracture mechanic approach should be taken for this assessment. The uncertainty of input used in the assessment, such as mechanical properties and physical environment, becomes a reason that the assessment is not sufficient if it is perfomed only by deterministic approach. Therefore, the uncertainty approach should be applied. The aim of this study is to analize the uncertainty of fracture mechanics calculations in evaluating the reliability of PWR`s reactor pressure vessel. Random character of input quantity was generated using probabilistic principles and theories. Fracture mechanics analysis is solved by Finite Element Method (FEM) with MSC MARC software, while uncertainty input analysis is done based on probability density function with Latin Hypercube Sampling (LHS) using python script. The output of MSC MARC is a J-integral value, which is converted into stress intensity factor for evaluating the reliability of RPV's 2D. From the result of the calculation, it can be concluded that the SIF from probabilistic method, reached the limit value of fracture toughness earlier than SIF from deterministic method. The SIF generated by the probabilistic method is 105.240 MPa m 0.5 . Meanwhile, the SIF generated by deterministic method is 100.

show abstract

Reactor pressure vessel integrity assessment by probabilistic fracture mechanics – A plant specific analysis

Cited by 6 publications

References 3 publications

Physical Ageing of The Research Reactor Core Structural Materials Due To Neutron Irradiation Exposure: A Review

Physical Ageing of The Research Reactor Core Structural Materials Due To Neutron Irradiation Exposure: A Review

Evaluation on Mechanical Fracture of PWR Pressure Vessel and Modeling Based on Neural Network

Fracture Mechanics Uncertainty Analysis in the Reliability Assessment of the Reactor Pressure Vessel: (2d) Subjected to Internal Pressure

Contact Info

Product

Resources

About