Creep-Damage Analysis: Comparison Between Coupled and Uncoupled Models

2000 ◽  
Vol 122 (4) ◽  
pp. 408-412 ◽  
Author(s):  
S. Bhandari ◽  
X. Feral ◽  
J.-M. Bergheau ◽  
G. Mottet ◽  
P. Dupas ◽  
...  
Keyword(s):  
Numerical Simulation ◽  
Internal Pressure ◽  
Validation Study ◽  
Pressure Vessel ◽  
Creep Damage ◽  
Damage Analysis ◽  
Hypothetical Accident ◽  
Computer Codes ◽  
Damage Theory ◽  
Reactor Pressure

Numerical simulation of creep rupture of a reactor pressure vessel in a severe hypothetical accident needs to perfectly take account of interactions between creep phenomena and damage. The continuous damage theory enables to formulate models strongly coupling elasto-visco-plasticity and damage. Such models have been implemented in various computer codes and, in particular, in ASTER at Electricite´ de France, CASTEM 2000 at Commissariat a` l’Energie Atomique and SYSTUS+® at SYSTUS International. The objective of this paper is to present briefly a validation study of the three different numerical implementations and to compare the coupled approach to an uncoupled one on an example of a cylinder of the program “RUPTHER,” under internal pressure and heated to a temperature of 700°C. [S0094-9930(00)01004-0]

Negligible Creep Conditions for Mod 9Cr1Mo Steel

2006 ◽  
Author(s):  
Bernard Riou ◽  
Claude Escaravage ◽  
Robert W. Swindeman ◽  
Weiju Ren ◽  
Marie-The´re`se Cabrillat ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Creep Damage ◽  
Steel Grade ◽  
Reactor Pressure Vessel ◽  
Surveillance Program ◽  
9Cr1mo Steel ◽  
Grade 91 ◽  
Creep Regime ◽  
Very High ◽  
Reactor Pressure

Mod 9Cr1Mo Steel (grade 91) is one of the materials envisaged for the Reactor Pressure Vessel of Very High Temperature Reactors. To avoid the implementation of a surveillance program covering the monitoring of the creep damage throughout the whole life of the reactor, it is recommended to operate the Reactor Pressure Vessel in the negligible creep regime. In this paper, the background of negligible creep criteria available in nuclear Codes is first recalled and their limitations analyzed. Then, guidance for deriving criteria more appropriate for mod 9Cr1Mo steel is provided. Finally, R&D actions in the U. S. and France to support the new approaches are discussed and recommended.

Investigation on Structural Behaviors of Reactor Pressure Vessel With the Effects of Critical Heat Flux and Internal Pressure

2016 ◽  
Vol 139 (2) ◽  
Author(s):  
Jianfeng Mao ◽  
Jianwei Zhu ◽  
Shiyi Bao ◽  
Lijia Luo ◽  
Zengliang Gao
Keyword(s):  
Heat Flux ◽  
Internal Pressure ◽  
Critical Heat Flux ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Wide Range ◽  
Depth Analysis ◽  
Structural Behaviors ◽  
Reactor Pressure

The so-called “in-vessel retention (IVR)” is a severe accident management strategy, which is widely adopted in most advanced nuclear power plants. The IVR mitigation is assumed to be able to arrest the degraded melting core and maintain the structural integrity of reactor pressure vessel (RPV) within a prescribed hour. Essentially, the most dangerous thermal–mechanical loads can be specified as the combination of critical heat flux (CHF) and internal pressure. The CHF is the coolability limits of RPV submerged in water (∼150 °C) and heated internally (∼1327 °C), it results in a sudden transition of boiling crisis from nucleate to film boiling. Accordingly, from a structural integrity perspective, the RPV failure mechanisms span a wide range of structural behaviors, such as melt-through, creep damage, plastic deformation as well as thermal expansion. Furthermore, the geometric discontinuity of RPV created by the local material melting on the inside aggravates the stress concentration. In addition, the internal pressure effect that usually neglected in the traditional concept of IVR is found to be having a significant impact on the total damage evolution, as indicated in the Fukushima accident that a certain pressure (up to 8.0 MPa) still existed inside the RPV. This paper investigates structural behaviors of RPV with the effects of CHF and internal pressure. In achieving this goal, a continuum damage mechanics (CDM) based on the “ductility exhaustion” is adopted for the in-depth analysis.

Creep Evaluation for a PWR Reactor Pressure Vessel Lower Head Under Severe Accident Conditions Considering Sustained Internal Pressure

2018 ◽  
Author(s):  
YongJian Gao ◽  
Ming Cao ◽  
YinBiao He
Keyword(s):  
Internal Pressure ◽  
Failure Modes ◽  
Structural Integrity ◽  
Creep Damage ◽  
Pressure Vessels ◽  
Mitigation Strategies ◽  
Field Analysis ◽  
Severe Accident ◽  
Lower Head ◽  
Reactor Pressure

In-Vessel Retention (IVR) is one of appropriate severe accident mitigation strategies for AP1000 Nuclear Power Plant (NPP), and assurance of prevention against to thermal failure and structural failure of Reactor Pressure Vessels (RPV) is the prerequisite of IVR. A Finite Element Model fora RPV considering lower head melting was established, the creep calculation was carried out after the temperature field analysis, and the stress-strain responses for different times were obtained. By means of choosing representative evaluation sections and applying the Accumulative Damage Theory based on Larson-Miller Parameter, the Creep Damage calculations and evaluations were conducted. The results showed that the failure modes associated with creep rupture would not happen under IVR condition when a certain amount of internal pressure sustained. The approaches employed in this paper could be utilized in structural integrity evaluation of RPV under IVR for other new type NPPs.

scholarly journals Creep and Damage Analysis of Reactor Pressure Vessel Considering Core Meltdown Scenario

2015 ◽  
Vol 130 ◽  
pp. 1148-1161 ◽  
Author(s):  
J.F. Mao ◽  
J.W. Zhu ◽  
S.Y. Bao ◽  
L.J. Luo ◽  
Z.L. Gao
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Damage Analysis ◽  
Core Meltdown ◽  
Reactor Pressure

scholarly journals Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity

2010 ◽  
Vol 2010 ◽  
pp. 1-11 ◽  
Author(s):  
Siniša Šadek ◽  
Srđan Špalj ◽  
Bruno Glaser
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Primary System ◽  
Coolant Pump ◽  
The Core ◽  
Hydraulic Behaviour ◽  
Computer Codes ◽  
Vessel Integrity ◽  
Reactor Pressure

RELAP5/SCDAPSIM and MAAP4 are two widely used severe accident computer codes for the integral analysis of the core and the reactor pressure vessel behaviour following the core degradation. The objective of the paper is the comparison of code results obtained by application of different modelling options and the evaluation of influence of thermal hydraulic behaviour of the plant on core damage progression. The analysed transient was postulated station blackout in NPP Krško with a leakage from reactor coolant pump seals. Two groups of calculations were performed where each group had a different break area and, thus, a different leakage rate. Analyses have shown that MAAP4 results were more sensitive to varying thermal hydraulic conditions in the primary system. User-defined parameters had to be carefully selected when the MAAP4 model was developed, in contrast to the RELAP5/SCDAPSIM model where those parameters did not have any significant impact on final results.

Study on Heavy Section Forging Steel for Reactor Pressure Vessel Applying Simulation Methods

2011 ◽  
Vol 194-196 ◽  
pp. 194-200
Author(s):  
Xue Xia Xu ◽  
Bing Zhe Bai ◽  
Wei You
Keyword(s):  
Numerical Simulation ◽  
Heat Treatment ◽  
Pressure Vessel ◽  
Small Sample ◽  
Reactor Pressure Vessel ◽  
Simulation Methods ◽  
Experimental Steel ◽  
Heavy Section ◽  
Numerical Simulation Methods ◽  
Reactor Pressure

The composition and heat treatment of heavy section forging steel for reactor pressure vessel were optimized applying physical and numerical simulation methods, including numerical simulation to calculate temperature distribution during quench of model forging, artificial neural network to predict CCT diagrams of steels and small sample control cooling to simulate specific heat treatment. And the influence of compositon and heat treatment on microstructure and properties were discussed. Results showed that the experimental steel obtained satisfactory properties based on optimization of chemical composition and heat treatment. It is estimated that the hardenability and temper stability of experimental steel were improved by tungsten alloyment and higher temperature temper was good for superior microstructure, proper strength and better toughness. In the present work, application of simulation methods is proved to be reasonable for study on heavy section forging steel.

scholarly journals FRACTURE MECHANICS UNCERTAINTY ANALYSIS IN THE RELIABILITY ASSESSMENT OF THE REACTOR PRESSURE VESSEL: (2D) SUBJECTED TO INTERNAL PRESSURE

2016 ◽  
Vol 18 (2) ◽  
pp. 55
Author(s):  
Entin Hartini ◽  
Roziq Himawan ◽  
Mike Susmikanti
Keyword(s):  
Fracture Mechanics ◽  
Uncertainty Analysis ◽  
Internal Pressure ◽  
Pressure Vessel ◽  
Probabilistic Method ◽  
Latin Hypercube Sampling ◽  
Reactor Pressure Vessel ◽  
J Integral ◽  
Deterministic Method ◽  
Reactor Pressure

ABSTRACT 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 m0.5. Meanwhile, the SIF generated by deterministic method is 100.876 MPa m0.5. Keywords: Uncertainty analysis, fracture mechanics, LHS, FEM, reactor pressure vessels   ABSTRAK ANALISIS KETIDAKPASTIAN FRACTURE MECHANIC PADA EVALUASI KEANDALAN BEJANA TEKAN REAKTOR: 2D DENGAN BEBAN INTERNAL PRESSURE. Bejana tekan reaktor (RPV) merupakan pressure boundary dalam reaktor tipe PWR yang berfungsi untuk mengungkung material radioaktif  yang dihasilkan pada proses reaksi berantai. Maka dari itu integritas bejana tekan reaktor harus senantiasa terjamin baik reaktor dalam keadaan operasi normal, maupun kecelakaan. Dalam melakukan analisis integritas RPV, khususnya yang berkaitan dengan pecahnya bejana tekan reaktor akibat adanya retak dilakukan analisis secara fracture mechanics. Adanya ketidakpastian input seperti sifat mekanik bahan, lingkungan fisik, dan input pada data, maka dalam melakukan analisis keandalan tidak hanya dilakukan secara deterministik saja. Tujuan dari penelitian ini adalah melakukan analisis ketidakpastian input pada perhitungan fracture mechanik pada evaluasi keandalan bejana tekan reaktor PWR. Pendekatan untuk karakter random dari kuantitas input menggunakan  teori probabilistik. Analisis fracture mechanics dilakukan berdasarkan metode elemen hingga (FEM) menggunakan perangkat lunak MSC MARC. Analisis ketidakpastian input dilakukan berdasarkan probability density function dengan Latin Hypercube Sampling (LHS) menggunakan python script. Output dari MSC MARC adalah nilai J-integral untuk mendapatkan nilai stress intensity factor pada evaluasi keandalan bejana tekan reactor 2D. Dari hasil perhitungan dapat disimpulkan bahwa SIF probabilistik lebih dulu mencapai nilai batas fracture tougness  dibanding  SIF deterministik. SIF yang dihasilkan dengan metode probabilistik adalah 105,240 MPa m0,5. Sedangkan SIF metode deterministik adalah 100,876 MPa m0,5. Kata kunci: Analisis ketidakpastian, fracture mechanics, LHS, FEM, bejana tekan reaktor

Sign in / Sign up

Export Citation Format

Share Document