Use of the Failure Assessment Diagram to Evaluate the Safety of the Reactor Pressure Vessel

2015 ◽  
Vol 137 (5) ◽  
Author(s):  
Mingya Chen ◽  
Feng Lu ◽  
Rongshan Wang
Keyword(s):  
Pressure Vessel ◽  
Failure Modes ◽  
Mechanical Load ◽  
Reactor Pressure Vessel ◽  
Plastic Collapse ◽  
Integrity Assessment ◽  
Failure Assessment Diagram ◽  
Single Criterion ◽  
Failure Assessment ◽  
Reactor Pressure

Analysis of multiple failure modes is the key element of the integrity evaluation of the nuclear reactor pressure vessel (RPV). While the simple single-criterion failure code provides the guidance for structural integrity, the guidance ignores the interaction between fast fracture and plastic collapse. In this paper, the differences between the reserve factor (RF) in the R6 two-criteria failure procedure and the safety coefficient (SC) in the single-criterion failure code were compared. Based on 3D finite element (FE) analyses, the option 3 failure assessment diagram (FAD) of the beltline of the RPV was established according to the R6 basic route and alternative approaches, respectively. Also, the nonconservation of the secondary stress correction parameter ρ was reviewed. In this paper, it was shown that the effect of crack sizes on the FAD is considered to be limited, and the influence of the thermal stress on the FAD is obvious in the transition region of the failure assessment curve (FAC). The FAD only considering the mechanical load encloses the FAD considering the thermal–mechanical load for the Lr smaller than 1, but it is contrary when the Lr is bigger than 1. It is not enough to just satisfy the requirement in the IWB-3612 of the ASME code because the risk of plastic-collapse failure is ignored. And in this study, the maximum nonconservation of the fracture toughness RF is more than 7% due to the approximate value of ρ. Accordingly, the accurate method in the R6 procedure should be used in the integrity assessment of the RPV under the faulted transient.

Reactor Pressure Vessel Integrity Assessment: French Simplified Analysis Method Applied to Underclad Postulated Defect in Nozzle

2016 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Stéphane Vidard
Keyword(s):  
Fracture Mechanics ◽  
Fracture Risk ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Core Area ◽  
Integrity Assessment ◽  
Elastic Plastic ◽  
The Core ◽  
Simplified Method ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main concerns regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Fast fracture risk is the main potential damage considered in the integrity assessment of RPV. In France, deterministic integrity assessment for RPV vis-à-vis the brittle fracture risk is based on the crack initiation stage. As regards the core area in particular, the stability of an under-clad postulated flaw is currently evaluated under a Pressurized Thermal Shock (PTS) through a dedicated fracture mechanics simplified method called “beta method”. However, flaw stability analyses are also carried-out in several other areas of the RPV. Thence-forward performing uniform simplified inservice analyses of flaw stability is a major concern for EDF. In this context, 3D finite element elastic-plastic calculations with flaw modelling in the nozzle have been carried out recently and the corresponding results have been compared to those provided by the beta method, codified in the French RSE-M code for under-clad defects in the core area, in the most severe events. The purpose of this work is to validate the employment of the core area fracture mechanics simplified method as a conservative approach for the under-clad postulated flaw stability assessment in the complex geometry of the nozzle. This paper presents both simplified and 3D modelling flaw stability evaluation methods and the corresponding results obtained by running a PTS event. It shows that the employment of the “beta method” provides conservative results in comparison to those produced by elastic-plastic calculations for the cases here studied.

Upgrading the Operational Safety of Nuclear Power Plants Through the TACIS Nuclear Safety Assistance Programme

2004 ◽  
Author(s):  
M. Bie`th ◽  
R. Ahlstrand ◽  
C. Rieg ◽  
P. Trampus
Keyword(s):  
European Commission ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Safety ◽  
Reactor Pressure Vessel ◽  
Nuclear Safeguards ◽  
Integrity Assessment ◽  
Reactor Pressure

The European Union’ TACIS programme was established for the New Independent States since 1991. One priority for TACIS funding is nuclear safety. The European Commission has made available a total of € 944 million for nuclear safety programmes covering the period 1991–2003. The TACIS nuclear safety programme is devoted to the improvement of the safety of Soviet designed nuclear installations in providing technology and safety culture transfer. The Joint Research Center (JRC) of the European Commission is carrying out works in the following areas: • On-Site Assistance for TACIS Nuclear Power Plants; • Design Safety and Dissemination of TACIS results; • Reactor Pressure Vessel Embrittlement for VVER in Russia and Ukraine; • Regulatory Assistance; • Industrial Waste Management and Nuclear Safeguards. This paper gives an overview of the Scientific and Technical support that JRC is providing for the programming and the implementation of the TACIS nuclear safety programmes. In particular, two new projects are being implemented to get an extensive understanding of the VVER reactor pressure vessel embritttlement and integrity assessment.

Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Direct Coupling Between Probabilistic and Finite-Element Codes to Model Sensitivity to Key Thermo-Hydraulic Variability

2004 ◽  
Author(s):  
Etienne de Rocquigny ◽  
Yoan Chevalier ◽  
Silvia Turato ◽  
Eric Meister
Keyword(s):  
Finite Element ◽  
Pressure Vessel ◽  
Mechanical Model ◽  
Analytical Approach ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Direct Coupling ◽  
Integrity Assessment ◽  
Key Variables ◽  
Reactor Pressure

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions such as loss-of-coolant accident (LOCA) is a major safety concern. Besides conventional deterministic calculations to justify as a nuclear operator the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Probabilistic analyses become most interesting when some key variables, albeit conventionally taken at conservative values, can be modelled more accurately through statistical variability. In the context of low failure probabilities, this requires however a specific coupling effort between a specific probabilistic analysis method (e.g. Form-Sorm method) and the thermo-mechanical model to be reasonable in computing time. In this paper, the variability of a key variable — the mid-transient cooling temperature, tied to a climate-dependent tank — has been modelled, in some flaw configurations (axial sub-clad) for a French vessel. In a first step, a simplified analytical approach was carried out to assess its sensitivity upon the thermo-mechanical phenomena; hence, a direct coupling had to be implemented to allow a probabilistic calculation on the finite-element mechanical model, taking also into account a failure event properly defined through minimisation of the instantaneous failure margin during the transient. Comparison with the previous (indirectly-coupled) studies and the simplified analytical approach is drawn, demonstrating the interest of this new modelling effort to understand and order the sensitivity of the probability of crack initiation to the key variables. While being noticeable in the cases studied, sensitivity to the safety injection temperature variability proves to be less than the choice of the toughness model. Finally, regularity of the thermo-mechanical model is evidenced by the coupling exercise, suggesting that a modified response-surface based method could replace direct coupling for further investigation.

Deterministic and Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Benchmark Comparisons

2003 ◽  
Author(s):  
Silvia Turato ◽  
Vincent Venturini ◽  
Eric Meister ◽  
B. Richard Bass ◽  
Terry L. Dickson ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Safety Concern ◽  
Nuclear Regulatory Commission ◽  
Reactor Pressure Vessel ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Reactor Pressure

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions, such as loss-of-coolant accident (LOCA), is a major safety concern. Besides Conventional deterministic calculations to justify the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Since in the USA the probabilistic fracture mechanics analyses are accepted by the Nuclear Regulatory Commission (NRC), a benchmark has been realized between EDF and Oak Ridge Structural Assessments, Inc. (ORSA) to compare the models and the computational methodologies used in respective deterministic and probabilistic fracture mechanics analyses. Six cases involving two distinct transients imposed on RPVs containing specific flaw configurations (two axial subclad, two circumferential surface-breaking, and two axial surface-braking flaw configurations) were defined for a French vessel. In two separate phases, deterministic and probabilistic, fracture mechanics analyses were performed for these six cases.

Using a Probabilistic Approach in the Brittle Fracture Deterministic Integrity Assessment of a Nuclear Reactor Pressure Vessel

2011 ◽  
Author(s):  
Romain Beaufils ◽  
Eric Meister ◽  
Emmanuel Ardillon
Keyword(s):  
Brittle Fracture ◽  
Pressure Vessel ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Probabilistic Method ◽  
Probabilistic Approach ◽  
Reactor Pressure Vessel ◽  
Deterministic Method ◽  
Integrity Assessment ◽  
Reactor Pressure

This work deals with the possibility of the life extension of nuclear power plants in France. The aim is to justify the resistance of the pressure vessel, which is non-replaceable. The brittle fracture deterministic integrity assessment of the nuclear Reactor Pressure Vessel (RPV) is based on the analysis of a flaw under the austenitic cladding of the RPV. The demonstration of the RPV resistance is controlled by the regulations. It is proposed here to use a probabilistic method by propagating uncertainties into the deterministic mechanical model in order to quantify conservatism of the deterministic method. The regulatory requirements must be respected and the purpose of the work presented here is thus to link the probabilistic result to the deterministic method.

Reactor Pressure Vessel Integrity in Light of the Evolution of Materials Science and Engineering

2005 ◽  
Vol 473-474 ◽  
pp. 287-292
Author(s):  
Péter Trampus
Keyword(s):  
Pressure Vessel ◽  
Structural Integrity ◽  
Materials Science ◽  
Reactor Pressure Vessel ◽  
Special Focus ◽  
Science And Engineering ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Materials Science And Engineering ◽  
Reactor Pressure

Structural integrity of the reactor pressure vessel of pressurized water reactors is one of the key safety issues in nuclear power operation. Integrity may be jeopardized during operational transients. The problem is compounded by radiation damage of the vessel structural materials. Structural integrity assessment as an interdisciplinary field is primarily based on materials science and fracture mechanics. The paper gives an overview on the service induced damage processes and associated changes of mechanical properties, the prediction of degradation and the assessment of the entire component against brittle fracture with a special focus on how the evolution of materials science and engineering has contributed to reactor vessel structural integrity assessment.

The Brittle Strength Assessment of WWER-1000 Reactor Pressure Vessel Nozzle With Cladding

2013 ◽  
Author(s):  
Andrii Oryniak ◽  
Sergii Radchenko ◽  
Igor Orynyak
Keyword(s):  
Pressure Vessel ◽  
Crack Surface ◽  
Reactor Pressure Vessel ◽  
Brittle Strength ◽  
Strength Assessment ◽  
Main Metal ◽  
Integrity Assessment ◽  
The Difference ◽  
Reactor Pressure

Brittle strength calculation of RPV nozzle is the central point of the integrity assessment of the reactor pressure vessel when extending its life. The important part of this calculation is a determination of the stress intensity factor, SIF, for the postulated crack of partly elliptical form in a nozzle under inner pressure, bending moments (from the main circulating pipe) and difference of temperatures. In this paper we use method of influence functions as the most convenient one for solution of similar tasks. Eight basic laws of the crack surface loading are introduced which account for real stress distribution in the depth and length direction of a crack including the jump of stress between cladding and main metal due to the difference in the thermal expansion factors. To determine the dimensionless SIF under chosen laws of loading were developed the FEM models of nozzle with crack of different ratios of axes. For all possible modes (regimes) of operation were carried the detailed calculations of the temperature field in the nozzle, which were used later for determining the stress state at each time point. The stress field defined in 120 discrete points of the crack surface was treated by the method of least squares for the presention as a linear combination of eight basic load laws with defined coefficients. The procedure for determination of the temperature brittle strength margin which employs the presentation of critical values of SIF (fracture toughness) in the exponential function form is described.

Ways for Extending the Reactor Pressure Vessel Lifetime

2008 ◽  
Author(s):  
Milan Brumovsky
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Low Leakage ◽  
Vessel Integrity ◽  
Direct Use ◽  
Reactor Pressure ◽  
Lifetime Extension

Reactor pressure vessels are components that usually determine the lifetime of the whole nuclear power plant and thus also its efficiency and economy. There are several ways how to ensure conditions for reactor pressure vessel lifetime extension, mainly: - pre-operational, like: • optimal design of the vessel; • proper choice of vessel materials and manufacturing technology; - operational, like: • application of low-leakage core; • increase of water temperature in ECCS; • insertion of dummy elements; • vessel annealing; • decrease of conservatism during reactor pressure vessel integrity assessment e.g. using direct use of fracture mechanics parameters, like “Master Curve” approach. All these ways are discussed in the paper and some qualitative as well as quantitative evaluation is given.

Nuclear Reactor Pressure Vessel Integrity Assessment: Enhancement Provided by 3D Modeling Flaw Stability Calculations

2013 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Henriette Churier
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Potential Damage ◽  
Vessel Integrity ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main considerations regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Brittle fracture risk associated with the embrittlement of RPV steel in irradiated areas is the main potential damage. In France, deterministic integrity assessment for RPV is based on the crack initiation stage. The stability of an under-clad postulated flaw in the core area is currently evaluated under a Pressurized Thermal Shock (PTS) through a fracture mechanics simplified method. One of the axes of EDF’s implemented strategy for NPP lifetime extension is the improvement of the deterministic approach with regards to the input data and methods so as to reduce conservatisms. In this context, 3D finite element elastic-plastic calculations with flaw modelling have been carried out recently in order to quantify the enhancement provided by a more realistic approach in the most severe events. The aim of this paper is to present both simplified and 3D modelling flaw stability evaluation methods and the results obtained by running a small break LOCA event.

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