Risk Analysis of Reactor Pressure Vessels Considering Modeling-Induced Uncertainties

2016 ◽  
Vol 1 (4) ◽  
Author(s):  
Matthew E. Riley ◽  
William M. Hoffman
Keyword(s):  
Risk Analysis ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Simulation Models ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Modeling Process ◽  
Model Form ◽  
Physical Response ◽  
Reactor Pressure

Uncertainties in simulation models arise not only from the parameters that are used within the model, but also due to the modeling process itself—specifically the identification of a model that most accurately predicts the true physical response of interest. In risk-analysis studies, it is critical to consider the effect that all forms of uncertainty have on the overall level of uncertainty. This work develops an approach to quantify the effect of both parametric and model-form uncertainties. The developed approach is demonstrated on the assessment of the fatigue-based risk associated with a reactor pressure vessel subjected to a thermal shock event.

Measurements of Stress During Thermal Shock in Clad Reactor Pressure Vessel Material Using Time-Resolved In-Situ Synchrotron X-Ray Diffraction

2018 ◽  
Author(s):  
Sam Oliver ◽  
Chris Simpson ◽  
Andrew James ◽  
Christina Reinhard ◽  
David Collins ◽  
...  
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
X Ray Diffraction ◽  
Time Resolved ◽  
Vessel Steel ◽  
X Ray ◽  
Reactor Pressure

Nuclear reactor pressure vessels must be able to withstand thermal shock due to emergency cooling during a loss of coolant accident. Demonstrating structural integrity during thermal shock is difficult due to the complex interaction between thermal stress, residual stress, and stress caused by internal pressure. Finite element and analytic approaches exist to calculate the combined stress, but validation is limited. This study describes an experiment which aims to measure stress in a slice of clad reactor pressure vessel during thermal shock using time-resolved synchrotron X-ray diffraction. A test rig was designed to subject specimens to thermal shock, whilst simultaneously enabling synchrotron X-ray diffraction measurements of strain. The specimens were extracted from a block of SA508 Grade 4N reactor pressure vessel steel clad with Alloy 82 nickel-base alloy. Surface cracks were machined in the cladding. Electric heaters heat the specimens to 350°C and then the surface of the cladding is quenched in a bath of cold water, representing thermal shock. Six specimens were subjected to thermal shock on beamline I12 at Diamond Light Source, the UK’s national synchrotron X-ray facility. Time-resolved strain was measured during thermal shock at a single point close to the crack tip at a sample rate of 30 Hz. Hence, stress intensity factor vs time was calculated assuming K-controlled near-tip stress fields. This work describes the experimental method and presents some key results from a preliminary analysis of the data.

Thermal Hydraulic Analysis of Pressurized Thermal Shock for Loviisa NPP Using Apros Simulation Code

2018 ◽  
Author(s):  
Gintaras Žemulis ◽  
Pasi Junninen ◽  
Petri Kytömäki
Keyword(s):  
Power Plant ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Large Set ◽  
Energy Company ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure ◽  
Emergency Operating

Loviisa Nuclear Power Plant consists of two VVER-440 type pressurized water reactor units and it is owned by a Finnish energy company Fortum Power and Heat Oy. Only some years after Loviisa Unit 1 start-up, during early 1980’s, some reactor pressure vessel embrittlement by fast neutron irradiation was recognized. Embrittlement of the reactor pressure vessel materials increases the risk of a fast fracture under emergency core cooling conditions. This led to an extensive study of pressurized thermal shock (PTS) scenarios, which included experimental research, large set of probabilistic and deterministic safety analyses, code development and plant modifications to both lower overcooling transient probability and to mitigate it’s consequences. Current operating licenses for Loviisa 1 and 2 Units reactor pressure vessels were applied by Fortum in 2012 to cover operation up to the years 2027 and 2030, respectively. The operating licenses were granted as applied by the Finnish Radiation and Nuclear Safety Authority (STUK) according to application. However a know-how of PTS scenarios shall be maintained. This becomes even more important in a light of any big power plant automation modernization projects targeting also safety systems. Currently Fortum is involved in an automation renewal project also known under its ELSA-project name. The project is scheduled to be ready in 2018. The project targets at replacing parts of old analogue automation with new digital systems and also extending their diversity by introducing new safety functions for emergency management scenarios. As a result, changes to some emergency operating procedures are inevitable. Due to the changes to emergency operating procedures PTS-scenarios for Loviisa NPP have been also updated. In 2016 Fortum launched PTS-project in connection with ELSA-project. Previous PTS-scenario related thermal hydraulic analyses were based on the original 80’s decade analyses, which were subsequently revised many times following other modernization projects that took place at Loviisa NPP. In this new round of PTS-analyses the most important cases were recalculated using Apros (Advanced Process Simulator). ELSA-project introduced changes are not only emergency procedures specific but also involve safety system related process and automation logic modifications. The changes affected therefore boundary conditions needed in the analyses. In total probabilistic safety analyses included over 160 different PTS-sequences. The most limiting sequences were selected for deterministic analyses. PTS-analyses were performed in co-operation with Platom Oy and plant owner Fortum Power and Heat Oy. The main results of PTS-analyses are presented in the paper.

Failure Probability Assessment for a Boiling Water Reactor Pressure Vessel Under Low Temperature Over-Pressure Event

2012 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang ◽  
Bo-Yi Chen ◽  
Hsien-Chou Lin ◽  
Ru-Feng Liu
Keyword(s):  
Low Temperature ◽  
Failure Probability ◽  
Pressure Vessel ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Reactor Pressure

The fracture probability of a boiling water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been numerically analyzed using an advanced version of ORNL’s FAVOR code. First, a model of the vessel beltline region, which includes all shell welds and plates, is built for the FAVOR code based on the plant specific parameters of the reactor pressure vessel. Then, a novel flaw model which describes the flaw types of surface breaking flaws, embedded weld flaws and embedded plate flaws are simulated along both inner and outer vessel walls. When conducting the fracture probability analyses, a transient low temperature over-pressure event, which has previously been shown to be the most severe challenge to the integrity of boiling water reactor pressure vessels, is considered as the loading condition. It is found that the fracture occurs in the fusion-line area of axial welds, but with only an insignificant failure probability. The low through-wall cracking frequency indicates that the analyzed reactor pressure vessel maintains sufficient stability until either the end-of-license or for doubling of the present license of operation.

Accounting for the Effects of Inservice Inspection Flaw Depth Sizing Errors in Developing Flaw Size Distribution Tables

2009 ◽  
Author(s):  
S. R. Gosselin ◽  
F. A. Simonen
Keyword(s):  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Expert Elicitation ◽  
Flaw Size ◽  
Physical Models ◽  
Asme Code ◽  
Valid Comparison ◽  
Vessel Material ◽  
Reactor Pressure

Probabilistic fracture mechanics studies have addressed reactor pressure vessels that have high levels of material embrittlement. These calculations have used flaw size and density distributions determined from precise and optimized laboratory measurements made and validated with destructive methods as well as from physical models and expert elicitation. The experimental data were obtained from reactor vessel material samples removed from cancelled plants (Shoreham and the Pressure Vessel Research Users Facility (PVRUF)). Consequently, utilities may need to compare the numbers and sizes of reactor pressure vessel flaws identified by the plant’s inservice inspection program to the numbers and sizes of flaws assumed in prior failure probability calculations. This paper describes a method to determine whether the flaws in a particular reactor pressure vessel are consistent with the assumptions regarding the number and sizes of flaws used in other analyses. The approach recognizes that ASME Code Section XI examinations suffer from limitations in terms of sizing errors for very small flaws. Direct comparisons of a vessel specific flaw distribution with other documented flaw distributions would lead to pessimistic conclusions. This paper provides a method for a valid comparison that accounts for flaw sizing errors present in ASME Code Section XI examinations.

Computational studies on pressurized thermal shock in reactor pressure vessel

2021 ◽  
Vol 152 ◽  
pp. 107987
Author(s):  
Rakesh Chouhan ◽  
Anuj Kumar Kansal ◽  
Naresh Kumar Maheshwari ◽  
Avaneesh Sharma
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Computational Studies ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure
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