A Reassessment of Fracture-Safe Operating Criteria for Reactor Vessel Steels Based on Charpy-V Performance

1971 ◽  
Vol 93 (2) ◽  
pp. 247-258 ◽  
Author(s):  
F. J. Loss ◽  
J. R. Hawthorne ◽  
C. Z. Serpan
Keyword(s):  
Fracture Analysis ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Ratio Analysis ◽  
Test Procedures ◽  
Level Data ◽  
Surveillance Programs ◽  
Safe Operating ◽  
F 33 ◽  
Mechanical Constraint

Fracture-safe operating criteria for commercial nuclear pressure vessels based on fracture analysis diagram procedures and Charpy-V energy trends are reappraised with respect to the effects of thick section mechanical constraint and low Charpy - V shelf energies resulting from neutron irradiation. Comparisons of the Charpy-V test with the more definitive dynamic tear test procedures indicate the former to be an acceptable means of assessing the fracture toughness of A533-B steel. The mechanical constraint associated with 12-in. thicknesses of this steel suggests the addition of 70 F (39 C) to the existing criterion requiring vessel operation above NDT + 60 F (33 C). Ratio analysis diagram procedures are shown to be useful in interpreting Charpy-V shelf level data obtained from vessel surveillance programs in terms of critical toughness levels relating to brittle fracture.

Brittle Fracture Assessments on Piping Systems: MSOT Curves

2019 ◽  
Author(s):  
Ishita Chakraborty ◽  
Kannan Subramanian ◽  
Jorge Penso
Keyword(s):  
Brittle Fracture ◽  
Case Studies ◽  
Pressure Vessel ◽  
Iterative Process ◽  
Pressure Vessels ◽  
Temperature Changes ◽  
End Conditions ◽  
Piping Systems ◽  
Plant Personnel ◽  
Safe Operating

Abstract Brittle fracture assessments (BFAs) of pressure vessels based on API 579-1/ASME FFS-1, Section 3 procedures are frequently easier and more straightforward to implement in comparison to the BFAs on piping systems. Specifically, the development of the MSOT curves. This is due to the complexities involved in the piping systems due to the branch piping interactions, end conditions of piping systems such as nozzle flexibilities at the pressure vessel connections, temperature changes in the length of piping especially when the piping is significantly long as seen in flare header piping systems. MSOT curves that are alternatively used for MAT curves provide a better picture to the plant personnel in understanding the safe operating envelope. Development of MSOT curves is an iterative process and therefore involves significant number of piping stress analyses during their development. In this paper, an approach to develop the MSOT curves is discussed with two case studies that are of relevance to olefin plants.

Modeling and Performance Analysis of the Rolls-Royce Fuel Cell Systems Limited: 1 MW Plant

2010 ◽  
Vol 133 (2) ◽  
Author(s):  
Francesco Trasino ◽  
Michele Bozzolo ◽  
Loredana Magistri ◽  
Aristide F. Massardo
Keyword(s):  
Fuel Cell ◽  
Characteristic Point ◽  
Pressure Vessels ◽  
Performance Model ◽  
System Behavior ◽  
Cell Systems ◽  
And Performance ◽  
Definition Of ◽  
Component Models ◽  
Safe Operating

This paper is focused on the performance of the 1 MW plant designed and developed by Rolls-Royce Fuel Cell Systems Limited. The system consists of a two stage turbogenerator coupled with pressure vessels containing the fuel cell stack, internal reformer, cathode ejector, anode ejector, and off-gas burner. While the overall scheme is relatively simple, due to the limited number of components, the interaction between the components is complex and the system behavior is determined by many parameters. In particular, two important subsystems such as the cathode and the anode recycle loops must be carefully analyzed also considering their interaction with and influence on the turbogenerator performance. The system performance model represents the whole, and each physical component is modeled in detail as a subsystem. The component models have been validated or are under verification. The model provides all the operating parameters in each characteristic point of the plant and a complete distribution of thermodynamics and chemical parameters inside the solid oxide fuel cell (SOFC) stack and reformer. In order to characterize the system behavior, its operating envelope has been calculated taking into account the effect of ambient temperature and pressure, as described in the paper. Given the complexity of the system, various constraints have to be considered in order to obtain a safe operating condition not only for the system as a whole but also for each of its parts. In particular each point calculated has to comply with several constraints such as stack temperature distribution, maximum and minimum temperatures, and high and low pressure spool maximum rotational speeds. The model developed and the results presented in the paper provide important information for the definition of an appropriate control strategy and a first step in the development of a robust and optimized control system.

Assessment of Large-Scale Pressurized Thermal Shock Experiments Using the FAVOR Fracture Mechanics Computer Code

2004 ◽  
Author(s):  
T. L. Dickson ◽  
M. T. Kirk
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Large Scale ◽  
Major Element ◽  
Computer Code ◽  
Fracture Analysis ◽  
Pressure Vessels ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
Evaluation Project

Large-scale experiments of pressure vessels performed at the Oak National Laboratory (ORNL)1 in the mid 1980s validated the applicability of the linear-elastic fracture mechanics (LEFM) computational methodology for application to fracture analysis of reactor pressure vessels (RPVs) in nuclear power plants. The current federal regulations to insure that nuclear RPVs maintain their structural integrity, when subjected to transients such as pressurized thermal shock (PTS) events, were derived in the early-mid 1980s from a comprehensive computational methodology of which LEFM is a major element. Recently, the United States Nuclear Regulatory Commission (USNRC) has conducted the PTS re-evaluation project that has the objective to establish a technical basis for a potential relaxation to the current PTS regulations which could have profound implications for plant license-extension considerations. The PTS re-evaluation project has primarily consisted of the development and application of an updated risk-based computational methodology that has been implemented into the Fracture Analysis of Vessels: Oak Ridge (FAVOR) computer code. LEFM continues to be a major element of the updated computational methodology. As part of the PTS re-evaluation program, there has been an extensive effort to validate that FAVOR has an accurate implementation of the LEFM methodology. This effort has consisted of the successful benchmarking of thermal analysis, stress analysis, and LEFM fracture analysis results between FAVOR and ABAQUS, a commercial general-purpose finite element computer code that has fracture mechanics capabilities, for a range of transient descriptions. The NRC has also participated in international round-robin benchmarking exercises in which FAVOR-generated solutions to well-specified PTS problems have been compared to solutions generated by other research institutions. A more fundamental aspect of the ongoing validation of FAVOR is demonstration that FAVOR can be used to successfully predict the results of large-scale fracture experiments. The objective of this paper is to document the FAVOR analysis of the first large-scale pressurized thermal shock experiment (PTSE) performed at ORNL. Results of these analyses provide validation that FAVOR accurately predicts the cleavage fracture initiation of a long surface breaking flaw in a large-scale thick-walled pressure vessel.

A procedure of dynamic fracture analysis for reactor pressure vessels

1991 ◽  
Vol 129 (3) ◽  
pp. 287-292
Author(s):  
Yoshio Urabe ◽  
Kiminobu Hojo ◽  
Kinji Baba
Keyword(s):  
Dynamic Fracture ◽  
Fracture Analysis ◽  
Pressure Vessels ◽  
Reactor Pressure

Stress and Fracture Analysis of Ceramic Lined, Composite or Steel Jacketed Pressure Vessels

2004 ◽  
Author(s):  
John H. Underwood ◽  
Anthony P. Parker
Keyword(s):  
Room Temperature ◽  
Fracture Analysis ◽  
Crack Size ◽  
Pressure Vessels ◽  
Interface Pressure ◽  
Temperature Stresses ◽  
Critical Crack ◽  
Steel Jacket ◽  
Differential Thermal Expansion ◽  
Critical Crack Size

Stress and fracture analysis of ceramic-lined cannon pressure vessels is described, for a Si3N4 or SiC liner and A723 steel or carbon-epoxy jacket and with an initial residual interface pressure between liner and jacket. Room temperature stresses for a steel jacket over ceramic are similar to those for a carbon-epoxy jacket, but both radial and hoop jacket stresses can exceed typical carbon-epoxy strength values. Elevated temperature liner stresses are reduced for a carbon-epoxy jacket, due to the effective increase in interface pressure caused by differential thermal expansion. Critical crack size for brittle fracture is larger for Si3N4 than SiC due to lower liner stresses and higher fracture toughness with a Si3N4 liner.

NOMAD: Non-Destructive Evaluation (NDE) System for the Inspection of Operation-Induced Material Degradation in Nuclear Power Plants — Overview of the Neutron Irradiation Campaigns

2020 ◽  
Author(s):  
Inge Uytdenhouwen ◽  
Rachid Chaouadi
Keyword(s):  
Neutron Irradiation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Material Heterogeneity ◽  
Non Destructive Evaluation ◽  
Surveillance Programs ◽  
Non Destructive

Abstract Worldwide there are more than 449 nuclear power plants (NPPs) in operation among which 329 reactors are older than 25 years and 94 will be operating for more than 40 years in 2020. Lifetime extensions are requested up to 50–60 years and sometimes even up to 80 years of operation for many existing NPPs. Long-term operation (LTO) of existing NPPs has therefore been accepted in many countries as a strategic objective to ensure supply of electricity for the coming decades. Within this strategy, the European Commission launched the NOMAD project, among others, through the Horizon-2020 programme. The reactor pressure vessels (RPVs) cannot be tested destructively in a direct way, neither can it be replaced. An indirect way is the use of Charpy samples from the so-called surveillance programs. The general strategy on the long term should focus on the ability to perform direct non-destructive evaluation (NDE) of the embrittlement of the vessel. NDE can be used to confirm that the data obtained by surveillance programs are being representative of the real state of the vessel for LTO. Moreover, a generic concern of large nuclear components such as the reactor vessel is the possible material heterogeneity such as macro-segregated regions which could eventually be located in the component but not in the baseline material used as surveillance material. Local non-destructive material inspection and comparison to reference materials in similar irradiation conditions would lead to a better assessment of the properties of the materials at any location of the vessel. The objective of NOMAD is to develop a tool that is capable of non-destructively evaluate the embrittlement of the vessel wall. The final system should be capable of inspecting the microstructure of the materials through the cladding. The tool that will be developed, will use existing and proven nondestructive testing techniques (NDT) with optimized and adjusted sensors. A combination of several techniques based on micro-magnetic, electrical and ultrasonic methods are investigated. Within NOMAD, they are calibrated and validated on a set of existing and newly irradiated samples consisting out of the most common RPV steels from Eastern and Western design, such as 22NiMoCr37, 18MND5, A533-B, A508 Cl.2, A508 Cl.3 and 15kH2NMFA. For the first time, a systematic study on a well-characterized set of samples that correlates the microstructure, mechanical properties, neutron irradiation conditions and non-destructive properties will be carried out. It will not only extend the existing database, but will include issues such as reliability, and uncertainty of the techniques as well as on material heterogeneity. The focus is laid on unbroken Charpy samples and large blocks with and without cladding to “simulate” the actual RPV inspection scenario. This paper gives an overview of the present status of the NOMAD project with focus on the outcome in WP1. The first preliminary NDE results from 6 set-ups and 28 parameters were compared with DBTT results from Charpy impact tests. They are very promising. Final results and detailed analysis will however only be available at the end of the project.

Fracture Toughness Characterization of Low Upper-Shelf Linde 80 Weld Using Mini-C(T) Specimens

2018 ◽  
Author(s):  
Michael R. Ickes ◽  
J. Brian Hall ◽  
Robert G. Carter
Keyword(s):  
Fracture Toughness ◽  
Direct Measurement ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Weld Metals ◽  
Surveillance Programs ◽  
Test Reactor ◽  
Reactor Pressure

The Charpy V-notch specimen is the most commonly used specimen geometry in reactor pressure vessel irradiation surveillance programs and there is an extensive stored inventory of irradiated broken Charpy specimens. The advantage of the mini-C(T) (4mm thick C(T)) specimen technique is that multiple specimens can be machined from each half of broken irradiated Charpy specimens. Fracture toughness specimens that can be machined from broken halves of standard Charpy specimens enable the direct measurement of fracture toughness which can be used for engineering evaluation of reactor pressure vessels. Work to validate the mini-C(T) specimens has been performed mostly on unirradiated reactor pressure vessel base and weld metals . In this study, mini-C(T) specimens were tested providing fracture toughness characterization of an irradiated low upper-shelf Linde 80 weld (WF-70). This weld was utilized in the Midland beltline and has been previously well characterized at ORNL with various types and sizes of fracture toughness specimens. The mini-C(T) specimens were machined from broken previously tested Charpy V-notch size specimens which were irradiated in a material test reactor. The effect of different methods of measuring the displacement on the results is assessed. The ASTM E1921 results are compared to previous test data produced from larger fracture toughness specimens. In addition, the sensitivity of T0 to the ASTM E1921 censoring value is discussed.

Deterministic and Probabilistic Fracture Mechanics Analysis for Structural Integrity Assessment of Pressurized Water Reactor Pressure Vessel

2016 ◽  
Vol 138 (3) ◽  
Author(s):  
Kuan-Rong Huang ◽  
Chin-Cheng Huang ◽  
Hsiung-Wei Chou
Keyword(s):  
Structural Reliability ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Benchmark Problems ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Reactor Pressure

Cumulative radiation embrittlement is one of the main causes for the degradation of pressurized water reactor (PWR) reactor pressure vessels (RPVs) over their long-term operations. To assess structural reliability of degraded reactor vessels, the FAVOR code from the Oak Ridge National Laboratories of the U.S. is employed to perform probabilistic fracture analysis for existing Taiwan domestic PWR reactor vessels with consideration of irradiation embrittlement effects. The plant specific parameters of the analyzed reactor vessel associated with assumed design transients are both considered as the load conditions in this work. Furthermore, two overcooling transients of steam generator tube rupture (SGTR) and pressurized thermal shock (PTS) addressed in the USNRC/EPRI benchmark problems are also taken into account. The computed low failure probabilities can demonstrate the structural reliability of the analyzed reactor vessel for its both license base and extended operations. This work is helpful for the risk assessment and aging management of operating PWR RPVs and can also be referred as its regulatory basis in Taiwan.

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