Use of Irradiated Fracture Toughness Values in Nuclear Vessel and Component Design Specifications for Fitness-For-Service Analysis Using the Unified Curve Method

2019 ◽  
Vol 5 (2) ◽  
Author(s):  
Waleed Ishaque
Keyword(s):  
Fracture Toughness ◽  
Reactor Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Design Specifications ◽  
Failure Assessment ◽  
Curve Method ◽  
Unified Curve Method

The American Society of Mechanical Engineers (ASME) Section III Rules for Construction of Nuclear Facility Components subsection NB-2331 Material for Vessels requires that the effects of irradiation shall be considered on material toughness properties in the core belt line region of the reactor vessel. The code also states that “the design specifications shall include additional requirements, as necessary, to ensure adequate fracture toughness for the service lifetime of the vessel.” In a design report of nuclear pressure vessel, the design and service loads do not include loads that are affected by fracture toughness of the material. However, in the cases of fitness-for-service assessment for component flaws (prevalent with age of component), irradiated material properties become highly relevant. An example of a fitness-for-service is that of a beyond design basis reactor vessel head drop accident in a pressurized water reactor with a nozzle junction flaw. As a case study, the critical size of a postulated external surface semielliptical circumferential crack in the combustion engineering three-loop pressurized water reactor nozzle–vessel junction is calculated using ansys Workbench (Academic version) with the applied impact load from the vessel head drop accident. Failure assessment diagrams for numerous crack depths and lengths were developed considering the fracture toughness properties of the irradiated reactor vessel steel. The mode I stress intensity results used in the failure assessment diagram were compared with the available finite element and the American Petroleum Institute (API) standard API 579 analytical solutions for validation, showing good agreement. From this case study, it is demonstrated that the effects of irradiation on fracture toughness become prominent at the same postulated crack size in the nozzle–vessel junction dispositioned as “safe” becomes “unsafe” in the fracture assessment diagram. Using the unified curve method, the irradiated fracture toughness data in design specification can be supplied so that it may be used in fitness-for-service analysis to account for component aging.

Application of the Failure Assessment Diagram to the Evaluation of Pressure-Temperature Limits for a Pressurized Water Reactor

1985 ◽  
Vol 107 (2) ◽  
pp. 192-196 ◽  
Author(s):  
K. K. Yoon ◽  
J. M. Bloom ◽  
W. A. Pavinich ◽  
H. W. Slager
Keyword(s):  
Fracture Mechanics ◽  
Reactor Vessel ◽  
Normal Operation ◽  
Coolant Temperature ◽  
Operating Pressure ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Failure Assessment Diagram ◽  
Failure Assessment

The failure assessment diagram approach, an elastic-plastic fracture mechanics procedure based on the J-integral concept, was used in the evaluation of pressure-temperature (P-T) limits for the beltline region of the vessel of a pressurized water reactor. The main objective of this paper is to illustrate the application of an alternate fracture mechanics method for the evaluation of pressure-temperature limits, as allowed by the Code of the Federal Regulation 10 CFR 50, Appendix G. The evaluation of P-T limits for the beltline region of a pressurized water reactor was based on the following assumptions: • ASME Pressure Vessel and Piping Code, Section III, Appendix G reference flaw • End-of-life fluence level in the beltline region • Longitudinal flaw in the beltline weld • J-resistance material toughness curves obtained from the U.S. Nuclear Regulatory Commission’s heavy section steel technology (HSST) program • Other material properties obtained from the Babcock & Wilcox Integrated Reactor Vessel Material Surveillance Program The maximum allowable pressure levels were calculated at 33 time points along the given reactor bulk coolant temperature history representing the normal operation of a pressurized water reactor. The results of the calculations showed that adequate margins of safety on operating pressure for the critical weld in the beltline of the pressurized water reactor vessel are assured.

Simulation of JAERI Downcomer Effective Water Head Experiments With WCOBRA/TRAC-TF2

2018 ◽  
Author(s):  
Jeffrey R. Kobelak ◽  
Jun Liao ◽  
Katsuhiro Ohkawa
Keyword(s):  
Hydraulic System ◽  
Large Scale ◽  
Reactor Vessel ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
System Pressure ◽  
Water Head ◽  
Effective Water

During the reflood phase of a postulated large break loss-of-coolant accident (LBLOCA), the liquid head in the reactor vessel downcomer provides the driving force to reflood the core. Since the reflood rate is a function of the downcomer inventory, the calculation of the downcomer liquid inventory is critical in simulating the reflood phase of a postulated LBLOCA accident in a pressurized water reactor. Since the reactor coolant system pressure decreases rapidly after the onset of a LBLOCA transient, the walls surrounding the downcomer become superheated for the duration of the transient. The Japan Atomic Energy Research Institute (JAERI) downcomer effective water head test facility was designed to study boiling and steam-water interaction in the reactor vessel downcomer under prototypical reflood conditions. A number of tests were conducted at this facility with varying degrees of wall superheating (among other things) that cover the expected degree of superheating in a pressurized water reactor. The wall superheating achieved at the JAERI facility is greater than that of other large-scale facilities that are typically simulated to validate thermal-hydraulic system codes. WCOBRA/TRAC-TF2 is the thermal-hydraulic system code utilized in the FULL SPECTRUM™ LOCA (FSLOCA™) evaluation model (EM). The ability of the WCOBRA/TRAC-TF2 code to predict phenomena occurring in the reactor vessel downcomer during the reflood phase of a postulated LBLOCA has been previously validated. However, only limited wall superheating was present in the existing validation basis. As such, two experiments conducted at the JAERI downcomer effective water head test facility are simulated to provide additional information on the capability of WCOBRA/TRAC-TF2 to predict the liquid inventory in the reactor vessel downcomer during the reflood phase of a postulated LBLOCA. The code captured all the trends observed in the experimental data for both Run 115 and Run 121. The various collapsed liquid levels tended to be well-predicted or under-predicted by the code after the initial simulated accumulator injection period.

Aging Management Strategies for Pressurized Water Reactor Vessel Internals

2004 ◽  
Author(s):  
Timothy J. Griesbach ◽  
Robert E. Nickell ◽  
H. T. Tang ◽  
Jeff D. Gilreath
Keyword(s):  
Reactor Vessel ◽  
Performance Criteria ◽  
Pressurized Water Reactor ◽  
Aging Effects ◽  
Pressurized Water ◽  
Water Reactor ◽  
Flaw Tolerance ◽  
License Renewal ◽  
Technical Basis ◽  
Reactor Internals

Management of materials aging effects, such as loss of material, reduction in fracture toughness, or cracking, depends upon the demonstrated capability to detect, evaluate, and potentially correct conditions that could affect function of the internals during the license renewal term. License renewal applicants in their submittals to NRC have identified the general elements of aging management programs for Pressurized Water Reactor (PWR) internals, including the use of inservice inspection and monitoring with the possibility of enhancement or augmentation if a relevant condition is discovered. As plants near the license renewal term, plant-specific aging management programs will be implemented focusing on those regions most susceptible to aging degradation. A framework for the implementation of an aging management program is proposed in this paper. This proposed framework is based on current available research results and state of knowledge and utilizes inspections and flaw tolerance evaluations to manage the degradation issues. The important elements of this framework include: • The screening of components for susceptibility to the aging mechanisms, • Performing functionality analyses of the components with representative material toughness properties under PWR conditions, • Evaluating flaw tolerance of lead components or regions of greatest susceptibility to cracking, loss of toughness, or swelling, and • Using focused inspections to demonstrate no loss of integrity in the lead components or regions of the vessel internals. The EPRI Material Reliability Program (MRP) Reactor Internals Issue Task Group (RI-ITG) is actively working to develop the data and methods to quantify an understanding of aging and potential degradation of reactor vessel internals, to develop materials/components performance criteria, and to provide utilities tools for extending plant operations. Under this MRP Program, the technical basis for the framework will be documented. Then, based on that technical basis, PWR internals inspection and flaw evaluation guidelines will be developed for plants to manage reactor internals aging and associated potential degradation.

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.

Heavy Components Replacement in Nuclear Power Plants: Guidelines and Experiences

2010 ◽  
Author(s):  
Ki Sig Kang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Vessel ◽  
Primary System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Heavy Water Reactor ◽  
Technical Innovations

Utilities are looking for ways to optimize plant lifetime, and must therefore prevent stress corrosion in primary components, while combating other phenomena, such as thermal fatigue or certain metallurgical weaknesses. The replacement of sections of the main primary system is one way of solving these problems. The increase in the number of the replacement of heavy components carried out in the reactor building on specific reactor geometries has called for major technical innovations on the replacement of heavy components. For above, the IAEA published a nuclear energy series (NES) on replacement of heavy components to propose guidance and share experiences. The major and heavy components to be considered are; 1) Steam generators for pressurized water reactor plants, 2) Reactor vessel head for PWR plants, 3) Reactor internal components for boiling water reactor plants, 4) Reactor vessel internals for PWR plants, 5) Pressurizer for PWR plants, 6) Reactor coolant piping/ recirculation piping PWR, and 7) Press Tube and feed piping for pressurized heavy water reactor. This paper is focused on heavy components replacement considered strategic aspects for nuclear power plants life management.

Sign in / Sign up

Export Citation Format

Share Document