Technical Basis for Structural Integrity Performance Criterion for Steam Generator Tubing

2006 ◽  
Author(s):  
Russell C. Cipolla ◽  
James A. Begley ◽  
Robert F. Keating
Keyword(s):  
Steam Generator ◽  
Nuclear Energy ◽  
Structural Integrity ◽  
Performance Criterion ◽  
Major Fraction ◽  
Reactor Coolant ◽  
Steam Generator Tubing ◽  
Accident Conditions ◽  
Technical Basis ◽  
Repair Techniques

General Design Criteria (GDC) 1, 2, 4, 14, 30, 31 and 32 of 10 CFR Part 50, Appendix A, define requirements for the reactor coolant pressure boundary (RCPB) with respect to structural and leakage integrity [1]. Steam generator tubing and tube repairs constitute a major fraction of the RCPB surface area. Steam generator tubing and associated repair techniques and components, such as sleeves, must be able to maintain reactor coolant inventory and pressure. The Structural Integrity Performance Criterion (SIPC) from Nuclear Energy Institute (NEI) 97-06 [2] was developed to provide reasonable assurance that a steam generator tube will not burst during normal or postulated accident conditions. This paper presents the SIPC and its technical basis.

A Review of Alloy 600 Cracking in Operating Nuclear Plants Including Alloy 82 and 182 Weld Behavior

2004 ◽  
Author(s):  
Warren Bamford ◽  
John Hall
Keyword(s):  
Weld Metal ◽  
Base Metal ◽  
Steam Generator ◽  
Structural Integrity ◽  
Reactor Vessel ◽  
Alloy 600 ◽  
Control Rod ◽  
Nuclear Plants ◽  
Steam Generator Tubing ◽  
Long Time

Service induced cracking in Alloy 600 has been known for a long time, having been first observed in the 1980’s in steam generator tubing and small bore piping, and later, in 1991, in reactor vessel control rod drive mechanism (CRDM) head penetrations. Other than steam generator tubing, which cracked within a few years of operation, the first Alloy 600 cracking was in base metal of Combustion Engineering small bore piping, followed closely by CE pressurizer heater sleeves. The first reactor vessel CRDM penetrations (base metal) to crack were in France, US plants found CRDM cracking several years later. Three plants have discovered weld metal cracking at the outlet nozzle to pipe weld region. This was the first known weld metal cracking. This paper will chronicle the development of service-induced cracking in these components, and compare the behavior of welds as opposed to base metal, from the standpoint of time to crack initiation, growth rate of cracks, and their impact on structural integrity. In addition, a discussion of potential future trends will be provided.

Evaluation of Interaction Effects on Part-Through Cracks in Steam Generator Tubes

2013 ◽  
Author(s):  
Yong-Seok Kang ◽  
Hong-Deok Kim ◽  
Kuk-Hee Lee ◽  
Jai-Hak Park
Keyword(s):  
Steam Generator ◽  
Interaction Effect ◽  
Structural Integrity ◽  
Structural Parameters ◽  
Interaction Effects ◽  
Multiple Cracks ◽  
Performance Limit ◽  
Steam Generator Tubing ◽  
Steam Generator Tubes

Degraded steam generator tubing can affect its safety functions. Therefore, its integrity should be maintained for each degradation form and all detected degradation must be assessed to verify that if adequate integrity is retained. Determination of tube integrity limits includes identifying acceptable structural parameters such as flaw length, depth, and amplitude of signals. If we consider just single-cracked tubes, short and deep flaws are not likely to threaten structural integrity of tubes. But if it has multiple-cracks, we have to consider interaction effects of multiple adjacent cracks on its burst pressure. Because adjacent multiple cracks can be merged due to the crack growth then it can challenge against the structural performance limit. There are some studies on the interaction effects of adjacent cracks. However, existing works on the interaction effect consider only through-wall cracks. No study has been carried out on the interaction effects of part-through cracks. Most cracks existing in real steam generator tubing are not through-wall cracks but part-through cracks. Hence, integrity of part-through cracks is more practical issue than that of through-wall cracks. This paper presents experimental burst test results with steam generator tubing for evaluation of interaction effects with axial oriented two collinear and parallel part-through cracks. The interaction effect between two adjacent cracks disappeared when the distance exceeds about 2 mm.

Fitness-for-Service Assessment of Steam Generator Tubing Subject to Dealloying Degradation

2017 ◽  
Author(s):  
Rosita Mousavi ◽  
Xinjian Duan ◽  
Michael Kozluk ◽  
Min Wang ◽  
Yihai Shi
Keyword(s):  
Steam Generator ◽  
Degradation Mechanism ◽  
Life Cycle Management ◽  
Steam Generators ◽  
Nickel Copper ◽  
Steam Generator Tubing ◽  
Management Plans ◽  
Monel 400 ◽  
Technical Basis ◽  
Steam Generator Tubes

A new degradation mechanism has been observed in Monel 400 Steam Generator tubing material, a nickel-copper alloy (63Ni-28Cu-2½Fe) with the ASME material designation SB-163/N04400. The location is above the top preheater support plate of the two re-circulating steam generator in one of the units of the Pickering Nuclear Generating Station. This paper provides a brief description of the regulatory environment, OPG’s steam generator life cycle management plans, the Canadian Industry’s fitness-for-service guidelines for steam generator tubes, and the afflicted steam generators. The paper then goes on to discuss the following activities that were conducted to support the technical basis to justify that the steam generators fit to be returned to service: • Inspection scope expansion, methods, and results. • Examination of removed tubes. • Condition monitoring assessment. • Operational assessment. • Burst-pressure tests of removed tubes and of fabricated test specimens. • Degradation specific flaw model and acceptance standards. • Flaw growth rate predictions. • Plugging limit adopted.

Structural Integrity Evaluation of High Strength Anchor Bolt for Supporting Steam Generator and Reactor Coolant Pump in Nuclear Power Plants

2013 ◽  
Author(s):  
Eun-Mo Lim ◽  
Nam-Su Huh ◽  
Hee-Jin Shim ◽  
Chang-Kyun Oh ◽  
Hyun-Su Kim
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
High Strength ◽  
Nominal Stress ◽  
Coolant Pump ◽  
Anchor Bolt ◽  
Reactor Coolant ◽  
Anchor Bolts

In Korea, a fitness-for service evaluation for assuring structural integrity of high strength anchor bolts which support nuclear components such as steam generator and reactor coolant pump, has been one of the important issues in nuclear industry. The main failure mechanism of high strength anchor bolts supporting nuclear components might be degradation due to stress corrosion cracking and brittle fracture. In the present work, the structural integrity of high strength anchor bolts which are used to support steam generator and reactor coolant pump of one of the Korean older vintage nuclear power plants is evaluated by adopting a procedure proposed by Electric Power Research Institute (EPRI) based on an elastic fracture mechanics concept. In this EPRI’s procedure, an accurate estimation of nominal stress acting on the cross section of the bolt is a crucial element since a structural integrity of an anchor bolt is evaluated in the EPRI’s procedure using this nominal stress incorporating reference flaw factors reflecting effects of stress concentration due to bolt thread and reference sized surface crack. In this context, detailed elastic finite element stress analyses are firstly performed on the anchor bolt assemblies to come up with nominal stress in the cross-section of anchor bolt. As for loading condition, bolt pretention as well as normal and faulted loads of the anchor bolts were considered. In addition, the structural integrity of the anchor bolts is demonstrated by comparing nominal stresses of anchor bolts with the maximum allowable stresses obtained by using the EPRI’s reference flaw factors and critical fracture toughness. Furthermore, the accuracy of EPRI’s reference flaw factors which are derived on the assumption that reference sized surface crack is existed on the thread roots is investigated using 3-dimensional elastic finite element fracture mechanics analyses.

Structural Integrity Assessment of the Pump Shaft Assembly of the APR1400 Reactor Coolant Pump

2010 ◽  
Author(s):  
Min-Chul Kim ◽  
Sang-Youn Bang ◽  
Ki-Won Lee ◽  
Young-Jin Choi ◽  
Yong-Su Kim
Keyword(s):  
High Speed ◽  
Critical Speed ◽  
Structural Integrity ◽  
Normal Operation ◽  
Working Fluid ◽  
Coolant Pump ◽  
Integrity Assessment ◽  
Reactor Coolant ◽  
Pump Shaft ◽  
Accident Conditions

The RCP (Reactor Coolant Pump) is operated in high speed and high pressure conditions. Therefore, the problem of vibration has arisen caused by the hydraulic forces of the working fluid. These forces can drastically alter the critical speeds and stability characteristics, and can act as significant destabilizing forces. In this study, the structural integrity of the pump shaft of APR1400 RCP estimated under normal operation and accident conditions. In order to predict the vibration behavior and dynamic characteristics, modal and critical speed analysis were performed. For evaluation due to the load cases, stress characteristics were investigated. This paper shows that the 1st critical speed occurs at over rotational speed range and the calculation and analysis result of the pump shaft satisfied the code requirement.

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