Structural Integrity Assessment of the Fusion Reactor First Wall Using the ASME Code

Author(s):  
M. Merola ◽  
R. Matera
Author(s):  
Sam Ranganath ◽  
Guy DeBoo

Structural integrity assessment of reactor components requires consideration of crack growth. A key input to this is the development of reference stress corrosion crack (SCC) growth rate curves for use in the structural evaluation. The ASME Section XI Task Group on SCC Reference Curve is looking into available SCC data for stainless steel and nickel based alloys and associated weldment in both pressurized water reactor (PWR) and boiling water reactor (BWR) environments. The test data show significant data scatter in crack growth rates (CGR). The conservative approach is to develop reference curves that bound all available data so that upper bound crack growth predictions. While this approach may be conservative, it may lead to excessive estimates of crack growth and result in unrealistic (and often meaningless) structural margin predictions. Selection of the appropriate SCC reference curves requires realistic interpretation of test data so that the predictions are consistent with field behavior and provide reasonable, but conservative assessment. This paper describes crack growth assessment for stainless steel piping and Alloy 600 safe end components with Alloy 182/82 welds in BWR environment. The results from the crack growth analysis for piping can be used to determine whether a proposed reference curve provides reasonable results. The objective is to use the piping and safe end crack growth predictions to develop optimal SCC Reference Curves for use in ASME Code evaluations.


Author(s):  
P. Babics ◽  
S. Ratkai ◽  
D. Szabo ◽  
P. Trampus

The owner of Paks NPP, Hungary’s nuclear generating facility, is aiming at adjusting the ISI program to meet ASME Code requirements. The objective is to achieve an internationally acceptable level in structural integrity assessment of long-lived and passive components, and to create the basis for a proper ageing management program for the operations period beyond design life of the units. Apart from this, it would allow to extend the current four-year inspection interval for Class 1 components up to an eight-year one, which would contribute to a more cost-efficient operation and maintenance. Hungarian nuclear regulatory regime gives an opportunity for this because the nuclear safety regulation does not determine explicitly the applicable codes neither for the design nor for the ISI. First, the basic regulatory principles related to ASME adaptation will be summarized. They focus on aspects of maintaining the current licensing basis as well as on the necessity to demonstrate the compliance with Section III requirements. The substantial part of the work is the construction review of selected Class 1 and 2 components. Then, the results of comparison of the current ISI program, mainly based on Russian normative documents, and the Section XI based one will be shown. These comparative studies have justified the feasibility of the project. The licensing of the ASME based ISI program is under way, and the regulator’s position will be presented as well.


Author(s):  
E. Ruedl ◽  
P. Schiller

The low Z metal aluminium is a potential matrix material for the first wall in fusion reactors. A drawback in the application of A1 is the rel= atively high amount of He produced in it under fusion reactor conditions. Knowledge about the behaviour of He during irradiation and deformation in Al, especially near the surface, is therefore important.Using the TEM we have studied Al disks of 3 mm diameter and 0.2 mm thickness, which were perforated at the centre by double jet polishing. These disks were bombarded at∽200°C to various doses with α-particles, impinging at any angle and energy up to 1.5 MeV at both surfaces. The details of the irradiations are described in Ref.1. Subsequent observation indicated that in such specimens uniformly distributed He-bubbles are formed near the surface in a layer several μm thick (Fig.1).After bombardment the disks were deformed at 20°C during observation by means of a tensile device in a Philips EM 300 microscope.


Author(s):  
Daigo Watanabe ◽  
Kiminobu Hojo

This paper introduces an example of structural integrity evaluation for Light Water Reactor (LWR) against excessive loads on the Design Extension Condition (DEC). In order to assess the design acceptance level of DEC, three acceptance criteria which are the stress basis limit of the current design code, the strain basis limit of the current design code and the strain basis limit by using Load and Resistance Factor Design (LRFD) method were applied. As a result the allowable stress was increased by changing the acceptance criteria from the stress basis limit to the strain basis limit. It is shown that the practical margin of the LWR’s components still keeps even on DEC by introducing an appropriate criterion for integrity assessment and safety factors.


1987 ◽  
Vol 12 (1) ◽  
pp. 104-113 ◽  
Author(s):  
K. Taghavi ◽  
M. S. Tillack ◽  
H. Madarame

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