Fracture Toughness of Highly Irradiated Pressure Vessel Steels in the Upper Shelf Temperature

Nine Reactor Pressure Vessel (RPV) Steels and four RPV weld were irradiated up to 1.2 × 1024n/m2 fast neutron fluence (E>1MeV), and their fracture toughness and Charpy impact energy were measured. As chemical compositions, such as Cu, are known to affect the fracture toughness reduction due to neutron exposure, the above steels were fabricated by changing chemical composition widely to cover the chemical composition of the RPV materials of the operating Japanese nuclear power plants. 2.7 mm thick compact specimens were used to measure the upper shelf fracture toughness of highly irradiated materials, and their Charpy upper shelf energy was also measured. By correlating Charpy upper shelf energy to fracture toughness, the upper shelf fracture toughness evaluation formulae for highly irradiated reactor pressure vessel steels were developed. Both compact and V-notched Charpy impact specimens were irradiated in a test reactor. The fast neutron flux above 1MeV was about 5 × 1016n/(m2s). Charpy impact specimens made of Japanese PWR reference material containing 0.09w% Cu were irradiated simultaneously. The upper shelf energy of the reference material up to the medium fluence level showed little difference in the reduction of upper shelf energy to that which had been in the operating plant and which was irradiated to the same fluence. The developed correlation formulae have been adopted in the Japan Electric Association Code as new formulae to predict the fracture toughness in the upper shelf region of reactor pressure vessels. They will be applied to time limited ageing analysis of low upper shelf reactor pressure vessels in Japan, on a concrete technical basis in very high fluence regions.

Fatigue of Stabilized SS and 316 NG Alloy in PWR Environment

Strain controlled constant and variable amplitude fatigue tests for 316NG and Titanium stabilized stainless steels in low oxygen PWR waters were performed. The stabilized steel has been plant aged for 100 000 hours. Constant amplitude test results at 0,01 Hz sinusoidal straining comply with predicted lives according to the Fen approach for both materials. Spectrum straining both in air and in environment caused predicted life reduction factors (about 3) for the stabilized steel, but for the 316NG steel spectrum straining in environment resulted to a larger reduction in life.

Computer Based Process Piping Stress Analysis: ASME B31.3 Appendix S — Example S2

The American Standards Association (ASA) B31.3-1959 Petroleum Refinery Piping Code [1] grew out of an ASA document that addressed all manner of fluid conveying piping systems. ASA B31.3 was created long before widespread engineering use of computer “mainframes” or even before the inception of piping stress analysis software. Also as B31.3 continued to pass thru standards organizations from ASA, ANSI, to ASME, the B31.3 Process Piping Code [2] (hereafter referred to as the “Code”) has remained ambiguous over the past few decades in several areas. The displacement stress range, SE, has always been explicitly defined by both verbiage and equation. Yet, the sustained condition(s) stress, SL, is mentioned not with an explicit equation but with a statement that the sustained stress shall be limited by the allowable stress at the corresponding operating temperature, Sh. Also one might infer from the vague verbiage in the Code that there is only one sustained condition; this would also be an incorrect inference. Also, assumptions would then have to be made as to which supports are allowed to be included in a sustained analysis based on whether the piping “lifts-off” any of the pipe supports during the corresponding operating condition. This paper describes the subtle yet possibly radical concepts that are included in the (recently approved for inclusion into) ASME B31.3-2006 Appendix S Example S2. This paper discusses: • when and in what manner the most severe set of operating temperature and pressure is to be used; • the concept of “sustained condition” and multiple “anticipated” sustained conditions; • determining the support scenario(s) for each anticipated sustained condition; • establishing the most severe sustained condition to evaluate and determine the stress due to sustained loads, SL; • utilizing an equation with sustained stress indices to evaluate SL; • establishing the least severe sustained condition and its effect in determining the largest displacement stress range, SE.

2004 Edition of Japanese Fitness-for-Service Code for Nuclear Power Plants

JSME (Japan Society of Mechanical Engineers) published the first edition of a FFS (Fitness-for-Service) Code for nuclear power plants in May 2000, which provided rules on flaw evaluation for Class 1 pressure vessels and piping, referring to the ASME Code Section XI. The second edition of the FFS Code was published in October 2002, to include rules on in-service inspection. Individual inspection rules were prescribed for specific structures, such as the Core Shroud and Shroud Support for BWR plants, in consideration of aging degradation by Stress Corrosion Cracking (SCC). Furthermore, JSME established the third edition of the FFS Code in December 2004, which was published in April 2005, and it included requirements on repair and replacement methods and extended the scope of specific inspection rules for structures other than the BWR Core Shroud and Shroud Support. Along with the efforts of the JSME on the development of the FFS Code, Nuclear and Industrial Safety Agency, the Japanese regulatory agency approved and endorsed the 2000 and 2002 editions of the FFS Code as the national rule, which has been in effect since October 2003. The endorsement for the 2004 edition of the FFS Code is now in the review process.

A Review of Fatigue and SCC Crack Growth Rate Relationships for Ferritic and Stainless Steels and Ni-Cr-Fe Materials in BWR Environment

When in-service inspection of a nuclear plant component reveals the presence of cracking, an engineering evaluation (typically called a justification for continued operation, or JCO) is required to demonstrate the structural suitability for continued operation. A key element in such a flaw evaluation is the projected crack growth over the period when the cracked component will be reinspected. The crack growth is expected to be a combination of stress corrosion cracking (SCC) and corrosion fatigue. The ASME Section XI Code is in the process of developing a full range of SCC and corrosion fatigue crack growth rate relationships (CGRs) for stainless steel and Ni-Cr-Fe materials. The objective of this paper is to summarize several available SCC and fatigue crack growth rate relationships for these materials exposed to boiling water reactor (BWR) water environments. For completeness, low alloy steel SCC and corrosion fatigue CGRs in BWR water environment are also briefly reviewed. Two example evaluations are provided that used some of these CGRs in developing the JCOs for BWR components. A detailed comparison of these CGRs along with a review of the underlying data will be part of a future effort undertaken by the ASME Section XI Task Group.

Use of Class 1 Analysis to Aid Inservice Inspection

This paper describes an experience of utilization of ASME Section III Code Class 1 analysis to aid in inservice inspection (ISI). The pressurizer surge line weld to the pressurizer nozzle was targeted to be inspected under the ASME Section XI program and also to address Alloy 600 issues. The pipe close to the nozzle was encased by a source shield to protect the pressurizer and the pressurizer support from the effect of postulated longitudinal rupture of the pressurizer surge line. Consequently, the shield did not permit the access needed to perform ISI. The encasement was segmented with bolted joints for removal and installation. With the passage of time, the environmental condition around the encasement made the removal and replacement complex and dose intensive. A modification to the shield by permanent removal of a portion was considered as an attractive option to help provide access for the current inspection, as well as anticipated future inspections. This modification exposed a portion of the pipe which may have longitudinal rupture. ASME Class 1 analysis was performed for the portion of the exposed pipe. Analysis was performed in detail to establish that stress and cumulative usage factor is below the threshold level at which a break has to be postulated. The paper concludes that detailed use of stress analysis results in elimination of the postulated break. As a result, permanent access could be created for current and future inservice inspections of the pressurizer nozzle.

Creep-Fatigue Life Assessment Under Multiaxial Strain State

This paper studies the multiaxial creep-fatigue life for type 304 stainless steel at elevated temperature. Strain controlled biaxial tension-compression creep-fatigue tests were carried out using cruciform specimens under four strain waves at three principal strain ratios. The strain wave and the principal strain ratio had a significant effect on creep-fatigue life of the cruciform specimen. The creep-fatigue life ratio decreased as the principal strain ratio increased which indicates that larger creep damage occurred at larger principal strain ratio. The effects of the strain wave and principal strain ratio were discussed in relation to the observations of surface crack and void area density in the gage part of the specimen. Creep-fatigue lives were discussed in relation to the principal stress amplitude calculated by finite element analysis and creep-fatigue damage was evaluated by linear damage rule.

Low-Cycle Fatigue Evaluation Using the Weld Master S-N Curve

In the context of fatigue evaluation in the low-cycle regime, the use of the master S-N curve in conjunction with elastic FE-based structural stress calculations is presented. An elastic pseudo structural stress estimation is introduced by assuming that Neuber’s rule applies in relating structural stress and strain concentration at a weld to the material’s cyclic stress-strain behavior. With the pseudo structural stress procedure, recent sources of recent full scale test data on pipe and vessel welds were analyzed as a validation of the proposed procedure. The estimated fatigue lives versus actual test lives show a reasonable agreement. Finally, the feasibility of using monotonic stress-strain curves as a first approximation is also examined for applications when cyclic stress-strain curve may not be readily found. The analysis results indicate that the life estimations using monotonic stress-strain curves are reasonable, with the recent test data falling within mean ± 2σ, where σ represents the standard deviation of the master S-N curve.

Revised and New Proposal of Environmental Fatigue Life Correction Factor (Fen) for Carbon and Low-Alloy Steels and Nickel Base Alloys in LWR Water Environments

Low cycle fatigue life of carbon and low alloy steels reduces remarkably as functions of strain rate, temperature, dissolved oxygen and sulfur in steel in high temperature water simulating LWR coolant. A model for predicting such fatigue life reduction was first proposed in the early 1980s and since then has been revised several times. The existing model established in 2000 is used for the MITI Guideline [6] and the TENPES Guideline [7] which stipulate procedures for evaluating environmental fatigue damage at LWR plants in Japan. This paper presents the most recent environmental fatigue evaluation model derived based on additional fatigue data provided by the EFT Project over the past five years. This model differs not significantly with previous version but does provide more accurate equations for the susceptibility of fatigue life to sulfur in steel, strain rate, temperature and dissolved oxygen. Test data on environmental fatigue of nickel base alloys are available only to a limited extent and there is yet no model for predicting fatigue life reduction in such an environment. The EFT Project has made available considerable environmental fatigue test data and developed a new model for calculating Fen of nickel base alloys. The contribution of environment to fatigue of nickel base alloy is much less compared to that in austenitic stainless steel.

Failure Assessment Curves (FACs) in Japanese FFS Code, HPIS Z101 Level 2

The High Pressure Institute of Japan has prescribed the Japanese FFS code, HPIS Z101 Level 1, which can evaluate a crack-like flaw without an extensive knowledge of fracture mechanics. Level 2 is currently being studied and will also be prescribed. In HPIS Z101 Level 2, a two-parameter method (FAD: Failure Assessment Diagram) will be adopted. This paper reveals that FACs dependent on the type of materials should be specified. Kr-Lr relations calculated from experimentally obtained Ramberg-Osgood constants of several steels are compared with the FACs of BS7910 Level 2A & API 579 Level 2, R6 Option 1 and ASME N494-3. It is found that the FAC of ferritic steel in ASME N494-3 is pertinent to carbon steels with a marked yield point plateau while the FAC of BS7910 Level 2A & API 579Level 2 is suitable for ferritic steels other than carbon steels such as 0.5Mo, 2.25Cr-1Mo and 2.25Cr-1Mo-0.25V steels. Further, cut-off values dependent on the strength of the material and the plasticity interaction factors for the FAC of carbon steels are proposed.