Supplementary Technical Basis for ASME Section XI Code Case N-597

Efforts to develop clear and conservative methods to measure and evaluate wall thinning in nuclear piping have been underway since the late 1980’s. The Electric Power Research Institute (EPRI) carried out a successful campaign to address programmatic issues, such as locating and predicting flow-accelerated corrosion (FAC) degradation. This included developing a computer code (CHECWORKS), a users group (CHUG), and a comprehensive program guideline document for the effective prediction, identification and trending of flow-accelerated corrosion degradation. U.S. Nuclear Regulatory Commission (NRC) guidelines are provided in the NRC Inspection Manual Inspection Procedure 49001. At the same time, committees under Section XI of the ASME Boiler and Pressure Vessel Code have addressed evaluation of structural integrity of piping subjected to wall thinning. Code Case N-480 of Section XI provided acceptance criteria that focused on primary piping stresses, with evaluation based on a uniform wall thinning assumption for evaluating the minimum wall thickness of the piping. However, when applying this methodology to low pressure piping systems, Code Case N-480 was very conservative. Code Case N-597 was first published in 1998, and supercedes Code Case N-480. The current version is N-597-2. Code Case N-597-2 provides acceptance criteria and evaluation procedures for piping items, including fittings, subjected to a wall thinning mechanism, such as flow-accelerated corrosion. Code Case N-597-2 is a significant improvement over N-480, containing distinct elements to be satisfied in allowing the licensee to operate with piping degraded by wall thinning. The Code Case considers separately wall thickness requirements and piping stresses, and maintains original design intent margins. The Code Case does not provide requirements for locations of inspection, inspection frequency or method of prediction of rate of wall thinning. As described in the original technical basis document published at the 1999 ASME PVP Conference, the piping stress evaluation follows very closely the Construction Codes for piping. Five conditions related to industry use of Code Case N-597-1 have been published by the NRC in Regulatory Guide 1.147, Revision 13. A number of these issues are related to a need for additional explanation of the technical basis for the Code Case, such as the procedures for evaluation of wall thickness less than the ASME Code Design Pressure-based minimum allowable wall thickness. This presentation addresses these NRC conditions by providing additional description of the technical basis for the Code Case.

An Assessment of Low Alloy Steel EAC Corrosion Fatigue Relationships for BWR Environment

The fatigue crack growth rates for ferritic steels in water environments given in A-4300 of Appendix A, Section XI, ASME Code, were developed from data obtained prior to 1980. Subsequently, updated assessments by Eason, et al. and recent laboratory test results from Seifert and Ritter demonstrated that under certain conditions, ferritic steels exposed to oxygenated water environments may be susceptible to high fatigue crack growth rates that exceed the current disposition curves. In the light of ASME adopting Code Case N-643 for PWRs, there is a need for a similar Code Case for the BWR water environments (for both the normal water chemistry and hydrogen water chemistry/NobleChem) that takes into account these findings. This could mean modification of current EAC curves in the ASME Code. A joint program of EPRI and GE was developed to address this need for updated evaluations of the corrosion fatigue. The program’s first task has been to re-assess the role of rise time, environment, alloy, heat treatment and impurity levels on the established ASME codified disposition curves/methodologies. The data was then used as a basis to assess the impact of on modified cyclic curves on the disposition approaches that are currently used to evaluate postulated flaws in the BWR reactor pressure vessel or RPV head and the feed water nozzle regions. The presentation would include a discussion of the appropriate BWR plant transients and the GE process for performing evaluations. The role of the evaluations on the establishment of inspection intervals currently determined using NUREG-0619 and the latest BWROG Report would also be presented. Finally, the relationship between cyclic load and constant load behavior in these steels are discussed in the context of the mechanisms for environmentally assisted cracking.

Crack Growth Rates of Alloy 600 From the Davis-Besse Reactor CRDM Nozzle #3 in PWR Environment

Reactor–vessel internal components made of nickel–base alloys are susceptible to environmentally assisted cracking. A better understanding of the causes and mechanisms of this cracking may permit less conservative estimates of damage accumulation and requirements on inspection intervals of pressurized water reactors (PWRs). This paper presents crack growth rate (CGR) results for Alloy 600 removed from nozzle#3 of the Davis–Besse (D-B) control rod drive mechanism (CRDM). The tests were conducted on 1/4-T or 1/2-T compact tension specimens in simulated PWR environment, and crack extensions were determined by DC potential drop measurements. The experimental CGRs under cyclic and constant load are compared with the existing CGR data for Alloy 600 to determine the relative susceptibility of the D-B CRDM nozzle alloy to environmentally enhanced cracking. The CGRs under constant load for the nozzle material are higher than those predicted by the best-fit curve for Alloy 600 at 316 °C. The results also indicate significant enhancement of CGRs under cyclic loading in the PWR environment. Characterization of the material microstructure and tensile properties is described.

Reinforcement Analysis of a Slab Tank

The objective of this paper is to reduce the stresses and deflection of an existing slab tank [2]. The slab tank is to store various criticality liquids used in today’s industry. The preliminary overall dimensions of the slab tank are determined from the capacity of the stored liquids. The slab tank design is broken up into (a) two long side members, (b) two short side members, (c) top head, and (d) bottom head. The slab tank is supported from the bottom at a height by a rectangular plate enclosure. The deflection of the linear space is a critical requirement. The deflection is controlled by providing external supports from the bottom at a height by adjustable bolts. The analysis of the slab tank showed excessive stresses at the concentrated supports. The slab tank was modified by providing reinforcement on the long side members. Several reinforcement arrangements were considered. The slab tank is subjected to two conditions. First, vacuum condition, the long side plates will deflect inwards. Second, internal pressure condition the design pressure consists of working internal pressure plus static head pressure. For this the long side plates will deflect outwards. The heads are designed for internal pressure at the bottom where the pressure is the maximum. The vacuum pressure is not critical. The dimensioned slab tank is modeled using STAAD III finite element software. The slab tank showed excessive stresses. The concentrated supports were removed. The long side member was reinforced by a Channel section. The slab tank analysis was simplified by modeling a single long side member and three cases of Channel section reinforcement were considered. The reinforced arrangement was analyzed by STAAD III finite element software. Further analysis by changing the Channel section by plate reinforcement was found to be better.

The ASME Section XI Code and the Future

This paper will describe the process under which the ASME Section XI Subcommittee works in addressing industry initiatives and setting of priorities and goals. Each calendar year an Operational Plan is developed which comprises the priorities and goals established for the different Subgroups and Working Groups that report directly to the Subcommittee XI. These goals include committee actions developed to respond to industry initiatives and concerns. This Operational Plan also helps establish the “work-list” of each Subgroup. This work-list provides the priority of the items to be worked on for the upcoming year. Just as any other business, the Subcommittee on Nuclear Inservice Inspection manages and prioritizes activities to meet the industry needs.

Fatigue Evaluation Incorporating Multiaxiality of Stress States

In this work the effects of multiaxiality on the fatigue evaluation by the ASME Boiler and Pressure Vessel Code procedures have been assessed. The conservatism associated with the Ke factor has been critically appraised for fatigue evaluation using a design example of a feed water nozzle subjected to pressure and thermal fluctuations. A fictitious stress concentrator is applied to account for the ratio of the peak stress to the stress linearized through the thickness of the section under consideration. The effect of the traxiality of the stress distribution has also been assessed using the same design example for fatigue evaluations. Additional analytical and experimental studies have been recommended to study these important critical factors for fatigue assessment.

Fatigue Assessment of a Weld Joint of a Pressure Vessel

A full-penetration weld on both sides between a shell and a flat head is evaluated for fatigue strength by four methods, two based on structural stress and two on notch stress. Internal pressure is cycled with constant amplitude. The allowable cycles are calculated by each method. The number of cycles for the same geometry and loading varies widely, ranging from 4,835 to 137,000.

Fatigue Crack Propagation of 304 Stainless Steel in High Temperature Water: Additional Tests and Data Correlation

Fatigue crack propagation (FCP) testing of 304 stainless steel (304 SS) specimens showed a strong acceleration of rates in high temperature water with 40–60 cc H2/kg H2O at 243°C and 288°C, with rates up to 20X the air rates [1]. However, FCP rates were markedly reduced for a second heat at long rise times and for both heats with addition of a constant load hold time of 1200 s at a high stress ratio [2]. Such behavior had not been previously reported in the literature and merited further investigation. Tests have been extended to include two additional heats and a wider set of loading conditions. FCP rates were accelerated at long rise times in the two additional heats, consistent with a large series of tests on wrought, weld, and cast austenitic stainless steel materials recently reported by Nomura, et al. [3]. Hold time tests at a lower stress ratio showed that small increases or decreases in rate occur with holds at minimum or maximum load, but the changes were within normal data scatter. The rate reductions are not a generic result of less frequent cycling, but are limited to specific loading parameters or heats. A time-based correlation successfully describes the accelerated rates observed on all four heats, and is nearly identical to fits of literature data in PWR water and hydrogen water chemistry (HWC). A power law fit with separate terms for rise time, ΔK, and stress ratio provides an equivalent correlation.

Environmental Fatigue Testing of Type 316 Stainless Steel in 310 °C Water

Low cycle fatigue tests were conducted to investigate fatigue behaviors of Type 316 stainless steel in 310 °C low oxygen water. In the tests, strain rates were 4 × 10−4, 8 × 10−5 s−1 and applied strain amplitudes were 0.4, 0.6, 0.8, and 1.0%. The test environment was pure water at a temperature of 310 °C, pressure of 15 MPa, and dissolved oxygen concentration of < 1 ppb. Type 316 stainless steel underwent a primary hardening, followed by a moderate softening for both strain rates in 310 °C low oxygen water. The primary hardening was much less pronounced and secondary hardening was observed at lower strain amplitude. On the other hand, the cyclic stress response in room temperature air exhibited gradual softening and did not show any hardening. The fatigue life of the studied steel in 310 °C low oxygen water was shorter than that of the statistical model in air. The reduction of fatigue life was enhanced with decreasing strain rate from 4 × 10−4 to 8 × 10−5 s−1.

Repair of Class 1 Control Rod Drive and In-Core Housing Bottom Head Penetrations in BWRs

The ASME Section XI committee is developing a code case to permit repair of leakage on Boiling Water Reactor (BWR) bottom head penetrations with a mechanical roll expansion process. This technology has been successfully utilized for this application over the last two decades to address leakage due to Control Rod Drive (CRD) stub tube cracking. The code case defines the technical and administrative requirements for use of the mechanical roll expansion process for repair of Class 1 Control Rod Drive and Incore housing penetrations in the bottom head of BWRs. The code case specifies the process qualification, essential variables, process application, examination and pressure testing requirements for this process. The technical basis of the proposed code case includes detailed fracture mechanics analysis to evaluate the structural consequences of the cracking, metallurgical assessment and extensive testing to determine the load capability of the roll repair joint. Based on this assessment, the successful BWR field experience and the inspections/ qualifications required under the code case, the mechanical roll expansion repair can be used as a permanent repair option for addressing leakage in BWR CRD and In-core housing penetrations, The application of this code case will provide significant reduction in facility down time and will offer a reduction in personnel radiation exposure as compared to welded repair options.