Non Destructive Testing of Unfired Pressure Vessels

This paper discusses the non destructive testing (NDT) of unfired pressure vessels made of ductile and tough steels, as contained in Part 5 of the European standard EN 13445:2002. The concept and use of testing groups along with “satisfactory experience” in welding are presented. Also the background and rationale for the determination of standards used for NDT methods, characterisation and acceptance criteria are discussed in detail. Benefits for the pressure equipment industry are emphasised.

Drop Testing of DOE Spent Nuclear Fuel Canisters

The National Spent Nuclear Fuel Program (NSNFP) at the Idaho National Engineering and Environmental Laboratory (INEEL) prepared four representative Department of Energy (DOE) spent nuclear fuel (SNF) canisters for the purpose of drop testing. The first two canisters represented a modified 24-inch diameter standardized DOE SNF canister and the second two canisters represented the Hanford Multi-Canister Overpack (MCO). The modified canisters and internals were constructed and assembled at the INEEL. The MCO internal weights were fabricated at the INEEL and assembled into two MCOs at Hanford and later shipped to the INEEL for drop test preparation. Drop testing of these four canisters was completed in August 2004 at Sandia National Laboratories. The modified canisters were dropped from 30 feet onto a flat, essentially unyielding surface, with the canisters oriented at 45 degrees and 70 degrees off-vertical at impact. One representative MCO was dropped from 23 feet onto the same flat surface, oriented vertically at impact. The second representative MCO was dropped onto the flat surface from 2 feet oriented at 60 degrees off-vertical. These drop heights and orientations were chosen to meet or exceed the Yucca Mountain repository drop criteria. This paper discusses the comparison of deformations between the actual dropped canisters and those predicted by pre-drop and limited post-drop finite element evaluations performed using ABAQUS/Explicit. Post-drop containment of all four canisters, demonstrated by way of helium leak testing, is also discussed.

Life Prediction of Furnace Heater Tube Using Omega Parameter and Life Fraction Rule

Recently, predicting the remaining life of petrochemical equipment that is used in high temperature and pressure service is an essential element. Specifically, predicting the life of furnace heater tubes used in the range of 850°C to 950°C is an important subject. For economic reasons, it is desirable to replace degraded tubes before tube fails occur. Remaining life assessments are used to help in making tube replacement decisions at proper time. This paper contains Ω parameter and life fraction rule used to life assessment factor to predict remaining furnace tube life, compares the two methods. As the result of assessment, two methods show that the general life of the furnace heater tubes are about 160,000hr and future remaining life is 50,000hr. Based on this result, the heater tubes are enabling to operate until 4years.

Training in the Application of the ASME Code to Transportation Packaging of Radioactive Materials

The Department of Energy has established guidelines for the qualifications and training of technical experts preparing and reviewing the safety analysis report for packaging (SARP) and transportation of radioactive materials. One of the qualifications is a working knowledge of, and familiarity with the ASME Boiler and Pressure Vessel Code, referred to hereafter as the ASME Code. DOE is sponsoring a course on the application of the ASME Code to the transportation packaging of radioactive materials. The course addresses both ASME design requirements and the safety requirements in the federal regulations. The main objective of this paper is to describe the salient features of the course, with the focus on the application of Section III, Divisions 1 and 3, and Section VIII of the ASME Code to the design and construction of the containment vessel and other packaging components used for transportation (and storage) of radioactive materials, including spent nuclear fuel and high-level radioactive waste. The training course includes the ASME Code-related topics that are needed to satisfy all Nuclear Regulatory Commission (NRC) requirements in Title 10 of the Code of Federal Regulation Part 71 (10 CFR 71). Specifically, the topics include requirements for materials, design, fabrication, examination, testing, and quality assurance for containment vessels, bolted closures, components to maintain subcriticality, and other packaging components. The design addresses thermal and pressure loading, fatigue, nonductile fracture and buckling of these components during both normal conditions of transport and hypothetical accident conditions described in 10 CFR 71. Various examples are drawn from the review of certificate applications for Type B and fissile material transportation packagings.

Thermal Properties of Fiberboard Overpack Materials in the 9975 Shipping Package

The 9975 shipping package incorporates a cane fiberboard overpack for thermal insulation and impact resistance. Thermal properties (thermal conductivity and specific heat capacity) have been measured on cane fiberboard and a similar wood fiber-based product at several temperatures representing potential storage conditions. While the two products exhibit similar behavior, the measured specific heat capacity varies significantly from prior data. The current data are being developed as the basis to verify that this material remains acceptable over the extended storage time period.

Second Phase Precipitation in As-Welded and Solution Annealed Alloy 22 Welds

The precipitation characteristics of tetrahedrally close-packed (TCP) phases during the welding and the subsequent solution annealing process of Alloy 22 1 1/2” thick plate double-U prototypical welds are investigated. Electron backscatter diffraction (EBSD) was used to provide large scale microstructural observation of the weld cross section, and scanning electron microscopy (SEM) was used to map the location of the TCP phases. Analysis shows that TCP precipitation occurs congruent to the weld passes, with the solution annealing reducing the sizes of coarser precipitates.

Numerical Analyses of Locomotive Impacts on a Spent Fuel Truck Cask and Trailer

During the transportation of spent nuclear fuel by truck, the possibility exists that a train could run into the spent fuel cask at a grade crossing. Sandia National Laboratories has conducted a numerical study to assess the possibility of cask breach or material release in the event of a high-speed, broadside locomotive collision. A numerical approach has the advantage over conducting a physical test as was done in the 1970s [1] in that varying parameters can be examined. For example, one of the criticisms of the 1970s test was the height of the cask. In the test, the centerline of the cask was above the main frame-rails of the locomotive. In this study the position of the cask with respect to the locomotive was varied. The response of the cask and trailer in different collision scenarios was modeled numerically with LS-DYNA [2]. The simulations were performed as a collaborative endeavor between Sandia National Laboratories (SNL), Applied Research Associates, Inc. (ARA) and Foster-Miller, Inc (FMI). ARA developed the GA-4 Spent Fuel Cask and Cask Transporter models described in this report. These models were then combined with two existing FMI heavy freight locomotive finite element models to create the overall simulation scenarios. The modeling effort, results, and conclusions are presented in this paper.

Probabilistic Analysis of a Preemptive Replacement of an NPP Component With a Component of Superior Material

The Westinghouse proactive aging management tool, PAM, considers three major sets of variables when calculating the NPV or economic value of age replacement: (a) the projected time to failure, (b) the economic consequences of unplanned failure and (c) the cost of the replacement. All of these variables will typically be uncertain; particularly the time to part failure. A not uncommon complication in evaluating whether and when to replace a degrading component or part in a plant is that the replacement part is thought to have a longer expected life (be more resistant to degradation) but, to date, there is little field experience to substantiate that belief. This paper shows how two different approaches for statistical estimation of a Weibull failure distribution can be used in tandem to surmount this problem, and illustrates it within the context of the replacement of a nuclear power plant component tube bundle with a tube bundle expected to provide superior corrosion resistance.

Economic Evaluation of Proactive Replacement of Upper Internals Guide Tube Support Pins (Split Pins)

Certain nuclear power plants have “Rev B” reactor vessel upper internals guide tube support pins, commonly referred to as split pins, made from material with properties similar to Alloy 600 and known to be susceptible to primary water stress corrosion cracking (PWSCC). This paper describes a rigorous probabilistic methodology for evaluating the economics of a preemptive replacement of these split pins, and describes an application at four of Exelon Generation’s nuclear plants. The method uses Bayesian statistical reliability modeling to estimate a Weibull time-to-failure prediction model using limited historical failures, and a Westinghouse proactive aging management simulation tool called PAM to select a split pin replacement date that would maximize the net present value of cash flow to a plant. Also in this study is a sensitivity evaluation of the impact of zinc addition on split pin replacement timing. Plant decisions made based in part on results derived from applying this approach are noted.

Aging Management Program of Large Diameter Safety Class Piping

Ageing management of Nuclear Power Plants is an essential issue for utilities, in term of safety and availability and corresponding economical consequences. Practically all nuclear countries have developed a systematic program to deal with ageing of components on their plants. This paper presents the ageing management program developed by EDF and that are compared with different other approaches in other countries (IAEA guidelines and GALL report). The paper presents an example of application to large diameter safety class piping. Different degradation mechanisms are considered like fatigue, corrosion or thermal aging. Maintenance and surveillance actions are discussed in the paper.