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Stress analysis of high-level waste canisters: methods, applications, and design data

Description: An overview of stress analysis methods, structural design procedures, and design data is presented for canisters used to package solidified wastes, particularly borosilicate glass. In addition, waste processing, canister materials, fabrication and inspection methods, and performance testing are summarized. Sources of stress in canisters are lifting and handling loads, internal pressure, high-temperature filling operations, transient heating and cooling, differential thermal expansions of canisters and glass, and impact loadings from low-probability accidents. Results of case studies that illustrate applicable methods of stress analyses are presented for these sources of stress. Existing sections of ASME Boiler and Pressure Vessel Code are applicable to canister fabrication, but the code does not cover many aspects of canister service loadings. Specialized criteria for minimum wall thicknesses to sustain filling stresses are proposed in this report. Results of a test program to measure the creep strength of candidate canister materials are described. Methods to predict residual stresses in the walls of waste canisters are described; predicted residual stress levels agree with measured stress levels. The consequences of these residual stresses are reviewed, and stress-corrosion cracking is identified as the mode of canister failure affected by residual stresses. Canister-closure design is covered in detail, particularly the welding and inspection of the final closure seal-weld. It is shown that the methods of fracture mechanics and fatigue-crack-growth analyses are valuable tools for evaluating the performance of closure welds in the presence of crack-like defects. Canister performance in process trials at PNL shows the ability of canisters to survive high temperatures and loadings during processing. Impact tests show that a suitably designed canister can sustain severe impacts without loss of intergrity.
Date: October 1, 1979
Creator: Simonen, F.A. & Slate, S.C.
Partner: UNT Libraries Government Documents Department

Collapse of experimental capsules under external pressure

Description: Stress analyses and developmental tests of capsules fabricated from thick-walled tubing were performed for an external pressure design condition. In the design procedure no credit was taken for the expected margin in pressure between yielding of the capsule wall and catastrophic collapse or flattening. In tests of AISI-1018 low carbon steel capsules, a significant margin was seen between yield and collapse pressure. However, the experimental yield pressures were significantly below predictions, essentially eliminating the safety margin present in the conservative design approach. The differences between predictions and test results are attributed to deficiencies in the plasticity theories commonly in use for engineering stress analyses. The results of this study show that the von Mises yield condition does not accurately describe the yield behavior of the AISI-1018 steel tubing material for the triaxial stress conditions of interest. Finite element stress analyses successfully predicted the transition between uniform inward plastic deformation and ovalization that leads to catastrophic collapse. After adjustments to correct for the unexpected yield behavior of the tube material, the predicted pressure-deflection trends were found to follow the experimental data.
Date: January 1, 1980
Creator: Simonen, F.A. & Shippell, R.J. Jr.
Partner: UNT Libraries Government Documents Department

Parametric calculations of fatigue-crack growth in piping. [PWR; BWR]

Description: A major objective of this program is to provide data that can be used to formulate recommended revisions to ASME Section XI and regulatory requirements for inservice inspection of piping and pressure vessels. This study presents calculations of the growth of piping flaws produced by fatigue. Flaw growth was predicted as a function of the initial flaw size, the level and number of stress cycles, the piping material, and environmental factors.
Date: March 1, 1983
Creator: Simonen, F.A. & Goodrich, C.W.
Partner: UNT Libraries Government Documents Department

Analyses of the impact of inservice inspection using a piping reliability model

Description: This report presents the results of a study of the impact of inservice inspection (ISI) programs on the reliability of specific nuclear piping systems that have actually failed in service. Two major factors are considered in the ISI programs: one is the capability of detecting flaws; the other is the frequency of performing ISI. A probabilistic fracture mechanics model is used to estimate the reliability of two nuclear piping lines over the plant life as functions of the ISI programs. Examples chosen for the study are the PWR feedwater steam generator nozzle cracking incident and the BWR recirculation line safe-end cracking incident. The results show that an effective inservice inspection requires a suitable combination of flaw detection capability and inspection schedule. An augmented inspection schedule is required for piping with fast-growing flaws to ensure that the inspection is done before the flaws reach critical sizes. Also, the elimination of poor inspection teams through training and qualification testing can produce significant benefits to ISI effectiveness.
Date: July 1, 1984
Creator: Simonen, F.A. & Woo, H.H.
Partner: UNT Libraries Government Documents Department

Stress analysis and testing of the outer capsule design for the Strontium Heat Source Development Program

Description: The objective of the Strontium Heat Source Development Program is to obtain the data needed to license /sup 90/SrF/sub 2/ heat sources - specifically the /sup 90/SrF/sub 2/ capsules produced in the Waste Encapsulation and Storage Facility (WESF) at Hanford. Toward this end, a high integrity outer capsule has been designed to replace the present outer capsule of the WESF /sup 90/SrF/sub 2/ capsule. The proposed design of a Hastelloy S outer capsule which features a mechanical interlock type of end closure is described. Qualification testing requirements are outlined, and stress analyses and developmental tests are described. These tests were performed on AISI-1018 steel stand-in capsules, and included both external pressure and impact tests. The external pressure tests showed that stress calculations seriously overestimated the pressure capability of the outer capsule. Possible reasons for the lack of agreement between the tests and the analyses are evaluated. The stress analyses and tests results indicate that the proposed outer capsule will meet the heat source qualification requirements. Future tests will be conducted to experimentally verify that the Hastelloy S outer capsule in an aged condition meets the structural integrity requirements.
Date: January 1, 1980
Creator: Simonen, F.A.; Shippell, R.J. Jr. & Atteridge, D.G.
Partner: UNT Libraries Government Documents Department

Fabrication Flaw Density and Distribution In Repairs to Reactor Pressure Vessel and Piping Welds

Description: The Pacific Northwest National Laboratory is developing a generalized fabrication flaw distribution for the population of nuclear reactor pressure vessels and for piping welds in U.S. operating reactors. The purpose of the generalized flaw distribution is to predict component-specific flaw densities. The estimates of fabrication flaws are intended for use in fracture mechanics structural integrity assessments. Structural integrity assessments, such as estimating the frequency of loss-of-coolant accidents, are performed by computer codes that require, as input, accurate estimates of flaw densities. Welds from four different reactor pressure vessels and a collection of archived pipes have been studied to develop empirical estimates of fabrication flaw densities. This report describes the fabrication flaw distribution and characterization in the repair weld metal of vessels and piping. This work indicates that large flaws occur in these repairs. These results show that repair flaws are complex in composition and sometimes include cracks on the ends of the repair cavities. Parametric analysis using an exponential fit is performed on the data. The relevance of construction records is established for describing fabrication processes and product forms. An analysis of these records shows there was a significant change in repair frequency over the years when these components were fabricated. A description of repair flaw morphology is provided with a discussion of fracture mechanics significance. Fabrication flaws in repairs are characterized using optimized-access, high-sensitivity nondestructive ultrasonic testing. Flaw characterizations are then validated by other nondestructive evaluation techniques and complemented by destructive testing.
Date: April 1, 2008
Creator: Schuster, G. J.; Simonen, F. A. & Doctor, S. R.
Partner: UNT Libraries Government Documents Department

Acceptance criteria for ultrasonic flaw indications in the inner liner of double-shell waste storage tanks

Description: Radioactive defense waste, resulting from the chemical processing of spent nuclear fuel, has been stored in double-shell tanks (DSTS) at the Hanford Site since 1970. As part of the program to assure that the DSTs maintain their structural integrity, an inspection plan is being developed and implemented. This report provides recommendations and technical bases for acceptance criteria for flaw indications detected during ultrasonic inspection of inner liners of the DSTS. The types of indications addressed are crack-like flaws, wall thinning, and pitting. In establishing acceptable flaw sizes, the evaluations have taken into consideration the potential for crack growth by the mechanism of stress corrosion cracking. Consideration was given to technical approaches used in ASME Codes, for reactor tanks at the Department of Energy Savannah River facilities, and in recommendations by the Tank Structural Integrity Panel. The goal was to ensure that indications discovered during inspections are not large enough to ever cause a leak or rupture of the tank inner liner. The acceptance criteria are intended to be simple to apply using a set of tables giving acceptable flaw sizes. These tables are sufficiently conservative to be applicable to all double-shell tanks. In those cases that a flaw exceeds the size permitted by the tables, it is proposed that additional criteria permit more detailed and less conservative evaluations to address specific conditions of stress levels, operating temperature, flaw location, and material properties.
Date: July 1, 1995
Creator: Simonen, F.A.; Graves, R.E. & Johnson, K.I.
Partner: UNT Libraries Government Documents Department

Pipe-to-pipe impact program

Description: This report documents the tests and analyses performed as part of the Pipe-to-Pipe Impact (PTPI) Program at the Pacific Northwest Laboratory. This work was performed to assist the NRC in making licensing decisions regarding pipe-to-pipe impact events following postulated breaks in high energy fluid system piping. The report scope encompasses work conducted from the program's start through the completion of the initial hot oil tests. The test equipment, procedures, and results are described, as are analytic studies of failure potential and data correlation. Because the PTPI Program is only partially completed, the total significance of the current test results cannot yet be accurately assessed. Therefore, although trends in the data are discussed, final conclusions and recommendations will be possible only after the completion of the program, which is scheduled to end in FY 1984.
Date: June 1, 1984
Creator: Alzheimer, J.M.; Bampton, M.C.C.; Friley, J.R. & Simonen, F.A.
Partner: UNT Libraries Government Documents Department

EVALUATION OF CONCRETE PROPERTY DATA AT ELEVATED TEMPERATURES FOR USE IN THE SAFE-CRACK COMPUTER CODE

Description: Design and analysis of Hanford double-shell waste storage tanks has made use of the finite element computer code SAFE-CRACK as a check of the concrete portion of the tank design after cmpletion of design. Rockwell Hanford Operations, the site contractor responsible for operation of the tanks, has requested Battelle Pacific Northwest Laboratory (PNL) to evaluate the use of the Hanford concrete property data at elevated temperatures by the SAFE-CRACK code. The purpose of this investigation is to evaluate the proper use of the mathematical expressions in SAFE-CRACK to best define the physical concrete properties extrapolated from the documented concrete property data when subjected to elevated temperatures and cyclic temperature variations.
Date: October 1, 1986
Creator: Henager, C. H.; Piepel, G. F.; Anderson, W. E.; Koehmstedt, P. L. & Simonen, F. A.
Partner: UNT Libraries Government Documents Department

A pilot application of risk-informed methods to establish inservice inspection priorities for nuclear components at Surry Unit 1 Nuclear Power Station. Revision 1

Description: As part of the Nondestructive Evaluation Reliability Program sponsored by the US Nuclear Regulatory Commission, the Pacific Northwest National Laboratory has developed risk-informed approaches for inservice inspection plans of nuclear power plants. This method uses probabilistic risk assessment (PRA) results to identify and prioritize the most risk-important components for inspection. The Surry Nuclear Power Station Unit 1 was selected for pilot application of this methodology. This report, which incorporates more recent plant-specific information and improved risk-informed methodology and tools, is Revision 1 of the earlier report (NUREG/CR-6181). The methodology discussed in the original report is no longer current and a preferred methodology is presented in this Revision. This report, NUREG/CR-6181, Rev. 1, therefore supersedes the earlier NUREG/CR-6181 published in August 1994. The specific systems addressed in this report are the auxiliary feedwater, the low-pressure injection, and the reactor coolant systems. The results provide a risk-informed ranking of components within these systems.
Date: February 1, 1997
Creator: Vo, T.V.; Phan, H.K.; Gore, B.F.; Simonen, F.A. & Doctor, S.R.
Partner: UNT Libraries Government Documents Department

VISA: a computer code for predicting the probability of reactor pressure-vessel failure. [PWR]

Description: The VISA (Vessel Integrity Simulation Analysis) code was developed as part of the NRC staff evaluation of pressurized thermal shock. VISA uses Monte Carlo simulation to evaluate the failure probability of a pressurized water reactor (PWR) pressure vessel subjected to a pressure and thermal transient specified by the user. Linear elastic fracture mechanics are used to model crack initiation and propagation. parameters for initial crack size, copper content, initial RT/sub NDT/, fluence, crack-initiation fracture toughness, and arrest fracture toughness are treated as random variables. This report documents the version of VISA used in the NRC staff report (Policy Issue from J.W. Dircks to NRC Commissioners, Enclosure A: NRC Staff Evaluation of Pressurized Thermal Shock, November 1982, SECY-82-465) and includes a user's guide for the code.
Date: September 1, 1983
Creator: Stevens, D.L.; Simonen, F.A.; Strosnider, J. Jr.; Klecker, R.W.; Engel, D.W. & Johnson, K.I.
Partner: UNT Libraries Government Documents Department

Modeling of time-variant concrete properties at elevated temperatures

Description: The elevated-temperature, time-variant concrete property equations used by a previous version of SAFE-CRACK were evaluated via comparison to the Portland Cement Association (PCA) data for Hanford concrete. For some properties, the agreement was reasonably good, while for others there were significant differences. Since the previous SAFE-CRACK equations for Hanford concrete properties were developed before completion of the PCA study, the property equations were re-evaluated, modified where necessary, and fitted to the full PCA data base. Statistical methods were applied to produce expressions describing the uncertainties in the modified equations. Uncertainty expressions for both expected property behavior (confidence bands) and ranges of property behavior (tolerance bands) were developed.
Date: April 1, 1988
Creator: Henager, C.H.; Piepel, G.F.; Anderson, W.E.; Koehmstedt, P.L. & Simonen, F.A.
Partner: UNT Libraries Government Documents Department

PNL Technical Review of Pressurized Thermal Shock Issues Supplement 1: Technical Critique of the NRC Near-Term Screening Criteria

Description: Pacific Northwest Laboratory (PNL) provided a technical critique of the draft report, NRC Staff Evaluation of Pressurized Thermal Shock, dated September 13, 1982. This report provided the basis for the NRC near-term regulatory position on pressurized thermal shock {PTS) and recommended a generic screening criteria for welds in the vessel beltline region. The PNL staff concluded that the screening criteria were adequate to meet the intent of the NRC safety goal and to retain past predictions of vessel reliability. The conclusion was based on selecting the plant-specific nilductility transition reference temperature (RT{sub NDT}) in the conservative manner described within the staff report. Conservative and unconservative factors were mentioned throughout the NRC staff report. The PNL staff has listed these factors together with unknown (may be either conservative or unconservative) factors and estimated, where possible, the range in °F RT{sub NDT}. The unknown factors were so widespread that the PNL staff recommended that specific conservatisms not be reduced until the unknowns are further resolved.
Date: May 1, 1983
Creator: Pederson, L. T.; Apley, W. J.; Bian, S. H.; Pelto, P. J.; Simonen, E. P.; Simonen, F. A. et al.
Partner: UNT Libraries Government Documents Department

Integration of NDE Reliability and Fracture Mechanics

Description: The Pacific Northwest Laboratory is conducting a four-phase program for measuring and evaluating the effectiveness and reliability of in-service inspection (lSI} performed on the primary system piping welds of commercial light water reactors (LWRs). Phase I of the program is complete. A survey was made of the state of practice for ultrasonic rsr of LWR primary system piping welds. Fracture mechanics calculations were made to establish required nondestrutive testing sensitivities. In general, it was found that fatigue flaws less than 25% of wall thickness would not grow to failure within an inspection interval of 10 years. However, in some cases failure could occur considerably faster. Statistical methods for predicting and measuring the effectiveness and reliability of lSI were developed and will be applied in the "Round Robin Inspections" of Phase II. Methods were also developed for the production of flaws typical of those found in service. Samples fabricated by these methods wilI be used in Phase II to test inspection effectiveness and reliability. Measurements were made of the influence of flaw characteristics {i.e., roughness, tightness, and orientation) on inspection reliability. These measurernents, as well as the predictions of a statistical model for inspection reliability, indicate that current reporting and recording sensitivities are inadequate.
Date: March 1, 1981
Creator: Becker, F. L.; Doctor, S. R.; Heas!er, P. G.; Morris, C. J.; Pitman, S. G.; Selby, G. P. et al.
Partner: UNT Libraries Government Documents Department

Revisiting the Integrated Pressurized Thermal Shock Studies of an Aging Pressurized Water Reactor

Description: The Integrated Pressurized Thermal Shock (IPTS) studies were a series of studies performed in the early-mid 1980s as part of an NRC-organized comprehensive research project to confirm the technical bases for the pressurized thermal shock (PTS) rule, and to aid in the development of guidance for licensee plant-specific analyses. The research project consisted of PTS pilot analyses for three PWRs: Oconee Unit 1, designed by Babcock and Wilcox; Calvert Cliffs Unit 1, designed by Combustion Engineering; and H.B. Robinson Unit 2, designed by Westinghouse. The primary objectives of the IPTS studies were (1) to provide for each of the three plants an estimate of the probability of a crack propagating through the wall of a reactor pressure vessel (RPV) due to PTS; (2) to determine the dominant overcooling sequences, plant features, and operator actions and the uncertainty in the plant risk due to PTS; and (3) to evaluate the effectiveness of potential corrective actions. The NRC is currently evaluating the possibility of revising current PTS regulatory guidance. Technical bases must be developed to support any revisions. In the years since the results of IPTS studies were published, the fracture mechanics model, the embrittlement database, embrittlement correlation, inputs for flaw distributions, and the probabilistic fracture mechanics (PFM) computer code have been refined. An ongoing effort is underway to determine the impact of these fracture-technology refinements on the conditional probabilities of vessel failure calculated in the IPTS Studies. This paper discusses the results of these analyses performed for one of these plants.
Date: August 1, 1999
Creator: Bryson, J. W.; Dickson, T. L.; Malik, S. N. M. & Simonen, F. A.
Partner: UNT Libraries Government Documents Department

PNL technical review of pressurized thermal-shock issues. [PWR]

Description: Pacific Northwest Laboratory (PNL) was asked to develop and recommend a regulatory position that the Nuclear Regulatory Commission (NRC) should adopt regarding the ability of reactor pressure vessels to withstand the effects of pressurized thermal shock (PTS). Licensees of eight pressurized water reactors provided NRC with estimates of remaining effective full power years before corrective actions would be required to prevent an unsafe operating condition. PNL reviewed these responses and the results of supporting research and concluded that none of the eight reactors would undergo vessel failure from a PTS event before several more years of operation. Operator actions, however, were often required to terminate a PTS event before it deteriorated to the point where failure could occur. Therefore, the near-term (less than one year) recommendation is to upgrade, on a site-specific basis, operational procedures, training, and control room instrumentation. Also, uniform criteria should be developed by NRC for use during future licensee analyses. Finally, it was recommended that NRC upgrade nondestructive inspection techniques used during vessel examinations and become more involved in the evaluation of annealing requirements.
Date: July 1, 1982
Creator: Pedersen, L.T.; Apley, W.J.; Bian, S.H.; Defferding, L.J.; Morgenstern, M.H.; Pelto, P.J. et al.
Partner: UNT Libraries Government Documents Department

Nondestructive examination (NDE) reliability for inservice inspection of light waters reactors

Description: Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from April 1988 through September 1988. 33 refs., 70 figs., 12 tabs.
Date: November 1, 1989
Creator: Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A. et al.
Partner: UNT Libraries Government Documents Department

Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors. Semiannual report, April 1992--September 1992: Volume 16

Description: The Evaluation and Improvement of NDE Reliability for Inservice inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs);using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel and other components inspected in accordance with Section XI of the ASME Code. This is a programs report covering the programmatic work from April 1992 through September 1992.
Date: November 1, 1993
Creator: Doctor, S. R.; Diaz, A. A.; Friley, J. R.; Greenwood, M. S.; Heasler, P. G.; Kurtz, R. J. et al.
Partner: UNT Libraries Government Documents Department

Progress in evaluation and improvement in nondestructive examination reliability for inservice inspection of Light Water Reactors (LWRs) and characterize fabrication flaws in reactor pressure vessels

Description: This paper is a review of the work conducted under two programs. One (NDE Reliability Program) is a multi-year program addressing the reliability of nondestructive evaluation (NDE) for the inservice inspection (ISI) of light water reactor components. This program examines the reliability of current NDE, the effectiveness of evolving technologies, and provides assessments and recommendations to ensure that the NDE is applied at the right time, in the right place with sufficient effectiveness that defects of importance to structural integrity will be reliably detected and accurately characterized. The second program (Characterizing Fabrication Flaws in Reactor Pressure Vessels) is assembling a data base to quantify the distribution of fabrication flaws that exist in US nuclear reactor pressure vessels with respect to density, size, type, and location. These programs will be discussed as two separate sections in this report. 4 refs., 7 figs.
Date: October 1, 1991
Creator: Doctor, S.R.; Bowey, R.E.; Good, M.S.; Friley, J.R.; Kurtz, R.J.; Simonen, F.A. et al.
Partner: UNT Libraries Government Documents Department