One of the most critical welds in the design of a large storage tank is the weld between the bottom annular plates and shell wall. The change in direction of the material introduces a biaxial stress, and possible stress concentration. This paper evaluates a method of determining a required inspection interval based on using a combination of conservative design modifications, risk based inspection (RBI) assessments, on site construction verification, and FEA analyses. In the case study presented, a design modification was made to specify this traditionally dual fillet weld as a full-penetration fillet weld. While this exceeds the typical design requirements, exceeding the minimum requirements, inspection codes do not provide acceptance criteria for the design. A more thorough volumetric inspection was conducted, in addition to the typical magnetic particle inspection, to confirm the full penetration of the weld was achieved. In making conservative changes to the design, more robust inspections, and performing RBI methodology, a longer inspection interval can be determined due to the increased information available on the outset. These small changes at the start of the construction can lead to increased operating time without an intrusive inspection through the lifespan of the tank.

Code Case N-513 provides evaluation rules and criteria for temporary acceptance of flaws, including through-wall flaws, in moderate energy piping. The application of the Code Case is restricted to moderate energy, Class 2 and 3 systems, so that safety issues regarding short-term, degraded system operation are minimized. The first version of the Code Case was published in 1997. Since then, there have been four revisions to augment and clarify the evaluation requirements and acceptance criteria of the Code Case that have been published by ASME. The technical bases for the original version of the Code Case and the four revisions have been previously published [1, 2, and 3]. There is currently work underway to incorporate additional changes to the Code Case and this paper provides the technical basis for the changes proposed in a fifth revision. These changes include clarification for buried piping, investigation of various radii used in the Code Case, removal of the 0.1 limit on the flexibility characteristic for elbow flaw evaluation, and an update of the stress intensity factor parameters for circumferential through-wall flaws. In addition, a new flaw evaluation procedure is given for through-wall flaws in gate valve body ends. This procedure evaluates flaws in the end of the valve body as if in straight pipe. These changes and their technical bases are described in this paper. Clarifications and changes deemed editorial are not documented in this paper.

Welds are inspected by various techniques which include visual examination, surface examination and volumetric examination. While the above techniques would qualify a weld to workmanship criteria, they would not necessarily be indicative of weld properties. Preparation and qualification of welding procedures and testing of production welds are indicative that the weldment would probably provide a safe and satisfactory service life. However, weldments have to operate at their design conditions which may include high temperatures and ASME Codes do not necessarily stipulate tests for verification of high temperature properties. In addition, defective welds are often repaired by removing the originally deposited weld metal and re-welding. The effects of double heat input are not necessarily evaluated. In this paper, an insight is provided into the factors which provide assurance that weldments will perform satisfactorily in service and the combination of non-destructive evaluation methods which would enable effective detection of imperfections. Non-destructive volumetric examination method for welds has traditionally be radiography. With the advent of automated data acquisition methods in Ultrasonics, like Time of Flight Diffraction and Phased Array Ultrasonic Testing, these methods are rapidly replacing radiographic methods for weld inspection. Ultrasonic acceptance criteria in ASME Section VIII Div. 1, ASME Section VIII Div. 2 and ASME Section IX do not include evaluation of porosity as ultrasonic methods do not easily detect porosity. The result of all this is that today we are accepting welders qualified using Ultrasonic examination as per ASME Section IX but on the job there is still the option of inspecting the weld using Radiography in which, excessive porosity can be a cause for weld repair. Considering this and various other criteria, a comprehensive weld evaluation methodology is proposed taking advantage of the strengths of each inspection technique while welding technology used would ensure that welds have required properties at service temperatures. A proposal is also made to improve the detectability of imperfections using modifications of existing Ultrasonic A-scan Techniques.

Requirements for pressure-temperature limits to protect against rupture of CANDU nuclear reactor Zr-Nb pressure tubes are provided in the Canadian Standards Association (CSA) Standard N285.8. The requirements are based on a stability evaluation of a postulated axial through-wall flaw for all ASME Service Level A, B, C and D loadings. The flaw stability evaluation is strongly dependent on the fracture toughness of the Zr-Nb pressure tube material. The fracture toughness of Zr-Nb pressure tubes is decreasing with operating hours. The decrease in fracture toughness as well as compounding conservatisms based on using bounding values make deterministic evaluations more challenging. The CSA Standard N285.8 permits probabilistic evaluations of fracture protection, but does not provide acceptance criteria. Proposed acceptance criteria that meet the intent of the design basis for Zr-Nb pressure tubes have been developed. The proposed acceptance criteria consist of a proposed maximum allowable conditional probability of pressure tube rupture for the entire reactor core, as well as a proposed maximum allowable conditional probability of rupture of a single pressure tube. The paper provides a description of the technical basis for the proposed acceptance criteria for probabilistic evaluations of fracture protection.

This paper presents an overview of added features in a new version of the FAVOR ( F racture A nalysis of V essels O ak R idge) computer code called FAVOR-OCI. The original FAVOR code was developed at the US Department of Energy’s Oak Ridge National Laboratory (ORNL) under the sponsorship of the US Nuclear Regulatory Commission (NRC). FAVOR is applied by US and international nuclear power industries to perform deterministic and probabilistic fracture mechanics analyses of commercial nuclear reactor pressure vessels (RPVs). Applications of FAVOR are focused on insuring that the structural integrity of aging, and increasingly embrittled, RPVs is maintained throughout their licensed service life. Based on the final ORNL release of FAVOR, v16.1, FAVOR-OCI extends existing deterministic features of FAVOR while preserving all previously-existing probabilistic capabilities of FAVOR. The objective of this paper is to describe new deterministic features in FAVOR-OCI that can be applied to analytical evaluations of planar flaws. These evaluations are consistent with the acceptance criteria of ASME Code, Section XI, Article IWB-3610, including Subarticles IWB-3611 (acceptance based on flaw size) and IWB-3612 (acceptance based on applied stress intensity factor). The linear elastic fracture mechanics (LEFM) capabilities of FAVOR-OCI also incorporate the analytical procedures presented in the Nonmandatory Appendix A, Analysis of Flaws , Article A-3000, Method of K I Determination , for both surface and subsurface (embedded) flaws. The paper describes a computational methodology for determining critical values of fracture-related parameters that satisfy ASME Code Section XI acceptance criteria for flaws exposed to multiple thermal-hydraulic transients. These compute-intensive analyses can be carried out with a single execution of FAVOR-OCI. The new methodology determines critical values by solving for either the point of tangency or point of intersection between applied K I versus time histories and a user-selected cleavage initiation toughness material property (e.g., ASME K Ic , FAVOR Weibull K Ic , or Master Curve Weibull K Jc ) for surface or subsurface flaws. Situations where warm prestress conditions apply can also be addressed. The paper highlights a need for this new capability via applications to a recent independent review of safety cases for RPVs in two Belgian nuclear power plants (NPPs). That review required ASME Section XI assessments of several thousand embedded, quasi-laminar flaws in the wall of each RPV Analysis results provided by the new capability contributed to the technical bases compiled from several sources by the Belgian nuclear regulatory agency (FANC) and eventually used by FANC to justify the restart of these NPPs.

Regions of higher-than-normal carbon content due to carbon macro-segregation have been found in large, pressure retaining forged ferritic steel components in some nuclear reactors. Higher carbon content in ferritic steel can decrease the resistance to fracture from the presence of flaws in the material. Acceptable margins against failure of pressurized components in nuclear safety systems must be maintained throughout their service life to ensure core integrity for all operational and postulated transient loading events. Should carbon macro-segregation substantially reduce the material resistance to fracture in safety components, then the margins against through-wall flaw propagation may fall below those specified by regulatory requirements to ensure adequate component and reactor core integrity. Probabilistic fracture mechanics (PFM) analyses were performed to assess the risk and structural significance of postulated carbon macro-segregation in large, forged pressure retaining components in pressurized water reactors (PWRs). The risk assessment was performed to evaluate several forged components and two classes of loading events. The forged components include the ring and head forgings in the reactor pressure vessel (RPV), steam generator (S/G) and pressurizer. The loading events used in the risk evaluation include pressurized thermal shock (PTS) transient events and a normal RPV cooldown event. The analyses included a range of component dimensions, surface and embedded flaw distributions, various levels of carbon macro-segregation up to and beyond the maximum measured values for the components, and the effects of neutron irradiation, including the effects of potential copper and phosphorus co-segregation. The PFM analyses were performed using the software, Fracture Analysis of Vessels, Oak Ridge (FAVOR). The results from the risk assessment indicate that: acceptable margins against failure are maintained through an 80-year operating interval even if carbon macro-segregation were to be present in RPV, S/G and pressurizer ring and head forgings in PWRs; and the risk associated with the presence of carbon macro-segregation in PWR ring and head forgings is significantly lower than regulatory risk related acceptance criteria.

A new degradation mechanism has been observed in Monel 400 Steam Generator tubing material, a nickel-copper alloy (63Ni-28Cu-2½Fe) with the ASME material designation SB-163/N04400. The location is above the top preheater support plate of the two re-circulating steam generator in one of the units of the Pickering Nuclear Generating Station. This paper provides a brief description of the regulatory environment, OPG’s steam generator life cycle management plans, the Canadian Industry’s fitness-for-service guidelines for steam generator tubes, and the afflicted steam generators. The paper then goes on to discuss the following activities that were conducted to support the technical basis to justify that the steam generators fit to be returned to service: • Inspection scope expansion, methods, and results. • Examination of removed tubes. • Condition monitoring assessment. • Operational assessment. • Burst-pressure tests of removed tubes and of fabricated test specimens. • Degradation specific flaw model and acceptance standards. • Flaw growth rate predictions. • Plugging limit adopted.

This paper proposes Piping Code rules to address the effects of hydrogen deflagrations inside piping. Previous work proposed a set of criteria for piping subject to detonation loading [PVP2012-78519, PVP2012-78525]. This paper provides criteria to evaluate the effect of deflagrations, which typically have a slower rise time and lower energy, inside the piping. These deflagration criteria, coupled with the previously cited detonation criteria, are being used at the Hanford Tank Waste Treatment and Immobilization Plant to evaluate piping systems subject to hydrogen accumulation. The previous papers did not investigate or propose criteria for deflagrations, as these were known to have lower pressures and slower pressure rise times, but are still of some significance for piping design. Recent work has shown that there exists a scenario in which the deflagration loading may be very significant: deflagrations in small gas pockets surrounded by large waste slugs. Depending on the assumptions used to develop the loading, the unbalanced forces on piping segments in a long piping system can become high during a deflagration event. Thus, for the set of criteria chosen for deflagration, the deflagration event may become the limiting event, especially if it is the more frequent event. The criteria proposed need to recognize this scenario and guide the user to possible solutions. This paper presents the original methodology for evaluating these “slug” events, briefly discusses the recent testing and theory being pursued to reduce the effect of the loading [PVP2015-45970, PVP2016-63260, PVP2016-63262], and then proposes criteria for evaluating deflagration induced stresses and loads.

Preventing brittle fracture is an essential part of instituting life-cycle management strategies for fixed pressure equipment. Using fracture mechanics principles to establish permissible minimum pressurization temperature (MPT) envelopes for components is one way to mitigate the potential for unstable flaw growth. In the refining industry, heavy-walled, low-alloy hydroprocessing reactors are designed to operate at elevated temperatures and high hydrogen partial pressures. Components that operate in high-pressure hydrogen environments require special treatment and necessitate guidance that falls outside the bounds of current pressure vessel construction codes. This operating environment results in two factors that affect the MPT envelope: long-term temper embrittlement and hydrogen embrittlement. Additionally, hydrogen charging can manifest damage in two ways: fast (brittle) fracture due to a reduction in fracture toughness and slow (subcritical) hydrogen-assisted crack growth. When developing a MPT envelope for a given component, both failure modes need to be considered in addition to residual stress effects from weld overlay or cladding. MPT envelopes provide insight into the permissible pressure-temperature combinations for specific locations and for chosen reference flaw sizes. From a reliability standpoint, understanding the risk of brittle fracture associated with heavy-walled reactors for all operating scenarios is crucial. Furthermore, taking the appropriate life-cycle management steps, such as establishing MPT envelopes, coupling MPT analysis predictions with targeted inspection, optimizing process operating conditions, and developing a flaw acceptance criteria to mitigate the risk of crack propagation and ultimately, brittle fracture is essential. In this study, a fracture-mechanics based methodology is summarized that is fully documented in upcoming Welding Research Council (WRC) Bulletin 562 [1] to determine MPT envelopes for all components (in any service environment) based on fast fracture with supplemental MPT requirements based on slow fracture for equipment that operates in high-pressure hydrogen environments. Finally, a finite element analysis-based case study of a 2-1/4-Cr-1-Mo hydrotreater reactor is summarized and practical life cycle-management guidance is offered based on analysis results. This example highlights how evaluation of start-up and shut-down procedures for heavy-walled reactors has the potential to save significant time and related cost per unit shut-down cycle, while maintaining an acceptable risk tolerance against subcritical crack growth and brittle fracture.

Drop test scenarios with cubic containers without impact limiters at interim storage sites or in a final repository have been investigated by numerical simulations. An ideally flat drop is impossible to conduct as a free fall of a container even under laboratory conditions. Dynamic stresses and strains inside the container structure are sensitive to the impact angle. Even very small impact angles cause remarkable changes in the experimental or numerical results when a flat bottom or wall of a container hits a flat target. For drop tests with transport packages the International Atomic Energy Agency (IAEA) regulations define an essentially unyielding target. In contrast, potential accident scenarios for storage containers are derived from site-specific safety analyses or acceptance criteria in Germany. Each interim storage site or repository has a yielding or so-called real target with individual structural and material properties. The real target acts as a kind of impact limiter. A more conservative container design is required if the impact limiting effect of the target is not considered.

According to the current ASME Code Section XI, IWB-3640 and Appendix C flaw evaluation procedure, cast austenitic stainless steel (CASS) piping with ferrite content less than 20% is treated as wrought stainless steel. For CASS piping with ferrite content equal or greater than 20%, there is currently no flaw evaluation procedure in the ASME Code. In this paper, the technical basis for a proposed flaw acceptance criteria for CASS piping is presented. The procedure utilizes the current rules in ASME Code Section XI, IWB 3640/Appendix C and the existing elastic-plastic correction factors (i.e., Z-factors) for other materials in the Code. The appropriate Z-factor to use for the CASS piping is determined based on the ferrite content (using Hull’s equivalent factor). Experimentally measured fully saturated fracture toughness and tensile data of the three most common grades of CASS material in the US (CF3, CF8 and CF8M) were used to determine the flaw acceptance criteria in the proposed method. The proposed method is conservative since it utilizes the fully saturated condition of CASS materials. In addition, it is simple and consistent with current regulatory guidance on aging management of CASS piping.

The part-through-wall crack perpendicular to the circumferential weld on the outside surface of a spent nuclear fuel (SNF) multiple-purpose canister (MPC) can be shown to be the most limiting fracture configuration driven by the welding residual stress (WRS). A series of semi-elliptical cracks of various sizes is chosen to calculate the stress intensity factors ( K ) under a bounding residual stress (i.e., the stress distribution that bounds all WRS in a canister). The threshold stress intensity factor ( K ISCC ) of the canister material in the storage environment is used to determine a critical flaw size, below which the stress corrosion cracking would be unlikely to take place. This result can be considered as the flaw disposition criterion should a surface flaw be detected during the inservice inspection as required by the aging management program (AMP), and can be proposed to American Society of Mechanical Engineers (ASME) Section XI Code Case N-860, “Examination Requirements and Acceptance Standards for Spent Nuclear Fuel Storage and Transportation Containment Systems.”

Mandatory Appendix XXVI in Section III of the ASME Code (2015 Edition) includes the requirements for the construction of Class 3 high density polyethylene (HDPE) pressure piping. Currently, there are no acceptance limits for subsurface voids in HDPE piping included in the Code. Data from recent EPRI tests conducted to determine the rate of growth of flaws in HDPE piping can be used to develop acceptance criteria for subsurface voids. The tests exposed specimens with surface flaws to sustained loads under elevated temperature and stress conditions. No failures were observed in the tests. Therefore, the results are useful to establish lower bound lifetime estimates for flawed piping under the maximum allowable temperature and stress conditions for buried pipe. The HDPE material has considerable ability to resist crack growth at temperatures well above the maximum allowable service conditions permitted in Appendix XXVI. The test data were used to establish allowable subsurface void sizes in HDPE piping. Correlations for rupture time that are based on the applied stress intensity and net section stress were used to determine an allowable subsurface void size. Because the creep failure mechanism in HDPE piping is time-dependent, an allowable void size was developed to ensure HDPE pipe integrity at the void location for a 50-year service life. The allowable flaw sizes can be used for the disposition of void indications identified in piping both before and after installation.

This paper will discuss the ASME Code Committee activities involved in the incorporation of surface stress improvement (SSI) into ASME Code Cases N-770-4 and N-729-5. ASME Code Cases N-770 [1] and N-770-1 introduced several mitigation approaches for dissimilar metal weld (DMW) locations in PWR primary system piping and provided inspection relief for locations that were mitigated. The initial approaches contained in N-770 and N-770-1 included mechanical stress improvement and weld overlay methods that have a global stress relief effect to achieve a very low tensile surface stress state or a compressive stress state at the weld inside surface to halt crack initiation, as well as growth of acceptably sized cracks. The weld overlay mitigation methods are also effective because they introduce PWSCC-resistant material, i.e., Alloys 52, 152, or their variants. (The initial approaches also included Alloy 52/152 weld inlay and weld onlay, methods that do not require stress improvement but do require access to the weld inside surface.) While the mechanical stress improvement and weld overlay methods address the majority of the DMW locations in the primary piping system, there are locations that cannot be treated by these approaches due to the weld geometry or access limitations for the needed equipment. Additionally the dissimilar metal J-groove welds in the reactor pressure vessel head penetration nozzles (RPVHPN) could not be addressed at all by the approaches developed for DMW locations. To address the industry need to mitigate the unfavorable DMW geometries and locations along with the RPVHPN locations, the use of surface stress improvement (SSI) was studied and documented in EPRI reports Materials Reliability Program (MRP)-267 [2], “Technical Basis for Primary Water Stress Corrosion Cracking by Surface Stress Improvement,” and MRP-335 [3], “Topical Report for Primary Water Stress Corrosion Cracking by Surface Stress Improvement.” These reports formed the technical basis for the SSI-related changes made in Code Cases N-770-4 and N-729-5. Along with the technical bases noted, support from the international community in terms of operational experience with SSI in their power plants was invaluable in providing the necessary understanding, context, and confidence to committee members. The ASME “Task Group High Strength Nickel Alloy Issues” (TGHSNAI) was assigned the task of revising the existing Code Cases, N-770 [1], “Alternate Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities” and N-729 [4], “Alternate Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds.” To incorporate the SSI approach into these Code Cases, the first action was to determine whether the SSI process was considered to be a peening process as defined by ASME Section III NB-4422 criteria. This required the submittal of an Interpretation of NB-4422 to determine if SSI techniques were considered a peening process under ASME Section III. The interpretation (Interpretation III-1-13-03), documented in ASME File 12-1192 [5], specified that SSI was not considered peening by Section III. This interpretation provided the framework by which SSI could be directly applied to ASME Section XI inspection criteria without the need to first revise ASME Section III NB-4422. SSI (peening) was first incorporated into Code Case N-770 [1] to provide a mitigation alternative for locations unable to be addressed by the methods addressed thus far. The revision to Code Case N-770 [1] does not provide guidance for the application of SSI activities but rather, it provides the process performance criteria and the inspection guidance following the application of SSI and establishes the pre-application inspection acceptance criteria. Following the approval of SSI in Code Case N-770 [1] addressing the DMW in the primary coolant piping system, the SSI approach was applied to the partial penetration dissimilar metal J-groove welds in RPVHPNs in Code Case N-729 [4]. The application to RPVHPNs provides the industry with a valuable asset preservation tool while significantly lowering the safety risks associated with primary water stress corrosion cracking (PWSCC) and degradation from borated water leakage for the RPVHPNs.

Flaws found during in-service inspection of CANDU (1) Zr-2.5Nb pressure tubes include fuel bundle scratches, debris fretting flaws, fuel bundle bearing pad fretting flaws and crevice corrosion flaws. These flaws are volumetric and blunt in nature. Crack initiation from in-service flaws can be caused by the presence of hydrogen in operating pressure tubes and resultant formation of hydrided regions at the flaw tips during reactor heat-up and cool-down cycles. Zr-2.5Nb pressure tubes in the as-manufactured condition contain hydrogen as an impurity element. During operation, the pressure tube absorbs deuterium, which is a hydrogen isotope, from the corrosion reaction of the zirconium with the heavy water coolant. In addition, deuterium ingresses into the pressure tube in the rolled joint region. The level of hydrogen isotope in pressure tubes increases with operating time. Over the years, Canadian CANDU industry has carried out extensive experimental and analytical programs to develop evaluation procedures for crack initiation from in-service flaws in Zr-2.5Nb pressure tubes. Crack initiation experiments were performed on pressure tube specimens with machined notches to quantify resistance to crack initiation under various simulated flaw geometries and operating conditions such as operating load and hydrogen concentration. Predictive engineering models for crack initiation have been developed based on understandings of crack initiation and experimental data. A set of technical requirements, including engineering procedures and acceptance criteria, for evaluation of crack initiation from in-service flaws in operating pressure tubes has been developed and implemented in the CSA Standard N285.8. A high level review of the development of these flaw evaluation procedures is described in this paper. Operating experience with the application of the developed flaw evaluation procedure is also provided.

For the assurance of fitness-for-service of CANDU Pressure Tubes (PTs), guidelines and acceptance criteria are provided in Canadian Standard Association (CSA) N285.8-15, Technical requirements for in-service evaluation of zirconium alloy pressure Tubes in CANDU reactors . With respect to the assessment of risk of operation associated with degradation mechanisms and aging of the PTs in the entire core of a given reactor Unit, Clause 7 of CSA N285.8 allows Licensee’s to use either a deterministic or probabilistic method to assess the likelihood of PT failures. When a probabilistic method is used, the Licensee is obligated to demonstrate that the combined frequency of PT failure(s) over the evaluation period, due to the various potential degradation mechanisms, is less than the maximum acceptable frequency provided in Table C.1 of CSA N285.8-15. The maximum acceptable frequency provided in Table C.1 of CSA N285.8-15 was developed in the early-1990’s based on reactor operating experience and knowledge at that time, Station Siting Guides and Consultative Regulatory Guide C-006 (Revision 1). A task group was established by the CSA N285.8 Technical Steering Committee to re-evaluate the allowable failure frequencies to confirm that they remain relevant given the current state of knowledge and the additional evaluation tools available. This paper provides Canadian Nuclear Safety Commission staff views regarding the technical basis for revisions to the allowable frequencies based upon current industry practices in conducting probabilistic core assessments.

The conditions of continued dry storage of the spent nuclear fuel in multipurpose canisters render the canisters, a component for confinement in dry storage cask systems, susceptible to chloride-induced stress corrosion cracking (SCC). The requisite conditions involve deposits of chloride-bearing marine salts and/or dust that deliquesce on the external surface of the cooling canister to create brine at weld residual stress regions. The subcritical crack growth rate at this “dry salt” condition, investigated by several researchers, has shown a relatively slow growth rate compared to chloride-cracking under aqueous conditions. A new SCC growth rate test specimen configuration has been developed to enable an initially dried salt assemblage to deliquesce under temperature and humidity conditions to load the fatigue pre-cracked, wedge-opening-loaded (WOL) specimen with the brine and enable measurements of crack growth rate (da/dt) under falling stress intensity factor, K J , conditions. The application of the results to a canister weldment with a residual stress profile to predict crack extension in time is described. The results are evaluated in terms of development of acceptance standards for this type of flaw, should SCC be identified and characterized through inservice inspection (ISI).

The use of non-destructive examination (NDE) for assessing the quality of butt fusion joints in polyethylene (PE) pipes has been included in the draft Mandatory Appendix XXVI to Section III of the ASME Boiler and Pressure Vessel Code (Rules for construction of Class 3 buried polyethylene pressure piping). However, currently, there are no acceptance criteria for flaws in butt fusion joints in PE pipes. There is an ASME Task Group on flaw evaluation for PE pipe, which is developing a code case using linear elastic fracture mechanics (LEFM) to determine critical flaw sizes. However, the initial experimental crack growth data generated suggests that linear elastic fracture mechanics is not able to adequately describe slow crack growth in PE materials. In addition, this work is only considering planar lack of fusion flaws in the joint; it is not considering other critical flaw types that can occur in butt fusion joints, such as particulate contamination and cold fusion. TWI has developed procedures using mechanical testing to develop flaw acceptance criteria for butt fusion joints in PE pipes. This is based on inserting lack of fusion flaws of known size and particulate contamination flaws of known concentrations into butt fusion joints and determining the effect of these flaws on both the short-term and long-term integrity of the joints. An important aspect of this work is to determine which of the wide array of mechanical tests available for assessing the integrity of butt fusion joints in PE pipes are the most discriminating. This paper describes the procedures developed for inserting simulated flaws into butt fusion joints in PE pipes, the experimental work to compare the results from different standard short-term and long-term tests on flawed and unflawed joints and the procedures developed to determine flaw acceptance criteria. Results have shown that the most discriminating short-term test for butt fusion joints in PE pipes is a tensile test using a waisted specimen, such as those defined in EN 12814-2, EN 12814-7 and ISO 13953, and the most discriminating property is the energy to break the specimen. The most appropriate long-term test for butt fusion joints in PE pipes is the whole pipe tensile creep rupture test, as defined in EN 12814-3; this is the only long-term whole pipe test that consistently generates slow crack growth in the fused joint, even if it contains no flaws.

Many stainless steel canisters for the dry storage of spent nuclear fuel are located in coastal regions. Because the heat treatment for relieving the welding residual stress is not required during fabrication, these canisters may be susceptible to chloride induced stress corrosion cracking due to the deliquescence of chloride-bearing marine salts or dust that enter the overpack system and deposit on the canister external surface. The NDE techniques and the associated delivery system are being developed to conduct periodic inservice inspections. The acceptance standards are needed to disposition findings should flaw-like indications be found. The instability crack lengths and depths for these flaws in the form of semi-elliptical shape near the welds are determined with R6 procedure. The cracks are subject to the canister design pressure and handling loads as well as the estimated welding residual stresses.

Seismic analysis and design of high-hazard nuclear facilities requires evaluation of soil-structure interaction (SSI) effects on structure and soil response due to earthquake ground motions. The industry-wide methodology of computing SSI response of buildings is through linear SSI techniques using the computer code SASSI. Technical issues were identified by users and regulators (Ref. 1) resulting in the U.S. Department of Energy (DOE) commissioning a large scale, multi-year Validation and Verification (V&V) Project for SASSI (Ref. 2). The project was a highly peer reviewed process that included an esteemed Participatory Peer Review Team, DOE oversight, as well as regulatory and stakeholder input. The project goal was to develop benchmark SSI solutions for the range of SSI problems associated with high-hazard facilities within the DOE complex. As per industry software quality control requirements, an acceptance limit for the benchmark solutions must be provided to define the acceptable accuracy of the results produced by SASSI relative to the benchmark. In order to define this limit, variation in the solution of foundation impedance must be related to an expected level of accuracy in structural design quantities of interest (e.g., response spectra, base shear, etc.). Therefore, extensive parametric analyses were performed for coupled soil-structure systems having a broad range of SSI parameters defined by foundation size, soil properties, building stiffness and mass properties, building height, etc. Based on the results of the parametric analyses, ASCE 4 code guidance, and engineering judgment of CJC&A and the peer review team, an acceptable level of accuracy in computed foundation impedance was determined for SASSI solutions. This supported the successful qualification of SASSI for use in two large-scale DOE projects.