Carbon Fiber Reinforced Polymer (CFRP) composites have been used for decades in various industries such as aerospace, oil and gas, mainly due to their high strength-to-weight-ratio and excellent corrosion resistance. However, the use of CFRP in nuclear industry is very limited. Recently, a new ASME Boiler and Pressure Vessel (BPV) Code Case N-871 has been accepted for internal repair of buried Class 2 and 3 nuclear safety related piping using CFRP for Service Levels A, B, C and D. The ASME BPV Code Case N-871 allows the use of both Allowable Stress Design (ASD) and Load and Resistance Factor Design (LRFD) methods to design the CFRP repair system. The USNRC has not yet accepted this code case. In 2016, a relief request was submitted to US Nuclear Regulatory Commission (US-NRC) by Surry Power Station, Virginia Electric and Power Company to perform internal repair of degraded ASME Class 2 and 3 safety related circulating and service water buried piping using LRFD method and CFRP, which was approved by US Nuclear Regulatory Commission (NRC). It is documented in the literature that CFRP materials experience property degradation when exposed to certain environments and creep behavior under sustained loading. From a limited number of tests, it has also been found that the overall strength of CFRP repair decreases with an increase in the number of plies, however, the extent of the strength reduction is not well-supported with test data. The statistical values of CFRP strength can be determined in three ways. The ASME BPV Code Case N-871 recommends using the characteristic value of CFRP strength at a 5 th percentile value with 80% confidence, whereas the A-basis (1 st percentile with 95% confidence) and B-basis (10 th percentile with 95% confidence) values of strength are used in other industries such as in aerospace applications. For a safety related nuclear application, it is therefore important to evaluate how these values compare to one another. The property degradation due to environmental exposure, creep behavior, multi-ply laminate as well as difference in various statistical analysis can be given a general term as strength reduction factor and These factors will have a direct effect on the safety margin of any CFRP repair using the ASD method and hence, it is essential to confirm these factors with an adequate number of experiments. In this work, mechanical testing and statistical analyses have been conducted to identify specific degradation mechanisms in CFRP as well as other design considerations that may affect the effective factor of safety of the repair system. These include the effect of multi-ply laminates, misalignment angle, width of the specimen as well as the characteristic values of the ultimate strength. Finally, the strength reduction factors are determined from test results followed by a discussion of how these factors affect the overall effective factor of safety of the repair system.

A modified 9-Cr alloy was developed by Oak Ridge National Laboratory, in early 1980s, to increase high temperature capabilities of ferritic steels for superheater tubing. The material improved high temperature creep properties by controlling alloying elements and microstructure. The material was added to ASME BPVC in 1983 (thru Code Case 1943) as Grade 91. Higher yield and tensile strength, in comparison to other low-Cr alloy steels (like Grade 22), allowed for fabrication of thinner component wall thickness. This in-turn reduced susceptibility to through-wall thermal stresses during transient events. Directionally this also reduced material costs. Consequently, petrochemical industries have utilized Grade 91 in applications at Heat Recovery Steam Generation units (HRSG), Steam Methane Reformers (SMR), CO2 Boilers and as convection coils in Ethane conversion units. Grade 91 material has complex microstructure and requires careful control of welding parameters to assure crack free welds that provide adequate creep ductility and retain creep strength at high temperatures. The current guidelines documented in API 582 and Technical Report 938 provide limited insights on success factors for weldability. Grade 91 material use has been growing in the recent past Petrochemicals Complex and in offshore applications, at once-throw steam generators (OTSG). The aim of this paper is to share experience on welding parameters. Guidance needs be adjusted to specific projects and repair activities.

Plasma Arc Welding (PAW) has been used for many critical applications due to its flexibility, reliability and high weld quality. In this paper, two code compliance plasma arc welding applications in the nuclear industry are discussed. The first application is an innovative welding process using PAW with specially designed dual hot wire feeding system, namely Dual Hot Wire Gas Metal Plasma Arc Welding (GMPAW). The GMPAW process offers unique advantages for high deposition and low dilution weld overlay application. The second application is a remote weld overlay repair from pipe inside diameter (ID) for a highly radiated nuclear component using PAW process with remote machining and NDE capability. In this paper, the benefits and advantages are provided for the aforementioned PAW applications. The versatility of plasma arc welding system configuration, as well as high quality and productivity can make plasma arc welding a good candidate for many critical code compliance applications.

Steam methane reforming is the most common method of hydrogen production relevant for plants in the petroleum upgrading, downstream refining, methanol, and ammonia industries. Owner-operators of steam methane reformer furnaces continue to make repair and replacement decisions that involve the cast outlet manifold fittings. One key part of these plans is assessment of the weldability and remaining life of the cast components. The 20Cr-32Ni-1Nb alloy casting materials typically used in the outlet manifolds are usually operated in the low end of their creep temperature range but are subject to metallurgical aging mechanisms which reduce their ductility, weldability, homogeneity, and fracture toughness. This paper covers the practices employed by several owner-users to optimize the lifecycle costs of the outlet manifold castings. These practices include but are not limited to controlled materials specifications, in-situ weldability tests, non-destructive testing in-situ and destructive testing post service, and repair practices such as annealing heat treatments. This paper also includes a limited survey of several owner-users and their fleets of reformer heaters. The details in the survey include the population of affected cast manifold components, alloy grades for the castings and welds, operating temperature ranges, number of startup and shutdown cycles, ranges of time in service, generic design details, and repair case studies. Also discussed are recent improvements in the state of the art for high temperature materials property data-gathering, as well as the structural modeling via Finite Element Methods. These new technologies are opportunities for future work to develop better strategies in the areas of condition assessment, repair planning, and remaining life prediction, taking into account the relevant parameters of installed manifold components, including: specific aging behavior of the casting chemistry, component mechanical design details, as well as the welding and heat treatment parameters during initial fabrication and subsequent maintenance activities.

As part of nuclear power plant ageing management, the increased probability of a need of repair welding must be taken into account along with the increase of plant lifetime. An essential prerequisite for successful and safe repair welding is that the applied welding procedures have been properly validated and qualified prior to their use. For instance, if no post-weld heat treatment can be performed and the desired tempering effect has to be based on temper-bead technique, a user needs to scan among several available repair welding procedures. A decision has to be made which of the procedures provides the maximum desired tempering effect with the case in question. This research is a part of a larger experimental effort developing repair welding techniques, and is a part of the Finnish Nuclear Power Plant Safety Research Programme SAFIR2022. The currently studied experimental repair welding case is a low-alloy steel mock-up with an austenitic cladding. Repair welding is assumed to represent a ‘worst-case’ scenario where a postulated linear crack-like defect exists beneath the cladding and might extend across the interface into the reactor pressure vessel steel side. This postulated defect will be removed by machining, and the thereby machined groove will be filled by repair welding using a nickel-base super alloy 52M filler metal by cold metal transfer-gas metal arc welding with a robotic arm. In this paper, different repair welding techniques and alternatives are shortly surveyed based on existing literature. Overall, published documentation was sparse. While only few studies were considered relevant in terms of established links to actual repair cases of under-cladding defects in reactor pressure vessels, others were mainly for modelling and simulation purposes without e.g. cladding groove preparation or the use of irradiation-embrittled material. Most of these procedures were based on the use of nickel-base alloy filler metal in the combination with temper-bead welding technique, with the aim at omitting both preheating and post-weld heat treatment. The main challenge in the repair weld design is to optimise all relevant welding parameters, including the thermal efficiency of temper-bead welding, in order to obtain a sound, defect-free weld with controlled reactor pressure vessel steel heat affected zone maximum hardness. In the simulations presented in the paper, the goal was to compute the resulting deformations, strains and stresses induced by the repair process and make a-priori estimates of the effectiveness of different repair techniques based on the numerical predictions. The numerical analyses allow the comparison of the procedures and enable selecting the one with most efficient combination of weld thermal cycles in terms of tempering and normalisation effects. The prediction of prevailing residual stresses is also important when further application of the component is considered. The paper is followed by Part II, in which the topics of experimental evaluation and material characterization of the repair weld are presented.

Aging management of the existing fleet of nuclear power plants is becoming an increasingly important topic, especially as many units are approaching their design lifetimes or are entering long-term operation. As these plants continue to age, there is an increased probability for the need of repairs due to extended exposure to a harsh environment. It is paramount that qualified and validated solutions are readily available. A repair method for a postulated through cladding crack into the low alloy steel of a nuclear power plant’s reactor pressure vessel has been investigated in this study. This paper is part of larger study that evaluates the current possibilities of such repair welds. The present paper documents the weld-trials and method selection. A parallel paper describes numerical simulations and optimization of weld parameters. The presented weld-trial represents a case where a postulated crack has been excavated and repaired using a nickel base Alloy 52M filler metal by gas metal arc welding-cold metal transfer with a robotic arm. A SA235 structural steel has been used as a base material in this weld-trial. No pre-heating or post-weld heat treatment will be applied, as it would be nearly impossible to apply these treatments in a reactor pressure vessel repair situation. While Alloy 52M presents good material properties, in terms of resistance to environmentally assisted degradation mechanisms, such as primary water stress corrosion cracking, it is notoriously difficult to weld. Some difficulties and challenges during welding include a sluggish weld puddle, formation of titanium and/or aluminium oxides and its susceptibility to lack of fusion defects and weld metal cracking, such as ductility dip cracking and solidification cracking. Moreover, gas metal arc welding-cold metal transfer is not traditionally used in the nuclear industry. Nonetheless, it presents some interesting advantages, specifically concerning heat input requirements and automation possibilities, as compared to traditional welding methods. The mechanical properties, in terms of indentation hardness, and microstructure of a weld-trial sample have been evaluated in this study. The fusion boundary and heat affected zone were the main areas of focus when evaluating the mechanical and microstructural properties. Detailed microstructural characterization using electron backscatter diffraction and nanoindentation were performed across the weld interface. Based on these results, the gas metal arc welding cold metal transfer is seen as a potential high-quality weld method for reactor pressure vessel repair cases.

The ASME Code, Section XI is working on guidance for application of weld overlay repairs to repair thermal fatigue cracking in nuclear piping systems. This new guidance will eventually be published as Code Case N-894 with the original technical basis in PVP2019-93360. Weld overlays have been extensively used in boiling water reactors (BWRs) and pressurized water reactors (PWRs) to mitigate stress corrosion cracking (SCC). The weld overlays applied to date mitigate SCC by putting the flaw into compression and they use materials that are resistant to SCC. Mitigation of thermal fatigue requires the crack to be in compression so that it does not achieve tensile cycling under the thermal fatigue loading condition. Code Case N-894 allows for the use of either stainless steel or nickel alloy filler metals to repair thermal fatigue flaws. This paper will evaluate the use of both filler metals for the weld overlay process to assess the performance difference between the two filler metals such that the welding advantages of stainless steel over nickel alloys can be quantified. Specifically, this paper assesses the residual stress state difference for nominal sized weld overlays on a six-inch pipe. Various cyclic thermal loading conditions are postulated, and the stress intensity factors are determined for both filler metals to assess the difference in mitigation of thermal fatigue flaws.

Compared with the steel cylinders for liquefied petroleum gas (LPG), fully-wrapped fiber reinforced cylinders with non-metallic liners (type IV cylinders) have been widely used due to their advantages of light weight, corrosion resistance, low maintenance cost and no explosion when exposed in fire. However, for the heterogeneity of raw material performances and instability of manufacturing process, the mechanical performances of type IV cylinders show a certain dispersion. Therefore, in this study, a series of burst tests and pressure cycle tests on 89 type IV LPG cylinders (26.2 L, 2.1 MPa) were conducted to examine the performances of burst pressure and number of pressure cycles, and analyze the distribution characteristics. The total length of the experimental cylinder was 465 mm, the outer diameter and the thickness of the liner was 304 mm and 2 mm, respectively. Two types of cylinders were tested, with the thickness of composite overwrap in 2 mm and 4 mm, corresponding to the designed burst pressure of 12 MPa and 15 MPa. Burst test and pressure cycle test were conducted according to the standard GB/T 15385 and GB/T 9252, respectively. Results showed that the 59 cylinders had the similar crevasses morphology along the axial wall of liners after burst tests. Burst pressure of two design types fluctuated within the range of 11.4 MPa-12.5 MPa, and 13.7 MPa-16.0 MPa, respectively. The statistical results of burst pressure showed approximately normal distribution with a standard deviation of 0.263 MPa and 0.529 MPa. Also, the number of pressure cycles of 30 cylinders were all above 12,000 with a cycle frequency no more than 15 cycles per minute. Above studies have provided a data support for the recently published Chinese association standard T/CATSI 02 003-2019.

Temper bead (TB) welding is routinely used as an alternative to post weld heat treatment (PWHT) when performing weld repairs on low alloy steels in the nuclear power industry. Historically, TB welding procedures have been qualified by Charpy V-notch (CVN) testing to demonstrate acceptable impact properties in the weld heat affected zone (HAZ). Part 1 of this work investigated a hardness drop criterion to demonstrate proper HAZ tempering using a vintage heat of SA-508 forged material (see PVP2019-93950). Optimization of hardness test parameters was also a major piece of the Part 1 scope of work. Part 2 of this work adds a more modern heat of SA-508 material and two heats of A-387 F22 material to the study. The overarching goal of this work is to support an alternative hardness test protocol for qualification of temper bead welding (see PVP2015-45663).

Since 1982 the nuclear industry has employed weld overlay repairs to address intergranular stress corrosion cracking (IGSCC) in boiling water reactors (BWR) and primary water stress corrosion cracking (PWSCC) in pressurized water reactors (PWR). The American Society of Mechanical Engineers (ASME) has created several documents to provide rules and guidelines for weld overlay repair of nuclear components that have experienced stress corrosion cracking (SCC). These documents include ASME Code Case N-504-4 and ASME Section XI, Nonmandatory Appendix Q which specifically address weld overlay repair of stainless steel components. Recently, stainless steel components that have experienced thermal fatigue cracking at the inner diameter surfaces have been repaired with structural weld overlays (SWOL) using the methodology of Code Cases N-504-4 and N-740-2. The SWOL is a good choice for repair of thermal fatigue cracks in piping because it provides structural reinforcement to the affected location and places the inside diameter (ID) surface into compression preventing, or significantly reducing, further flaw growth. However, the rules of Case N-504-4 and N-740 were not specifically written to address thermal fatigue cracking as the primary cause and may not adequately address design, analysis and examination requirements when thermal fatigue is the active mechanism because it is very different in nature than SCC. For example, SCC is driven by a combination of environment, steady state operating stresses, residual stresses from welding and fabrication processes, and operating temperature, whereas thermal fatigue is driven by thermal stress cycles resulting from fluid thermal cycling or stratification. The source of the thermal events that result in cracking may not be as well understood or predictable as SCC degradation. In addition, weld overlays applied to address SCC are constructed of SCC resistant material but are not resistant to thermal fatigue. Therefore, ASME Section XI recognized that alternative rules were needed for repair of piping damaged by thermal fatigue. This paper provides a technical basis for weld overlay repair of components that have experienced thermal fatigue cracking. It addresses design, analysis and examination requirements considering the nature of thermal fatigue in nuclear piping systems. The Code Case was originally drafted based on the industry accepted rules of Case N-504-4 and Appendix Q but includes appropriate modifications needed to address thermal fatigue cracking. These modifications include removing restrictions such as the delta ferrite limit for PWRs that is only applicable to address SCC in BWR environments, and enhancements to the examination requirements to ensure that the repaired location is adequately monitored throughout the remaining service life of the plant. The purpose of this paper is to document the technical basis for Code Case N-894, which is currently still under development by ASME Section XI.

Coke drums are pressure vessels used in the delayed coking process at oil refineries, which transform heavy residual oil into light-weight hydrocarbon molecules and solid coke through thermal cracking. Due to the severe thermal and mechanical loadings during operation, these vessels experienced low-cycle fatigue failure, which led to shell and skirt damage such as bulging and cracking. External weld repairs using the temper bead technique have been widely applied to repair damaged regions caused by bulging and cracking for preventing the leaks of the residual oil contents. However, a substantial proportion of the external repairs have been reported to experience post-repair cracking issues. In this study, an external weld repair mockup with a backing plate was evaluated through metallurgical characterizations and isothermal low-cycle fatigue tests. The micro-hardness mapping identified the mismatching interfaces from base metal to weld metal (WM) and from root passes to internal clad. Four types of dog-bone samples were extracted from the weld: weld metal, heat affected zone, internal clad and backing plate. These samples were used to evaluate the fatigue resistance of weld metal and transition zone under low-cycle fatigue tests. Failure analysis showed that weld metal samples were susceptible to multiple-crack initiations, while other transition samples failed at mismatching interfaces or stress concentration points at weld toes.

In recent time, solid state crack repair techniques in spent nuclear fuel dry storage canisters (SNF-DSC) have garnered significant interest. Plates of austenitic stainless steels are usually arc welded to fabricate the cylindrical canisters which are prone to chloride-induced stress corrosion cracking. Friction stir welding (FSW) is considered to be a potential candidate for damage repair. In this work, electrical discharge machining was used to simulate cracks in 304L stainless steel plates of 12.7 mm thickness. Subsequently, isothermal FSW was carried out at two different temperatures (725 and 825°C) in order to repair the simulated crack. Microstructure-mechanical property correlations in FSWed plates were studied in detail. Significant grain refinement observed in the stir zone (SZ) was attributed to dynamic recrystallization occurring during FSW. A comparison of Vickers microhardness showed higher hardness and homogeneity in the 725°C SZ than its 825°C counterpart. The fraction of Σ3 boundary is found to be low in the SZ as compared to base metal (BM). Yield strength of the joints was found to be approximately 100 MPa higher than that of the BM accompanied by a drop in ductility by a factor of 2. Residual stress profile across 825°C weld was measured using x-ray diffraction.

Engineered enclosures for the repair of piping, often referred to as clamps, are used to restore the integrity of in-service piping components as an interim measure until the damaged components can be repaired. They are typically used in the petroleum refining and chemical process industries, nuclear and non-nuclear utilities, and other on-shore and offshore facilities in the oil and gas industry. This type of enclosure has typically been viewed as a short term repair with a limited design life. However, it is the authors’ experience that properly engineered and installed enclosures perform in a manner consistent with the design criteria covering the original piping component(s) to which they have been installed. To investigate this further, a series of tests were conducted on several engineered enclosures that had been recently removed to replace the damaged component. Tests were also conducted on several new constructed enclosures of a similar configuration and size to those that had been recently removed. An analytical approach was also performed on one such enclosure-pipe configuration to support and study the experimental results in detail. This paper presents and summarizes the tests, analyses, and results from this work. The results provide information that can benefit users, standards developers, and regulatory authorities in better understanding the considerations affecting the design life of engineered enclosures.

For many nuclear power plants worldwide the operation period will be extended to 60 or 80 years in the coming years. As the operation period increases, the importance of knowledge of ageing mechanisms increases. In the framework of LTO there is limited knowledge about ageing and structural integrity of concrete structures. Recent developments have shown that ageing of civil structures receive more attention internationally (for example concrete degradation in bunker building Doel and buried piping attention in topical peer review EU). In order to increase the knowledge in the field of civil structures, this work focusses on investigation of ageing of civil structures and determining an ageing management strategy. Knowledge of the ageing mechanisms of civil structures and especially concrete, will lead to improvement of ageing management and assessment methods of concrete. As a first step international information was gathered on civil structures ageing issues and management thereof (see PVP2019-93029). In addition a very high level assessment methodology was proposed. The goal of the project is to obtain more knowledge on ageing management of civil structures and especially concrete. It will lead to an assessment method for civil ageing management and ageing management programs dealing with the relevant mechanisms at the various locations in a practical manner. The results of this ongoing work are presented in this report. The initial proposed assessment methodology has been tested by application to the HFR research reactor. The resulting list of relevant AMPs has been verified with the outcome for another PWR with a steel containment. With this experience the assessment methodology is improved. In addition each civil SSC in the scope of the Continued Safe Operation program is linked to the relevant AMP(s). The improved, but not finalized assessment method of ageing management for civil structures can be seen in figure 2. The proposed assessment method for ageing of civil structures has been tested, compared and improved. The results indicated that the method can be used to obtain a list of plant specific AMPs. The comparison of the list of relevant AMPs for a steel containment PWR, showed similar results. What is added to the assessment method is the link to the TLAAs for civil structures. The detailing to a level of practical work instructions for the maintenance of the plant has to be made in the near future. In the near future the step will be made from a high level of IGALL AMP to a practical AMP that will deal with the relevant mechanisms at the various locations. Therefore further steps are in studying of relevant degradation mechanisms, relevant AMPs (like AMP305 [9],AMP306 [10]) and applicable literature (e.g.[21]) in combination with the practical knowledge from operation of a reactor. The international developments on ageing management of concrete will be included. It is foreseen that the future report will contain information on concrete degradation mechanisms relevant for nuclear reactors. If findings requires calculations the assessment method will be verified with the finite element modelling techniques.

Composite repair systems have been gaining each time more space in industry, especially when it comes to repairing through-wall defects in pipes. They are simpler to apply, have no costly downtime and provide lower risks to the environment when compared to metallic repairs. ASME PCC-2 and ISO 24817 standards are responsible for defining the parameters necessary to a successful repair, however neither of them addresses a very common practice in such repairs, which is the addition of a bonded metallic patch over the defect. Several companies are adepts of such practice and it has already been proven that is actually the metallic patch and not the composite sleeve itself that sustains most of the load applied on the repair, and for that reason it becomes necessary to conduct further studies regarding the behavior of the patch alone. One important issue is to understand why the strength of similar repairs due to operation errors with very similar amplitude of pressure transients seems to vary randomly, with unexplained early failures. The present paper is concerned with an experimental study about how pressure variations can generate cyclic inelastic strains in the pipe, which can weaken the adhesion between pipe and patch, leading the repair to fail prematurely.

Nuclear power plants use lots of valves in different applications. Each valve type has specific tasks to perform. The vast majority of safety related valves are monitored during plant operation to ensure they are operable and can be relied upon to do their design functions. Unfortunately, check valves do not fall in the category of valves that can be readily monitored because of how they are designed; usually with all operating components internal to the pressure boundary. Since they are not easily monitored externally, plant operators typically disassemble and inspect them to assess the internal conditions and determine what actions to take, which can be either repair, replace, or use as found. However, not all check valves get inspected for a number of reasons resulting in some valves running to failure. In this paper, the authors present actions the nuclear industry has used to comply with ASME OM Code requirements (8) on check valves and US NRC regulatory activities in this area. The paper also covers the role the nuclear industry and other organizations have played to improve check valve reliability.

The nuclear industry has recently been shifting to value-based maintenance in order to keep nuclear power competitive in the power generation market. A key challenge in value-based maintenance is the optimization of a maintenance schedule. With most components having ten to fifteen available maintenances, the complexity of the optimization grows quite quickly. This paper presents a methodology for combining maintenance effectiveness, cost estimates, failure impacts, and overall reliability data to estimate an expected life cycle cost (LCC) for a component. The maintenance types are categorized into three types: monitoring, wear-rate reducing (e.g. oil change), or life-restoring (e.g. refurbishment). Each maintenance type has a different effect on problem detection, degradation rate, and future life expectancy. A Markov model uses the maintenance effects to estimate the distribution of what state of degradation a component is in at a specific time and concurrently, the component failures, maintenance costs, and failure impacts are tallied up in order to provide an expected life cycle cost for a given maintenance schedule. Optimization of the maintenance schedule is performed using the genetic algorithm where multiple maintenance schedules are simultaneously calculated, compared, and evolved in order to find the lowest expected life cycle costs. The genetic algorithm was selected as a suitable optimization algorithm for its ability to find relatively close approximations to the global optimum with relative ease while concurrently being able to handle non-smooth objective functions.

The initial assembly of a bolted flanged joint (BFJ) commonly includes several ramped torque steps using various bolt torque patterns, final circular pass(es) at 100% of target torque, and then the choice of whether and when to come back and retorque the just-assembled joint. The value of retorquing a newly assembled bolted flanged joint varies significantly by which gasket material and type is used in the joint. The additional maintenance costs of labor, tool rental, scaffolding, crane usage, etc. to perform a retorque on a joint can be significant, but the larger cost is generally the additional process downtime. Gaskets based on sheet materials are often retorqued. Conventional practice indicates that BFJs with gaskets made from flexible graphite or fiber sheets typically do not benefit from retorques, or at least do not benefit enough to offset the additional costs. Conversely, industry experience with many polytetrafluoroethylene (PTFE) based gaskets indicates these joints benefit significantly from retorquing a BFJ after initial assembly. This paper explores some practical methods to estimate potential retorque benefit as a function of retained gasket stress without the retorque at temperatures normally “in range” for the gasket material being considered. That is, at temperatures that would not cause thermal damage to the gasket material. For flexible graphite and fiber materials known to experience mechanical property degradation with time and/or temperature (“aging”), a typical 5-year life is used. For PTFE gaskets known to relax primarily with temperature and that do not appreciably age, use of the Hot Blowout Thermal Cycling test (HOBTC) can provide a quick evaluation of gasket stress loss as a function of temperature. Retained gasket stress in both cases can then be used to estimate tightness for the gasket material using the Room Temperature Tightness (ROTT) data. If the joint is still at or above required tightness, then retorquing the joint soon after initial assembly is of little benefit. If the gasket has lost enough stress, then a retorque can bring the gasket stress back into range to provide sufficient tightness. Knowing this, the end user can make better choices for both gasket material selection and assembly procedures in order to reduce overall job costs.

SA508 low-alloy steel for a reactor vessel was exposed to primary water in a pressurized water reactor (PWR) plant because the cladding layer of type 309 stainless steel for a reactor pressure vessel (RPV) was peeled off owing to an accident in which the thermal sleeve was detached. The advantage of electrochemical deposition (ECD) Ni plating techniques is that the RPV can be repaired without significant thermal effects, and Ni has sound corrosion resistance that can withstand exposure in primary water. The corrosion rate of the damaged part was assessed, and its trend was analyzed. The essential variables of Ni plating for repair of the damaged part were derived. These conditions are applicable variables for the repair plating device and have been carefully adjusted using the repair plating device. The process for establishing ASME technical standards called code case N-840 is described. The process for developing Ni-plating devices and the electrodeposition procedure specifications are described as well.

ASME Boiler Pressure Vessel Code, Section III, Division 1, Subsection NC, Class 2 components, and Subsection ND, Class 3 components, have significant technical and administrative similarities. The ASME BPV III Standards Committee has a long-standing goal of combining these two subsections (NC and ND). Consolidating Subsections NC and ND will simplify, reduce repetitions and make the Code easier to use. Additionally, a combined Subsection NC/ND will simplify Code maintenance. To facilitate this consolidation, the Subgroup on Component Design, under the BPV III Standards Committee assigned a Task Group to develop a strategy to combine the two subsections into a single subsection while maintaining both Class 2 and Class 3 as separate classes of construction. Both Subsections NC and ND of the Code have been completely reviewed, compared and the technical bases for the differences have been established. The conclusion of this review is that there are only a few major technical differences between the two code class rules; however, there are a significant number of editorial differences. Based on the review, the Task Group developed a strategy that completes the consolidation within two publishing cycles of Code edition. For the Code Edition 2019, two separate Subsections NC and ND books will be published to resolve editorial differences and otherwise align the two subsections. For the Code Edition 2021, a single merged subsection will be published. This paper provides the background for the proposed code change, discusses the detailed result of the NC/ND comparison, and provides the basis for the major technical differences. The paper will also update the status of the project and code actions needed to consolidate to a single subsection.