A design for flaw placement in a full-scale pipe test was developed to both detect crack initiation and measure crack growth rate upon internal pressurization of a pipe exposed to a sulfide stress cracking (SSC) environment. The objective of this work was to model different sizes of longitudinally oriented, inside diameter (ID) surface flaws and lay them out on the pipe in such a manner that (1) the flaws experience their target stress intensity factor (K) value at a chosen value of internal pressure, (2) the stress interaction between flaws is minimized, and (3) the flaw layout is optimized for detecting both crack initiation and growth using a direct-current (d-c) electric potential (EP) technique. The approach to the flaw design and layout used finite element analysis (FEA) modeling and consisted of optimizing K-profiles. First, the K-profiles were optimized by designing curved-bottom flaws such that the target K along the flaw front occurred along most of the flaw length. Then, stress interactions between the flaws were checked to confirm minimum interactions were achieved and that the proposed flaw layout around the pipe circumference was acceptable. In addition, the FE models were used to predict strains on the pipe outside surface. Finally, global (single large current supply), local (individual small current supplies) and hybrid (individual medium-sized current supplies at larger distance) approaches to the d-c EP measurements were evaluated to select which methodology would be most appropriate to detect both crack initiation and growth of the flaws. The results of the analysis show that optimizing all these design factors provides a solid basis for achieving experimental success.

Heat resistant Iron-Nickel-Chrome alloys, such as Alloy 800H, are designed to be used at elevated temperatures. At temperatures ranging from 550 to 750°C (1022 to 1380°F) these alloys can be susceptible to stress relaxation cracking (SRC). For these and other high temperature alloys, API Technical Report 942-B provides recommended ranges for material composition, as well as fabrication and heat treatment measures to mitigate Stress Relaxation Cracking. However, it has been shown to be difficult to obtain material that satisfies the API recommended prescriptions. Recommended alloy composition, welding and fabrication steps can become complex for special situations found on projects (i.e. thicker wall piping, dissimilar metal welds, etc.). This paper discusses example applications, and the reasons for selecting Alloy 800H. It also describes the SRC mechanism, the technical justifications for the API 942-B recommendations, and the challenges that have been experienced trying to meet these recommendations.

In order to ensure continued safe operation of European nuclear reactors, it is necessary to solve specific issues that arise from irradiation induced reactor pressure vessel (RPV) embrittlement under long term operation. This requires an extension of the RPV surveillance programs to cover longer operation times than originally planned. The limited availability of surveillance materials poses a challenge for the feasibility of such programs. Among others, the use of the small specimen technology is a promising option to overcome the lack of materials. For example, a number of not less than 16 sub-sized 0.16 C(T) specimens (4 mm thickness) can be manufactured from two tested Charpy sized (10 mm × 10 mm × 55 mm) specimens, allowing a reliable determination of the reference temperature T 0 . Such Charpy sized fracture mechanics specimens are currently widely used in the RPV surveillance programs. To establish the methodology for fracture mechanics testing of irradiated and unirradiated RPV steels using sub-sized specimens, a joined R&D project was launched partly financed by the German Federal Ministry for Economic Affairs and Energy. Moreover the following points shall be addressed: • Manufacturing, pre-cracking procedure, measurement of the crack opening displacement and load line displacement under hot cell conditions • Demonstration of the transferability of fracture mechanics data The purpose is to demonstrate that the results measured on sub-sized specimens can safely be used in the safety assessment of RPVs. In addition, the results will establish a basis to assess results from international projects regarding sub-sized fracture mechanics specimens.

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.

Tensile residual stresses associated with stainless steel pipe welds can promote in-service cracking and influence the need for inspections. Previous research via finite element analysis (FEA) [1] and experimental characterization [2] has shown that welds in thick wall pipe can produce compressive residual stresses at the inner diameter (ID) surface. However, research that has evaluated the relationship between the number of weld layers, stemming from different weld bead sizes, and the resulting pipe residual stress profiles is limited. This investigation used two-dimensional (2D) FEA to evaluate the influence of the number of weld layers (resulting from different weld bead sizes) on the ID surface and through-wall residual stress profiles for varying stainless steel pipe radii, thicknesses, and weld joint geometries. The findings herein are compared to previous experimental results [2]. The results demonstrated that for the larger pipe sizes and the welding conditions investigated, increasing the number of weld layers (reducing individual weld bead sizes) reduced the ID surface tensile axial residual stresses. In the larger pipe sizes, the magnitude of the tensile residual stresses extending through (into) the pipe wall is also reduced with an increased number of weld layers. The FEA results show that the weld joint geometry may not affect the residual stress profiles as strongly as do the number of weld layers, based on the similarities in the tensile stress values for the joint geometries that were evaluated.

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.

LMPH (Radiation heat exchange tube), the key section of ethylene cracking furnace, has various failure behaviors (e.g., carbonation, vulcanization, oxidation, and deformation) because of its rigorous working conditions and complicated structure [1]. In this paper, the ultrasonic guided wave is applicable to inspect the interior and surface of the LMPH tube. The results demonstrate that large components in the similar structure of the heat exchange bent tube bundle can be inspected effectively, efficiently, conveniently, compared with the traditional inspection method. One inspection example of an ethylene cracking furnace tube bundle is forwarded. The paper also provides an engineering application/case and valuable crafting process for the detection of similar structures by the ultrasonic guided wave method.

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.

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 weld overlays using the methodology of Case N-504-4 and Appendix Q. This repair technique is appropriate to address thermal fatigue cracking 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 Appendix Q 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 steady state operating stresses, residual stresses from welding and fabrication processes, and 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 predicable as SCC degradation. Therefore, alternative rules are 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 basis begins with the industry accepted rules of Case N-504-4 and Appendix Q and discusses the appropriate modifications needed to address thermal fatigue cracking. These modifications include removing restrictions such as the delta ferrite limit that is only applicable to address SCC, 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 outline the basis for a new ASME Code Case that will contain the appropriate rules for weld overlay repair of Class 1, 2 and 3 stainless steel piping degraded by thermal fatigue cracking. The new Case is currently in draft form and the requirements and specific details are still evolving. Thus, it is envisioned that this technical basis will be revised to include updates and revisions to the Case.

The integrity of transmission oil pipelines are often managed through in-line inspections (ILI) at regular intervals. For the last two decades, such ILI-based integrity programs along with excavations and field non-destructive testing (NDE) have proven their effectiveness in terms of reliability. In a few cases, some pipes contain; for example, a unique cracking mechanism exhibited by short, deep axial cracks located in the vicinity of girth welds. These attributes pose sizing difficulties for ultrasonic crack ILI tools. Accordingly, operators may lean on supplemental integrity activities to prove the safety of the pipelines such as; but not limited to, hydrostatic testing, laboratory testing of cut-outs, qualitative ranking of features, borehole leak detection analysis, Just-Missed-Flaw (JMF) or Just Surviving Flaw (JSF) analysis, discharge and/or point pressure restrictions, and/or a mix between all the previous techniques. Moreover, it is the operators’ responsibility to evaluate the risk associated with their integrity plans. Hence, it is important to be able to analyze the reliability of such integrity activities quantitatively. This paper presents an event-tree approach which can augment standard ILI or hydrostatic test results and probabilistic analysis with non-ILI integrity measures under one umbrella. In this approach, the likelihood of failure for both leak and rupture modes can be comprehensively estimated. The event tree approach is used herein as an inductive analytical diagram in which failure events are analyzed using Boolean logic to examine a chronological series of subsequent integrity actions and consequences. The proposed approach is also designed to capture subject matter experts’ opinion into the analysis as part of the integrity management program. The work discusses a real practical application along with verification and validation elements of the proposed integrated approach.

High density polyethylene (HDPE) is widely used in city water and gas piping system and even in nuclear engineering for its excellent properties in flexibility, impact resistance, aging and corrosion resistance, reliable connection, long service life. It is generally believed that the typical failure mode of PE pipe under long-term static pressure load is the slow crack growth (SCG) caused by creep crack initiation (CCI) and creep crack growth (CCG), which is a kind of quasi-brittle failure. Scratching often occurs during PE pipe installation due to the drag of pipe without any protection, which would leave axial cracks as the initial defect. This would harm the long term performance of pressurized HDPE pipe. In the present work, the creep crack kinetics was experimentally determined in term of C * integral, and then CCG behavior of PE pipe with axial semi-elliptical crack was discussed. With the help of a CARE electronic universal testing machine and DIC displacement measuring system, creep tests are performed with smooth bar specimen of PE100 pipe parent material. And the stationary creep law parameters were determined. The crack opening displacement ( COD ) rates were determined by creep cracking test with cracked round bar specimen. With the help of compliance calibration tests, the COD curves were changed into crack propagation curves and the creep crack kinetics was eventually determined in term of C * integral. By assuming an semi-elliptical crack front with nearly the same C * integral, the creep crack propagation was predicted with the determined crack kinetics. And the allowable initial crack size was also determined for the PE pipe discussed. Compared with that determined in terms of SIF, current study yields much margin for the allowable initial crack.

A hydrogen manufacturing plant experienced circumferential cracking at the dissimilar weld on the outlet header. The outlet header was a cold wall design and the dissimilar weld was between HP40 modified and carbon ½ Mo steels. The resultant failure investigation found the cause to be hydrogen induced cracking of the dissimilar weld at the fusion boundary zone. The hydrogen was generated from the CO 2 corrosion which occurred due to operating the tubes below the dew point and the hydrogen was trapped in the steel by the CO chemisorption onto the steel. The following paper outlines the failure investigation and the fitness for service conducted to maintain the running of the plant.

Making light water reactor (LWR) components using additive manufacturing (AM) provides a high degree of design freedom to create complex near net shape geometries, with significant reduction in the deployment time. Since most of the current AM research focus on the refining of the fabrication variables, little information exists on the actual corrosion behavior (including general corrosion, localized corrosion and environmentally assisted cracking - EAC) of AM components. The most frequent operational material failure modes in LWR is EAC and debris fretting of fuel rods, therefore the objective of this work was to evaluate the environmental performance of AM type 316L SS in relation to the traditionally melted and forged wrought material. Stress corrosion cracking results show similar behavior between AM and wrought 316L SS in high temperature water. Ambient temperature electrochemical testing showed that the AM 316L SS was slightly more resistant to localized corrosion than the traditional wrought material and that AM material did not suffer sensitization during fabrication or heat treatment processes.

This paper provides a technical basis for a crack growth rate (CGR) for use in performing evaluations of cracking in stainless steel canister materials. The source of crack initiation and growth is deposition of chloride aerosols on the canister surface followed by deliquescence leading to a brine solution. The brine solution attacks the stainless steel surface, leading to pitting; in the presence of tensile stress (such as residual tensile stress due to welding), stress corrosion cracking can occur. The CGR will be used for evaluating flaw growth under a proposed ASME Boiler and Pressure Vessel Code Case covering stainless steel canisters.

The welding task focuses on development of advanced welding technologies for repair and maintenance of nuclear reactor structural components to safely and cost-effectively extend the service life of nuclear power reactors. This paper presents an integrated research and development effort by the Department of Energy Light Water Reactor Sustainability Program through the Oak Ridge National Laboratory (ORNL) and Electric Power Research Institute (EPRI) to develop a patent-pending technology, Auxiliary Beam Stress Improved Laser Welding Technique, that proactively manages the stresses during laser repair welding of highly irradiated reactor internals without helium induced cracking (HeIC). Finite element numerical simulations and in-situ temperature and strain experimental validation have been utilized to identify candidate welding conditions to achieve significant stress compression near the weld pool during cooling. Preliminary welding experiments were performed on irradiated stainless-steel plates (Type 304L). Post-weld characterization reveals that no macroscopic HeIC was observed.

Temper bead (TB) welding is often used as an alternative to post weld heat treatment (PWHT) for repair of pressure vessels and piping in the nuclear power industry. Historically, qualification of TB welding procedures has employed the Charpy V-notch test to ensure acceptable heat-affected-zone (HAZ) impact properties. The 2004 Edition of ASME Section IX provided a new provision in QW-290 that allows temper bead qualification using a peak hardness criterion. The peak hardness provision is appropriate for industries such as oil and gas, where peak allowable hardness is specified to ensure adequate resistance to sulfide stress cracking in sour service environments. However, a peak hardness criterion is not appropriate where impact properties are specified for resistance to brittle fracture during low temperature conditions that can occur during certain postulated accident scenarios at a nuclear power plant. Work at the Electric Power Research Institute (EPRI) and The Ohio State University (OSU) show that a hardness drop protocol can be used to demonstrate acceptable impact properties in the HAZ of a temper bead weld. This paper presents a quantitative correlation between hardness measurements and HAZ microstructures with presumed optimum impact properties using a hardness drop approach. The overarching goal is to develop a hardness test protocol for temper bead weld procedure qualification for applications where impact properties are specified.

Radiation-induced degradation of reactor pressure vessels and internals is a concern to the aging nuclear fleet and welding solutions will be required if repair of these irradiated components is deemed necessary. However, the weldability of highly irradiated austenitic materials is significantly diminished due to the presence of irradiation induced helium in the material matrix. Helium-induced weld cracking is a complex phenomenon that is related to the concentration of helium, the heat input from the welding process, and stresses generated during cooling of the weld. During conventional high heat input welding methods such as gas tungsten arc welding, helium bubbles can coalesce and grow on base metal grain boundaries within the heat-affected zone which subsequently causes intergranular cracking. The objective of this work was to obtain weldability data by characterizing welds made on highly activated, neutron irradiated Type 304 stainless steel containing both radiation-induced helium and microstructural damage such as void swelling. All irradiated materials welding was performed inside a Westinghouse hot cell utilizing a pulsed Nd:YAG laser with welds made on three rectangular samples of highly activated Type 304 stainless steel. The rectangular samples were cut and milled in-cell from sections previously obtained from two neutron reflector hex blocks. The hex blocks are U.S. Department of Energy owned material and were irradiated for approximately 13 years in the EBR-II sodium cooled fast reactor from 1982 until 1995. The three samples selected for welding have nominal damage doses of approximately 0.4, 11, and 28 dpa with corresponding estimated helium contents of 0.2, 3 and 8 appm helium, respectively. A number of different weld parameter sets were utilized and included variations of travel speed, wire feed speed and lens-to-work distances. The parameter sets allowed for a range of effective weld heat input levels to be compared. Single pass and multiple pass as well as wire fed and autogenous welds were made. This paper presents the results from post-weld evaluations performed on the three welded irradiated samples, focusing on the reduced tendency for cracks to form adjacent to the weld as a function of weld parameters, lens-to-work distance and helium content.

A typical Fluid Catalytic Cracking Unit (FCCU) generates high temperature flue gas in the process of regenerating the catalyst. This flue gas is diverted to a stack after removal of catalyst fines and excess heat using a Waste Heat Recovery Unit (WHRU) or CO boiler. This flue gas line is a large diameter (1.2 m /2 m) piping and is a combination of Hot Wall (bare SS304H piping with external insulation) upstream of Orifice Chamber and Cold Wall (Carbon steel piping with internal refractory lining) for the downstream side. In a major revamp project, large portion of flue gas line was replaced with some dimensional and design changes. A crack was noticed at the SS304H side of hot wall to cold wall transition joint downstream of Orifice Chamber after approximately 2 years in operation. The line operates around 700 °C and 0.15 Bar(g) at the location of the crack. The initial crack was measured to be approximately 250 mm to 300 mm and grew to a full circumference crack in a short time resulting in minor flue gas leaking with catalyst fines. This paper discusses the details on how the issue was addressed on site and a temporary repair (i.e. welding of a box on high temperature piping) was carried out online safely, while the unit remained in operation. Further, the paper presents the root cause assessment and design modifications implemented for hot wall to cold wall transition joint during a scheduled turnaround.

Coke drums are subjected to severe thermal cycling with the skirt to shell connection weld being vulnerable to fatigue cracking. It is essential this connection is well designed to ensure a long life before repairs are inevitably required. Much has been written on coke drum skirt design with the aim of reducing the thermal stresses and strains encountered at the skirt connection weld, some designs have removed the weld completely allowing the drum to sit in an “egg-in-cup” arrangement. This paper includes a short literature review discussing Coke drum skirt designs and explains skirt behaviour during the drum cycle that results in eventual skirt cracking. A case study is reviewed in detail for a new pair of coke drums, where the predicted fatigue life of the chosen welded connection is assessed using axisymmetric, quarter symmetry and half symmetry finite element analysis supported by thermocouple data. The optimised design focuses on a conventional tangential design where the effects of the essential variables such as skirt thickness, skirt connection location, skirt-to head-gap and slot design (length, location & spacing) have been modelled and optimised to obtain a skirt design that produces the longest fatigue life for the intended duty cycle. Coke drum skirts must be installed onto the shell to exacting tolerances during manufacture to ensure concentricity and minimal gap between the skirt and shell. A brief overview of how this is achieved will be presented.

Bottom head flange leaks on coke drums are a too common occurrence for coke drum operators. This paper discusses why these complex joints leak, exploring phenomena such as bolt relaxation, flange rotation, bolt hole cracking, plastic deformation of a flange face due to cyclic thermal transients, as well as investigating the effects of gasket stress variation which can lead to gasket movement and distortion during successive drum cycles. The industry trend is to install automatic bottom un-heading devices (BUDs) to facilitate safe, reliable coke removal and to increase production through shortening un-heading operations. A case study is reviewed which shows how a new drum flange, coupled to a BUD, has been optimized using Finite Element Analysis (FEA). The findings show adequate flange thickness and optimized hub dimensions are required to combat plastic deformation of the flange ring and reduce gasket stress variation. Five designs are modelled, varying the flange thickness, outside diameter and hub geometry. Due to the close proximity of a new side nozzle, the full lower section of the drum has been modelled using quarter and half symmetry FEA with applied temperature distributions from each phase of a typical coke drum cycle; heating, coking and quench. This has allowed for nozzle loadings to be evaluated and the location of the flange weld to be optimized to give the greatest fatigue life. An explanation into why periodic re-tightening is required to keep these joints tight is provided along with recommendations on suitable joint assembly techniques using a combination of bolt load verification and alternative bolting patterns from ASME PCC-1.