The purpose of this paper is to describe the changes of the structure and the creep performance for Cr35Ni45Nb alloy after varying carburization. Carburizing layers with different thickness were produced in Cr35Ni45Nb alloy directly by vacuum carburizing experiment. The surface microstructure of Cr35Ni45Nb alloy was essentially changed after varying carburization. Creep performance of carburized only and carburization-creep treatment alloys were compared to study the effect of the carburization on creep for Cr35Ni45Nb heat resistant alloy. Mechanical properties of carburizing layer, carburizing transition layer and base alloy were characterized by micro-hardness, respectively. Results of creep rupture testing for the carburized and non-carburized specimens showed that carburization can influence the rupture time, failure strain and creep rate. Carburization also been proved to affect the constant of C MG and C MMG for different stress or carburizing level.

Thin wall test articles were made using Fusion Powder Bed processing of Type 316L stainless steel. This hollow structure was fabricated using a Renishaw AM250. The material was characterized in the as fabricated condition using optical, scanning electron microscopy, and tensile testing. The tensile behavior indicates high initial residual stress due to the higher than annealed material yield strength. A series of thermal treatments were investigated to determine the effect on mechanical properties and microstructural evolution. In order to determine these effects, miniature flat tensile test samples and metallographic samples were wire electrical discharge machined from the structure. These samples were vacuum heat treated at 870, 1040 and 1080°C for 30 to 120 minutes and metallographically examined to determine the microstructural evolution. The samples were tensile tested and the data are evaluated based on the power-law type relationships. AM Type 316L SS exhibits less strain hardening than wrought stainless steel. In addition, the ductility is somewhat lower in the as-fabricated condition compared to the heat-treated condition. The results of the testing and the rationale for these different behaviors will be discussed.

The objective of this work is an experimental and numerical investigation for a bol Richard Cullen ted composite flange connection for composite pipes, which are used in the oil and gas applications, and obtain a joint with high strength and high corrosion resistance. For the experimental part, we have designed and manufactured the required mould, which ensures the quality of the composite materials and controls its surface grade. Based on the ASME Boiler and Pressure Vessel Code, Section X, this GFRP flange has been fabricated using biaxial glass fibre braid and polyester resin in a vacuum infusion process. Numerically, an investigation is carried out using 3D finite element analysis (FEA) of a bolted GFRP flange joint including flange, pipe, gasket and bolts. This model has taken into account the orthotropy of the GFRP material and the non-linear behaviour of the rubber gasket material for both the loading and non-loading conditions. Furthermore, the leakage propagation between the flange and the gasket has also been simulated in this investigation by using the pressure-penetration criteria PPNC in ANSYS. Finally, the flange has been tested under the internal pressure and the agreement between the experimental and numerical results is excellent.

The 2018 edition of the RCC-MRx Code (1) will be issued, by the end of the 2018, in French and English versions by AFCEN (Association Française pour les règles de Conception et de Construction des Matériels des Chaudières Electro-nucléaires). This Code set up rules applicable to research reactor components (coming from the RCC-MX 2008 developed in the context of the Jules Horowitz Reactor project), to reactor components operating at high temperature (coming from the RCC-MR 2007) and to the Vacuum Vessel of ITER (also coming from the RCC-MR 2007). This edition takes the benefits of an important feedback of the users, such as Jules Horowitz Reactor or ASTRID project, but also from ITER and MYRRHA projects. In parallel, in compliance with the EC’s objectives and its own policy of openness, AFCEN proposes to make its codes evolve, taking into account the needs and expectations of European stakeholders (operators, designers, constructors, suppliers ...) through a workshop called CEN Workshop 64 phase 2. The end of the workshop, planned for 2018, will allow to integrate recommendations issued from this work in the code. This paper gives an overview of the performed work and also identifies the work to be done for a development of a standard such as RCC-MRx code.

RCC-MRx Code is the result of the merger of the RCC-MX 2008, developed in the context of the research reactor Jules Horowitz Reactor project, and the RCC-MR 2007, which established rules applicable to the design of components operating at high temperature and to the Vacuum Vessel of ITER. This code has been issued in French and English versions by AFCEN (Association Française pour les règles de Conception et de Construction des Matériels des Chaudières Electro-nucléaires) in 2012, 2015, and a new edition will be published by the end of 2018. This new edition integrates a significant evolution for the rules dedicated to the ratchetting evaluation through the so-called “efficiency diagram rule”. This rule was initially developed to present a less penalizing rule than the classical “3 Sm” rule and to analyze the interaction between creep and ratchetting in the high temperature domain. Since the first edition of RCC-MR code, several modifications have been made to this rule to improve its representativeness for this kind of damage. These modifications were motivated and justified by numerous experimental tests. The last significant evolution is thus the occasion to present through this article, the background of this rule, the major evolutions already incorporated in the code and also the on-going developments.

The main failure mode of a large vacuum spherical tank, subjected to uniformly external pressure, is buckling of thin shell. Taking a large vacuum spherical tank with volume of 18000m 3 for research objects, the different calculation methods of buckling pressure between the ASME standard, analytic method and finite element method, were compared and analyzed. Using ANSYS software, the buckling pressure of stiffened spherical tanks were analyzed, and the optimum design of the thickness, height and quantity of stiffeners are studied. The results of analysis showed that the anti-instability ability of the large vacuum spherical tank can be improved significantly by a reasonable arrangement of stiffeners.

A failure mode and effect analysis is performed on 58 centrifugal pumps of Atmospheric and Vacuum Distillation Unit in one of the plants of SINOPEC. Firstly, the failure modes of the pumps are classified. Then, the failure causes of the pumps are analyzed. After that, the influence of the failure and the risk level are determined. Finally, the maintenance strategies to reduce risk are put forward according to the equipment failure mode and risk level, which is the basis for the maintenance of pumps. The results show that the main failure modes of the centrifugal pumps are process medium leakage, unexpected stop, vibration, damage, noise, low output and of useful medium leakage. The mean time to failures of centrifugal pumps is 14533 hours, about 20 months. Nine centrifugal pumps are in the risk of high or medium-high, which accounts for 15.52% of the total number of pumps.

This paper presents the design and analysis of the Hydrogen Transfer Lines (HTL) in the Spallation Neutron Source Facility (SNS), Oak Ridge National Laboratory in Oak Ridge, Tennessee. SNS is a Neutron Accelerator providing intense pulsed neutron beams for scientific research into the behavior of materials. The use of supercritical Hydrogen provides moderation of these neutron beam lines. This is facilitated by the HTL, which consists of six coaxial tubes transporting supercritical hydrogen to and from the Moderators surrounding the Mercury Target inside the Core Vessel (CV). Each Hydrogen line is surrounded by a vacuum tube in order to preserve the supercritical Hydrogen temperature of −425°F (−254° C/19 K). An inert Helium tube and a Water Jacket surround the Outer Vacuum Jacket. The Hydrogen boundary is considered the primary containment boundary, with the vacuum tubing being the secondary containment boundary both considered to be safety significant. The HTL assembly tubing must be qualified as Performance Category 3 to the requirements of Department of Energy Standard 1020-1994 for Natural Phenomena Hazards. This requires a detailed dynamic analysis. The HTL must also be qualified to ASCE 43-05 for Limit State consideration for abnormal events and ASME B31.3 for sustained and occasional (upset) loadings. The design basis includes consideration of normal sustained loadings such as deadweight and pressure, as well as thermal expansion/contraction, a Design Basis Earthquake (DBE) with associated Anchor Movements (SAM) and the potential physical interaction between coaxial tubes. Intermittent spacers are provided between some of the tubes, providing a functional interspace between the Hydrogen tubing and the Vacuum Jackets. The analytical methodology involves supplementing unaddressed criteria per ASME B31.3 thus requiring the application of more detailed rules found in ASME Section III, NC-3600, ASCE/SEI 43 – 05 as well as Regulatory Requirements such as USNRC Regulatory Guide 1.61 and 1.92. To assess the potential of a postulated failure of the primary Hydrogen boundary inside the Core Vessel, NUREG 0800, Section 3.6.3, Leak-Before-Break Evaluation Procedures are employed.

One of the vacuum column made of Carbon steel (SA 285 Gr. C) was found with severe corrosion under insulation during turnaround inspection. The equipment was in service for the past 46 years. The corrosion was around the full circumference of the column at a location immediately above the stiffener ring. The wall loss was extensive, which mandated a welded full circumferential patch to encapsulate the corroded region. A full circumferential lap patch with reinforcement plug welds was designed as per ASME PCC-2 [2], fabricated and installed at site. During the dye penetrant examination of the patch plate welds, fine cracks were noted at the toe of the fillet welds. A subsequent PAUT (Phased Array Ultrasonic Testing) of patch plate welds, showed that the cracks were in column shell (base material) surface initiating from toe of the fillet weld of the patch and beneath the fillet weld. The cracks found in first few sections of welds were ground and re-welded with appropriate pre-heating and post-heating. However, some cracks were still observed in base material, although with reduced magnitude compared to previous welding. Despite proper controls in place during the welding, the challenge was, some minor cracks were re-occurring at the base material surface below fillet welds. A multi discipline review was carried out to understand the potential root cause for the cracks, detailed assessment of risks and impact due to the cracks. A risk based approach was followed, in evaluating cracks for acceptance based on API 579-1/ASME FFS-1 [1], for continued service of the column and to minimise the extension of turn around duration. Later, a long-term plan was made to replace the equipment.

During vacuum drying of used nuclear fuel (UNF) canisters, helium pressure is reduced to as low as 67 Pa to promote evaporation and removal of remaining water after draining process. At such low pressure, and considering the dimensions of the system, helium is mildly rarefied, which induces a thermal-resistance temperature-jump at gas–solid interfaces that contributes to the increase of cladding temperature. It is important to maintain the temperature of the cladding below roughly 400 °C to avoid radial hydride formation, which may cause cladding embrittlement during transportation and long-term storage. Direct Simulation Monte Carlo (DSMC) method is an accurate method to predict heat transfer and temperature under rarefied condition. However, it is not convenient for complex geometry like a UNF canister. Computational Fluid Dynamics (CFD) simulations are more convenient to apply but their accuracy for rarefied condition are not well established. This work seeks to validate the use of CFD simulations to model heat transfer through rarefied gas in simple two-dimensional geometry by comparing the results to the more accurate DSMC method. The geometry consists of a circular fuel rod centered inside a square cross-section enclosure filled with rarefied helium. The validated CFD model will be used later to accurately estimate the temperature of an UNF canister subjected to vacuum drying condition.

In this work, a geometrically-accurate two-dimensional (2D) computational fluid dynamic (CFD) model of a used nuclear fuel cask, that can contain up to 32 pressurized water reactor (PWR) used nuclear fuel (UNF) assemblies, is constructed. This model is similar to the TN-32 cask employed in the ongoing high-burnup (HBU) Spent Fuel Data Project lead by the Electric Power Research Institute (EPRI). This model is used to predict the peak cladding temperature under vacuum drying conditions. Due to the symmetry of the cask, only one-eighth of the cross-section is modeled. Steady-state simulations that include the temperature-jump boundary conditions at the gas-solid interfaces are performed for different heat generation rates in the fuel regions and a range of dry helium pressures, from ∼10 5 to 100 Pa. These simulations include conduction within solid-gas regions and surface-to-surface radiation across all gas regions. The peak cladding temperatures are reported for various heat generation rates and rarefaction conditions, along with the maximum allowable heat generation that brings the cladding temperatures to the radial hydride formation limit. The results showed that the decrease of helium pressure significantly increased the temperature of the cladding material compared to the atmospheric pressure condition.

Vacuum drying of nuclear fuel canisters may cause the temperature of fuel assemblies to considerably increase due to the effect of gas rarefaction at low pressures. This effect may induce a temperature-jump at the gas-solid interfaces. It is important to predict the temperature-jump at these interfaces to accurately estimate the maximum temperature of the fuel assemblies during vacuum drying. The objective of this work is to setup a concentric cylinders experimental apparatus that can acquire data to benchmark rarefied gas heat transfer simulations, and determine the temperature-jump coefficient at the interface between stainless steel surface and helium gas. The temperature-jump is determined by measuring the temperature difference and heat flux across a 2-mm gap between the concentric cylinders that contains rarefied helium and compare the results to analytical calculations in the slip rarefaction regime.

Nowadays, fiber reinforced composites are widely used in variety of industrial applications such as aircraft structures, automotive, pressure vessels and piping, etc. Aircraft standard fabrication process requires certain level of vacuum compaction (debulking) during the lay-up process, and a standard bagging method for curing in an autoclave. Every compacted component cured in an autoclave needs to be vacuum bagged employing edge breathers or bleeders and surface breathers. This process is repetitive and time consuming, and therefore needs further investigation. In this research, the combined effect of the removal of compaction and edge breathing on the thermomechanical behavior of plain weave woven laminated composites is studied. Tests have been conducted on 12 lamina plain weave composite specimens. Results indicated an insignificant difference on the thermomechanical properties between compacted and non-compacted specimens. Tensile, ILSS, flexural and DSC tests confirm that for the 12 lamina specimens compaction and edge breathing are not needed.

During vacuum drying of used nuclear fuel canister, helium pressure is decreased to as low as 67 Pa to promote evaporation and removal of water remaining in the canister following draining operation. At low pressures associated with vacuum drying, there is a temperature jump (thermal resistance) between the solid surfaces and helium in contact with them. This temperature jump increases as the pressure decreases (rarefied condition), which contributes to the fuel assembly’s temperature increase. It is important to keep the temperature of the fuel assemblies below 400°C during vacuum drying to ensure their safety for transport and storage. In this work, an experimental apparatus consisting of a 7×7 array of electrically heated rods maintained between two spacer plates and enclosed inside a square cross-section stainless steel pressure vessel is constructed to evaluate the temperature of the heater rods at different pressures. This geometry is relevant to a BWR fuel assembly between two consecutive spacer plates. Thermocouples are installed in each of the 49 heater rods, spacer plates and enclosure walls. They provide a complete temperature profile of the experiment. Different pressures and heat generation relevant to vacuum drying conditions are tested. The results showed that the maximum temperature of the heater rods increases as the pressure decreases. The results from these experiments will be compared to computational fluid dynamics simulations in a separate work.

Computational fluid dynamics simulations of a 7×7 array of heated rods within a square-cross-section enclosure filled with rarefied helium are performed for heat generation rates of 50 W and 100 W and various helium pressures ranging from 10 5 to 50 Pa. The model represents a section of nuclear fuel assembly between two consecutive spacer plates inside a nuclear canister subjected to during vacuum drying process. A temperature jump model is applied at the solid-gas interface to incorporate the effects of gas rarefaction at low pressures. The temperature predictions from simulations are compared to measured temperatures. The results showed that when helium pressure decreased from 10 5 to 50 Pa, the maximum temperature of the heater rod array increased by about 14 °C. The temperatures of the hottest rod predicted by simulations are within 4°C of the measured values for all pressures. The random difference of simulated rod temperatures from the measured rod temperatures are 3.33 °C and 2.62 °C for 100 W and 50 W heat generation rate.

In the process units with Polythionic Acid Stress Corrosion Cracking (PTA-SCC) such as hydrotreater or hydrocracker, the neutralization treatment like soda ash washing and/or PWHT, Post Weld Heat Treatment, are conducted for stabilized austenitic stainless steels of TP321H or TP347H to mitigate the potential of PTA-SCC. 347AP, the proprietary version of Type 347LN, is remarkable for the possibility to save the fabrication and turnaround cost by the elimination of the neutralization treatment and/or PWHT and its reliability to PTA-SCC resistance without PWHT and neutralization was proven by the long-term service experience. However it has not been investigated the difference between 347AP and TP347H for the Stress Relaxation (SR) cracking resistance of welded joint required for the long term operation at elevated temperature. On the other hand, in atmospheric distillation units and vacuum distillation units using the crude containing high Total Acid Number (TAN), the resistances to both Naphthenic Acid Corrosion (NAC) and PTA-SCC are required by elevating the design temperature. Therefore, Fit-for-purpose austenitic stainless steels with high PTA-SCC resistance without stabilization treatment are required for the complicated environment with the variation of temperature, pressure and fluid in petroleum refinery process using the crude containing high S content. In this study, SR cracking susceptibility of 347AP is investigated and it is found that 347AP has lower SR cracking susceptibility than TP347H without stabilization treatment. Next, in order to improve NAC resistance compared with 347AP, the effect of Mo, Cu and W contents to NAC are investigated based on 347AP, and it is clarified that the increase of Mo content greater than 3% is the most effective to improve NAC resistance. Moreover, the conventional welding filler with the similar NAC and PTA-SCC resistances to base metal is investigated, and it is reported that ER NiCrCoMo-1 filler was recommended for the welding fabrication of Fit-for-purpose austenitic stainless steels with high PTA-SCC resistance.

Latent heat of vaporization of water is very large compared to the latent heat of vaporization of other fluids. Therefore, when evaporation is promoted by decompressing, water lowers the temperature and freezes finally. We froze without heat exchange by exposure to vacuum the water using a vacuum pump. Despite violent boiling occurs, pressure in the vessel was going down. But, finally, at freeze process, the pressure in the vessel was increased. This pressure rise is caused by that amount of evaporation increases temporarily rapidly. The rapid increase in the amount of evaporation is due to the release of latent heat during freezing. This paper was focused on the pressure temperature change and mass change when the phase change of the water is switched to sublimation from the evaporation. This is likely to provide useful information when considering a phase-change model in a special state.

Hydrogen desorption from 6061 aluminum alloy during deformation was investigated by a new type of ultrahigh vacuum apparatus. The new apparatus which combined tension and outgassing experiments can be used to measure hydrogen desorption from the specimen during deformation. Outgassing corresponded to the deformation process, which initially released surface-adsorbed hydrogen. After elastic deformation, the as-received specimen released hydrogen because of the rupturing of the surface oxide layer. The hydrogen-charged specimen showed lower hydrogen concentration and hydrogen release rate than the as-received specimen. The second hydrogen release peak was due to the hydrogen accumulation at the interface between the oxide layer and matrix. The irregular sharp hydrogen release peaks resulted from the hydrogen bubble exposed to the surface during the rupture of the surface oxide layer.

RCC-MRx Code (ref 1) is the result of the merger of the RCC-MX 2008 developed in the context of the research reactor Jules Horowitz Reactor project, in the RCC-MR 2007 which set up rules applicable to the design of components operating at high temperature and to the Vacuum Vessel of ITER. This code has been issued in French and English versions by AFCEN (Association Française pour les règles de Conception et de Construction des Matériels des Chaudières Electro-nucléaires) in 2012, and a new edition was published at the end of 2015. A significant work has been performed for this edition to improve the code in order to facilitate its use and understandability, and also to have a better fit with the feedbacks of the users. In parallel, in compliance with the EC’s objectives and its own policy of openness, AFCEN proposes to make its codes evolve, taking into account the needs and expectations of European stakeholders (operators, designers, constructors, suppliers...) through a workshop called CWA phase 2. This paper gives an overview of the realized work and also will identify the work to be done for an opening of a standard such as RCC-MRx code.

The 2015 edition of the RCC-MRx Code will be issued, by the end of the year, in French and English versions by AFCEN (Association Française pour les règles de Conception et de Construction des Matériels des Chaudières Electro-nucléaires). This Code set up rules applicable to research reactor components (coming from the RCC-MX 2008 developed in the context of the Jules Horowitz Reactor project), and to components operating at high temperature and to the Vacuum Vessel of ITER (coming from the RCC-MR 2007). A significant work has been performed since 2012, first edition of the RCC-MRx, in order to improve the rules, to facilitate its use and understandability, and also to have a better fit with the feedbacks of the main projects (ASTRID, JHR, ITER). This paper presents the technical evolutions in the 2015 edition and also the other initiatives led upstream (prenormative works) and downstream (publication of the rules background - criteria) of the publication of rules.