bluuebunny/crossref_metadata_2025 huggingface dataset:
Corrosion, Corrosion Barrier Development and Mechanical Properties of Steels Foreseen as Structural Materials in Liquid Lead Alloy Cooled Nuclear Systems
Authors: Weisenburger Alfons , Mu¨ller Georg , Heinzel Annette , Jianu Adrian , Muscher Heinrich , Kieser Martin
Published: February 2010

A key problem in development of heavy liquid metal cooled nuclear energy and transmutation reactors is the corrosion of structural and fuel cladding materials in contact with the liquid metal. Lead and lead bismuth attack unprotected steel surfaces by dissolution of the metallic components into the liquid metal. It is common understanding that oxide scales on the surface provide the best protection against dissolution attack. However, at temperatures above 500°C austenitic steels suffer from severe dissolution attack, while martensitic steels form thick oxide scales, which hinder heat transfer from the fuel pins and which may break off and eventually lead to a blocking of the coolant channel. Above 500°C steels have to be protected by stable, thin oxide scales. A well understood measure is alloying of stable oxide formers into the surface. Al has shown its ability to form such oxide scales. In the range of 4–10 wt% Al on the surface a stable thin alumina scale is formed by Al diffusion to the surface and selective oxidation. The alumina scale grows only very slowly and prevents migration of oxygen into the steel as well as migration of steel components onto the surface. A number of corrosion experiments showed the good protective behaviour of Al scales in LBE with 10−6 wt% oxygen up to 650°C and for exposure times up to 10000 h. Alloying Al into the surface was done by diffusion processes and also by pulsed electron beam (GESA) melting of a thin surface layer on which Al or an Al containing alloy was precipitated before. This presentation gives an overview on investigations of the steel behaviour in HLM environment carried out to explore their suitability for systems with Pb/LBE coolants. Results of experiments with static and flowing LBE are discussed. The behaviour of steels examined and their respective application ranges are described. Part of the presentation deals with protective barrier development on the steel surface by alloying of Al and its effect on the corrosion resistance. Furthermore the influence of parameters like stresses in the cladding wall, creep behaviour, different flow velocities of the LBE and changing temperatures and oxygen concentrations in LBE is discussed.
Study on Nano-Oxide Powder Dispersed Ni-Electroplated Composites for Components Materials of NPPs
Authors: Kim Joung Soo , Kim Myeng Jin , Kim Dong Jin , Kim Hong Pyo
Published: February 2010

In this study, Ni composite layers embedding nano-oxide powders were electroplated on an alloy plate to mitigate and/or prevent various corrosion damages such as stress corrosion cracking, pitting, erosion-corrosion, etc.. Al2O3, TiO2, Y2O3, and ZrO2 nano-oxide powders were used to make Ni composite deposits in a sulfamate solution. After electrodeposition, their microstructures were examined using OM, SEM and TEM, and their mechanical properties were evaluated. According to the results, the amount of the co-deposited oxide powders in the composite increased regardless of the types of the nano-oxide, as the amount of oxide powder in the electrolyte bath increased. The hardness was measured to be relatively higher for the specimens produced in the electrolyte bath with higher concentrations of the nano-oxide powders and to decrease with increasing aging temperatures. even after aging at high temperatures. However, the behavior of the tensile properties was observed to be different with the different types of the oxide powders, possibly due to heavy agglomeration and inhomogeneous distribution of the oxide powders in the composites. Their corrosion properties can’t be presented this time, but they will be open at the next meeting.
Aging Management of Nuclear Power Plant Reactor Internals
Authors: Marx Frank J. , Lott Randy G. , Boggess Cheryl L. , Demma Anne
Published: February 2010

As U.S nuclear power plants seek to extend operation beyond their 40 year license, aging management of reactor vessel internals is of prime concern for continued safe and reliable plant operation. A framework for aging management has been developed by screening all components for potential degradation, performing detailed finite element analyses of critical components, identifying components that are most likely to fail, and developing a strategy to monitor and inspect these components. Aging of reactor internals components can be attributed to specific damage mechanisms that include stress corrosion cracking, irradiation-associated stress corrosion cracking, wear, fatigue, thermal embrittlement, irradiation embrittlement, void swelling, irradiation-induced stress relaxation and irradiation creep. An understanding of the relationships among these damage mechanisms is the basis for developing a tiered approach to inspection and evaluation, i.e., evaluation of primary components with the earliest expected manifestations of damage followed, if necessary, by evaluation of expansion components that have been categorized to be subsequently susceptible to similar damage mechanisms. Sixty-year aging finite element simulations are conducted on the most highly irradiated components to identify trends in the predicted component behavior. They are used to develop broad-based recommendations for inspection and evaluation by considering the complex interactions between the structure, loading, and aging degradation in the reactor internal components. The results can be used to identify critical locations within the structure and indicate the time frame and scale of degradation in the component. This paper examines the aging management strategy, the detailed finite element analyses of highly irradiated components, the categorization of inspection components, and the inspection recommendations for life extension beyond 40 years for the Westinghouse and CE-designed reactor internals.
Advanced First Core Design for the Westinghouse AP1000
Authors: Fetterman Robert J.
Published: February 2010

As the nuclear renaissance is now upon us and new plants are either under construction or being ordered, a considerable amount of attention has also turned to the design of the first fuel cycle. Requirements for core designs originate in the Utilities Requirements Document (URD) for the United States and the European Utilities Requirements (EUR) for Europe. First core designs created during the development of these documents were based on core design technology dating back to the 1970’s, where the first cycle core loading pattern placed the highest enrichment fuel on the core periphery and two other lower enrichments in the core interior. While this sort of core design provided acceptable performance, it underutilized the higher enriched fuel assemblies and tended to make transition to the first reload cycle challenging, especially considering that reload core designs are now almost entirely of the Low Leakage Loading Pattern (LLLP) design. The demands placed on today’s existing fleet of pressurized water reactors for improved fuel performance and economy are also desired for the upcoming Generation III+ fleet of plants. As a result of these demands, Westinghouse has developed an Advanced First Core (AFCPP) design for the initial cycle loading pattern. This loading pattern design simulates the reactivity distribution of an 18 month low leakage reload cycle design by placing the higher enriched assemblies in the core interior which results in improved uranium utilization for those fuel assemblies carried through the first and second reload cycles. Another feature of the advanced first core design is radial zoning of the high enriched assemblies, which allows these assemblies to be located in the core interior while still maintaining margin to peaking factor limits throughout the cycle. Finally, the advanced first core loading pattern also employs a variety of burnable absorber designs and lengths to yield radial and axial power distributions very similar to those found in typical low leakage reload cycle designs. This paper will describe each of these key features and demonstrate the operating margins of the AFC design and the ability of the AFC design to allow easy transition into 18 month low leakage reload cycles. The fuel economics of the AFC design will also be compared to those of a more traditional first core loading pattern.
Progress and Experiences From the Decommissioning of the Eurochemic Reprocessing Plant
Authors: Walthe´ry Robert , Lewandowski Patrick , Ooms Bart , Reusen Nancy , Van Laer Wim
Published: February 2010

Belgoprocess started the industrial decommissioning of the main process building of the former EUROCHEMIC reprocessing plant in 1990, after completion of a pilot project in which two buildings were emptied and decontaminated to background levels. The remaining structures were demolished and the concrete debris was disposed of as industrial waste and green field conditions restored. The Eurochemic reprocessing plant operated from 1966 to 1974 to process fuel from power reactors and research reactors. The main building is a large concrete structure, comprising a surface area of 55,000 m2, concrete volume 12,500 m3, and 1,500 Mg of metal components. The building is divided into multiple cells. About 106 individual cell structures have to be dismantled, involving the removal and decontamination of equipment from each cell, the decontamination of the cell walls, ceilings and floors, the dismantling of the ventilation system. Most of the work involves hands-on operations under protective clothing tailored to each specific task. Tool automation and automatic positioning systems are successfully applied. In view of the final demolition of the main process building, the main process building has been divided into three parts — each part is isolated from the others. In September 2008 the eastern part of the building has been demolished. The paper presents a status overview of the decommissioning and decontamination activities at the main process building of the former Eurochemic reprocessing plant on the nuclear site of Dessel in Belgium. The specific BELGOPROCESS strategy will be highlighted, in which the decommissioning activities are carried out on an industrial scale with special emphasis on cost minimisation, the use of technology on an industrial representative scale and the specific alpha contamination of equipment and building surfaces, requiring that the decommissioning work is done with adequate protective clothing. Also specific breathing and cooling air systems have been provided to allow the operators to carry out the decommissioning tasks in acceptable working conditions.
Heavy Liquid Metal Natural Circulation in a One-Dimensional Loop
Authors: Coccoluto G. , Gaggini P. , Labanti V. , Tarantino M. , Ambrosini W. , Forgione N. , Napoli A. , Oriolo F.
Published: February 2010

The ENEA Brasimone Research Centre since 1999 is strongly involved in the national and European research programmes performed in the field of heavy liquid metal technology aiming at the development of critical (LFR) and subcritical (ADS) nuclear systems. In particular, in the frame of the IP-EUROTRANS, (6th Framework Program EU), ENEA assumed the commitment to perform an integral experiment with the aim to reproduce the primary flow path of a HLM pool-type nuclear reactor, cooled by Lead Bismuth Eutectics (LBE). This new experimental activity, named ICE “Integral Circulation Experiment”, will be performed by an appropriate test section designed to be installed in the CIRCE facility. In order to support the ICE activity, as well as characterise the natural and gas enhanced circulation flow regimes in a HLM loop, qualify test procedures, components nuclear relevant, a new facility was designed and built up by Brasimone Research Centre, named NACIE “NAtural CIrculation Experiment”. The paper reports a detailed description of the loop and the main experimental results carried out from the natural circulation tests already performed on the NACIE loop. Numerical simulations have been performed in collaboration with the University of Pisa, adopting the RELAP5/Mod3.3 system code modified to allow for LBE as a cooling fluid. The aim of the performed post-test calculations is to compare the code response with the experimental results under the natural circulation flow regime, allowing to qualify the adopted nodalisation as well as the performance of the code when employed on HLM loop.
Subcooled Boiling Heat Transfer in a Short Vertical SUS304-Tube at High Liquid Reynolds Number
Authors: Hata Koichi , Masuzaki Suguru
Published: February 2010

The subcooled boiling heat transfer and the steady state critical heat fluxes (CHFs) in a short vertical SUS304-tube for the flow velocities (u = 17.28 to 40.20 m/s), the inlet liquid temperatures (Tin = 293.30 to 362.49 K), the inlet pressures (Pin = 842.90 to 1467.93 kPa) and the exponentially increasing heat input (Q = Q0 exp(t/τ), τ = 10 s) were systematically measured by the experimental water loop comprised of a multistage canned-type circulation pump with high pump head. The SUS304 test tubes of inner diameters (d = 3 and 6 mm), heated lengths (L = 33 and 59.5 mm), effective lengths (Leff = 23.3 and 49.1 mm), L/d (= 11 and 9.92), Leff/d (= 7.77 and 8.18), and wall thickness (δ = 0.5 mm) with average surface roughness (Ra = 3.18 μm) are used in this work. The inner surface temperature and the heat flux from non-boiling to CHF were clarified. The subcooled boiling heat transfer for SUS304 test tube was compared with our Platinum test tube data and the values calculated by other workers’ correlations for the subcooled boiling heat transfer. The influence of flow velocity on the subcooled boiling heat transfer and the CHF is investigated into details and the widely and precisely predictable correlation of the subcooled boiling heat transfer for turbulent flow of water in a short vertical SUS304-tube is given based on the experimental data. The correlation can describe the subcooled boiling heat transfer coefficients obtained in this work within 15% difference. Nucleate boiling surface superheats for the SUS304 test tube become very high. Those at the high liquid Reynolds number are close to the lower limit of Heterogeneous Spontaneous Nucleation Temperature. The dominant mechanisms of the flow boiling CHF in a short vertical SUS304-tube are discussed.
Effect of In-Core Temperature on the Life Evaluation of Graphite Component
Authors: Fang Xiang , Yu Suyuan , Wang Haitao
Published: February 2010

Graphite is widely used as a major internal structural material in high temperature gas-cooled reactors (HTRs). In order to evaluate service lives of graphite components, both the fast neutrons and temperature distributions inside the reactor core are required as input data for the irradiation-induced stress analysis. Since the fast neutron distributions may vary along the vertical direction due to the movement of the control rods, the corresponding location of the maximum neutron fluence changes with different in-core temperature levels. In this paper, the effect of the in-core temperature on both the stresses of graphite components due to irradiation and the corresponding life evaluations are studied numerically. The associated constitutive law for the simulation of irradiated graphite covering properties, dimensional changes and creep is briefly reviewed. A FORTRAN code based on user subroutines of MSC.MARC is developed in INET in order to perform three-dimensional finite element analysis of irradiated behavior of the graphite components for the HTRs. A side reflector graphite model is established and analyzed under a prescribed fast neutron distribution and several temperature levels. The results of irradiation-induced stresses and failure probability of the side reflector are obtained and the impact of temperature levels is discussed.
Application of Wavelet Scaling Function Expansion to Continuous-Energy Resonance Calculation
Authors: Yang Weiyan , Wu Hongchun , Cao Liangzhi
Published: February 2010

Wavelet expansion method can simulate functions oscillating violently. It has been successfully applied to simulate the violent variety of the angular flux. Neutron spectrum has fierce oscillation within the resonance energy range. This paper introduces a method to discretize the energy variable by coupling the multi-group method and wavelets expansion method, in which Daubechies’ wavelet scaling functions are used to expand the energy-dependent angular flux within the resonance energy range while traditional multi-group method is used within fast and thermal neutron ranges. Preliminary calculation results indicate that this method can obtain precise k-inf, as well as determine the fine-structure energy distribution of the angular flux, which is significant to the resonance self-shielding calculation.
Features of Design and Results of Function Tests at Tomari Unit 3
Authors: Sakai Ichiro , Yano Takashi , Takemura Takashi , Yamada Kousei
Published: February 2010

Tomari Unit 3 is not APWR but many advanced and latest technologies such as integrated I&C system including advanced main control board, advanced steam turbine and zinc injection are introduced into it. These three technologies are the fist application in Japan or in the world. Advanced main control board consists of large display panel, operator console and shift supervisor console and improved alarm system is introduced into it. Operator’s mental workload and human error probability is each improved 27% and 59% by introducing advanced main control board. Also maintainability and construction cost is improved by introducing integrated I&C system. MTTR is less than 30 minutes. 54-inch last blade in LP turbine, 3-dimensional hydraulic design blade, integral shroud blade and advanced bearing are applied to advanced steam turbine. Improvement of turbine efficiency by introducing 54-inch and 3-dimensional hydraulic design blade is evaluated to be approximately 0.3%. Vibration amplitude of bearings in turbine-generator during HFT was very small due to advanced bearing. Zinc injection was started from HFT for the first time in the world. Measured zinc consumption is consistent with anticipated. According to result of Zn injection during HFT, radiation sources are approximately 10–20% reduced in comparison with no injection.
Addressing Boron Dilution Scenario Through RELAP5/3.3 Analysis of PWR SB LOCA
Authors: Pla Patricia , Galetti Regina , D’Auria Francesco , Parisi Carlo , Giannotti Walter , Del Nevo Alessandro , Muellner Nikolaus , Cherubini Marco , Galassi Giorgio , Revento´s Francesc
Published: February 2010

Reactivity accident scenarios can occur originated by internal boron dilution in the primary system of a nuclear pressurized water reactor type (PWR or VVER). In essence the problem is caused by boron dilution following vaporization and condensation of the primary system coolant in case of decrease of primary system mass inventory, for example during a small-break loss of coolant accident (SB-LOCA) that may include boiling in the core with condensation of steam in the steam generators. When the liquid level in the reactor vessel decreases below the hot leg elevation, steam begins to flow to the steam generators and condenses there. This steam carries no boron and thus boron concentration in the cold leg loop seals begins to decrease. If for some reason this water plug with low boron concentration begins to flow towards the core and enters it without any major mixing with the borated coolant, the result is a positive reactivity insertion. The paper presents an analysis by RELAP5 Mod 3.3 code [1] of a small break LOCA of 20 cm2 area in the lower plenum of a four-loop PWR nuclear reactor. The boundary conditions of the calculations consider the eight accumulator tanks available, two/four low pressure injection systems (LPIS) available, and two of the four high pressure injection systems (HPIS) available. Sensitivity calculations were performed, regarding among other things, the boron concentration in the Emergency Core Cooling Systems (ECCS) and reactor cooling system (RCS) from Design Basis Accident (DBA) to beyond DBA conditions. From the results obtained, in some calculations boron dilution is observed in more than one loop seal. The situation in which the plugs in the loop seals are transported to the core without mixing with other borated water led to a potentially hazardous situation for four calculations in which initial conditions were far from DBA. It is important to emphasize that the present study has not the objective of a safety analysis of the NPP involved, but it should be considered inside research activities regarding the boron dilution issue.
Towards European Regulations for Nuclear Installations: Contribution of Nuclear Licensees
Authors: Zaiss Werner
Published: February 2010

The European nuclear industry recognises that the liberalisation of the European energy market has led to the deregulation of electricity generation and supply and that diversity of national regulations could seriously distort competition. Undoubtedly, harmonizing regulations is the best way of ensuring that the industry can evolve within a stable legal framework. Consequently, nuclear license holders supported the work of the Western European Nuclear Regulators Association (WENRA) on the harmonization of European safety standards for existing nuclear power plants, as well as for waste and decommissioning. This support led to the creation, within FORATOM, of the ENISS (European Nuclear Installations Safety Standards) Initiative, in May 2005, in Brussels. The principal mission of ENISS is to bring together decision-makers, operators and specialists from the nuclear industry with national regulators in order to identify and possibly agree upon the scope and substance of harmonized safety standards. ENISS currently represents the nuclear utilities and operating companies from 17 European countries with nuclear power programme. ENISS above all provides the nuclear industry with the platform that it needs to express its views, provide expert input and interact fully with regulators throughout the harmonization process. ENISS first task has been to present a common industry position with regards to the Safety Reference Levels that WENRA has proposed. By engaging in constructive debate with WENRA and playing a dynamic role in the process, ENISS also defends the industry’s interests in a proactive way. The work of ENISS is a good example of how dialogue and results-oriented participation with stakeholders can help identify optimal solutions to the problems that our industry faces today. Another task of ENISS is to strengthen the industry influence in the revision work of the IAEA Safety Standards as well as in the European Directive on Nuclear Safety.
Modeling of the Upper Head of a Pressurized Water Reactor During Transient Operation
Authors: Dubiel Klaus , Macken Nelson
Published: February 2010

A mathematical modeling approach is used to describe the thermal hydraulic behavior of the reactor pressure vessel upper head of a pressurized water reactor. We specifically describe vapor formation in the upper head during a shutdown in the emergency power mode. The multiphase fluid in the upper head is modeled mathematically. Lumped modeling is used, i.e., the properties of the fluid within the head do not vary spatially. Heat transfer occurs between the fluid and the structures, which are also modeled as lumped systems. The ordinary differential equations are solved numerically using standard techniques. The model has been tested and it performs qualitatively as expected. Results show the temporal distribution of temperature, structure heat transfer and void fraction. The model is suitable for different designs of the upper head, including “open dome” and “closed dome”.
The Impact of Flow Accelerated Corrosion (FAC) on Feeder Life Cycle Management
Authors: Pandey Mahesh D. , Jyrkama Mikko I. , Lehockey Edward M.
Published: February 2010

Wall thinning of outlet feeder piping by flow accelerated corrosion (FAC) is a serious form of degradation affecting some CANDU® stations. The general and localized loss of wall thickness is typically highest at or near welds and changes in pipe geometry due to increased velocity or turbulence. While the process is not a high safety concern because catastrophic failure is unlikely, feeder wall thinning may result in significant economic losses as a result of forced shutdowns for repair and replacement. Accurate modelling and prediction of feeder replacements and the probability of feeder failure is not only important for continued fitness-for-service, but essential for feeder life cycle management (LCM). This paper discusses some of the key issues related to feeder FAC, and presents a probabilistic method for modelling the wall thinning process. The wall thickness loss due to FAC is modelled using a random rate model, while the probability of feeder failure is based on an empirical approach. The developed methodology allows the estimation of the remaining life of both inspected and uninspected feeder populations, while methodically accounting for the uncertainties in the problem.
LBE and Pb as Coolants for Generation IV Reactors: The Impurities Problem
Authors: Gessi A. , Agostini P. , Agostini M. , Coccoluto G.
Published: February 2010

Pb and the eutectic alloy Pb 55.5Bi (LBE) have been chosen as possible reference coolants for Generation IV fast reactors. Originally, in the frame of ADS systems research activities, the LBE was considered to be the best possible choice as reference coolant, and, in that case, as spallation material, due to not only the very favourable thermal characteristics and low chemical reactivity of the alloy, but also due to the very low melting point (125°C), resulting in relatively low temperatures of the whole nuclear system. During the last years, the attention of the scientific community switched its attention to pure Pb. The formation of a large amount of Po, when Bi is irradiated, together with several technological issues were the major reasons for this switch. At ENEA Brasimone laboratories several years of experience both with LBE and with pure flowing Pb have been gathered, in the frame of the V, VI and the recently started VII framework Programmes of the EU. This paper focuses on this experience, highlighting the major issues and open points on the choice of pure Pb for Gen IV applications, focussing on the impurities control and filtering techniques.
Measuring Surface Indications in Polyethylene Piping
Authors: Schaaf Frank J. , McElroy Ernie W. , Lefler S. Steven , Doss Rachel L.
Published: February 2010

Replication techniques offer a simple, accurate and effective way to measure attributes of indications discovered in polyethylene piping. Replication involves making an impression of an indication with a quick curing compound. After the compound has cured the impression can be peeled off of the surface of the pipe to provide a negative image of the indication. This technique is very valuable when indications are discovered during fabrication and installation of piping. Characterization of indications is important for piping installed in accordance with Code Case N-755 to assure compliance with limits imposed for indication depth. This technique is easily adaptable to field use. The replicant sample also provides a permanent record of the indication for future reference. This paper summarizes the replication technique for determination of indication attributes in polyethylene piping. It also provides a summary of typical attributes of indications found in polyethylene piping.
Mini-Channel Supercritical CO2 Heat Transfer Measurements for Brayton Cycle Regenerators
Authors: Kruizenga Alan , Anderson Mark , Corradini Michael
Published: February 2010

Recently, it has become increasingly important to improve efficiency and reduce capital costs in nuclear power plants. This prompted significant work in studying advanced Brayton cycles for high temperature energy conversion. A particular improvement in the operation of an advanced carbon dioxide cycle, is the use of compact, highly efficient, diffusion bonded heat exchangers for the recuperators. These heat exchangers operate near the pseudo-critical point of liquid carbon dioxide, making use of the drastic variation of the thermo-physical properties. This paper focuses on the experimental measurements of heat transfer and pressure drop characteristics within mini-channels. Two test section channel geometries are studied: a straight channel and a zig-zag channel. Both configurations are 0.5m in length and constructed out of 316 stainless steel with a series of nine parallel 1.9mm semi-circular channels. The zig-zag configuration has an angle of 115 degrees with an effective length of ∼0.58m. Heat transfer measurements are conducted for varying ranges of inlet temperatures, pressures, and mass flow rates. Local and average heat transfer coefficients near the critical point are determined from measured wall temperatures and calculated local bulk temperatures.
S-Prism Reactor: The Fast Reactor for GNEP
Authors: Stratton Charles
Published: February 2010

GNEP (Global Nuclear Energy Partnership) has the objectives for a recycling program to reduce nuclear waste and reduce the risk of proliferation. These goals are important for the perception of the nuclear industry and for the industry to have a successful future. A fast reactor is much better suited for these goals than the existing Light Water Reactors or the development of an Advanced Light Water Reactor. The S-PRISM reactor is a fast reactor design by GE that is cost-effective and feasible for implementation, due to its low cost for design certification. The GNEP program has the ability to achieve its objectives for a recycling program by implementing the S-PRISM reactor.
A New Shape of Expansion Loop to Compensate Thermal Expansion of Piping Systems
Authors: Alessandrescu Aurel M.
Published: February 2010

To provide the piping system self-compensation under the conditions in which lenticular thermal pipe expansion devices cannot be used due to the crack induced corrosion (e.g.in the Heavy Water Plant) is accomplished by special devices (e.g. U,L,Z type). Under such circumstances a novative alternative compensation solution was implemented, namely, the employment of ROTATING Z type or AA (VΛ). Such a type of compensation consists in 2× 90° pipe bends and 2× 45° pipe bends. Calculations of 15 alternative compensation solutions with more over 120 variants of materials, nominal diameter, operating conditions (Po, To) were made by running CaePipe 5.15 analysis program for 3 cases: sustained, thermal expansion and operating. The stress and distorsion conditions of the new shape satisfied the requirements in ASME Code B31.3 evidencing that this shape provides the most favorable results.
35 Years of Operating Experience of PHENIX NPP Sodium Cooled Fast Reactor
Authors: Dall’Ava Didier , Martin Laurent , Vray Bernard
Published: February 2010

Design features: Reactor core. The reactor block is of an integrated design (pool) except for a few auxiliary circuits. The entire primary sodium system, containing 800 tons of radioactive sodium, is enclosed in the main reactor vessel, which is 11.8 m in diameter (Fig. 1).
Progress in Developing an Improved Empirical Heat Transfer Equation for Use in Connection With Advanced Nuclear Reactors Cooled by Water at Supercritical Pressure
Authors: Jackson J. Derek
Published: February 2010

Consideration of advanced power plant such as nuclear reactors cooled by water at pressures above the critical value has stimulated a renewed interest in heat transfer to supercritical pressure fluids. Severe deterioration in the effectiveness of heat transfer can be encountered as a result of the extreme dependence on temperature of the physical properties of such fluids, particularly near the pseudocritical temperature where their molecular structure changes from being liquid-like to gaseous. This deterioration arises mainly as a result of the non-uniformity of density, which can lead to significant influences of bulk flow acceleration and fluid buoyancy. A good physical understanding has been arrived at of the mechanisms by means of which such influences can modify the mean flow and turbulence fields and thereby the advection and turbulent diffusion of heat and effectiveness of heat transfer. However, this progress in understanding the physics has so far not resulted in such effects being reliably accounted for in the empirical equations which are available for thermal design. With a view to addressing this matter, the author has recently attempted to update and improve an existing physically-based semi-empirical model of variable property heat transfer. The aim has been to combine it with a soundly-based empirical forced convection equation to extend the applicability and reliability of currently available thermal design procedures. In the present paper, progress in validating this approach and optimising the performance of the extended equation is reported.
Hydraulic Cylinders for Offshore Splash Zone Operation: A Review of Piston Rod Failure Cases and Alternative Concepts
Authors: von der Ohe Christian B. , Johnsen Roy , Espallargas Nuria
Published: February 2010

Hydraulic cylinders are used for many vital applications on offshore installations. For the long stroke cylinders used in the direct acting riser tensioning application the piston rod surface is exposed to severe degradation; combining corrosion, wear and mechanical stress. Direct acting riser tensioning systems are often used during ultra-deepwater offshore drilling, when exploring for oil and gas. These direct acting riser tensioning systems usually consist of 6 long stroke hydraulic cylinders located at the splash zone level. These cylinders ensure constant tensioning of the drilling riser as the drilling vessel is operated in ultra deep waters up to 3000 meters compensating for heave due to sea waves and tide. The direct acting tensioning systems have been in operation since the late 1990’s, but the system integrity and service life has so far not been acceptable. Operational experience shows that the thermal sprayed coatings applied on these piston rods have dramatically shorter life than predicted by qualification tests programs. The consequence of the failing rod surface is sudden coating damage such as pits and flaking which further damages the seals in the packing flange. As a result hydraulic oil leakages occurs, thus polluting the environment and in the worst case leads to loss of the tensioning in the riser causing subsequent drilling down time and eventual riser breakage and blowout. This paper will provide a generic summary of operational experience, development and degradation issues of direct-acting riser tensioning cylinder piston rod surfaces. Several case histories are included to highlight piston rod failures and an evaluation of alternative concepts to improve service life of piston rods is given.
Creating Multi-Sectoral Ocean Industry Leadership in Marine Spatial Management
Authors: Holthus Paul F.
Published: February 2010

Marine spatial management is emerging as a key tool for managing and conserving marine ecosystems. The development and implementation of ocean zoning, especially Marine Protected Area (MPA) networks, is expanding rapidly around the world. At a global scale, ocean zoning and MPAs are being pursued in many processes through a variety of governmental, inter-governmental (e.g. UN agencies, regional bodies), with significant input from non-governmental organizations (NGOs). Ocean industries, e.g. oil and gas, shipping, fisheries, aquaculture, etc. are the principle user group affected by ocean zoning. Important new ocean uses are under development as well, e.g. offshore wind, ocean energy, carbon sequestration. The creation of ocean zones, such as MPA networks, will have major effects on industry access to marine space and resources, with significant implications for exploration, development, production and transport. It is essential that the industry constructively engage with marine spatial management efforts and stakeholders to ensure that the process is well informed and balanced. This will increase the potential for industry to respect the management regulations of ocean areas often far removed from the reach of governments. Unfortunately, industry is often not present at key ocean zoning developments, especially at the international level, and is not engaged in a constructive, coordinated manner that brings together the range of industries operating in the marine environment. Barriers to industry involvement in marine spatial management include: 1) Lack of understanding of the movement and momentum behind ocean zoning efforts; 2) Limited engagement in the multi-stakeholder processes where zoning is moving most rapidly, e.g. Convention on Biological Diversity, because industry is engaged in sectoral processes; 3) Lack of means for engaging the broader ocean business community on marine management and sustainability issues, particularly at a global level. This paper will outline the need, opportunity and progress in creating proactive, constructive industry leadership and collaboration on ocean zoning, including: 1) Developing an understanding of marine spatial management issues, stakeholders and process; 2) Actively engaging in key international multi-stakeholder processes that are pursuing MPAs and other zoning efforts; 3) Building constructive relationships among ocean industries and other ocean zoning stakeholders; 4) Creating practical experience by constructively engaging in ocean zoning developments in a specific regions.
Assessment of Transient Sloshing Due to Encounter of an LNG Carrier With a Steep Wave
Authors: Schreier Sebastian , Godderidge Bernhard , Paschen Mathias , Turnock Stephen , Tan Mingyi , Cowlan Nick
Published: February 2010

This study investigates a transient sloshing flow in a rectangular tank by comparing model tests and unsteady Reynolds-Averaged Navier-Stokes CFD simulation. The sloshing flow is excited by a normally periodic surge motion, which is then subjected to a transient akin to an LNG carrier encountering a particularly steep wave. It is found that the impact pressure recorded in the transient motion is several times greater than that observed at resonance. High speed video data and CFD simulation is used to analyse the sloshing flow. The severity is explained by the absence of entrained air bubbles prior to impact and the second mode response one half period before impact.
Experimental Study on Dynamic Hydrocyclone for Thin Oil Dewatering
Authors: Lv Fengxia , Wang Zunce , Li Sen , Xu Yan , Zhang Shuang
Published: February 2010

The special dynamic hydrocyclone for the thin oil dewatering is developed combining with its advantage in creating high strength swirl field, aiming at the character of the thin oil. By lab experiment, the curves, between the input water cut, viscosity, speed of rotating wall, the ratio and the efficiency, are obtained. Then, the operating parameters are optimized for the dynamic hydrocyclone with specified construction. Subsequently, one process of series/parallel connection of several dynamic hydrocyclones is designed. The result of the experiments shows that applying the dynamic hydrocyclone in thin oil dewatering can lessen the water cut to under 5%, simultaneously, the manufacturing period of the product oil is reduced while the post disposal cost is also reduced since the oil concentrate-on has been controlled under 200mg/l in the sewerage. The research illustrates that the application of the dynamic hydrocyclone in thin oil dewatering is absolutely feasible, which can serve as the valuable experience of the dynamic hydrocyclone in extending application.
Preliminary Instability-Analysis of Deepwater Riser With Fairings
Authors: Khorasanchi Mahdi , Huang Shan
Published: February 2010

Instability of deepwater riser with fairings is investigated in this study. Despite the advantages over other devices for suppressing vortex-induced-vibration (VIV), fairings may be susceptible to flutter type instability. A two-body mathematical model is established for the coupled transverse-torsion motion of a top tensioned riser with fairings. The inner part (riser) can only move transversely while the outer part (fairing) has transverse-torsion motion. The effect of the transverse velocity on the angle of attack is taken into account and damping is considered for both degrees of freedom. An eigenvalue analysis is employed to examine the issue of stability. The emphasis is on identifying the critical current speed for a given riser and fairing configuration. The effects of key parameters are investigated and the results indicate that the section hydrodynamic characteristics of the fairings have a significant impact on the instability.
Tracking a Wave Power Buoy Using a Network Camera: System Analysis and First Results
Authors: Tyrberg Simon , Gravrakmo Halvar , Leijon Mats
Published: February 2010

An observation system has been set up on a small islet on the Swedish west coast. The purpose of the system is to monitor the wave buoys in The Lysekil Project. The project is an attempt to harvest wave energy using linear generators and point absorbing buoys. The observation system is self-sufficient and uses a network camera to follow the buoy motions. The first results from the camera, which has been operating since July 2008, have been analyzed to examine the motion tracking capabilities of the system. The motion tracking would work as a complement to the other measurements that are being done on the buoy. The method for extracting motion data from the two-dimensional pictures is presented. The results are graphs of translative buoy motion in two dimensions, and rotational motion about two different axes. The vertical buoy motion for the studied sequence is in the range of ±0.5 m.
Model Test Study on Vortex-Induced Motions of a Floating Cylinder
Authors: Wang Ying , Yang Jianmin , Peng Tao , Li Xin
Published: February 2010

Vortex-Induced Motions (VIM) under current flow is an important issue for surface piercing cylinders, such as Spar platforms and floating buoys, since it affects the motion performance of these structures greatly. In recent years this phenomenon attracts much attention and many researchers have been making efforts to deal with this problem. VIM is such a complicated phenomenon that more fundamental studies are needed to understand the essence behind VIM. This paper mainly concentrates on a circular cylinder, aiming to eliminate outside influences and reveal the inherent characteristic of vortex-induced motion mechanism. A circular cylinder with an aspect ratio of 1:2.4, which could be considered as a scale model for the hard tank of a typical Truss Spar, is studied by experimental method to investigate the surrounding fluid field, the excitation forces and Vortex-Induced Motion characteristics under various governing parameters, such as the current velocity and direction, the mooring stiffness and distribution, the use and efficiency of helical strakes, and so on. By using a simple flow visualization system, the unsteady flow passing the circular cylinder and the vortices in the wake are captured and recorded. The cylinder is tested respectively under fixed, forced-motion and elastically moored conditions. The fluid field, the vortex structures, and the lift and drag forces under fixed and forced-motion conditions are measured, the VIM performance of the cylinder with two different mooring distributions are studied, and strake efficiency is studied considering current directionality and strake height influence.
Ultra Heavy Wall Linepipe X65 for the Most Stringent Applications: Metallurgical Design and Industrial Development
Authors: Izquierdo Alfonso , Quintanilla Hector , Richard Gilles , Garcia Enrique , Armengol Mariano , Novelli Paolo , Di Vito Luigi , Porcu Giorgio , Mannucci Gianluca , Anelli Ettore , Mannucci Aldo
Published: February 2010

Offshore industry has evolved to meet numerous challenges, e.g. deep water, high currents, high pressure and high temperature (HPHT), and sour reservoirs, facing deepwater exploration. The trend in flowline specifications for deepwater offshore fields is a consequence of complex oil-gas field conditions, such as HPHT and developments in design criteria (i.e. limit state design), welding and laying technologies. The technological evolution exhibits a trend towards an increasing wall thickness (WT) to provide sufficient resistance for the very high operating pressures. Furthermore, the pipelay operations, especially when linepipes are installed by means of the reel laying method, cause repeated plastic bending and straightening deformation cycles. These cyclic loads affect final material stress-strain properties. Reeling is currently applied to an increasing range of pipe geometries, being the present limit given by pipes with 16″ outer diameter (OD) and 30 mm wall thickness (WT). Other pipeline installation techniques, for example, J-lay, S-lay and steep S-lay also introduce plastic strain. All previous factors mentioned before and adding one more variable when exploring and producing in regions alike to the Artic where low temperatures implied several material challenges calls for high performance seamless pipes tailored to the specific application required by the oil and gas industry. In this paper, a description is given of the results of latest fundamental studies on high-strength heavy-wall steel materials manufactured by Q&T processing. This work is part of an on-going development program on high performance heavy wall seamless pipes for special applications such as HPHT, low temperature design criteria, sour requirements and studying the material under the strain based design criteria involving metallurgical modeling, laboratory tests, industrial trials and advanced metallographic examinations. The most recent findings and overall conclusions are reported hereafter, these results have been exploited by Tenaris to manufacture a limited production seamless pipes in a wall thickness range from 40 mm to 48 mm in steel grade X65 Sour Service.
Experimental Study of Down Hole Gas Liquid Separators
Authors: Wang Zunce , Li Sen , Lv Fengxia , Xu Yan , Zhang Jinlong
Published: February 2010

The technology of Down-hole Gas Liquid Separation and Water Re-injection (DGLSWR) is an economical and effective method to solve gas well fluid accumulation. The separation performance of designed Down-hole Gas Liquid Separator (DGLS) is very important for DGLSWR systems applications. The principle of work and Characteristics of DGLSWR systems are introduced in this paper. Separation performance of DGLS was studied using computational fluid dynamics (CFD) simulation combining laboratory experiment. Relations of main operating parameters, such as flow rate and gas liquid ratio with pressure drop were studied. The effect of flow rate, gas liquid ratio and main structural parameters such as cone angle and exhaust on DGLS separation performance was also studied. Appropriate structure and operating parameters were determined. Field tests indicated satisfactory results as well.