bluuebunny/crossref_metadata_2025 huggingface dataset:
Mixing Behavior of Argon Jet With Liquid Sodium Around a Single Rod: A Basic Study on Sodium-Water Reaction
Authors: Tsuruoka Hokuto , Tamura Takeshi , Sugiyama Ken-Ichiro , Narabayashi Tadashi , Ohshima Hiroyuki
Published: July 2009

The occurrence of secondary heat transfer tube failure due to overheating by sodium-water reaction in LMFBR steam generators has been concerned from the viewpoint of public acceptance. To evaluate the phenomena for the secondary heat transfer tube failure, a sophisticated computer code SERAPHIM has been developed by JAEA. The comparison of simulation results with experimental data to verify the adequacy of SERAPHIM code or to upgrade it are now required. In our paper at ICONE15, we report the void fraction and the entrainment pattern of water around a single rod set in a water pool using the probes developed for sodium experiment. In addition to these results, the reliability of the developed probe was verified by the experiment with gas jet velocity more than 340m/s accompanying oscillation. In the present paper, we reported the void fraction around a single rod in a diameter of 20mm set in a sodium pool without sodium-water reaction. The velocity of argon gas jet changed from 17.3m/s to 173m/s, which was lower velocity than that in the water experiment from the safety consideration and the void fraction was measured at every 30°. The void fraction was observed to somewhat increase with increasing the gas jet velocity. The increase rate was clearly smaller compared with that in the water experiment. The void fraction also showed more monotonous distribution from the stagnation point to the rear point than that in water pool. These results reflect the difference of surface tension between water and sodium. Because sodium has about three times surface tension as large as water, the argon gas jet column with oscillation in the sodium pool causes easily break-up than that in the water pool. It is concluded that the entrainment of ambient sodium is easily caused and this leads monotonous distribution of void fraction in the sodium pool.
Suppression Features of a Vapor Explosion With Prototypic Reactor Materials
Authors: Hong S. W. , Kim J. H. , Min B. T. , Park I. K. , Kim H. D.
Published: July 2009

The suppression of a vapor explosion is reviewed from a void fraction point of view from previous research results and the results of an experiment and analysis for TROI using a prototypic reactor material. In a tin-water system, a high fraction of air which played the role of a steam reduced the peak pressure of a steam explosion. According to the sensitivity analysis that was carried out with an increase in the vapor volume fraction, an energetic vapor explosion hardly took place in the mixture with a high void fraction. In higher vapor fraction conditions (αv> 0.3), the vapor explosion was very weak. A prototypic corium shows a relatively high void fraction compared to the ZrO2 which is known as an explosive material because the corium system generated many smaller particles compared to the ZrO2 system. The corium system shows a relatively low explosivity compared to the ZrO2 system because the high void fraction of the corium system plays the role of preventing a contact between the water and the hot melt drops in the triggering stage. When considering the experimental results for the role of air instead of steam, an air supply system to provide a high volume fraction during a premixing process can radically prevent and/or mitigate a steam explosion.
Study on Velocity Field in a Deformed Fuel Pin Bundle: Influence of Pin Deformation and Wrapping Wire on Velocity Distribution
Authors: Sato Hiroyuki , Kobayashi Jun , Miyakoshi Hiroyuki , Kamide Hideki
Published: July 2009

A sodium cooled fast reactor is designed to attain a high burn-up core in a feasibility study on commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, deformation of fuel pin due to the swelling and thermal bowing may decrease flow rate via change of flow area in the sub-assembly and influence the heat removal capability. A 2.5 times enlarged 7-pin bundle water model was applied to investigate the influence of pin bowing and wrapping wire. The test section consisted of a hexagonal acrylic duct tube and fluorinated resin pins which had a refractive index (1.336) nearly the same with water (1.333) and a high light transmission rate (93%). This enabled to visualize around the central pin through the outer pins. In this experiment, the bowing deformation of a fuel pin was simulated by compressing the central pin in the axial direction. Velocity distribution was measured by using Particle Image Velocimetry (PIV) in sub-channels around the central pin in reference and deformation conditions. The Laser Doppler Velocimetry (LDV) was used for the cross-check, the velocity distributions were consistent between PIV and LDV results. The measured velocity distribution around the wrapping wire revealed that the wire influenced to the velocity and RMS in the wide region above the pin surface.
Probabilistic Model for Debris-Induced Loss of Long Term Core Cooling
Authors: Detar Heather L. , McLaughlin Daniel T. , Lutz Robert J.
Published: July 2009

Generic Safety Issue (GSI) 191 deals with the potential for generation and transport of debris following a design basis accident that is in excess of quantities assumed in the original design basis and licensing of Pressurized Water Reactor (PWR) plants. In addition to physical modifications to the sump screens to comply with the Generic Letter requirements, some plants have also changed Emergency Operating Procedures (EOPs) to include contingency actions to prevent debris-induced loss of long term core cooling. ASME Probabilistic Risk Assessment (PRA) standard RA-Sb-2005 requires that the plant PRA be maintained and updated to reflect the current plant design and operation. Development of PRA models to quantify the potential for debris-induced loss of long term core cooling supports the PRA updated to reflect the as-built as-operated plant. The PWR Owners Group (PWROG) has undertaken a program to develop a generic PRA model for this issue. The generic PRA model was developed to address the overall plant risk, including new physical and procedural modifications. The model also addresses applications, such as Maintenance Rule screening and the assessment of the risk significance of deviations from the licensing basis analyses. The new PRA model probabilistically treats several facets of the potential for debris-induced challenges to long term core cooling; including debris generation and transport as a function of Reactor Coolant System (RCS) break size and location. The PRA model will permit plant operators to easily incorporate the potential for inadequate core cooling during emergency core cooling recirculation from the containment sump into their PRA Level 1 and Level 2 models. The methodology is based on realistically modeling the conditions that may lead to a debris-induced loss of long term core cooling. The PWROG model also includes consideration of water management strategies being implemented by several PWR plant operators.
Analysis of PSA Conservative Assumptions’ Impact on Risk Profile for Armenian NPP Unit 2
Authors: Poghosyan Shahen , Amirjanyan Armen , Bznuni Surik
Published: July 2009

PSA is a powerful tool for detailed risk assessment and risk-informed decision making. The main deliverables from PSA models is comparative risk assessment of different risk contributors including systems, components, IEs, etc. Risk-informed decision making process is based on obtained risk profile. In comparison with deterministic approach conservatism included in PSA models could lead to deformation of risk profile which will bring to inappropriate decision making based on PSA results. This paper presents analysis done for Armenian NPP PSA study in order to assess the impact of conservative assumptions on final risk profile coming from potential internal initiating events at full power operation.
Dresden Units 2 and 3 Isolation Condenser Waterhammer Prevention During Restart Operations
Authors: Freeman John , Ramsden Kevin , Kovacs Paul
Published: July 2009

The Dresden Units 2 and 3 have a history of hydraulic transients occurring in the Isolation Condenser System piping. One of the most significant transients occurred following a spurious trip and isolation of the Isolation Condenser when operators restarted the system. Piping support damage occurred on supports in the condensate return line. A detailed RELAP5 M3 analysis of the Isolation Condenser and reactor system was performed to study the thermal hydraulic behavior of the system through system initiation and subsequently to system isolation and restart. Several post isolation restart times were analyzed to provide guidance for operating procedure changes. Isolation valve operation sequence was also considered. Piping segment force time histories in the condensate return line were generated to evaluate the pipe loadings. The loads computed compared favorably with the historical damage observed and demonstrated that allowing sufficient time for post isolation void transients to complete will allow system restart with minimal loading on the system.
The Visualization of the Flow in Pressurizer Spray Line Piping and Estimation of Thermal Stress Fluctuation Caused by the Swaying of Water Surface
Authors: Oumaya Toru , Nakamura Akira , Takenaka Nobuyuki
Published: July 2009

In a PWR plant, on top of the pressurizer, there’s spray line that cools the reactor coolant. The amount of minimum flow rate is considered to be insufficient to fulfill the pipe completely during commercial operation of the plant, so it is concerned that there is a surface of water in the pipe and it may periodically sway due to condensation of the steam from the pressurizer during commercial operation. In order to identify the flow conditions in the spray line piping and assess its impact on thermal stress, flow visualization experiment were conducted. In the experiment, air is used in substitute of steam to simulate the gas phase of the pressurizer. With a full scale mock-up made of acrylic, flow under room temperature and atmospheric pressure condition was visualized and the locations where periodical swaying of water surface may be observed were identified. The period of swaying and its extent is roughly estimated based on the experiment. Based on the visualization experiment, 3 patterns of swaying of water surface were assumed and the amplitude of thermal stress fluctuation when the layer swayed instantaneously was calculated. Thus, with the assumed 3 patterns of swaying based on the visualization experiment, it was confirmed that the amplitude of thermal stress would not exceed the endurance limit defined in the Japanese design and construction code.
Current Endorsement of Consensus Codes for Regulatory Technical Standards in Japan
Authors: Tada Kenji , Kanda Tadao , Sakuma Takayuki , Tanimoto Masataka , Kitahara Kazuhide , Kurihara Mikio , Goto Yukinori , Ogiso Zenichi
Published: July 2009

In Japan, on January 1, 2006, “Ordinance of the Ministry for Establishing Technical Standards of Nuclear Power Generation Systems” (so called Ordinance No.62) was revised. The revised Ordinance No. 62 provides only functional requirements or performance requirement for nuclear power facilities, and detailed specifications — such as the detailed requirement of material and structure for a pressure vessel — are used to endorse consensus codes, like the Codes of the Japan Society of Mechanical Engineers (JSME). To endorse consensus codes, the Nuclear and Industrial Safety Agency (NISA), the regulatory body, and the Japan Nuclear Energy Safety Organization (JNES), a technical support organization, have done technical evaluations on several of the required consensus codes, and these results have been provided in the Regulatory Guide of Ordinance No.62. The activities of technical evaluation of the consensus codes were reported at the ICONE14th. This paper shows their subsequent situation, and the activities of proposing new codes that are not yet in the Societies’ code.
Performance of DBA-Qualified/Acceptable Coatings When Subjected to Jet Flows From High Energy Pipe Breaks
Authors: Andreychek Timothy S. , Dingler Maurice E. , DeBarba John , Sadowski William F.
Published: July 2009

A postulated Loss-of-Coolant Accident (LOCA) for a Pressurized Water Reactor (PWR) results in the escape of high-pressure, high-temperature subcooled coolant from the pipe break location. As the coolant escapes from the break, it forms a two-phase jet. The impingement of a two-phase jet may damage protective coatings used on surfaces inside the reactor containment building, resulting in the generation of coatings debris. The coatings debris may be transported to the pool forming on the containment floor. When the Emergency Core Cooling (ECC) and the Containment Spray (CS) Systems are realigned to draw suction from the containment sump, the coatings debris can be transported to the containment sump screen. Large amounts of coatings debris at the sump screen may contribute to undesirable head losses across the sump screen and have undesirable effects on components downstream of the screen. This paper presents the experimental program undertaken to study the performance of Design Basis Accident (DBA) Qualified/Acceptable coating systems typically used inside PWR containments under jet impingement loads, temperatures and pressures associated with a postulated large break LOCA. The coating systems were applied to both carbon steel and concrete substrates consistent with the manner in which the coating systems were applied for DBA testing. A total of eight (8) coupons were subjected to the thermal and hydraulic processes and phenomena associated with jet impingement that would result from the coatings being in close proximity to a postulated large break LOCA. A description of the test protocol, the test facility, and the test conditions is presented, along with post-test inspection summaries for the individual test coupons. This paper will also discuss the extension of the test data to other DBA Qualified/Acceptable epoxy coating systems and untopcoated IOZ coating systems.
Non-Linear Characteristics of Subcooled Water Flow Boiling CHFS Versus Outlet Subcoolings for Flow Velocities at Pressures in Various Tubes
Authors: Fukuda Katsuya , Liu Qiusheng
Published: July 2009

Non-linear characteristics of the subcooled water flow boiling CHF versus outlet subcooling for a flow velocity with outlet pressure as a parameter in a horizontal or vertical tube with a twisted tape having flowing subcooled water was clarified using the existing databases expanding quite recently derived the CHF correlations based on two CHF mechanisms for the bare tubes with various inside diameters and length-to-diameter ratios for a flow velocity with outlet pressure as a parameter. The databases were applied measured using, the horizontal tubes with inside diameters of 8 and 12 mm, having length-to-diameter ratios of 6.25 and 4.17 respectively with twisted-tapes having a twist ratio of 2.0, for flow velocities of 7 and 10 m/s, at pressures ranging from 0.2 to 1.1 MPa. The nonlinear characteristics of subcooled water flow boiling CHFs for outlet subcoolings in the tubes having helically coiled 1 mm diameter wires of the coil pitches of 12, 24 and 36 mm, with flowing subcooled water of a flow velocity of 7 m/s at outlet pressures of 0.2, 0.3 and 1.1 MPa were investigated based on the subcooled water flow boiling CHF correlations for bare tubes previously derived. The following results were derived that (1) the non-linear characteristic of CHF versus outlet subcooling, for a horizontal or vertical tube having a twisted tape, divided into three regions for middle, transition, and high outlet subcooling was clarified. The characteristics were similar to that for a identical bare tube, with higher CHF values for middle and high outlet subcooling values respectively, and (2) the enhanced CHF values for middle and high outlet subcooling values being dependent on outlet pressuret and being independent of one respectively, were described by the same CHF correlations derived for a identical bare tube, with higher coefficients of the correlations depending on the tubes with twisted tapes of twist ratios.
Minor Actinides-Loaded FBR Core Concept Suitable for the Introductory Period
Authors: Fujimura Koji , Sasahira Akira , Yamashita Junichi , Fukasawa Tetsuo , Hoshino Kuniyoshi
Published: July 2009

We proposed the “Flexible Fuel Cycle Initiative” (FFCI), which has flexibility for the uncertainties like the introduction speed of FBRs. On the other hand, during the FBR introduction period, Pu from LWR spent fuel is used for startup of FBRs. But the FBR core being loaded with Pu from LWR spent fuel has larger burnup reactivity due to its larger isotopic fraction of Pu-241 than the core being loaded with Pu from the FBR multi-recycling core. The increased burnup reactivity may reduce the cycle length of the FBR. In this paper, an FBR transitional core concept to handle the issues of the FBR introductory period was investigated. Core specifications are based on the compact type sodium-cooled MOX-fueled core designed in the Japanese FBR cycle feasibility studies, because the lower Pu inventory should be better for the FBR introductory period in view of its flexibility for the required reprocessing amount of LWR spent fuel to start up the FBR. The reference specifications are selected as follows. Output is 1500MWe and the average discharge fuel burnup is about 150GWd/t. Minor Actinides (MAs) recovered from LWR spent fuels which provide Pu to startup FBR are loaded to the initial loading fuels and exchanged fuels during some cycles until equilibrium. We set a kind of MA fraction rate of the initial loading fuel with 4 as the number of the fuel exchange batches. The average of the MA fraction of the initial loading fuel assumed is 3%, and the MA fraction of the exchange fuel is set as 5%. This 5% maximum of the MA fraction is based on the irradiation results of the experimental fast reactor Joyo. The core performance including burnup characteristics and reactivity coefficient were evaluated, and we confirmed that the transitional core from the initial loading until equilibrium cycle loaded Pu from LWR spent fuels could keep the resemble performance with the FBR multi-recycling core.
A Two-Phase CFD Approach to the PTS Problem Evaluated on COSI Experiment
Authors: Coste Pierre , Pouvreau Je´roˆme , Lavie´ville Je´roˆme , Boucker Marc
Published: July 2009

The study is done in the context of Pressurized Thermal Shock (PTS) CFD investigations related to PWR life duration safety studies. In the simulations of such situations direct contact condensation on free surfaces much larger than cells size is a key phenomenon. Those large interfaces require a special full set of models which has been recently implemented in the NEPTUNE_CFD code. Out of large interfaces regions, some dispersed or churn flows can take place. Both situations can be taken into account in the computational domain. The approach includes an interface recognition but not an interface reconstruction. It is evaluated on COSI experiment. COSI facility is scaled 1/100 for volume and power from a 900 MW PWR under LOCA thermal hydraulic conditions. Measurements include temperature profiles at various axial positions in the pipe. The paper focuses on one test corresponding to a situation with a low water level in the leg. It is a demanding case for two-phase CFD because -a- the liquid layer which is of prime importance in PTS studies can be quite thin in comparison with the whole computational domain -b- the emergency core cooling jet plays an important role. Results are within a reasonable range even with a coarse mesh. Calculations with finer meshes quantify the effect of a better simulation of the jet fall and impact on the wall.
AP1000 Fuel Design and Core Operations
Authors: Miller R. Wade
Published: July 2009

This paper discusses the key features of the Fuel design and Core Operations of the AP1000 reactor. Core Design and thermal parameters are compared to those of existing Westinghouse reactors. Details of the Fuel Assembly design are provided, along with relevant operating experience. Finally, differences in core operation and reactivity control between the AP1000 and previous generation Westinghouse reactors are discussed in the context of overall plant design and operation.
RCS Depressurization and Loss of Flow No Trip 3D Neutronic Transient Analyses
Authors: Gu¨ler Cenk , Risher Daniel H.
Published: July 2009

To be able to correctly analyze an RCS Depressurization or Loss of Flow accident with no reactor trip, the Westinghouse 3D Transient Neutronics tool RAVE™ was used. Plants which utilize all-digital Instrumentation and Control (I&C) systems may be required to consider a complete failure of the digital protection system. A backup analog system is provided in the event of a failure of the digital system; however the backup analog systems do not carry all safety trip signals. Therefore, it is postulated that the plant might experience an accident event in which no immediate trip signal is generated and protection would rely on operator action. To allow adequate time for the operator to recognize the malfunction and take action to manually trip the plant, it is assumed that operator action will not occur for ten minutes from the initiation of the transient. These kinds of transients may be analyzed with point kinetic transient codes; however, due to the lack of 3D neutronic feedback with a core height dependent moderator temperature coefficient and its effect on the axial and radial power distributions, it may not be possible to show that Departure from Nucleate Boiling (DNB) does not occur. Using 3D Neutronic transient analyses methods (RAVE™), it was shown that neither of these transients (RCS Depressurization or Loss of Flow) would generate DNB violations, even though the point kinetics version is showing the opposite.
The Health of Risk-Informed Regulation
Authors: Lutz Robert J. , Schneider Raymond
Published: July 2009

Ten years ago, risk-informed regulation was proclaimed by many to be the future of regulatory activity. The three options for using risk in the regulatory arena had been defined in SECY-98-300 and efforts were initiated to develop the first regulatory changes. However, regulatory change has been slow. The revised 50.44 rule, which was supposed to be non-controversial, was not finalized until September of 2003, five years after it was started. The new 50.69 rule, which embodied the STP graded quality assurance principles, was not finalized until November of 2004 and the Pressurized Water Reactor Owners Group (PWROG) pilot plant application is still under review by the Nuclear Regulatory Commission (NRC). The 50.46a rule change, admittedly a very difficult and controversial undertaking, is still not finalized. On the other hand, the NRC has increased the use of plant risk insights and results in the Regulatory Oversight Process, particularly through the Significance Determination Process (SDP) and the Mitigating System Performance Index (MSPI). While the industry has been working on the major regulatory initiatives described above, they have also successfully initiated a significant number of risk-informed programs within the existing regulations. Significant among these industry efforts are the risk-informed changes to the Technical Specifications and the risk-informed changes to in-service inspection and in-service testing. One bright spot on the horizon is the use of NFPA-805, which uses fire risk insights, as an alternative to the Appendix R requirements for fire protection. The industry and the NRC are working together to develop an acceptable methodology for implementing this alternative. The increasing requirements for scope and quality of the probabilistic risk assessments (PRAs) that are used as the basis for many risk-informed activities has resulted in a major slow-down in licensee implementation of new risk-informed activities. The publication of industry consensus standards is resulting in significant resource expenditures to upgrade the scope and quality of the PRAs. Once the scope and quality of the plant-specific PRAs is completed, there should again be significant implementation of risk-informed applications. In summary, risk-informed regulation has already produced significant increases in safety and has potentially reduced licensee regulatory burdens. However, the promise of risk-informed changes to regulations is still an unfulfilled expectation and will likely remain so in the near future. Once the scope and quality of PRAs is upgraded to satisfy the industry consensus standards, there should be renewed activity in implementing risk-informed programs.
A Thermal-Hydraulics Analysis of Pipe Breaks in the Pebble Bed Modular Reactor Main Power System
Authors: Mazana Naison
Published: July 2009

This paper presents observations from analyses of pipe breaks presumed to occur in the Pebble Bed Modular Reactor’s (PBMR’s) Main Power System (MPS), and a brief insight into the implications of evaluated shear ratio values on dust lift-off and dust deposition in the MPS. The PBMR is a multibillion rand project tasked with the design, construction, commissioning and marketing of first-of-a-kind pebble-bed, high-temperature gas-cooled nuclear technology. The main loop of the envisaged nuclear power plant comprises a single Brayton thermodynamic cycle that integrates the pebble-packed reactor to a system of heat exchangers and to a single-shaft turbine-generator-compressor system. The reactor coolant is helium which, under 9 000 kPa, exits the reactor at approximately 900 °C. This drives the downstream turbine in a direct action set-up, supplying the grid with roughly 165 MWe of electricity. The PBMR’s MPS and support systems have progressed to advanced design phases. Thermal-hydraulics analysis has played an integral role in providing input data to design and systems engineering functions. Twenty-eight critical positions around the MPS have been identified for the capture of parameters used to calculate shear ratios, which are used to estimate dust lift-off fractions that could result from the pipe break scenarios, and to predict nuclear doses likely to affect the reactor’s adjacent compartments and the immediate environment.
Methodology and Principle for Emergency Operation Computerization in Loviisa NPP Automation Renewal Project
Authors: Kernei¨s Claude , Korhonen Jarmo , Martinsuo Mikko
Published: July 2009

Loviisa NPP has two VVER 440-type pressurised-water reactors. The automation renewal project started in the beginning of 2005 when Fortum signed a contract with a consortium of AREVA-NP and Siemens AG. The functional concept of the new I&C is defined in terms of task categories. In normal operational conditions the nuclear power unit is controlled by normal process control functions. Also safety-related preventive protection functions may be activated. Reactor protection functions are activated in operational transients and in accidents. Also diverse manual and automatic backup of the Reactor Protection System is foreseen. After an operational transient or an accident the operators steer the plant to a safe state using plant shutdown or accident management functions. The new control systems and Man Machine Interface (MMI) will be implemented using safety and non-safety digital control platforms and user interfaces. In addition to these conventional non-digital platform with dedicated MMI will be provided to control selected critical functions. Loviisa NPP is using Emergency Operating Procedures (EOP) designed together with AREVA-NP, EdF and Fortum consisting of event and state oriented procedures. This set of procedures is being adapted to a functional concept of the new I&C together with AREVA-NP and Fortum. The rules for organisation of operation with different EOPs and MMIs and the principles and methods of computerization of the EOPs are defined. The rules for organisation of operation with different EOPs and MMIs are designed taking into account requirements set for different task categories and criticality of the actions to be performed by the operator. Priority to the safety MMI is given in case of critical actions. Conventional MMI is used if fast operator response is required. The principles and methods of computerization of the EOPs are consistent with the concept developed for the French EPR and also for the new Chinese plant. This concept takes into account the N4 plant feedback. One of the main targets of the EOP computerizations is to keep the design of operation and support screens close to the present ones.
Flow-Induced Vibration Model for Steam Generator Tubes in Two-Phase Flow
Authors: Mureithi Njuki W. , Shahriary Soroush , Pettigrew Michel J.
Published: July 2009

While steam generators operate in two-phase flow, the complex nature of the flow makes the prediction of flow-induced fluidelastic instability of steam generator tubes a challenging problem yet to be solved. In the work reported here, the quasi-static fluid force-field, which is the important unknown for two-phase flows, is measured in a rotated-triangle tube bundle for a series of void fractions and flow velocities. The forces are shown to be strongly dependent on void fraction, flow rates and relative tube positions. The fluid force field is then employed along with quasi-steady vibration stability models, originally developed for single phase flows, to model the two-phase flow problem and predict the critical instability velocity. The results are compared with dynamic vibration stability tests and are shown to be in good agreement. The present work uncovers some of the complexities of the fluid force field in two-phase flows. The database provides new potential to designers to estimate expected fluid dynamic loads under operating conditions. The force field data may also be applied in dynamic computations for tube wear simulations, replacing the simple Connors’ model which is currently used.
Hydrogen Distribution in a Nuclear Reactor Containment
Authors: Prabhudharwadkar Deoras M. , Iyer Kannan N. , Mohan Nalini , Bajaj S. S. , Markandeya S. G.
Published: July 2009

The management of hydrogen in nuclear reactor containment after LOCA is of practical importance to preserve the structural integrity of the containment. This paper presents the results of systematic work carried out using the commercial software FLUENT to assess the concentration distribution of hydrogen in a typical Indian Nuclear Reactor Containment. Accurate turbulence modelling is important to predict the concentration distribution correctly. The turbulence models which were most commonly cited in the literature for modelling buoyancy driven flows were assessed for their suitability and it was found that the buoyancy modified Standard k-ε model is adequate for the purpose by comparing with some experimental data available in the literature. Subsequently, unstructured meshes were generated to represent the containment of a typical Indian nuclear reactor. Analyses were carried out to quantify the hydrogen distribution for three cases. These were (1) Uniform injection of hydrogen for a given period of time at room temperature, (2) Time varying injection as has been computed from an accident analysis code, (3) Time varying injection (as used in case (2)) at a high temperature. A parametric exercise was also carried out in case (1) where the effect of various inlet orientations and locations on hydrogen distribution was studied. Results of all these cases have been presented in this paper.
3-D Deterministic Burnup With High Performance Computing
Authors: Plower Thomas J. , Manalo Kevin , Rowe Mireille A.
Published: July 2009

Current 3-D reactor burnup simulation codes typically utilize either transport-corrected diffusion theory or Monte Carlo methods to perform flux calculations necessary for fuel depletion. Monte Carlo codes, particularly the Monte Carlo N-particle Transport Code (MCNP) from Los Alamos, have become increasingly popular with the growth of parallel computing. While achieving a criticality eigenvalue is relatively straight forward, run times for large models requiring converged fission sources from proper burnup computation quickly becomes very time consuming. Additionally, past analyses have shown difficulties in source convergence for lattice problems using Monte Carlo [1]. To invoke an alternative means of computing core burnup and decreasing computation time for large models, a deterministic tool such as the PENTRAN/PENBURN suite is necessary. PENTRAN is a multi-group, anisotropic Sn code for 3-D Cartesian geometries; it has been specifically designed for distributed memory, scalable parallel computer architectures using the MPI (Message Passing Interface) library. Automatic domain decomposition among the angular, energy, and spatial variables with an adaptive differencing algorithm and other numerical enhancements make PENTRAN an extremely robust solver with a 0.975 parallel code fraction (based on Amdahl’s law). PENBURN (Parallel Environment BURNup), a recently developed fuel depletion solver, works in conjunction with PENTRAN and performs 3-D zone based fuel burnup using the direct Bateman chain solution method. The aim of this paper is to demonstrate the capabilities and unique features of the PENTRAN/PENBURN suite through a fuel burnup study on a 3 wt% enriched UO2 fuel pin and 17×17 Westinghouse OFA assembly.
Study on Correlation of Orientation and Impact Properties in HTP Line-Pipe Steel
Authors: Zhu Lixia , Wang Yalong , Li Jinfeng , Chen Hongda , He Xiaodong , Han Xinli
Published: July 2009

The correlation of impact properties, orientation and temperature in HTP70 line-pipe steel were studied by use of OM, TEM and macroscope fractography based on series temperature Charpy notch impact test of sample with different orientation. The results showed that: HTP70 steel line-pipe has seriously anisotropic properties, the mechanical behaviors vary with the orientation of parallel, 45° and vertical to rolling direction. The samples which parallel to rolling direction have higher impact energy and demixing crack incline. The demixing crack was caused by banding defect, it was affect by dislocation configuration and has certain direction character, and was related to temperature.
Digital Imaging of an ERW Pipe Forming Process
Authors: Wiskel J. B. , Deng H. , Rieder M. , Ben-Zvi A. , Henein H.
Published: July 2009

A deformation analysis of an ERW forming operation for a 114.3 mm (NPS 4) diameter steel pipe was undertaken using digital imaging technology. Digitals images taken of the inside profile of the pipe were calibrated and converted into Cartesian coordinates and a profile map along length of the forming operation was generated. A numerical curve fitting technique was applied to each individual profile to calculate the circumferential strain around the inside diameter. The circumferential strains calculated using the digital imaging technique corresponded well with strain gauge measurements. The calculated strains were combined into a 3D contour plot of strain along the length of the ERW process.
Evaluating the Effects of Wrinkle Bends on Pipeline Integrity
Authors: Alexander Chris , Kulkarni Satish
Published: July 2009

Concerns exist among the pipeline industry about the effects of wrinkle bends on the long-term integrity of pipelines. For this reason, a study was sponsored to assess the relative severity of wrinkle bends present in the El Paso pipeline system. The study involved a combination of full-scale cyclic pressure fatigue tests, along with finite element analysis, to determine cycles to failure. Strain gages were installed on select samples to determine alternating stresses. Also included in the study was installation of E-glass composite repair materials (Armor Plate® Pipe Wrap) on selected wrinkles to determine the potential for life extension considering the presence of reinforcement. The study helped in developing “in-the-ditch” evaluation criterion and a tool to determine the severity of a specific wrinkle bend based on geometric parameters including wrinkle height and length. The effects of metal loss due to corrosion were also considered. Additionally, the experimental results demonstrated that composite materials can extend the fatigue life of wrinkle bends.
Influence of Geotechnical Loads on Local Buckling Behavior of Buried Pipelines
Authors: Mahdavi Hiva , Kenny Shawn , Phillips Ryan , Popescu Radu
Published: July 2009

Buried pipelines can be subjected to differential ground movement events. The ground displacement field imposes geotechnical loads on the buried pipeline and may initiate pipeline deformation mechanisms that exceed design acceptance criteria with respect to serviceability requirements or ultimate limit states. The conventional engineering approach to define the mechanical performance of pipelines has been based on combined loading events for “in-air” conditions. This methodology is assumed to be overly conservative and ignores soil effects that imposes geotechnical loads and also provides restraint, on buried pipelines. The importance of pipeline/soil interaction and load transfer mechanisms that may affect local buckling of buried pipelines is not well understood. In this study a three-dimensional continuum finite element (FE) model, using the software package ABAQUS/Standard, was developed and calibrated based on large-scale tests on the local buckling of linepipe segments for in-air and buried conditions. The effects of geotechnical boundary conditions on pipeline deformation mechanism and load carrying capacity were examined for a single small diameter pipeline with average diameter to thickness ratio and deep buried condition. The calibrated model successfully reproduced the large-scale buried test results in terms of the local buckling location, pipeline carrying load capacity, soil deformation and soil failure mechanism.
CTOA Measurement of Pipe Steels Using DWTT Specimen
Authors: Xu S. , Tyson W. R.
Published: July 2009

A major issue of applying the crack-tip opening angle (CTOA) approach to fast ductile fracture control is to characterize material toughness (CTOAc) using laboratory-scale specimens and to recommend a mill test. In this report, existing direct and indirect CTOA measurement methods are comprehendsively reviewed. Direct surface measurement methods as standardized in ASTM E2472-06 are mainly based on experience with sheet materials, and the measurement distance from the crack tip recommended in the ASTM standard should be increased for thicknesses typical of pipe steels. Methods using data on load (P) vs. load-line displacement (LLD) from instrumented drop-weight tear test (DWTT) specimens are given special attention, including original and modified two-specimen methods, original and modified single-specimen methods (SSM) and a recent simplified single-specimen method (S-SSM). The results of DWTT/CTOA tests at CANMET-MTL on six pipe steels (five modern high-strength steels and one older X52 steel) are briefly presented. The simplified single-specimen method (S-SSM) is repeatable and may be suitable for a mill test. However, there remains a discrepancy in CTOA values between surface measurements and those from the S-SSM, and it was found that the CTOA decreased somewhat with increasing ligament size from 66 mm to 95 mm. Transferability of results from lab to field will be discussed briefly.
Transporting the Next Generation of CO2 for Carbon, Capture and Storage: The Impact of Impurities on Supercritical CO2 Pipelines
Authors: Seevam Patricia N. , Race Julia M. , Downie Martin J. , Hopkins Phil
Published: July 2009

Climate change has been attributed to greenhouse gases with carbon dioxide (CO2) being the major contributor. Most of these CO2 emissions originate from the burning of fossil fuels (e.g. power plants). Governments and industry worldwide are now proposing to capture CO2 from their power plants and either store it in depleted reservoirs or saline aquifers (‘Carbon Capture and Storage’, CCS), or use it for ‘Enhanced Oil Recovery’ (EOR) in depleting oil and gas fields. The capture of this anthropogenic (man made sources of CO2) CO2 will mitigate global warming, and possibly reduce the impact of climate change. The United States has over 30 years experience with the transportation of carbon dioxide by pipeline, mainly from naturally occurring, relatively pure CO2 sources for onshore EOR. CCS projects differ significantly from this past experience as they will be focusing on anthropogenic sources from major polluters such as fossil fuel power plants, and the necessary CO2 transport infrastructure will involve both long distance onshore and offshore pipelines. Also, the fossil fuel power plants will produce CO2 with varying combinations of impurities depending on the capture technology used. CO2 pipelines have never been designed for these differing conditions; therefore, CCS will introduce a new generation of CO2 for transport. Application of current design procedures to the new generation pipelines is likely to yield an over-designed pipeline facility, with excessive investment and operating cost. In particular, the presence of impurities has a significant impact on the physical properties of the transported CO2 which affects: pipeline design; compressor/pump power; repressurisation distance; pipeline capacity. These impurities could also have implications in the fracture control of the pipeline. All these effects have direct implications for both the technical and economic feasibility of developing a carbon dioxide transport infrastructure onshore and offshore. This paper compares and contrasts the current experience of transporting CO2 onshore with the proposed transport onshore and offshore for CCS. It covers studies on the effect of physical and transport properties (hydraulics) on key technical aspects of pipeline transportation, and the implications for designing and operating a pipeline for CO2 containing impurities. The studies reported in the paper have significant implications for future CO2 transportation, and highlight a number of knowledge gaps that will have to be filled to allow for the efficient and economic design of pipelines for this ‘next’ generation of anthropogenic CO2.
A Geographic Content Management System for Pipeline Projects
Authors: Kneller Geoffrey R.
Published: July 2009

A content management system provides some key capabilities that can transform project management for pipeline surveying and engineering practices. These include methods to share and track the information used by project managers, engineers, project technical staff, and the pipeline owner / operator. The business need for content management is explained, along with a unique insight on the key ingredient that makes content management relevant to pipeline project managers: geographic referencing of project documents. Content management features, such as versioning and checkout, are presented, and a working web-based geographic content management system, developed entirely using free open source software, is demonstrated for a sample pipeline project. The working system integrates project documents, written communications, and CAD data through open standards. Conclusions include lessons learned in the development of the system and its use on real-world projects.
External Corrosion Control of Northern Pipelines
Authors: Papavinasam S. , Doiron A. , Panneerselvam T.
Published: July 2009

Based on one-year laboratory experiments in which samples were exposed to temperatures as low as −45°C and two-year data from field experiments, it is concluded that Canadian Standards Association (CSA) standards CSA Z662, CSA Z245.20 and CSA Z245.21 adequately address evaluation of coatings for northern pipelines. However, in order to evaluate the effects of low temperatures, specimens should be exposed for at least four months. Coatings qualified by CSA Z245.21 (System B1 and B2) are less affected by exposure to low temperatures than those qualified by CSA Z245.21 (System A1) and CSA Z245.20. Corrosion potentials measured at lower temperatures are in the range of potentials to use the cathodic protection (CP) potential criteria of −0.85 V vs. copper-copper sulphate for northern pipelines.
Implementing a Project Management Information System at a Major Pipeline Corporation
Authors: Greaves Thomas H.
Published: July 2009

A project management information system (PMIS) reports project cost and schedule information. Between 2006 and 2007 Pembina Pipeline Corporation’s Engineering Department saw the number of projects it managed leap from 89 to 147, an increase of 65%. Project managers found that their ability to report project information was being strained. This paper chronicles Pembina’s efforts in developing a PMIS to respond to the challenge.
Strain-Based Design Methods for Composite Repair Systems
Authors: Alexander Chris
Published: July 2009

Composite materials are commonly used to repair corroded and mechanically-damaged pipelines. Most of these repairs are made on straight sections of pipe. However, from time to time repairs on complex geometries such as elbows, tees, and field bends are required. Conventional design methods for determining the amount of required composite materials are not conducive for these types of repairs. Over the past several years, the author has developed a methodology for assessing the level of reinforcement provided by composite materials to damaged pipelines using finite element methods. Instead of stress as the design basis metric, the method employs a strain-based design criteria that is ideally-suited for evaluating the level of reinforcement provided to non-standard pipe geometries. The finite element work has been validated using experimental methods that employed strain gages placed beneath the composite repair to quantify the level of reinforcement provided by the repair. This paper provides a detailed description of the strain-based design method along with appropriate design margins for both the reinforced steel and long-term performance of the composite materials.
Optimum Operation of a Gas Export System: Model Validation
Authors: No̸rstebo̸ Vibeke Stærkebye , Paulsen Elisabeth Alne , Bakken Lars Erik
Published: July 2009

The gas transport system on the Norwegian continental shelf (NCS) consists of 7800 km of pipelines and is the largest offshore network of its kind in the world. Natural gas is processed at treatment plants in Norway, and dry gas exported through pipelines to customers in the UK and continental Europe. These customers have the opportunity to make varying gas delivery nominations, and meeting such sales gas commitments is important. A major challenge is to operate the network at minimum cost, with minimum environmental emissions, fulfilling variations in contractual nominations and maintaining a sufficient pipeline inventory to provide operational flexibility. A model of optimum operation of a gas export system, including customer nominations, system integration, and operation of pipelines and compressor stations has been established in earlier studies. This model is based on analyses of actual operational data from the gas export system on the NCS. The scope of the work presented in this paper has been to validate this model so that it can be implemented in planning and operation of the gas export system. The purpose of the validation process is first to confirm that the model results are correct in that they represent actual system performance. Second, it aims to confirm that the model results satisfy all system requirements and the main objective of the work. The process includes developing and executing procedures, documenting validation results and considering possible modifications to the model.