bluuebunny/crossref_metadata_2025 huggingface dataset:
Heat Transfer From Novel Target Surface Structures to a Normally-Impinging, Submerged and Confined Water Jet
Authors: Jeffers N. , Punch J. , Walsh E.
Published: February 2010

Contemporary electronic systems generate high component-level heat fluxes. Impingement cooling is an effective way to induce high heat transfer coefficients in order to meet thermal constraints. The objective of this paper was to experimentally investigate the heat transfer from five novel target surface structures to a normally-impinging, submerged and confined water jet. The five target structures were: a 90° vane; a 4×4 pin fin array; and three geometries which turn the flow away from, and back towards, the surface to be cooled. The experiments were conducted for Reynolds numbers of 500 ≤ Re ≤ 24 000. The confined impinging jet was geometrically constrained to a round, 8.5mm diameter, square-edged nozzle at a jet exit-to-target surface spacing, of H/D = 0.5. The heat transfer characteristics of the five novel target surfaces were non-dimensionally compared to a flat surface, and enhancements of up to 120% were recorded. Increases of up to 45% was noted when the surface area augmentation of the target surface structures was factored out. The findings of the paper are of practical relevance to the design of primary heat exchangers for high-flux thermal management applications.
Practical Application of a Compact, High-Effectiveness, Gas-To-Gas, Compound Recuperator With Liquid Intermediary (CRLI)
Authors: Doty F. David , Shevgoor Siddarth , Staab John
Published: February 2010

Gas-to-gas recuperation with both high thermal effectiveness and order-of-magnitude improvement in cost effectiveness is critical to addressing global energy needs, especially via fuel synthesis from waste CO2 and renewable H2. An uncommon compound recuperator with liquid intermediary (CRLI) has been simulated for high-ε heat exchange between a first shell-side gas stream and a second shell-side gas stream of similar thermal capacity rates (W/K). The compound recuperator uses a first Gas-to-Liquid (GL) recuperator for a nearly complete transfer of available energy from a shell-side gas to an intermediary tube-side heat transfer liquid (HTL), followed by a second GL recuperator to transfer the heat from the liquid intermediary to the second gas stream. Each GL recuperator resembles an arrangement of thermally isolated, serially connected, adjacent, cross-flow, finned-tube cores, such as used in AC condensers. They are arranged so to effectively achieve counterflow exchange between the HTL and the shell-side gas stream. The HTL may be water, an organic liquid, a molten alloy, or a molten salt. Minimization of exergy destruction for the case where there is a substantial temperature difference between the hot and cold sources requires (1) a fairly large number of series connected, thermally isolated cores, (2) similar thermal capacity rates in all three streams (the two gases and the liquid intermediary), (3) a relatively large value for the number of transfer units (NTU), and (4) no phase change. The simulations show that the optimized CRLI recuperator can achieve effectiveness above 97% at very low pumping losses and has the potential for order-of-magnitude reduction in manufacturing costs compared to current technologies for clean gases at pressures above 0.3 MPa at heat transfer rates above ∼200 kW.
A Novel Methodology for Combined Parameter and Function Estimation Problems
Authors: Hakkaki-Fard Ali , Molavi Hosein , Rahmani Ramin K.
Published: February 2010

This article presents a novel methodology, which is highly efficient and simple to implement, for simultaneous retrieval of a complete set of thermal coefficients in combined parameter and function estimation problems. Moreover, the effect of correlated parameters on convergence performance is examined. The present methodology is a combination of two different methods: The Conjugate Gradient Method with Adjoint Problem (CGMAP) and Box-Kanemasu method (BKM). The methodology uses the benefit of CGMAP in handling function estimation problems and BKM for parameter estimation problems. One of the unique features about the present method is that the correlation among the separate unknowns does not behave as a limiting factor to the convergence of the problem. Numerical experiments using measurement errors are performed to verify the proposed method in solving the combined parameter and function estimation problems. The obtained results show that the combined procedure can efficiently and reliably estimate the values of the thermal coefficients.
Steady-State Segmented Thermofluid Network Simulations of a Loop Heat Pipe Operating With Four Different Working Fluids
Authors: Jesuthasan Nirmalakanth , Baliga B. Rabi
Published: February 2010

A segmented network thermofluid model for the simulation of a loop heat pipe (LHP) operating under steady-state conditions is presented, with special emphasis on quasi one-dimensional models and semi-empirical correlations for the related multiphase phenomena. Attention is focused on an LHP with one flat-evaporator, a vapor transport line, one condenser, a liquid transport line, and a compensation chamber. The evaporator consists of the following parts: An upper piece, machined out of a stainless steel plate, with vapor-transport grooves of rectangular cross section on its bottom face; a lower piece, also machined out of a stainless steel plate, with a cavity of rectangular cross section that serves as the liquid pool in the evaporator during the operation of the LHP; and a rectangular wick sandwiched between the upper and lower pieces. The wick is a sintered powder metal plate made of stainless steel. The condenser is a horizontal tube that is fitted with excellent thermal contact inside a large high thermal conductivity metallic sleeve that is maintained at a fixed sink temperature. The vapor-transport line, the condenser, and the liquid-transport line are divided into control volumes or cells. Quasi one-dimensional models are used to impose balances of mass, momentum, and energy on each of these cells. The variation of fluid thermophysical properties and multiphase phenomena, such as the change in quality and pressure drops in the two-phase regions, are suitably accounted for in this model. Four different working fluids, ammonia, distilled water, ethanol and isopropanol, are considered, and the results obtained for a representative range of steady-state operating conditions of the LHP are presented and comparatively discussed.
Water Spray Cooling of High-Temperature Steam Flow Through a Steam Turbine Bypass Valve Line
Authors: Amano R. S.
Published: February 2010

The objective of the present study is to investigate the steam flow behavior through the high-pressure turbine bypass valve. Efforts have mainly been directed at investigating the process of steam flow and property variations aforementioned bypass valve as well as to obtain correlations between the flow rate and the valve opening ratio. Modeling of the high-pressure turbulent steam flow was performed on a three-dimensional non-staggered (co-located) grid system by employing the finite volume method and by solving the three-dimensional, turbulent, compressible Navier-Stokes, and energy equations. Through this research, numerous data have been acquired and analyzed. These efforts enable us to obtain a correlation data set for the flow rate coefficient as a function of valve opening. One of the significant accomplishments is to use the model presented here for further improve a design of a turbine bypass flow valve.
Effect of PWHT Cycles on Impact Toughness of Austenitic Stainless Steel Weldments
Authors: Ibrahim M. M. , Mohamed H. G. , Tawfik Y. E.
Published: February 2010

Austenitic stainless steels are widely used welding materials in nuclear reactors and power plants because of their high strength, good ductility, excellent corrosion resistance and a reasonable weldability. These properties make austenitic stainless steels attractive candidate materials for use in the fabrication of piping systems, automotive exhaust gas systems and in a variety of equipment associated with the chemical and nuclear power industries. PWHT is a stress relieving process whereby residual stresses are reduced by typically heating to 550–650 °C for a set time depending upon plate thickness. It concerns have emerged about possible effects on the mechanical properties of the base (parent) and weld plates (PM and WM). The 6 mm AISI 304L, 316L, and 347 austenitic stainless steels were used for this work. These welds were produced by SMAW and GTAW techniques using a single vee preparation and multiple weld beads, and welded by various types of consumables. The fracture surfaces of the Charpy V-notch PM and WM (before and after PWHT) samples were examined by SEM. Scanning electron fractographs was critical in this study, in that valuable information regarding the mechanism and nature of failure could be determined. This paper reports work on the impact toughness of the three types of austenitic stainless steels. The parent and weld regions were examined for all types of steels used, and then exposed to temperature in the PWHT range. The effect of exposure to multiple PWHT cycles on these properties is discussed. A decrease in impact energy and fracture toughness with an increase in the number of heat treatments was evident in the parent metal. Similary, the weld metal showed a decrease in impact energy after two PWHT cycles.
Modeling of Sodium Two-Phase Flow With the TRACE Code
Authors: Chenu Aurelia , Mikityuk Konstantin , Chawla Rakesh
Published: February 2010

In the framework of PSI’s FAST code system, the TRACE thermal-hydraulics code is being extended for representation of sodium two-phase flow. As the currently available version (v.5) is limited to the simulation of only single-phase sodium flow, its applicability range is not enough to study the behavior of a Sodium-cooled Fast Reactor (SFR) during a transient in which boiling is anticipated. The work reported here concerns the extension of the two-fluid models, which are available in TRACE for steam-water, to sodium two-phase flow simulation. The conventional correlations for ordinary gas-liquid flows are used as basis, with optional correlations specific to liquid metal when necessary. A number of new models for representation of the constitutive equations specific to sodium, with a particular emphasis on the interfacial transfer mechanisms, have been implemented and compared with the original closure models. As a first application, the extended TRACE code has been used to model experiments that simulate a loss-of-flow (LOF) accident in a SFR. The comparison of the computed results, with both the experimental data and SIMMER-III code predictions, has enabled validation of the capability of the modified TRACE code to predict sodium boiling onset, flow regimes, dryout, flow reversal, etc. The performed study is a first-of-a-kind application of the TRACE code to two-phase sodium flow. Other integral experiments are planned to be simulated to further develop and validate the two-phase sodium flow methodology.
Development and Validation of a Computational Fluid Dynamics Model for the Simulation of Two-Phase Flow Phenomena in a Boiling Water Reactor Fuel Assembly
Authors: Tentner Adrian , Pointer W. David , Lo Simon , Splawski Andrew
Published: February 2010

This paper presents the current status in the development and validation of an advanced Computational Fluid Dynamics (CFD) model, CFD-BWR, which allows the detailed analysis of the two-phase flow and heat transfer phenomena in Boiling Water Reactor (BWR) fuel assemblies under various operating conditions. The CFD-BWR model uses an Eulerian Two-Phase (E2P) approach, and is also referred to as the E2P modeling framework. It is being developed as a customized module built on the foundation of the commercial CFD-code STAR-CD which provides general two-phase flow modeling capabilities. The integral validation efforts have focused on the analysis of the NUPEC Full-Size Boiling Water Reactor Test (BFBT) within the framework of the OECD/NRC benchmark exercise. The paper reviews the two-phase models implemented in the CFD-BWR code, and emphasizes recently implemented models of inter-phase and coolant-cladding momentum and energy exchanges. Results of recent BFBT experiment simulations using these models are presented and the effects of the new models on the calculated void distribution are discussed. The paper concludes with a discussion of future model development and validation plans.
Maanshan PWR Loss of Flow Transients Analysis With TRACE
Authors: Wang Jong-Rong , Lin Hao-Tzu , Wang Wei-Chen , Cheng Yi-Hsiang , Shih Chunkuan
Published: February 2010

TRACE model of Maanshan Nuclear Power Plant (three-loop PWR) was used to analyze Loss of Flow transient as defined in FSAR Chapter 15. The results were compared with those from RETRAN02 and LOFTRAN/THINC licensing analysis of Westinghouse Inc. Three different initiation events were involved in this analysis: Partial Loss of Flow (PLOF), Complete Loss of Flow-Under Voltage (CLOF-UV) and Complete Loss of Flow-Under Frequency (CLOF-UF). This paper compared important thermal hydraulic parameters at steady state, such as the pressure of pressurizer, cold-leg temperature, and the pressure of steam generator, etc.. It also compared system parameters under transient conditions, such as core thermal power, core flow rate, and pressure of pressurizer, etc.. It is concluded that the steady state results of TRACE calculations are in general good agreements with those from RETRAN02 and have a largest error of 3.03% in the steam generator flow. For transient condition, TRACE results are also comparable with those from LOFTRAN and RETRAN02. In summary, our studies show that Maanshan TRACE model is correct and accurate enough for future safety analysis applications.
Automated Operation With Redundant Ultrasonic Reactor Water Level Monitoring System
Authors: Askari Vahid
Published: February 2010

The proposed ultrasonic Reactor water Level Monitoring system is intended for monitoring water level in the hot leg of a nuclear reactor during Operation at Reduced Inventory (ORI) with the core in place as may be needed for certain maintenance activities. Since the existing fuel bundles continue the production of heat thorough the decay process, it is critical to maintain a certain water flow rate through the reactor. The Residual Heat Removal System (RHRS) handles this task during the reactor shutdown. The free water surface during ORI subjects the RHRS to the risk of air intake. Such an event would trigger a set of emergency procedures for priming the RHRS and bringing it back online. Some of the possible outcomes of such an event are: extension of outage, handling of the extra regulatory reporting tasks, potential requirements for extra testing after system recovery, possible damage to RHR pump, boiling of reactor water, generation of airborne contamination, and fuel damage. Another risk during ORI is reactor overflow. The major immediate concern is the safety of personnel in case of accidental overflow. This event could result in contamination and the related outage delays. The previously existing systems have always had problems with transducer reliability. This has lead to the need for removal and re-installation of their transducers from outage to outage. However, the proposed system uses a redundant design with permanently mounted transducers. The system uses advanced digital signal processing techniques to determine the water level.
Defense-in-Depth Approach for Use in a Risk-Informed, Technology-Neutral Design and Licensing Framework for New Nuclear Plants
Authors: Chapman Jim , Hess Stephen M.
Published: February 2010

The regulatory framework for the current generation of operating plants and advanced light water reactors (ALWRs) planned for near term construction has evolved over several decades to permit effective regulation of the light water reactor designs. To address other reactor types, development of a framework that possesses the attributes of being technology neutral, risk-informed and performance-based with corresponding processes (regulations and guidance) is ongoing by several U.S. and international organizations. A key design and operating principle which is applied to existing plants and will continue to be applied to future plants is defense-in-depth. The advent of advanced reactor designs, some of which are not based on light water reactor technology, provides incentive for changes in the regulatory framework in several areas, including defense-in-depth practices. To support development of an integrated framework, the Electric Power Research Institute (EPRI) conducted research to identify and assess specific elements of possible technology neutral, risk-informed, performance based frameworks that had been proposed by others. The intent was to develop a preliminary framework based on the results of this review and evaluation and to provide recommendations in areas where additional development and testing would appear to be most beneficial. “Technical Elements of a Risk-Informed, Technology-Neutral Design and Licensing Framework for New Nuclear Plants”, EPRI Report 1016150 documents this research (Reference [1]). For defense-in-depth (D-in-D) existing viewpoints from various sources were compared and an alternative integrated approach which addresses key issues was developed. These alternative views are contained in publications such as NUREG-1860 [2], Regulatory Guide 1.174 [6], IAEA Safety Standards Series No. NS-R-1 [3], IAEA 75-INSAG-3 Revision 1 [4], INSAG-12 [4], and IAEA INSAG-10 [5]. The results of this research support the ongoing efforts to develop standards and guidance for advanced plants with safety characteristics which differ from existing and advanced LWRs.
US-APWR Human Systems Interface System Verification and Validation Results: Impact on Digital I&C Design
Authors: Hall Robert E. , Easter James , Roth Emilie , Kaban Leonard , Mashio Kenji , Yugami Kunio , Takahashi Koichi , Clouser Timothy
Published: February 2010

The US-APWR, currently under Design Certification Review by the US Nuclear Regulatory Commission, is a four-loop evolutionary pressurized water reactor with a four-train active safety system by Mitsubishi Heavy Industries. The Instrumentation and Control (I&C) System and Human Systems Interface (HSI) platform applied to the US-APWR is provided by Mitsubishi Electric Corporation. This design is currently being applied to the latest Japanese PWR plant under construction and to the nuclear power plant I&C modernization program in Japan. The US-APWR’s fully digital I&C system and HSI platform utilizes computerized systems, including computer based procedures and alarm prioritization, relying principally on an HSI system with soft controls, console based video display units and a large, heads up, overview display panel. Conventional hard controls are limited to system level manual actions and a Diverse Actuation System. The overall design philosophy of the US-APWR is based on the concept that operator performance will be enhanced through the integration of safety and non-safety display and control systems in a robust digital environment. This philosophy is augmented, for diversity, by the application of independent safety-only soft displays and controls. As with all the advanced designs, the digital systems open as many questions as they answer. To address these new questions, for an eight-week period during the months of July and August 2008, an extensive verification and validation program was completed with the objective of assessing US operators’ performance in this digital design environment. Over this time period, eight operating crews were subjected to a four-day exercise in which seven scenarios were run on Mitsubishi’s dynamic simulator. In all, twenty-two US-licensed operators and senior operators took part in the program. Additionally, a PC verification software package was developed to independently evaluate video display screen designs. Subjective and objective data were collected on each crew for each scenario and an extensive convergent measures analysis was performed, resulting in the identification of both specific design as well as generic conclusions. This paper discusses the USAPWR design, the V&V program data collection and analysis, and the study results related to the ongoing discussion of the impacts of digital systems on human performance, such as workload, navigation, situation awareness, operator training and licensing.
Thermal-Hydraulic Evaluation of Joyo Fuel Subassembly With Local Blockage
Authors: Ohira Hiroaki , Takamatsu Misao
Published: February 2010

In the incident of the experimental fast reactor Joyo in June 2007, little amount of metal powder was estimated to be produced by the contact between the bottom of the upper core structure and the test section of the material testing rig with temperature control. Up to now, no foreign materials were detected by the in-vessel observations. However, a small amount of metal powder which could be placed into the fuel pin bundles was assumed in the present study. Preliminary safety evaluation of the wire-wrapped fuel pin bundles of Joyo with local blockage were performed in the rated power operational condition, an anticipated transient during operation and an accident. A single-phase transient subchannel analysis code ASFRE, which was verified by various local blockage experiments in France and in Japan, was applied to the present evaluations. From these results, it was concluded a small amount of metal powder assumed to be produced by the incident would not affect the safety operations in rated power conditions, in anticipated transients and in accidents.
Axial Development of Local Flow Parameters in Bubbly Two-Phase Flow in Normal and Microgravity Conditions
Authors: Takata Yutaka , Xing Dong Chang , Fukuhara Yutaka , Hazuku Tatsuya , Takamasa Tomoji , Hibiki Takashi
Published: February 2010

In relation to the development of the interfacial area transport equation, a precise database of the axial development of void fraction profile, interfacial area concentration and Sauter mean bubble diameter in an adiabatic nitrogen-water bubbly flow in a 9 mm-diameter pipe was constructed for normal and microgravity conditions using stereo image-processing. The flow measurements were performed at four axial locations (axial distance from the inlet normalized by the pipe diameter, z/D = 5, 20, 40 and 60) and with various flows: superficial gas velocity of 0.00840–0.0298 m/s, and superficial liquid velocity of 0.138–0.914 m/s. The effect of gravity on radial distribution of bubbles and the axial development of two-phase flow parameters is discussed in detail based on the obtained database and visual observation.
Effect of Gravity on Axial Development of Phase Distribution Patterns in Bubbly Two-Phase Flow
Authors: Nishioji Takashi , Kato Taichi , Fukuhara Yutaka , Hazuku Tatsuya , Takamasa Tomoji , Hibiki Takashi
Published: February 2010

The axial development of the void fraction profile, interfacial area concentration and Sauter mean bubble diameter of adiabatic nitrogen-water bubbly flows in a 9 mm-diameter pipe were measured using stereo image processing in normal and microgravity conditions. The effect of gravity and flow conditions on the radial distribution of bubbles and the axial development of the two-phase flow parameter is discussed in detail based on the obtained data. By taking into account normalized parameters based on void peak fraction and void peak intensity in the pipe cross-section, the phase distribution patterns were classified into three types: a wall peak, a core peak and an intermediate peak. Phase distribution pattern maps are presented for vertical upward bubbly flows in normal and microgravity conditions.
The Hot Function Flow Characteristics of a Main Cooling Water System for the Nuclear Fuel Test Loop Installed in HANARO
Authors: Park Young-Chul , Lee Young-Seob , Chi Dae-Young
Published: February 2010

A nuclear fuel test loop (after below, FTL) is installed in IR1 of an irradiation hole in HANARO for testing neutron irradiation characteristics and thermo hydraulic characteristics of a fuel loaded in a light water power reactor or a heavy water power reactor. There is an in-pile section (IPS) and an out-pile section (OPS) in this test loop. When HANARO is normally operated, the fuel loaded in the IPS has a nuclear reaction heat generated by a neutron irradiation. To remove the generated heat and to maintain an operating conditions of the test fuel, a main cooling water system (MCWS) is installed in the OPS of the FTL. The MCWS is composed of a main cooler, a pressurizer, two circulation pumps, a main heater, interconnection pipe line and instruments. The interconnection pipeline is a closed loop which is connected respectively to an inlet and an outlet of the IPS. To absorb the nuclear reaction heat, there is a higher elevation pipeline with a reverse U letter type pipeline installed in upstream of the cooler. And there is another higher elevation pipeline for sucking a fluid by a vertically installed circulation pump with a top suction and a side discharge. Therefore, we predicted that it is difficult to continuously suck a fluid and pressurize the fluid due to air pockets in the two higher elevation pipelines under a high temperature operation. To verify the hot function flow characteristics of the MCWS including air pockets of the higher elevation pipelines, a flow net work analysis has been conducted under a high temperature operation. When the two higher elevation pipelines wholly filled with coolant, it was confirmed through the results that the pump pressurizes the coolant normally. And it was confirmed through the analysis results that the system hot function characteristics met the system design requirements.
Numerical Flow Simulation of Spacer Grids With Mixing Vanes in a 5 × 5 PWR Rod Bundle
Authors: Navarro Moyse´s Alberto , Campagnole dos Santos Andre´ Augusto
Published: February 2010

The spacer grids exert great influence on the thermal hydraulic performance of the PWR fuel assembly. The presence of the spacers has two antagonistic effects on the core: an increase of pressure drop due to constriction on the coolant flow area and increase of the local heat transfer downstream the grids caused by enhanced coolant mixing. The mixing vanes, present in most of the spacer grid designs, cause a cross and swirl flow between and in the subchannels, enhancing even more the local heat transfer at the cost of more pressure loss. Due to this important hydrodynamic feature the spacer grids are often improved aiming to obtain an optimal commitment between pressure drop and enhanced heat transfer. In the present work, the fluid dynamic performance downstream a 5 × 5 rod bundle with spacer grids is analyzed with a commercial CFD code (CFX 11.0). Eleven different split vane spacer grids with angles from 16° to 36° and a spacer without vanes were evaluated. The computational domain extends from ∼10 Dh upstream to ∼50 Dh downstream the spacer grids. The standard k-ε turbulence model with scalable wall functions and the total energy model were used in the simulations. The results show a considerable increase of the average Nusselt number and secondary mixing with the angle of the vane up to ∼20 Dh downstream the spacer, reducing greatly the influence of the vane angle beyond this region. As expected, the pressure loss through the spacer grid also showed considerable increase with the vane angle.
Aging Evaluation for the Extension of Qualified Life of Nuclear Power Plant Equipment Through Modulated Poisson Processes
Authors: Saldanha Pedro L. C. , Frutuoso e Melo P. F.
Published: February 2010

This paper presents an application of the Modulated Point Process (MPPP) to the service water pumps of a typical pressurized water reactor, as a model for the rate of occurrence of failures (ROCOF) of a repairable system, in order to decide for an extension of the qualified life in the context of a nuclear power plant license renewal. The analysis was carried considering some field data spanning a 2,300 calendar day period, which is approximately equivalent to four burn-up cycles and refueling periods. The reliability is estimated, and maintenance strategies are discussed. As a conclusion, the MPPP is adequate for modeling the rate of occurrence of failures that are time dependent, and can be used where aging mechanisms are present in the operation of repairable systems. This means that equipment characteristics that are important may be inserted into the model and the results can help make decisions in the context of maintenance programs, as is the case with the maintenance rule concept that has been proposed by the Nuclear Regulatory Commission and is about to be implemented in nuclear power plants in Brazil, for instance. Trade-offs on the difficulty with data acquisition for applying point processes are discussed.
Selection of Metal Electrodes for and Effect of Welding Treatment on Hardness and Microstructure of Some Type of Austenitic Stainless Steels
Authors: Ibrahim M. M. , Mohamed H. G. , Tawfik Y. E.
Published: February 2010

Austenitic stainless steels have been the focus of considerable research recently because of their high strength, good ductility, excellent corrosion resistance and a reasonable weldability. These properties make austenitic stainless steels attractive candidate materials for use in the fabrication of piping systems, automotive exhaust gas systems and in a variety of equipment associated with the chemical and nuclear power industries. PWHT is a stress relieving process whereby residual stresses are reduced by typically heating to 550–650 °C for a set time depending upon plate thickness. The effect of PWHT on mechanical properties such as hardness, ultimate tensile strength, yield strength, impact energy and ductile to brittle transition temperature are of great concern to the pressure vessel industry and pressure vessel codes. This paper reports on the effect of multiple PWHT on hardness and microstructure of austenitic stainless steels. The 6 mm AISI 304L, 316L, and 347 austenitic stainless steels were used for this work. This welds were produced by SMAW and GTAW techniques using a single vee preparation and multiple weld beads, and welded by various types of consumables. Selection of a suitable consumables metals for joining those weldment sample joints are an important criterion in view of the differences in physical, chemical, and mechanical properties of the base materials involved.
Transport of High-Temperature Molten Salt Slurry for Pyro-Reprocessing
Authors: Hijikata Takatoshi , Koyama Tadafumi
Published: February 2010

Pyro-reprocessing is one of the most promising technologies for advanced fuel cycle with favorable economic potential and intrinsic proliferation resistance. The development of transport technology for molten salt is a key issue in the industrialization of pyro-reprocessing. As for pure molten LiCl-KCl eutectic salt at approximately 773 K, we have already reported the successful results of transport using gravity and a centrifugal pump. However, molten salt in an electrorefiner mixes with insoluble fines when spent fuel is dissolved in porous anode basket. The insoluble consists of noble metal fission products, such as Pd, Ru, Mo, and Zr. There have been very few transport studies of a molten salt slurry (metal fines - molten salt mixture). Hence, transport experiments on a molten salt slurry were carried out to investigate the behavior of the slurry in a tube. The apparatus used in the transport experiments on a molten salt slurry consisted of a supply tank, a 10° inclined transport tube (10 mm inner diameter), a valve, a filter, and a recovery tank. Stainless steel (SS) fines with diameters from 53 to 415 μm were used. To disperse these fines homogenously, the molten salt and fines were stirred in the supply tank by an impeller at speeds from 1200 to 2100 rpm. The molten salt slurry containing 0.2 to 0.4 vol.% SS fines was transported from the supply tank to the recovery tank through the transportation tube. In the recovery tank, the fines were separated from the molten salt by the filter to measure the transport behavior of molten salt and SS fines. When the velocity of the slurry was 0.02 m/s, only 1% of the fines were transported to the recovery tank. On the other hand, most of the fines were transported when the velocity of the slurry was more than 0.6 m/s. Consequently, the molten salt slurry can be transported when the velocity is more than 0.6 m/s.
A Once-Through Fuel Cycle for Fast Reactors
Authors: Weaver Kevan D. , Gilleland John , Ahlfeld Charles , Whitmer Charles , Zimmerman George
Published: February 2010

A paradigm shift has recently altered the design targets for advanced nuclear energy systems that use a fast neutron spectrum. A previous emphasis on extending fissile fuel reserves has been supplanted by a desire for reactor technologies that are “cleaner, more efficient, less waste-intensive, and more proliferation-resistant.” [1] This shift, along with recent advances in fast-reactor designs that enable high fuel burn-up even with fuels that have been minimally enriched, creates an opportunity to employ fast reactors in an open nuclear fuel cycle. These goals now appear feasible as a result of recent design work exploiting a phenomenon, known as a traveling wave, that can attain high burn-ups without reprocessing. A traveling-wave reactor (TWR) breeds and uses its own fuel in place as it operates. Fueled almost entirely by depleted or natural uranium, such reactors would also require little initial enrichment. We have performed calculations demonstrating that TWRs can achieve burn-ups of ≥20%, which is four to five times that realized in current LWRs. Burn-ups of up to 50% appear feasible. The factors that contribute to these high burn-ups and the implications for materials design will be discussed.
The Role of Local Governments in the Japanese Nuclear Governance: Case Analysis of “Safety Agreements” in Siting Areas
Authors: Sugawara Shin-etsu , Kimura Hiroshi
Published: February 2010

Non-institutional measures often play an important role in the governance between technology and institutional system which have legal basis. In this study, we analyze the role of the “safety agreements” in Japan’s current regulation system of nuclear power plants (NPPs). In Japan, the national government regulates exclusively the safety of nuclear facilities based on the regulaions. But local governments also involve in the nuclear safety regulation de facto, to protect the publics’ safety and welfare. All the local governments in the siting areas concluded “safety agreements”, a gentleman-like agreement between local governments and power companies on safety issues. These Agreements contain various provisions, such as, the immediate notification procedures of accidents in NPPs, the access and inspection of NPPs after accidents, etc... However, these provisions don’t have a legal-binding of the power companies but, the local governors and local governments sometimes strongly assert the control for the restart operation of NPPs after the accidents utilizing the “safety agreements” as the pretext. This situation has sometimes been criticized by some experts who argue that the “safety agreements” have become an obstacle for stable NPPs operation and electricity supply. Since there are no studies which analyze the actual situations of local governments’ control on nuclear safety based on the “safety agreements,” we analyzed the situation by performing a qualitative survey and conducting a series of interview with personnel of the power companies and local governments. We found from our analysis that the “safety agreements” were utilized positively. For example, the inspection of NPPs by local government functions as an endorsement for the national regulation and assures public confidence. In this case the “safety agreements” promote and provide public acceptance on nuclear governance. However, the measures in these “safety agreements” could be over-extended in the political contexts because of the ambiguity and flexibility in the interpretation. In our paper, we would illustrate the merits and demerits of the role of local governments based on these “safety agreements” and suggest to improve the nuclear safety governance within the local governments’ capacity.
Extraction of Technetium as [Tc(II)(NO)(AHA)2H2O]+ Species in the UREX Process
Authors: Paviet-Hartmann Patricia , Nunez Gomez-Aleixandre Ana , Pak Joshua , Glez Espartero Amparo , Poineau Frederic , Wright Amber , Mausolf Edward , Czerwinski Kenneth R.
Published: February 2010

As it is envisioned today, the first segment of the UREX+ process uses low nitric acid concentrations for U(VI) extraction where pertechnetate anion, TcO4−, can be co-extracted with the uranyl and nitrate into TBP-hydrocarbon solutions. A reductant complexant, acetohydroxamic acid (AHA) is added to the process through the scrub to limit the extractability of plutonium and neptunium. Recent work performed in our laboratory (Ref. 1) demonstrated that TcO4− undergoes reductive nitrosylation by AHA under a variety of conditions. The resulting divalent technetium is complexed by AHA to form the pseudo-octahedral trans-aquonitrosyl-(diacetohydroxamic)-technetium(II) complex ([Tc(II)(NO)(AHA)2H2O]+). In this paper, we are reporting the extraction of [Tc(II)(NO)(AHA)2H2O]+ complex by new designed macrocompounds as well as commercially available crown ethers from 18-crown-6 to 24-crown-8 in ring size and of varying derivatization. Several organic diluents with different dielectric constants are used to enhance the distribution coefficient of technetium (II). The experimental efforts are focused on determining the best extraction conditions by varying the macrocomponds nature and concentration, and the organic phase composition.
Next Generation Safety Analysis Methods for SFRs—(6) SCARABEE BE+3 Analysis With SIMMER-III and COMPASS Codes Featuring Duct-Wall Failure
Authors: Uehara Yasushi , Shirakawa Noriyuki , Naitoh Masanori , Okada Hidetoshi , Yamano Hidemasa , Tobita Yoshiharu , Yamamoto Yuichi , Koshizuka Seiichi
Published: February 2010

Governing key phenomena in core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs) are supposed to be (1) fuel pin failure and disruption, (2) molten pool boiling, (3) melt freezing and blockage formation, (4) duct wall failure, (5) low-energy disruptive core motion, (6) debris-bed coolability, and (7) metal-fuel pin failure with eutectics between fuel and steel [1]. Although the systematic assessment program for SIMMER-III [4–7] has provided a technological basis that SIMMER-III is practically applicable to integral reactor safety analyses, further model development and validation efforts should be made to make future reactor calculations more reliable and rational. For mechanistic model development, a mesoscopic approach with the COMPASS code [1, 2, 3] is expected to advance the understanding of these key phenomena during event progression in CDAs. The COMPASS code has been developed since FY2005 (Japanese Fiscal Year, hereafter) to play a complementary role to SIMMER-III. In this paper, the overall analysis of SCARABEE-BE+3 test with the SIMMER-III and those with COMPASS, focusing the duct wall failure in a small temporal and spatial window cut from the SIMMER-III analysis results of the test, are described.
Development of Alternative Reductant Application in PWR Primary Systems
Authors: Domae Masafumi , Kawamura Hirotaka , Ohira Taku
Published: February 2010

In primary coolant of pressurized water reactors (PWRs), high concentration dissolved hydrogen (DH) has been added, to prevent generation of oxidizing species through radiolysis of coolant. Recently, number of ageing plants is increasing and utilities are concerned about primary water stress corrosion cracking (PWSCC). Although mechanism of PWSCC is not fully clarified, some researchers consider that occurrence of PWSCC and crack propagation rate are affected by DH concentration. The authors consider that one of possible mitigation methods toward PWSCC is use of alternative reductant for hydrogen. From the radiation chemical aspect, aliphatic alcohols are typical scavengers of oxidizing radical generated through the radiolysis of water. The aliphatic alcohols are promising candidates of the alternative reductant. In the present work, possible alternatives of hydrogen were screened, and methanol was selected as the best candidate. Corrosion tests of type 304 stainless steels were carried out at 320°C in two conditions: (1) DH 1.5 ppm (part per million) and (2) methanol 2.9 ppm. Under two conditions, electrochemical corrosion potential was measured during the immersion tests. After the immersion tests, surface morphology of the stainless steel specimens was observed by scanning probe microscope. Major component of oxide film was analyzed by X-ray diffraction. From comparison of the test results, it is concluded that addition of methanol 2.9 ppm has almost the same effect as addition of DH 1.5 ppm.
Remote-Controlled Inspection Robot for Nuclear Facilities in Underwater Environment
Authors: Okihana Harunobu , Iwata Keiji , Miwa Yasuhiro
Published: February 2010

A remote-controlled inspection robot for nuclear facilities in underwater environment was developed. The Array system of Eddy Current Testing (Array-ECT) is installed on the existing robot that consists of the driving unit and the inspection unit with enlarged Charge-Coupled Device (CCD) camera. It can swim in the tank, crawl on the inner surface of the tank, stay on the inspection area and by its two devices, Array-ECT and enlarged CCD camera, start inspection. Array-ECT can record the color contour image, and also can display oscillograph. This technology had been developed in order to reduce worker’s radiation exposure and improve inspection-ability in nuclear power plants for some Radioactive Waste (RW) tanks. With conventional manual inspection, RW tanks must be examined with the sequence of (a) full drainage and decontamination, (b) scaffolding set-up with workers under high dose, and (c) manual inspection. As such, the inspection is time consuming and workers should have high radiation exposure. With the original robot that has only enlarged CCD camera, inspection can be performed without tank drainage and with minimum worker’s radiation exposure. However, it needs experienced inspector for interpretation of visual examination using enlarged CCD camera. Newly developed robot with Array-ECT and enlarged CCD camera can achieve the same performance using the original robot and an experienced inspector for visual inspection using enlarged CCD camera. The new feature added to the original design (i.e. Array-ECT) eliminates the need for an experienced inspector.
Heat Transfer Assessment of Calandria Tube Enhancement Design Options
Authors: Hobbs Daniel M. , Freeburn Allan M. , Scrannage Richard C. , Sani Farshad , Harvel Glenn D.
Published: February 2010

The moderator in the Calandria of a CANada Deuterium Uranium (CANDU) reactor serves as a backup heat sink in a Loss of Coolant Accident (LOCA) combined with a Loss of Emergency Coolant Injection (LOECI) accident scenario. This requires that the heat from the fuel be transferred through the ballooned/sagged Pressure Tube and the contacted Calandria Tube (CT) to the moderator. Increasing the heat transfer to the moderator will enhance the safety of the reactor by increasing the margin to Calandria Tube Rupture and ensuring the fuel is cooled. Conceptual designs were assessed to determine their potential for increasing heat transfer to the moderator in accident scenarios. Three different design options for a CT have been investigated using numerical simulation techniques. The CT design options include fins or ribs of different sizes and combinations. The fins and ribs provide improved heat transfer. Analyses results have shown that the design options yield an increased heat transfer with a minimal impact on moderator flow. The analysis has determined that the ribbed design option is superior in terms of heat transfer enhancement.
Degradation of Alloy 800 Under Steam Generator Secondary Side Crevice Conditions
Authors: Lu Y. C. , Goszczynski G. , Ramamurthy S.
Published: February 2010

Alloy 800 is the preferred steam generator (SG) tube materials for CANDU™ reactors and is also used extensively in SGs in some pressurized water reactor (PWR) systems. Degradation of Alloy 800 SG tubing has only been found in a few tubes at a limited number of stations despite the large number of SG tube operating years accumulated to date. Recently, underdeposit corrosion was detected in a few ex-service tubs removed from some CANDU SGs. Pits like wall loss of about 5% to 10% through-wall depth were found in these ex-service tubes. Evidence of intra-tubesheet cracking of Alloy 800 tubes was detected in a few European PWR SGs. There is no degradation in mechanical properties of these ex-service CANDU SG tubes. In addition, the degradation of Alloy 800 tubes observed so far is not a safety issue. However, the findings suggest that Alloy 800 tubing may have some aging degradation susceptibility after many years of service. Whether the degradation of Alloy 800 tubing is due to imperfections in its compositional or metallurgical properties inherent from manufacturing, or due to the aggressive chemistry conditions that should have been precluded by modern chemistry control strategy require clarification. Comprehensive examinations, including metallurgical examinations, orientation imaging microscopy (OIM), surface analyses and electrochemical measurements were performed on the removed ex-service CANDU SG tubes that had some underdeposit corrosion. The results were compared with a reference nuclear grade Alloy 800 tubing and with archive Alloy 800 new SG tubes from several CANDU stations. High-temperature electrochemical tests, scanning vibrating electrode Technique (SVET) measurements as well as C-ring autoclave tests were performed to determine the possible factors leading to Alloy 800 SG tubing degradation. SCC was initiated in a few C-ring specimens in the presence of artificial cold work flaws under simulated acidic SG secondary-side crevices chemistry conditions. OIM and surface analysis were also performed to characterize the degradation initiated in Alloy 800 tubing under the influence of cold work flaws. The possible factors leading to Alloy 800 SG tubing degradation under SG secondary crevices conditions are discussed.
Optimization of Test Interval of Ignalina Nuclear Power Plant Auxiliary Feedwater Pumps
Authors: Voronov Roman , Alzbutas Robertas
Published: February 2010

Some safety systems of the Ignalina Nuclear Power Plant (NPP) operate in standby mode. An equipment of such systems is periodically tested and that allows timely detect and repair equipment failures. The periodic testing is an important measure of ensuring systems’ operability and reliability. However, during the test and repair the equipment cannot perform it’s safety function, therefore too often testing decreases the availability of the system. This paper describes the mathematical model that represents how availability and reliability of the systems and their components depend on testing interval, taking into account different failure modes of the equipment. This model allows to find the optimal testing interval for the safety. As an example, the auxiliary feedwater pumps, that are a part of the Ignalina NPP Reactor Emergency Core Cooling System, are analysed. The model parameters calculation is based on Ignalina NPP data regarding pumps operation and failure as well as on general Nordic NPPs reliability data (T-Book) appling Bayesian approach for parameters updating. The analysed safety system is a redundant system that consists of six pumps and other equipment. Therefore a model for multiple components failure was developed. The model accounts for actual operational requirements of the system. The results of this model are compared with usually used binominal model.