bluuebunny/crossref_metadata_2025 huggingface dataset:
A Perspective on the U.S. Nuclear Fuel Cycle
Authors: Rodwell Ed , Machiels Albert
Published: September 2008

There has been a resurgence of interest in the possibility of processing the US spent nuclear fuel, instead of burying it in a geologic repository. Accordingly, key topical findings from three relevant EPRI evaluations made in the 1990–1995 timeframe are recapped and updated to accommodate a few developments over the subsequent ten years. Views recently expressed by other US entities are discussed. Processing aspects thereby addressed include effects on waste disposal and on geologic repository capacity, impacts on the economics of the nuclear fuel cycle and of the overall nuclear power scenario, alternative dispositions of the plutonium separated by the processing, impacts on the structure of the perceived weapons proliferation risk, and challenges for the immediate future and for the current half-century. Currently, there is a statutory limit of 70,000 metric tons on the amount of nuclear waste materials that can be accepted at Yucca Mountain. The Environmental Impact Statement (EIS) for the project analyzed emplacement of up to 120,000 metric tons of nuclear waste products in the repository. Additional scientific analyses suggest significantly higher capacity could be achieved with changes in the repository configuration that use only geology that has already been characterized and do not deviate from existing design parameters. Conservatively assuming the repository capacity postulated in the EIS, the need date for a second repository is essentially deferrable until that determined by a potential new nuclear plant deployment program. A further increase in technical capacity of the first repository (and further and extensive delay to the need date for a second repository) is potentially achievable by processing the spent fuel to remove the plutonium (and at least the americium too), provided the plutonium and the americium are then comprehensively burnt. The burning of some of the isotopes involved would need fast reactors (discounting for now a small possibility that one of several recently postulated alternatives will prove superior overall). However, adoption of processing would carry a substantial cost burden and reliability of the few demonstration fast reactors built to-date has been poor. Trends and developments could remove these obstacles to the processing scenario, possibly before major decisions on a second repository become necessary, which need not be until mid-century at the earliest. Pending the outcomes of these long-term trends and developments, economics and reliability encourage us to stay with non-processing for the near term at least. Besides completing the Yucca Mountain program, the two biggest and inter-related fuel-cycle needs today are for a nationwide consensus on which processing technology offers the optimum mix of economic competitiveness and proliferation resistance and for a sustained effort to negotiate greater international cooperation and safeguards. Equally likely to control the readiness schedule is development/demonstration of an acceptable, reliable and affordable fast reactor.
Globalization of ASME Nuclear Codes and Standards
Authors: Swayne Rick , Erler Bryan A.
Published: September 2008

With the globalization of the nuclear industry, it is clear that the reactor suppliers are based in many countries around the world (such as United States, France, Japan, Canada, South Korea, South Africa) and they will be marketing their reactors to many countries around the world (such as US, China, South Korea, France, Canada, Finland, Taiwan). They will also be fabricating their components in many different countries around the world. With this situation, it is clear that the requirements of ASME Nuclear Codes and Standards need to be adjusted to accommodate the regulations, fabricating processes, and technology of various countries around the world. It is also very important for the American Society of Mechanical Engineers (ASME) to be able to assure that products meeting the applicable ASME Code requirements will provide the same level of safety and quality assurance as those products currently fabricated under the ASME accreditation process. To do this, many countries are in the process of establishing or changing their regulations, and it is important for ASME to interface with the appropriate organizations in those countries, in order to ensure there is effective use of ASME Codes and standards around the world.
Assessment of the TRACE Reactor Analysis Code Against Selected PANDA Transient Data
Authors: Zavisca M. , Ghaderi M. , Khatib-Rahbar M. , Bajorek S.
Published: September 2008

The TRACE (TRAC/RELAP Advanced Computational Engine) code is an advanced, best-estimate thermal-hydraulic program intended to simulate the transient behavior of light-water reactor systems, using a two-fluid (steam and water, with non-condensable gas), seven-equation representation of the conservation equations and flow-regime dependent constitutive relations in a component-based model with one-, two-, or three-dimensional elements, as well as solid heat structures and logical elements for the control system. The U. S. Nuclear Regulatory Commission is currently supporting the development of the TRACE code and its assessment against a variety of experimental data pertinent to existing and evolutionary reactor designs. This paper presents the results of TRACE post-test prediction of P-series of experiments (i.e., tests comprising the ISP-42 blind and open phases) conducted at the PANDA large-scale test facility in 1990s. These results show reasonable agreement with the reported test results, indicating good performance of the code and relevant underlying thermal-hydraulic and heat transfer models.
Characteristics of Downcomer Boiling Phenomena During the Reflood Phase of a Postulated Large Break LOCA for the APR1400
Authors: Yun Byong-Jo , Euh Dong-Jin , Park Won-Man , Youn Young-Jung , Song Chul-Hwa
Published: September 2008

Downcomer boiling phenomena in a conventional pressurized water reactor have an important effect on the transient behavior of a postulated large-break LOCA (LBLOCA), because it can degrade the hydraulic head of the coolant in the downcomer and consequently affect the reflood flow rate for a core cooling. To investigate the thermal hydraulic behavior in the downcomer region, a test program for a downcomer boiling is being progressed in the reflood phase of a postulated LBLOCA. For this, the test facility was designed as a one side heated rectangular test section which adopts a full-pressure, full-height, and full-size downcomer-gap approach, but with the circumferential length reduced 47.08-fold. The test was performed by dividing it into two-phases: (I) visual observation and acquisition of the global two-phase flow parameters and (II) measurement of the local two-phase flow parameters on the measuring planes along five elevations. In the present paper, the test results of Phase-I and parts of Phase-II were introduced.
CFD Simulation of 3D Flowfield in a Gas Centrifuge
Authors: Jiang Dongjun , Zeng Shi
Published: September 2008

A CFD method was used to study the whole flow field in a gas centrifuge. In this paper, the VSM (Vector Splitting Method) of the FVM (Finite Volume Method) was used to solve the 3D Navier-Stokes equations. An implicit second-order upwind scheme was adopted. The numerical simulation was successfully performed on a parallel cluster computer and a convergence result was obtained. The simulation shows that: in the withdrawal chamber, a strong detached shock wave is formed in front of the scoop; as the radial position increases, the shock becomes stronger and the distance to scoop front surface is smaller. An oblique shock forms in the clearance between the scoop and the centrifuge wall; behind the shock-wave, the radially-inward motion of gas is induced because of the imbalance of the pressure gradient and the centrifugal force. In the separation chamber, a countercurrent is introduced. This indicates that CFD method can be used to study the complex three-dimensional flow field of gas centrifuges.
A Model for Surface Induced Growth of Inert Gas Bubbles in Irradiated Copper-Boron Alloys
Authors: Tiwari G. P. , Ramadasan E.
Published: September 2008

A matrix containing inert gas bubbles dilates in direct proportion to the growth experienced by the gas bubbles. This phenomenon is termed as swelling. A model for the swelling induced by the growth of the helium gas bubbles in irradiated copper-boron alloys is presented. The bubbles grow by acquiring vacancies from the external surface, which acts as a source of vacancies. The vacancies reach the surface of the bubbles mainly via lattice diffusion and to a limited extent via diffusion through short-circuiting paths such as grain boundaries and dislocation pipes. The model predicts that overall swelling of the matrix varies as 1.5th power of time. Another consequence of the present model is that the growth rate of a gas bubble varies inversely as the cube of its distance from the external surface. The model has been applied to the data on irradiated copper-boron alloys and found to be in accord with the experimental results. The model is general and can be applied to the growth of all kinds of stationary inert gas bubbles trapped within a crystalline matrix.
An Innovative Non-Destructive and Computational Method for Uranium Activity and Enrichment Verification of UF6 Cylinder
Authors: El-Mongy Sayed A. , Allam K. M. , Farid Osama M.
Published: September 2008

Verification of 235U enrichment in uranium hexaflouride (UF6) cylinders is often achieved by destructive and non-destructive assay techniques. These techniques are time consuming, need suitable and similar standard, in addition to loss of the nuclear material in the case of destructive analysis. This paper introduce an innovative approach for verifying of 235U enrichment in UF6 cylinder. The approach is based on measuring dose rate (μSv/h) resulted from the emmitted gamma rays of 235U at the surface of the cylinder and then calculating the activity of uranium and enrichment percentage inside the cylinder by a three dimensional model. Attenuation of the main 235U gamma transitions due to the cylinder wall (5A Type of Ni alloy) was also caculated and corrected for. The method was applied on UF6 cylinders enriched with 19.75% of 235U. The calculated enrichment was found to be 18% with 9% uncertainty. By the suggested method, the calculated total uranium activity inside one of the investigated UF6 cylinder was found close to the target (certified) value (5.6 GBq) with 9% uncertainty. The method is being developed by taking into consideration other parameters.
Using Center Hole Heat Transfer to Reduce Formation Times for Ceramic Waste Forms From Pyroprocessing
Authors: Bateman Kenneth J. , Solbrig Charles W.
Published: September 2008

The waste produced from processing spent fuel from the EBR II reactor must be processed into a waste form suitable for long term storage in Yucca Mountain. The method chosen produces zeolite granules mixed with glass frit, which must then be converted into a solid. This is accomplished by loading it into a can and heating to 900 C in a furnace regulated at 915 C. During heatup to 900 C, the zeolite and glass frit react and consolidate to produce a sodalite monolith. The resultant ceramic waste form (CWF) is then cooled. The waste form is 52 cm in diameter and initially 300 cm long but consolidates to 150 cm in length during the heating process. After cooling it is then inserted in a 5-DHLW/DOE SNF Long Canister. Without intervention, the waste takes 82 hours to heat up to 900 C in a furnace designed to geometrically fit the cylindrical waste form. This paper investigates the reduction in heating times possible with four different methods of additional heating through a center hole. The hole size is kept small to maximize the amount of CWF that is processed in a single run. A hole radius of 1.82 cm was selected which removes only 1% of the CWF. A reference computation was done with a specified inner hole surface temperature of 915 C to provide a benchmark for the amount of improvement which can be made. It showed that the heatup time can potentially be reduced to 43 hours with center hole heating. The first method, simply pouring high temperature liquid aluminum into the hole, did not produce any noticeable effect on reducing heat up times. The second method, flowing liquid aluminum through the hole, works well as long as the velocity is high enough (2.5 cm/sec) to prevent solidification of the aluminum during the initial front movement of the aluminum into the center hole. The velocity can be reduced to 1 cm/sec after the initial front has traversed the ceramic. This procedure reduces the formation time to near that of the reference case. The third method, flowing a gas through the center hole, also works well as long as the product of heat capacity and velocity of the gas is equivalent to that of the flowing aluminum, and the velocity is high enough to produce an intermediate size heat transfer coefficient. The fourth method, using an electric heater, works well and heater sizes between 500 to 1000 Watts are adequate. These later three methods all can reduce the heatup time to 44 hours allowing production to be doubled and a more uniform heating.
Separation of Technetium in Nitric Acid Solution With an Extractant Impregnated Resin
Authors: Moon Jei-Kwon , Lee Eil-Hee , Jung Chong-Hun , Lee Byung-Chul
Published: September 2008

An extractant impregnated resin (EIR) was prepared by impregnation of Aliquat 336 into Amberlite XAD-4 for separation of technetium from rhodium in nitric acid solution. The prepared EIR showed high preference for rhenium (chemical analogue of technetium) over rhodium. The adsorption isotherms for rhenium were described well by Langmuir equation in both the single and multi-component systems. Maximum adsorption capacities obtained by modelling the isotherms of rhenium were 2.01 meq g−1 and 1.97 meq g−1 for the single and the multi-component systems, respectively. Column tests were also performed to confirm the separation efficiency of rhenium using a jacketed glass column (φ11 × L 150). The EIR column showed successful separation of rhenium with the breakthrough volume of about 122 BV for the breakthrough concentration of 0.08. Also the breakthrough data were modelled successfully by assuming a homogeneous diffusion model in the particle phase. The diffusivities obtained from the modelling were in the order of 10−7 cm2 min−1 for a rhenium. The rhenium adsorbed on the bed could be eluted with a high purity by using a nitric acid solution.
Experimental Study on Heat Transferring Augmentation in Corrugated Tubes
Authors: Wu Jianhua , Jin Zhihao , Jin Wen , Wen Bangchun
Published: September 2008

The heat transferring characteristics of corrugated tubes were investigated experimentally, with the water as cool media inside the tubes and vapor as hot media outside the tubes. The heat transferring coefficient was measured within the Reynolds Number from 1800 to 24000 and a correlation formula was obtained. It was shown that the heat transferring characteristics of corrugated tubes are better than corresponding smooth tubes and the heat transferring coefficient is about 3 times greater than that of the smooth tubes. The heat transferring characteristics are not sensitive to the contour tolerance of the corrugated tubes.
Thermal Stress Evaluation of a Closed Branch Pipe Connected to Reactor Coolant Loop
Authors: Oumaya Toru , Nakamura Akira , Takenaka Nobuyuki , Onojima Daisuke
Published: September 2008

Several small bore pipes are branched off from main coolant loop in a nuclear power plant for drain or letdown system in CVCS. These pipes are often bended and connected to a horizontal pipe that leads to the closed valve. The main flow usually initiates a cavity flow in the branch pipe and hot water penetrates into it. In some case, large temperature fluctuations with a rather long period were detected at the elbow when the cavity flow reaches down to the elbow. This phenomenon is considered to be the cause of thermal fatigue crack. The purpose of this study is to clarify the mechanism of the large temperature fluctuations at the elbow of the branch pipe and evaluate thermal stress that lead to the fatigue crack. A series of mock-up experiments were conducted with vertical straight pipe (the straight-pipe) and a vertical pipe with an elbow connected to a horizontal pipe (the bent-pipe) made of acrylic resin or stainless steel. Vortex form and flow pattern were visualized using a tracer method for various main flow velocities. The liquid temperature was measured by thermo-couples and thermally stratified layer was visualized with liquid crystal. The inner wall temperature distribution around the elbow was measured by thermo-couples. The temperature distribution/fluctuation was then converted to the input of FEA code and time-dependent stress fluctuations were simulated to evaluate the thermal stress fluctuation.
Development of Surveillance and In-Service Inspection Programme for Indian Research Reactors Cirus and Dhruva
Authors: Shukla D. K.
Published: September 2008

Many safety requirements for research reactors are quite similar to those of power reactors. For research reactors with a higher hazard potential, the use of safety codes and guides for power reactors is more appropriate. However, there are many important differences between power reactors and research reactors that must be taken into account to ensure that adequate safety margins are available in design and operation of the research reactor. Most research reactors may have small potential for hazard to the public compared to power reactors but may pose a greater potential hazard to the plant operators. The need for greater flexibility in use of research reactors for individual experiments requires a different safety approach. Safety rules for power reactors are required to be substantially modified for application to specific research reactor. Following the intent of the available safety guides for surveillance and In-Service Inspection of Nuclear Power Plants, guidelines were formulated to develop surveillance and In-Service Inspection programme for research reactors Cirus and Dhruva. Based on the specific design of these research reactors, regulatory requirements, the degree of sophistication and experience of the technical organization involved in operating the research reactor, guidelines were evolved for developing and implementing the surveillance and In-Service Inspection programme for research reactors Cirus (40 MWt) and Dhruva (100 MWt) located at Bhabha Atomic Research Centre, Trombay, Mumbai, India. Paper describes the approach adopted for formulation of surveillance and In-service Inspection programme for Dhruva reactor in detail.
On the Traceability of Accuracy of Ultrasonic Flowmeter
Authors: Takeda Yasushi
Published: September 2008

Time-of-flight ultrasonic flowmeters have been widely used these days in industry. It is however in suspicion if its high accuracy is tracable to the national standard. It was made clear why traceability cannot be guaranteed from a fluid mechanical point of view. The main reason is a difference of flow configuration between the flow standard and the measurement position on-site. The concept of ‘Facility Factor’ is introduced and it is concluded that the ‘Profile Factor’ is not sufficient for correcting the meter reading. It is discussed that measurement of velocity profile on-site is essentially required.
Rank Statistics and Bootstrap: A More Precise Evaluation of the 95th Percentile in Nuclear Safety LB-LOCA Calculations
Authors: Joucla Je´roˆme , Probst Pierre
Published: September 2008

In 1988, since the revision of the 10 CFR50.46, the best estimate codes may be used in safety demonstration and licensing, provided that uncertainties are added to the relevant output parameters before comparing them with the acceptance criteria. The uncertainty of output parameters is coming principally from the lack of knowledge of the input parameters (initial and boundary conditions of the calculated transient, empirical models of the code, etc.). In the safety analysis of the large break loss of coolant accident (LB-LOCA), for a best-estimate analysis, it was decided that the acceptance criteria should not be exceeded by the 95th percentile estimated with a high degree of confidence. Best-estimate plus uncertainty (BEPU) analysis of the LB-LOCA allows significant margin gains. Therefore the nuclear industry develops and uses widely this type of analysis for power uprating, longer fuel cycles, fuel design etc. It appeared necessary to IRSN, technical support of the French Safety Authority, to get more insight to these strategies. The assessment of the 95th percentile can be made in three steps: 1. The definition and the evaluation of uncertainties of the input parameters. 2. The modelling and understanding of the output parameters. 3. The evaluation of the 95th percentile with a high degree of confidence. This paper presents and shows two new methods of how to deal with the third point. The standard approach is the use of a non-parametric approach called the Wilks’ formula which requires only 59 calculations to get a majorant of the 95th percentile with a 95% degree of confidence. This method was applied in the 90’s when the calculation resources were not sufficient to perform many calculations in a reasonable time. However, at the present time, the growth of computer power and the fall in prices allow to run many more calculations in the same physical time. We propose to use rank statistics or bootstrap to estimate the 95th percentile with a high confidence level, using some hundreds of calculations. This paper presents the results of the LOFT (Loss-of-Fluid Test) loss of coolant experiment L2-5 using these new methodologies. LOFT L2-5 was the subject of the former ISP 13 and the ongoing BEMUSE international problem. The output parameter is the second maximum peak cladding temperature of the fuel. The best estimate code used is CATHARE2 V2.5. Rank statistics and bootstrap improve the information on the estimation of the 95th percentile: a higher degree of confidence, a more precise estimation etc.
Economic Dispatch Using Genetic Algorithm Based Hybrid Approach
Authors: Malik Tahir Nadeem , Ahmad Aftab , Khushnood Shahab
Published: September 2008

Power Economic Dispatch (ED) is vital and essential daily optimization procedure in the system operation. Present day large power generating units with multi-valves steam turbines exhibit a large variation in the input-output characteristic functions, thus non-convexity appears in the characteristic curves. Various mathematical and optimization techniques have been developed, applied to solve economic dispatch (ED) problem. Most of these are calculus-based optimization algorithms that are based on successive linearization and use the first and second order differentiations of objective function and its constraint equations as the search direction. They usually require heat input, power output characteristics of generators to be of monotonically increasing nature or of piecewise linearity. These simplifying assumptions result in an inaccurate dispatch. Genetic algorithms have used to solve the economic dispatch problem independently and in conjunction with other AI tools and mathematical programming approaches. Genetic algorithms have inherent ability to reach the global minimum region of search space in a short time, but then take longer time to converge the solution. GA based hybrid approaches get around this problem and produce encouraging results. This paper presents brief survey on hybrid approaches for economic dispatch, an architecture of extensible computational framework as common environment for conventional, genetic algorithm & hybrid approaches based solution for power economic dispatch, the implementation of three algorithms in the developed framework. The framework tested on standard test systems for its performance evaluation.
Thermal-Hydraulic Performance of Cross-Shaped Spiral Fuel in High-Power-Density BWRs
Authors: Conboy Thomas , Hejzlar Pavel
Published: September 2008

Power up-rating of existing nuclear reactors promises to be an area of great study for years to come. One of the major approaches to efficiently increasing power density is by way of advanced fuel design, and cross-shaped spiral-fuel has shown such potential in previous studies. Our work aims to model the thermal-hydraulic consequences of filling a BWR core with these spiral-shaped pins. The helically-wound pins have a cross-section resembling a 4-petaled flower. They fill an assembly in a tight bundle, their dimensions chosen carefully such that the petals of neighboring pins contact each other at their outer-most extent in a self-supporting lattice, absent of grid spacers. Potential advantages of this design raise much optimism from a thermal-hydraulic perspective. These spiral rods possess about 40% larger surface area than traditional rods, resulting in increased cooling and a proportional reduction in average surface heat flux. The thin petal-like extensions help by lowering thermal resistance between the hot central region of the pin and the bulk coolant flow, decreasing the maximum fuel temperature by 200°C according to Finite Element (COSMOS) models. However, COSMOS models also predict a potential problem area at the “elbow” region of two adjoining petals, where heat flux peaking is twice that along the extensions. Preliminary VIPRE models, which account only for the surface area increase, predict a 22% increase in critical power. It is also anticipated that the spiral twist would provide the flowing coolant with an additional radial velocity component, and likely promote turbulence and mixing within an assembly. These factors are expected to provide further margin for increased power density, and are currently being incorporated into the VIPRE model. The reduction in pressure drop inherent in any core without grid-spacers is also expected to be significant in aiding core stability, though this has not yet been quantified. Spiral-fuel seems to be a favorable alternative to traditional pins from a thermal-hydraulic standpoint, though further study of the trends shown in this paper are required.
CFD Analysis of Vortex Shedding in a Circular Cylinder: Flow Induced Vibration Design
Authors: Sheikh Nadeem Ahmed , Malik M. Afzaal , Qureshi Arshad Hussain , Khan M. Anwar , Khushnood Shahab
Published: September 2008

Flow past a blunt body, such as a circular cylinder, usually experiences boundary layer separation and very strong flow oscillations in the wake region behind the body at a discrete frequency that is correlated to the Reynolds number of the flow. The periodic nature of the vortex shedding phenomenon can sometimes lead to unwanted structural vibrations. The effect of vibrating instability of a single cylinder is investigated in a uniform flow using the power of computational methods. Fluid structure coupling procedure predicts the fluid forces responsible for structural vibrations. An implicit approach to the solution of the unsteady two-dimensional Navier-Stokes equations is used for computation of flow parameters. Calculations are performed in parallel using a domain re-meshing/deforming technique with efficient communication requirements. Results for the unsteady shedding flow behind a circular cylinder are presented with experimental comparisons, showing the feasibility of accurate, efficient, time-dependent estimation of shedding frequency and resulting vibrations.
Natural Cooling System Design and Forced Draft Test for Highly Radioactive Vitrified Waste Packages Storage Pits
Authors: Ono Isamu , Jinba Masaki , Ono Akira , Nagai Kiyoyuki , Kaku Yoshiyuki , Okubo Tetsuro , Mino Yoshitaka , Miyakawa Masamichi
Published: September 2008

Vitrified Waste Storage Facilities, which store high-level radioactive vitrified waste packages with natural air-cooling system, have been constructed at JNFL’s Rokkasho Reprocessing Site. [1] Construction work for the first vitrified waste storage facility which stores the waste packages vitrified in Rokkasho site was completed in December in 2005. After the final inspection, a forced draft test was conducted to verify actual fluid resistance throughout the flow passage. The test results show that the measured pressure loss is less than 70Pa at design flow rate and that it has approximately 10% margin to the design value. This confirms that the facility has the adequate cooling performance.
Safety Design and Mock-Up Tests on the Combustion of Hydrogen-Air Mixture in the Vertical CNS Channel of the CARR-CNS
Authors: Yu Qingfeng , Feng Quanke
Published: September 2008

A two-phase thermo-siphon loop is applied to the Cold Neutron Source (CNS) of China Advanced Research Reactor (CARR). The moderator is liquid hydrogen. The two-phase thermo-siphon consists of the crescent-shape moderator cell, the moderator transfer tube, and the condenser. The hydrogen is supplied from the buffer tank to the condenser. The most characteristic point is that the cold helium gas is introduced into the helium sub-cooling system covering the moderator cell and then flows up through the tube covering the moderator transfer tube into the condenser. The helium sub-cooling system also reduces the void fraction of the liquid hydrogen and takes a role of the helium barrier for preventing air from intruding into the hydrogen system. We call the two-phase thermo-siphon the hydrogen cold system. The main part of this system is installed in the CNS channel made of 6061 aluminum alloy (6061A) of 6 mm in thickness, 270 mm in outer diameter and about 6 m in height. For confirming the safety of the CNS, the combustion tests were carried out using the hydrogen-air mixture under the conditions in which air is introduced into the tube at 1 atmosphere, and then hydrogen gas is supplied from the gas cylinder up to the test pressures. And maximum test pressure is 0.140 MPa Gauge (G). This condition includes the design accident of the CNS. The peak pressure due to combustion is 1.09 MPa, and the design strength of the CNS channel is 3 MPa. The safety of the CNS was thus verified even if the design basis accident occurs. The pressure distribution, the stress, and the displacement of the tube were also measured.
Fast and Robust Computation of the Matrix Absolute Value Function: Application to Roe Solver for the Numerical Simulation of Two-Phase Flow Models
Authors: Ndjinga Michae¨l , Kumbaro Anela , Laurent-Gengoux Pascal , De Vuyst Florian
Published: September 2008

The application of the generalized Roe scheme to the numerical simulation of two-phase flow models requires a fast and robust computation of the absolute value of the system matrix. In several models such as the two-fluid model or a general multi-field model, this matrix has a non trivial eigenstructure and the eigen decomposition is often ill conditioned. We give two general algorithms avoiding the diagonalization process: an iterative computation, which turns out to be an exact computation, and an interpolation algorithm which is faster and can handle the case of complex eigenvalues. The knowledge of the characteristic polynomial gives us an easy access to the eigenvalues but however, the iterative scheme can be used with only estimates of the eigenvalues, using for example Gershgorin’s disk localization. We finally show some numerical results of two-fluid model simulations involving interfacial pressure and virtual mass force models.
Studies of the Deteriorated Turbulent Heat Transfer Regime for the Gas-Cooled Fast Reactor Decay Heat Removal System
Authors: Lee Jeong Ik , Hejzlar Pavel , Kazimi Mujid S. , Saha Pradip
Published: September 2008

Increased reliance on passive emergency cooling using natural circulation of gas at elevated pressure is one of the major goals for the Gas-cooled Fast Reactor (GFR). Since GFR cores have high power density and low thermal inertia, the decay heat removal (DHR) in depressurization accidents is a key challenge. Furthermore, due to its high surface heat flux and low velocities under natural circulation in any post-LOCA scenario, three effects impair the capability of turbulent gas flow to remove heat from the GFR core, namely: (1) Acceleration effect (2) Buoyancy effect (3) Properties variation. This paper reviews previous work on heat transfer mechanisms and flow characteristics of the Deteriorated Turbulent Heat Transfer (DTHT) regime. It is shown that the GFR’s DHR system has a potential for operating in the DTHT regime by performing a simple analysis. A description of the MIT/INL experimental facility designed and built to investigate the DTHT regime is provided together with the first test results. The first runs were performed in the forced convection regime to verify facility operation against well-established forced convection correlations. The results of the three runs at Reynolds numbers 6700, 8000 and 12800 showed good agreement with the Gnielinsky correlation [4], which is considered the best available heat transfer correlation in the forced convection regime and is valid for a large range of Reynolds and Prandtl numbers. However, even in the forced convection regime, the effect of heat transfer properties variation of the fluid was found to be still significant.
Swing Check Valve Design Criteria and CFD Validation
Authors: Dallstream Brian E. , Fricke Brian A. , Becker Bryan R.
Published: September 2008

This paper provides information on swing check valve selection criteria suitable for nuclear power plant applications. In this project, four swing check valves were analyzed to demonstrate the implementation and application of this information. In this example, swing check valves were selected according to “ASME Boiler and Pressure Vessel Code, Section III” [1] and “ASME B16.34, Valves Flanged, Threaded, and Welding End” [2]. This paper also discusses the utilization of Computational Fluid Dynamics Software (CFD) as a means to analyze valve design. The use of CFD is a relatively new approach for validation of valve design that is becoming invaluable due to the high cost of physical bench testing. The Instrument Society of America (ISA) Analysis Division and the American Society of Mechanical Engineers (ASME) Computational Fluid Dynamics Technical Committee have taken a proactive approach in setting standards and practices for the use of CFD in design and validation.
Recent IAEA Activities in Support of Safety Assessment Capabilities
Authors: Modro S. Michael , El-Shanawany Mamdouh , Lee Sukho
Published: September 2008

One of the key missions of the International Atomic Energy Agency (IAEA) is to develop nuclear safety standards and, based on these standards, to promote the achievement and maintenance of high levels of safety in the applications of nuclear energy. In the context of this mission the IAEA conducts programmes that support the safety assessment capabilities of its Member States. This paper focuses on two new long term technical activities specific to safety assessment. These concern the issue of uncertainties in deterministic evaluation of safety, and the role of computational fluid dynamics (CFD) methods in safety assessments. In addition, a new IAEA initiative involving the collaboration of Member States in the area of advanced safety assessment tools is presented.
Modeling Cladding-Coolant Heat Transfer of High-Burnup Fuel During RIA
Authors: Liu Wenfeng , Kazimi Mujid S.
Published: September 2008

This paper describes a model for the cladding-coolant heat transfer of high burnup fuel during a Reactivity Initiated Accident (RIA) which is implemented in the fuel performance code FRAPTRAN 1.2. The minimum stable film boiling temperature, affected by the subcooling and the clad oxidation, is modeled by a modified Henry correlation. This accounts for the effects of thermal properties of the cladding surface on the transient temperature drop during liquid-solid contact. The transition boiling regime is described as the interpolation of the heat flux between two anchor points on the boiling curve: the Critical Heat Flux (CHF) and minimum stable film boiling. The CHF correlation is based on the Zuber hydrodynamic model multiplied by a subcooling factor. Frederking correlation is chosen to model the film boiling regime. The heat conduction through the oxide layer of the cladding surface of high burnup fuel is calculated by solving heat conduction equations with thermal properties of zirconia taken from MATPRO. This model is validated in the FRAPTRAN code for test cases of both high burnup and fresh test fuel rods including the burnup level (0–56 MW d/kg), peak fuel enthalpy deposit (70–190 cal/g), degree of subcooling (0–80 °C), and extent of oxidation (0–25 micron). The modified code demonstrates the capability of differentiating between the departure from nucleate boiling (DNB) and none-DNB cases. The predicted peak cladding temperature (PCT) and duration of DNB achieves generally good agreement with the experimental data. It is found that the cladding surface oxidation of high burnup fuel causes an early rewetting of cladding or suppresses DNB due to two factors: 1) Thick zirconia layer may delay the heat conducted to the surface while keeping the surface heat transfer in the most effective nucleate boiling regime. 2) The transient liquid-solid contact resulting from vapor breaking down would cause a lower interface temperature for an oxidized surface, essentially raises the minimum stable film boiling temperature.
Preliminary Experimental Analysis of Soil Stabilizers for Contamination Control
Authors: Lagos L. , Varona J. , Zidan A. , Gudavalli R. , Wu Kuang-His
Published: September 2008

A major focus of Department of Energy’s (DOE’s) environmental management mission at the Hanford site involves characterizing and remediating contaminated soil and groundwater; stabilizing contaminated soil; remediating disposal sites; decontaminating and decommissioning structures, and demolishing former plutonium production process buildings, nuclear reactors, and separation plants; maintaining inactive waste sites; transitioning facilities into the surveillance and maintenance program; and mitigating effects to biological and cultural resources from site development and environmental cleanup and restoration activities. For example, a total of 470,914 metric tons of contaminated soil from 100 Areas remediation activities were disposed at the Environmental Restoration Disposal Facility (ERDF) during 2004 [5]. The Applied Research Center (ARC) at Florida International University (FIU) is supporting the Hanford’s site remediation program by analyzing the effectiveness of several soil stabilizers (fixatives) for contamination control during excavation activities. The study is focusing on determining the effects of varying soil conditions, temperature, humidity and wind velocity on the effectiveness of the candidate stabilizers. The test matrix consists of a soil penetration-depth study, wind tunnel experiments for determination of threshold velocity, and temperature and moisture-controlled drying/curing experiments. These three set of experiments are designed to verify performance metrics, as well as provide insight into what fundamental forces are altered by the use of the stabilizer. This paper only presents the preliminary results obtained during wind tunnel experiments using dry Hanford soil samples (with 2.7% moisture by weight). These dry soil samples were exposed to varying wind speeds from 2.22 m/sec to 8.88 m/sec. Furthermore, airborne particulate data was collected for the dry Hanford soil experiments using an aerosol analyzer instrument.
Design and Manufacture of the Storage Cask for the Old Reactor Internals
Authors: Tomiita Yasuhiro
Published: September 2008

Mitsubishi Heavy Industries, Ltd. (MHI) completed replacement work of upper reactor internals (UCI) and lower reactor internals (LCI) of the pressurized water reactor in Shikoku Electric Power Company’s Ikata Unit No.1 by “the all-in-one-piece extraction method” introduced in the document of [ICONE14-89233]. In the pressurized water reactor (PWR) plant, the UCI are usually removed from the reactor vessel (RV) independently and reinstalled into the RV again every refueling outage. The LCI are independently able to be removed from the RV and reinstalled again during in-service inspection, too. In the boiling water reactor (BWR) plant, there are several cases of replacing BWR shrouds by cutting small and containing in a container. But no replacement of all reactor internals (CI) for the PWR, in one piece without splitting or cutting, has been reported. The purpose of this paper is to introduce the key points about the design and manufacture of the storage cask for old reactor internals in the replacement work by “the all-in-one-piece extraction method”.
Transient Forced Convection Heat Transfer for Helium Gas Flowing Over a Horizontal Plate
Authors: Liu Qiusheng , Shibahara Makoto , Fukuda Katsuya
Published: September 2008

Transient heat transfer coefficients for helium gas flowing over a horizontal plate (ribbon) were measured under wide experimental conditions. The platinum plate with a thickness of 0.1 mm was used as test heater and heated by electric current. The heat generation rate was exponentially increased with a function of Q0exp(t/τ). The gas flow velocities ranged from 4 to 10 m/s, the gas temperatures ranged from 313 to 353 K, and the periods of heat generation rate, τ, ranged from 50 ms to 17 s. The surface superheat and heat flux increase exponentially as the heat generation rate increases with the exponential function. It was clarified that the heat transfer coefficient approaches the quasi-steady-state one for the period τ longer than about 1 s, and it becomes higher for the period shorter than around 1 s. The dependence of transient heat transfer on the gas flowing velocity becomes weaker when the period becomes very shorter. The gas temperature in this study shows little influence on the heat transfer coefficient. Empirical correlation for quasi-steady-state heat transfer was obtained based on the experimental data.
Application of CFX-10 to the Investigation of RPV Coolant Mixing in VVER Reactors
Authors: Moretti Fabio , Melideo Daniele , Terzuoli Fulvio , D’Auria Francesco
Published: September 2008

Coolant mixing phenomena occurring in the pressure vessel of a nuclear reactor constitute one of the main objectives of investigation by researchers concerned with nuclear reactor safety. For instance, mixing plays a relevant role in reactivity-induced accidents initiated by deboration or boron dilution events, followed by transport of a deborated slug into the vessel of a pressurized water reactor. Another example is constituted by temperature mixing, which may sensitively affect the consequences of a pressurized thermal shock scenario. Predictive analysis of mixing phenomena is strongly improved by the availability of computational tools able to cope with the inherent three-dimensionality of such problem, like system codes with three-dimensional capabilities, and Computational Fluid Dynamics (CFD) codes. The present paper deals with numerical analyses of coolant mixing in the reactor pressure vessel of a VVER-1000 reactor, performed by the ANSYS CFX-10 CFD code. In particular, the “swirl” effect that has been observed to take place in the downcomer of such kind of reactor has been addressed, with the aim of assessing the capability of the codes to predict that effect, and to understand the reasons for its occurrence. Results have been compared against experimental data from V1000CT-2 Benchmark. Moreover, a boron mixing problem has been investigated, in the hypothesis that a deborated slug, transported by natural circulation, enters the vessel. Sensitivity analyses have been conducted on some geometrical features, model parameters and boundary conditions.
Preparation and Execution of the GIS First Overhaul for Qinshan NPP Phase One
Authors: Lou Kaihong , Yan Jiapeng
Published: September 2008

This paper addresses the necessity and feasibility of the first major overhaul on the GIS based on the analysis of the special conditions and the issues we confronted; After the comparison of various schemes, the optimized scheme is put forward; the paper also expounds the proper preparation and cautious practice which led to the hard but final accomplishment of the initial overhaul on the GIS; this article further explains the necessity of the major overhaul on the GIS through the disposal of abnormalities during the execution of this major overhaul.
Prediction of Severe Accident Counter Current Natural Circulation Flows in the Hot Leg of a Pressurized Water Reactor
Authors: Boyd Christopher F.
Published: September 2008

During certain phases of a severe accident in a pressurized water reactor (PWR), the core becomes uncovered and steam carries heat to the steam generators through natural circulation. For PWR’s with U-tube steam generators and loop seals filled with water, a counter current flow pattern is established in the hot leg. This flow pattern has been experimentally observed and has been predicted using computational fluid dynamics (CFD). Predictions of severe accident behavior are routinely carried out using severe accident system analysis codes such as SCDAP/RELAP5 or MELCOR. These codes, however, were not developed for predicting the three-dimensional natural circulation flow patterns during this phase of a severe accident. CFD, along with a set of experiments at 1/7th scale, have been historically used to establish the flow rates and mixing for the system analysis tools. One important aspect of these predictions is the counter current flow rate in the nearly 30 inch diameter hot leg between the reactor vessel and steam generator. This flow rate is strongly related to the amount of energy that can be transported away from the reactor core. This energy transfer plays a significant role in the prediction of core failures as well as potential failures in other reactor coolant system piping. CFD is used to determine the counter current flow rate during a severe accident. Specific sensitivities are completed for parameters such as surge line flow rates, hydrogen content, as well as vessel and steam generator temperatures. The predictions are carried out for the reactor vessel upper plenum, hot leg, a portion of the surge line, and a steam generator blocked off at the outlet plenum. All predictions utilize the FLEUNT V6 CFD code. The volumetric flow in the hot leg is assumed to be proportional to the square root of the product of normalized density difference, gravity, and hydraulic diameter to the 5th power. CFD is used to determine the proportionality constant in the range from 0.11 to 0.13 and termed a discharge coefficient. The value is relatively unchanged for typical surge line flow rates as well as the hydrogen content in the flow. Over a significant range of expected temperature differences for the steam generator and reactor vessel upper plenum, the discharge coefficient also remained consistent. The discharge coefficient is a suitable model for determining the hot leg counter current flow rates during this type of severe accident.