bluuebunny/crossref_metadata_2025 huggingface dataset:
On Voltage-Sensitive Managing the Redox-Potential of MSR Fuel Composition
Authors: Alekseev Pavel N. , Shimkevich Alexander L.
Published: July 2009

The principles for optimal managing a composition of base solutions for the molten-salt reactor are formulated here for ensuring the given properties and exchange processes as a selective extracting of salt components. The correction of melt properties can be carried out by means of impurity additives parallel with the forced and controllable variation of reduction-oxidation (redox) potential of the non-stoichiometric salts. The accent is done on a possible application of the potentiometer for monitoring and managing of the properties of MSR fuel compositions. For this, one can use the precision methods of e.m.f and the coulomb-metric titration of sodium (lithium) in a galvanic cell upon the base of Na+(Li+)-β″-Al2O3 solid electrolyte with cation conductivity.
On Nanoceramic Fuel for Perspective Nuclear Reactors
Authors: Alekseev Pavel N. , Proshkin Alexander A. , Shimkevich Alexander L.
Published: July 2009

In developing methods for designing reactor materials with reference to nuclear fuel, a task for micro-structural suppressing the processes of swelling, corrosion, and forming macroscopic defects (phase inclusions, pores, gaseous bubbles, cracks) in the fuel is formulated. For managing the defective structure of materials under irradiation, it is offered to choose their composition that fuel oxide ceramics have consisted of steady-state fractal tetrahedral clusters as the dense part of amorphous matrix (the nano-ceramic structure). In this connection, the search of multi-component systems providing this stable state can become a perspective direction for designing fuel with small swelling and corrosion factor. Such the materials will have high density of equilibrium vacancies that effectively absorb solid and gaseous fission products, and reduce the fuel swelling factor, physical, chemical, and thermal properties.
Investigation of Combined Neutron Scattering and Neutron Radiography Techniques for Measurement of Gas-Liquid Two-Phase Flow
Authors: Khaial A. M. , Harvel G. D. , Chang J. S.
Published: July 2009

Neutron radiography has the ability to differentiate between a gas and liquid in two phase flow due both to the density difference and the affinity of Hydrogen to scatter the neutrons. The neutron radiography technique has the additional benefit of visualizing through metal walls due to the low probability of neutron interactions with most metals which has resulted in significant studies using neutron radiography for the measurement of two-phase flow. The studies have shown that both real-time and high-speed imaging of two-phase flow is possible for the measurement of void distribution in air-water, steam-water, and gas-liquid metal flow. However, since neutron radiography is a planar method, full three dimensional capability of two-phase flow has not yet been demonstrated. The purpose of this work is to investigate if neutron scattering could provide additional information to supplement the information obtained in real-time neutron radiography. Experiments are performed with a real-time neutron radiography system which provides the two-dimensional information. A turntable is used to evaluate the 3D structure of a non-moving two-phase system to identify the three dimensional structure. Additional thermal neutron detectors are placed at the periphery of the neutron beam to detect the scattered neutrons. The detector response as a function of the location of a gas-liquid interface rotated through the third dimension can be obtained. The neutron radiography technique was able to resolve the interface. The location of the interface could be observed to move as the object was rotated in the beam yet the location of the interface along the neutron beam axis could not be determined by radiography alone as expected. The neutron scattering technique did provide additional information in the third dimension and thus it was possible by using both techniques to locate the gas-liquid interface.
On the Pressure Dependency of Physical Parameters in Case of Heat Transfer Problems of Supercritical Water
Authors: Hazi Gabor , Farkas Istvan
Published: July 2009

Studying heat transfer problems of supercritical water, the pressure dependency of thermophysical parameters (density, specific heat, viscosity, thermal conductivity) and the work done by pressure are often neglected. Here we show that the variations of some physical parameters as functions of pressure have the same order of magnitude than their variations as functions of temperature in supercritical water. Therefore, pressure dependency of physical parameters should be taken into account in heat transfer calculations of supercritical water. It is also pointed out that the work done by pressure should not be neglected in supercritical water, since the pressure work term has the same order of magnitude than the convective term near the pseudocritical temperature.
Oil Recovery From Shale With Nuclear Generated Heat
Authors: Sandquist Gary M. , Kunze Jay F. , Rogers Vern C.
Published: July 2009

Shell Oil Corporation has developed an in-situ process for shale oil recovery that uses electric heaters to heat oil shale deposits and produce chemical reactions within the shale that can liberate the shale-oil. The major production expense is electrical power used to heat the shale. Significantly, small mobile nuclear reactors are now under development and testing that could provide high-temperature working fluids (both gaseous and liquid) at lower unit energy cost to replace current electrical heating. Nuclear generated steam is particularly cost effective and technically attractive for oil shale recovery. Estimates are that US oil shale deposits could be made to produce about 2 million barrels of oil per acre ($200 million/acre of oil at $100/barrel) if properly processed using high temperature steam. Furthermore, a these small nuclear reactors could be delivered by heavy haul truck, carefully buried for adequate shielding and safety, remotely operated, and moved as needed to process large oil shale fields.
The Advanced Fuel Cycle Facility (AFCF) Role in the Global Nuclear Energy Partnership
Authors: Griffith Andrew , Boger John , Perry Jeffrey
Published: July 2009

The Global Nuclear Energy Partnership (GNEP), launched in February, 2006, proposes to introduce used nuclear fuel recycling in the United States (U.S.) with improved proliferation-resistance and a more effective waste management approach. This program is evaluating ways to close the fuel cycle in a manner that introduces the most advanced technologies of today and builds on recent breakthroughs in U.S. national laboratories while drawing on international and industry partnerships. Central to moving this advanced fuel recycling technology from the laboratory to commercial implementation is the development and siting of three proposed GNEP facilities: the Consolidated Fuel Treatment Center (CFTC), the Advanced Burner Reactor (ABR), and the Advanced Fuel Cycle Facility (AFCF). These three projects are envisioned to introduce used fuel separations, advanced fuel fabrication, and fast reactor technology in a manner that efficiently recycles material, produces the most energy out of the existing inventory of used fuel, and improves our ability to manage nuclear waste. The CFTC and ABR will depend on industry involvement and will not be covered by this paper. This paper will cover key considerations for the AFCF. The AFCF will provide the U.S. with the capabilities required to evaluate technologies that separate used fuel into reusable material and waste in a proliferation-resistant manner. The separations technology demonstration capability is coupled with a remote transmutation fuel fabrication demonstration capability in an integrated manner that demonstrates advanced safeguard technologies.
Validation of Fluid-Structure Interaction Calculations in a Large-Break Loss of Coolant Accident
Authors: Timperi Antti , Pa¨ttikangas Timo , Karppinen Ismo , Lestinen Ville , Ka¨hko¨nen Jukka , Toppila Timo
Published: July 2009

In the hypothetical Large-Break Loss Of Coolant Accident (LBLOCA), rapid depressurization of the reactor primary circuit causes loads on the reactor internals. This paper presents numerical simulations of a HDR experiment, where LBLOCA of a pressurized water reactor due to a sudden pipe break in the primary loop was studied. In the experiment, Fluid-Structure Interaction (FSI) phenomena caused by the flexibility of the core barrel were studied in particular. Star-CD Computational Fluid Dynamics (CFD) code and ABAQUS structural analysis code were used for three-dimensional calculations. The MpCCI code was used for two-way coupling of the CFD and structural analysis codes in order to take FSI into account. Two-way FSI calculation was also performed with ABAQUS only by modeling water as an acoustic medium. Pressure boundary condition at the pipe break was evaluated with the system code APROS as a two-phase calculation. Comparisons with the experiment were made for fluid pressures and break mass flow as well as for structural displacements and strains. Fairly good agreement was found between the experiment and simulation when coupling of the CFD and structural analysis codes was used. For the acoustic calculation, the results showed good agreement in the early phase of the simulation. In the late phase, structural loads were over-predicted by the acoustic calculation due to the effect of bulk flow of water which is not included in the acoustic model.
Solution to Pakistan Electrical Power Crisis
Authors: Akbar Sajjad , Khusnood Shahab
Published: July 2009

Electricity is the engine for the growth of economy of any country. Total installed electricity generation capacity of Pakistan is presently approx 20,000 MW as given in Table-1. Despite this, almost 40% of the population is without electricity. Pakistan has been blessed with tremendous resources for electrical power generation with hydel, coal, renewable energy resources and Nuclear power. Hydel, coal potential of more than 40,000 MW and 10,000 MW are available but only 15% of hydroelectric potential has been harnessed so for where as only 150 MW power plant on indigenous coal has been set up. To exploit Pakistan hydel and coal resources for power generation large investments are needed which Pakistan economy can not afford. Govt. of Pakistan has created an organization of private power and infrastructure board (PPIB) to facilitate private sector in the participation of power generator. PPIB is tapping the resources and facilitating the private sector for establishment of power projects. Pakistan is collaborating with China for establishment of Nuclear Power Plants and plan to generate up to 10,000 MW by year 2025. Renewable energy resources are also required to be tapped. This paper will focus on the Pakistan power generation potential by utilizing local resources keeping in view the next 20 year supply and demand position.
Evaluation Analysis of Prediction Methods for Two-Phase Flow Pressure Drop in Mini-Channels
Authors: Sun Licheng , Mishima Kaichiro
Published: July 2009

2092 data of two-phase flow pressure drop were collected from 18 published papers of which the working fluids include R123, R134a, R22, R236ea, R245fa, R404a, R407C, R410a, R507, CO2, water and air. The hydraulic diameter ranges from 0.506 to 12mm; Relo from 10 to 37000, and Rego from 3 to 4×105. 11 correlations and models for calculating the two-phase frictional pressure drop were evaluated based upon these data. The results show that the accuracy of the Lockhart-Martinelli method, Mishima and Hibiki correlation, Zhang and Mishima correlation and Lee and Mudawar correalion in the laminar region is very close to each other, while the Muller-Steinhagen and Heck correlation is the best among the evaluated correlations in the turbulent region. A modified Chisholm correlation was proposed, which is better than all of the evaluated correlations in the turbulent region and its mean relative error is about 29%. For refrigerants only, the new correlation and Muller-Steinhagen and Heck correlation are very close to each other and give better agreement than the other evaluated correlations.
Study on Probabilistic Fracture Mechanics Analysis Codes for Pipes With Stress Corrosion Cracks
Authors: Machida Hideo , Arakawa Manabu , Yamashita Norimichi , Yoshimura Shinobu
Published: July 2009

Risk-Informed integrity management methodologies have been developed in Japanese nuclear power plants. One of the issues of concern is the reliability assessment of piping with flaws due to stress corrosion cracking (SCC). Therefore, the probabilistic fracture mechanics analysis code have been developed, which can perform the reliability assessment for the austenitic stainless steel piping with flaws due to SCC. This paper describes technical basis of this code. This method is based on Monte-Carlo technique considering many sample cases in a piping section, where the initiation and growth of cracks are calculated and piping failures, including leaks and rapture, are evaluated. A notable feature is that multiple cracks can be treated, consequently, assessment of coalescence of cracks and intricate break evaluation of piping section have been included. Moreover, the in-service inspection (ISI) and integrity evaluation by Fitness-for-Service (FFS) code are integrated into the analysis, and the contribution to failure probability decrease can be assessed. Key parameters are determined on a probability basis with the designated probability type throughout the procedure. Size, location and time of crack initiation, coefficients of crack growth due to SCC and factors for piping failure are included in those parameters. With this method the reliability level of the piping through the operation periods can be estimated and the contribution of various parameters including ISI can be quantitatively evaluated.
Effect of Temperature on Damping in Shell and Tube Heat Exchangers
Authors: El-Ghalban Ali Roheim , Iqbal Qamar , Khushnood Shahab , Qureshi M. Arshad , Shahid Khalil M.
Published: July 2009

Flow-induced vibration in heat exchangers has been a key source of concern in the process, power generation and nuclear industry for several decades. Many incidents of failure of heat exchangers due to apparent flow-induced vibration have been reported. Design of tube bundles with loosely supported tubes in baffles for process shell and tube heat exchanger and steam generator needs estimation of energy dissipation mechanisms or damping for a safer and long term operation. Damping has a major influence on the flow induced vibrations and is dependant on a variety of factors such as mechanical properties of the tube material, geometry and number of intermediate supports, the physical properties of shell-side fluid, type of tube motion, tube frequency, shell-side temperature etc. Various damping mechanisms have been identified and quantified such as Friction damping, Viscous damping, Squeeze film damping, Support damping and Two-Phase damping which affect the performance with respect to flow induced vibration design, including standard design guidelines. But generally the effects of the higher operating temperatures on the various damping mechanisms are neglected in the general design procedure. The operating temperatures play significant role on the contribution of various damping mechanisms. The current paper focuses on the thermal aspects of damping mechanisms subjected to single phase cross-flow in process heat exchangers and formulates the design guidelines for safer design based on experimental and empirical formulations. The research results show that he increase in the temperature results in the increase of the damping. Moreover it found that the natural frequency is higher for lower mass flow rate and lower working pressures and lower temperatures.
Application of Acoustic Technique for Surveillance and Anomaly Detection in LMFBRs
Authors: Prakash V. , Anandaraj M. , Thirumalai M. , Kalyanasundaram P. , Vaidyanathan G.
Published: July 2009

Acoustic techniques find wide application in Liquid Metal Fast Breeder Reactors (LMFBRs) for ensuring its high reliability, safety and plant availability. Various surveillance methods based on acoustic technique can be employed in these reactors to detect deviations from normal operating conditions. This could be used for the measurement of drop time of Diverse Safety Rods (DSRs) in the core, detection of in-sodium water leaks in Steam Generators, cavitation detection in sodium pumps and reactor core components. An active R&D program is being pursued in these areas at Indira Gandhi Centre for Atomic Research. Acoustic measurement technique has been used to determine the drop time of Diverse Safety Rods in sodium. 3 nos of Diverse Safety Rods (DSRs) are provided in Prototype Fast Breeder Reactor (PFBR) for its safe shut down in case of a SCRAM. An online drop time measurement system using acoustic technique is planned to detect the proper insertion of DSRs into their corresponding Subassemblies. Experiments were conducted during the performance testing of DSRs in sodium using accelerometer instrumented wave-guide system and free fall time and braking time of DSR have been measured. For detection of in-sodium water leaks in Steam Generators, acoustic method serves as a supplementary monitoring technique with an intermediate sensitivity and instantaneous response. To develop an acoustic leak detection system for Steam Generators of Prototype Fast Breeder Reactor, preliminary studies on the behavior of micro leak and its propagation has been carried out in Sodium Water Reaction Test Rig, injecting steam into sodium. Acoustic technique was employed to characterize the onset of leak. Cavitation in LMFBRs can occur in fuel subassemblies, pressure drop devices, pumps etc. It is important to minimize cavitation to reduce the risk of damage from erosion. Acoustic technique was extensively used in qualifying Prototype Fast Breeder Reactor components against cavitation phenomenon. This paper discusses the various experiments carried out towards the development of the acoustic surveillance methods for FBRs, instrumentation involved, results obtained from experiments and brief details of the future programme.
The Study of Instrument Channel Precision and Application in Nuclear Power Qinshan Joint Venture Company Limited
Authors: Yi Tang
Published: July 2009

This paper studies the factors that effect instrument channel precision in Nuclear Power Qinshan Joint Venture Company Limited (NPQJVLC) and the methods to improve the precision. Also, it describes the attempts to improve channel precision that have been used in NPQJVLC and the effect on the maintenance strategy. It aims at helping other plants to improve instrument channel precision and get an optimum maintenance plan for nuclear power engineering.
Developing a Standard Code for Spent Fuel and High Level Waste Containment Baskets
Authors: Lewis Donald Wayne
Published: July 2009

ASME Section III, Division 3, “Containments for Transportation and Storage of Spent Nuclear Fuel and High Level Radioactive Material and Waste” currently addresses the design of transportation and storage containment shells but it has yet to address the containment internal support structure that holds the spent fuel or high level waste in place. However, the code for internal support structures, hereafter referred to by its common name “basket”, has been under development by ASME for the past 2 years. Development of the new code, to be known as Subsection WD, “Internal Support Structures” was deemed necessary because current containment system basket construction is a piecemeal approach using ASME Section III, Division 1, Subsection NF, “Supports” and/or ASME Section III, Division 1, Subsection NG, “Core Support Structures” or some other engineering method. Approvals for the various combinations are granted from the regulatory authority. The piecemeal approach tries to capture the critical elements important for a containment basket. However, Subsections NF and NG are based on nuclear power plant design which has different design goals than for a spent fuel or high level waste containment. The issuance of Subsection WD will ensure standardization of future containment baskets, assist the regulatory agency in the review and approval of the baskets, and ensure that the essential criteria in the basket related to spent fuel and high level waste storage transportation and disposal is adequately addressed. The purpose of the basket is primarily to ensure that the radioactive components in the containment are supported in a way as not to create a criticality event. Current acceptance is typically based on a no yield design that the containment manufactures all say is too conservative and based on unreasonable criteria. What should the basket design be based on, how should Subsection WD address them, etc.? The purpose of this paper is to inform interested parties of the progress that has been made in development of Subsection WD, what construction provisions it will initially include and what is planned for it, and when is it scheduled to be issued.
Effects of Mesh Density and Flow Conditioning in Simulating 7-Pin Wire Wrapped Fuel Pins
Authors: Smith Jeffrey G. , Babin Bruce R. , Pointer W. David , Fischer Paul F.
Published: July 2009

In response to the goals outlined by the U.S. Department of Energy’s Global Nuclear Energy Partnership program, Argonne National Laboratory has initiated an effort to create an integrated multi-physics multi-resolution thermal hydraulic simulation tool package for the evaluation of nuclear power plant design and safety. As part of this effort, the applicability of a variety of thermal hydraulic analysis methods for the prediction of heat transfer and fluid dynamics in the wire-wrapped fuel-rod bundles found in a fast reactor core is being evaluated. The work described herein provides an initial assessment of the capabilities of the general purpose commercial computational fluid dynamics code Star-CD for the prediction of fluid dynamic characteristics in a wire wrapped fast reactor fuel assembly. A 7-pin wire wrapped fuel rod assembly based on the dimensions of fuel elements in the concept Advanced Burner Test Reactor [1] was simulated for different mesh densities and domain configurations. A model considering a single axial span of the wire wrapped fuel assembly was initially used to assess mesh resolution effects. The influence of the inflow/outflow boundary conditions on the predicted flow fields in the single-span model were then investigated through comparisons with the central span region of models which included 3 and 5 spans. The change in grid refinement had minimal impact on the inter-channel exchange within the assembly resulting in roughly a 5 percent maximum difference. The central span of the 3-span and 5-span cases exhibits much higher velocities than the single span case,, with the largest deviation (15 to 20 percent) occurring furthest away from the wire spacer grids in the higher velocity regions. However, the differences between predicted flow fields in the 3-span and 5-span models are minimal.
Fundamental Mechanisms of Component Degradation by Corrosion in Nuclear Power Plants of Russian Design
Authors: Saji G. , Yurmanov V. A. , Baranenko V. I. , Fedorova V. A. , Karzov G. , Timofeev B. T.
Published: July 2009

By focusing on NPPs of Western design (e.g. PWR and BWR), the first author (G. Saji) has established that ‘long cell action’ corrosion plays a pivotal role in practically all unresolved corrosion issues for all types of nuclear power plants as presented in a series of papers already published (1–9). The authors believe that a similar study of NPPs of Russian design, with their unique scientific and technological basis compared to Western plants, are important to illustrate that this mechanism can occur even with different materials, welding technology or operation (e.g. water chemistry control). Among all the differences, it is important to note that PWSCC per se does not seem to be occurring in VVER plants, although no specific reason has yet to be identified. In this paper, a detailed electrochemical assessment is first made on the behavior of ammonia-potassium water chemistry and structural materials at the normal operational temperature in the primary water of VVERs. The chemical and electrochemical characteristics of the ammonia in VVERS were found to be significantly different from those of PWRs which use the hydrogen water chemistry. However, the water chemistry of RBMK is not fundamentally different from that of the Western BWR and therefore the previous studies on SCC of BWRs are generally applicable. On the bases of these studies, various corrosion issues commonly experienced in NPPs of Russian design (VVER and RBMK) are briefly reviewed. They include: (i) pitting corrosion in un-clad VVER-440 RV; (ii) corrosion cracking at the transition welding joints of RV nozzles and piping; (iii) corrosion issues in PGV-440 steam generator collectors; (iv) steam generator tube and collector corrosion; (v) IGSCC in RBMK with austenitic steel piping; (vi) FAC (E-C) in the secondary system of VVERs; and (vii) Anomalous corrosion products sedimentation in the core region in some VVERs. Since the long cell action hypothesis does not seem to contradict the various corrosion activities being experienced in NPPs of Russian design, the first author invites further study on the potential involvement of this mechanism since this hypothesis provides new insight into many of the unresolved corrosion issues. More specifically, the VVERs’ ammonia-potassium water chemistry has theoretically been identified as playing a key role in the prevention of PWSCC, which is one of the most troublesome mechanism of corrosion degradation in many Western PWRs. In view of this significance, the authors proposed an urgent international joint initiative to prove or disprove this mechanism’s existence in nuclear power systems.
Flow Regime Identification in Large Diameter Pipe
Authors: Sawant Pravin , Schelegel Joshua , Paranjape Sidharth , Ozar Basar , Hibiki Takashi , Ishii Mamoru
Published: July 2009

Air-water vertical two-phase flow experiments were performed in a 0.15 m diameter and 4.4 m long test section. Superficial liquid velocity was varied from 0.05 m/s to 2.0 m/s and superficial gas velocity was varied to obtain the area averaged void fraction range of 0.1 to 0.7. Exit pressure was close to the atmospheric pressure. In order to study the development of flow structure over the length of test section, area averaged void fraction was measured using impedance meters at four different measuring ports. Pressure drop was also measured between these ports. Since the temporal variation of void fraction signal obtained from the impedance meter and its distribution are characteristic of the flow regime, a Cumulative Probability Distribution Function (CPDF) of the void fraction signal was utilized for the identification of flow regime at each port. The CPDFs of the impedance probe void fraction signal were supplied as an input to the Kohonen Self Organized neural network or the Self Organized Map (SOM) for the identification of the patterns by employing self organized neural network technique. The three flow regimes identified by the neural network are subjectively named as bubbly flow, cap-bubbly flow and cap-turbulent flow.
The Study on the Applicability of Laser Surface Modification Technology to Irradiated Stainless Steel
Authors: Sato Masatoshi , Kanno Masanori , Nakata Kiyotomo , Takahashi Hidenori , Sakamoto Hiroshi
Published: July 2009

Recently, occurrence of stress corrosion cracking has been reported at core shrouds in Boiling Water Reactor (BWR) nuclear power plants. Yttrium aluminum garnet (YAG) laser surface modification technologies (i.e. Laser Surface Melting Technology (LSM), Laser Cladding Technology (LC)) have been developed as promising preventive maintenance technologies to stress corrosion cracking (SCC) of austenitic stainless steel structures and components. On the other hand, it has been also well-known that the helium transmuted from nickel and boron is accumulated to neutron irradiated stainless steel, and that helium related cracks may occur at weld heat affected zone which were attributed to nucleation along grain boundaries, coalescence and growth of helium bubbles due to thermal cycle and thermal stress during welding. Then, the laser surface modification technologies to the irradiated stainless steels was developed and the applicability of these technologies was evaluated based on the results of various tests (e.g. dye-penetrant test, micro structure observation and bending test) to the laser surface modified Type 304 and Type 316L specimens containing up to about 10 appm helium. The laser surface modification applicability diagram was developed as a function of weld heat input and helium concentration, which was supported by numerical simulation on helium bubble formation and growth during welding for irradiated stainless steels.
Conceptual Thermal-Design Options for Pressure-Tube SCWRs With Thermochemical Co-Generation of Hydrogen
Authors: Mokry Sarah , Naidin Maria , Baig Farina , Gospodinov Yevgeniy , Zirn Udo , Bakan Kurtulus , Pioro Igor , Naterer Greg
Published: July 2009

Currently there are a number of Generation IV SuperCritical Water-cooled nuclear Reactor (SCWR) concepts under development worldwide. The main objectives for developing and utilizing SCWRs are: 1) To increase gross thermal efficiency of current Nuclear Power Plants (NPPs) from 33–35% to approximately 45–50%, and 2) To decrease the capital and operational costs and, in doing so, decrease electrical-energy costs (∼$1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to current NPPs (i.e., pressures of about 25 MPa and outlet temperatures up to 625°C). Additionally, SCWRs will have a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. will be eliminated. Furthermore, SCWRs operating at higher temperatures can facilitate an economical co-generation of hydrogen through thermo-chemical cycles (particularly, the copper-chlorine cycle) or direct high-temperature electrolysis. To decrease significantly the development costs of a SCW NPP and to increase its reliability, it should be determined whether SCW NPPs can be designed with a steam-cycle arrangement that closely matches that of mature SuperCritical (SC) fossil power plants (including their SC turbine technology). The state-of-the-art SC steam cycles in fossil power plants are designed with a single-steam reheat and regenerative feedwater heating and reach thermal steam-cycle efficiencies up to 54% (i.e., net plant efficiencies of up to 43% on a Higher Heating Value Basis). It would be beneficial if SCWRs could involve a regenerative feedwater heating and nuclear steam reheat to be able to adapt the current SC turbine technology and to achieve similar high thermal efficiencies as the advanced fossil steam cycles. The nuclear steam reheat is easier to implement inside pressure-tube or pressure-channel reactors compared to pressure-vessel reactors. Atomic Energy of Canada Limited (AECL) and Research and Development Institute of Power Engineering (RDIPE or NIKIET in Russian abbreviations) are currently developing concepts of the pressure-tube SCWRs. Therefore, no-reheat, single-reheat, and double-reheat cycles of future SCW NPPs were analyzed in terms of their thermal efficiencies. On this basis, several conceptual steam-cycle arrangements of pressure-tube SCWRs, their corresponding T-s diagrams and steam-cycle thermal efficiencies are presented in this paper together with major parameters of the copper-chlorine cycle for the co-generation of hydrogen. Also, bulk-fluid temperature and thermophysical properties profiles were calculated for a non-uniform cosine Axial Heat-Flux Distribution (AHFD) along a generic SCWR fuel channel, for reference purposes.
The Applicability of NFI-1 DNB Correlation and Fluid-to-Fluid Similarities to Freon DNB Test
Authors: Kakinoki Shumpei , Matsuura Keizo , Kitagawa Kenichi , Kataoka Isao
Published: July 2009

Freon thermal hydraulic test is expected to be one of the workable methods to develop high thermal hydraulic performance PWR fuel. That is, high pressure water and high heat flux condition in PWR core can be substituted with lower pressure Freon and lower heat flux by applying appropriate fluid-to-fluid similarity and modeling parameters. Freon DNB tests and mixing tests were carried out against a 4×4 rod bundle configuration where R-134A flowed vertically upwardly. The tests were carried out at Freon thermal hydraulic test loop in Korea Atomic Energy Research Institute (KAERI). The spacer grid used in these tests was modeled on that of conventional PWR fuel, that is, square lattice grid with split type mixing vanes. Diameter of heater rod simulating PWR fuel rod is about 10.7mm and heating length is about 2000 mm. Freon mixing tests were carried out to estimate Turbulence Diffusivity Coefficient (TDC), which was normally used in conventional thermal hydraulic design of nuclear reactor. Freon CHF test results showed that parametric trends agreed with those of existing CHF data. To predict CHF of 4×4 rod bundle, subchannel analysis code Modified COBRA-3C and NFI-1 DNB correlation were applied. TDC value used in subchannel analysis was determined by fitting Freon mixing test data. NFI-1 DNB correlation was developed for predicting DNB heat flux in rod bundle configuration by using water CHF test results at HTRF test loop at Columbia University. The design of spacer grids used in KAERI Freon DNB test was similar to that used in water CHF test at HTRF. Water equivalent flow condition of this R-134A test was estimated using fluid-to-fluid similarities. NFI-1 DNB correlation was applied to this water equivalent condition to estimate water equivalent DNB heat flux. Then R-134A equivalent DNB heat flux was estimated reversely, and compared to Freon DNB test result. The test results were predicted well and applicability of NFI-1 DNB correlation and fluid-to-fluid similarities in 4×4 rod bundle is discussed.
Reliability of Water Jet Peening as Residual Stress Improvement Method for Alloy 600 PWSCC Mitigation
Authors: Okimura Koji , Konno Takao , Narita Marekazu , Ohta Takahiro , Toyoda Masahiko
Published: July 2009

As a countermeasure against high residual stress, we have developed some residual stress improvement methods, as Water Jet Peening (WJP) [1] [2] for components installed in water, Shot Peening by Ultrasonic-wave vibration (USP) for components installed in air, and outer surface irradiated Laser Stress Improvement Process (L-SIP) [3] for components being able to approach from outer surface only. WJP is applied to Reactor Vessel (RV) outlet/inlet nozzle safe-end joints (Alloy600 weld metal), RV Bottom Mounted Instrument (BMI) inner surface and J-weld. Especially, it is difficult to apply BMI because BMI inner surface is very narrow space (inner diameter; approximately 10–15mm) and BMI J-weld is complicated 3-dimensional form. On the occasion of actual application, we carry out the verification tests and check that a stress improvement was effective as one of PWSCC mitigation. And the compressive stress induced by WJP is verified to continue to exist under actual plant operation conditions. Thus, in addition to replacing the material with Alloy 690, converting the residual stress to the compressive can prevent the occurrence of PWSCC.
R&D of the Next Generation Safety Analysis Methods for Fast Reactors With New Computational Science and Technology: 4 — Experimental Analyses by SIMMER-III for the Integral Verification of COMPASS
Authors: Yamano Hidemasa , Tobita Yoshiharu
Published: July 2009

This paper describes experimental analyses using the SIMMER-III computer code, which were precedently carried out to give boundary conditions for the integral verification of the new COMPASS code, which is based on MPS method. Two topics of key phenomena in core disruptive accidents were presented in this paper: molten fuel freezing and dispersion; and boiling behavior of molten fuel pool. Related experimental database are reviewed to select appropriate experiments. To analyze the fuel freezing behavior, the GEYSER out-of-pile and the CABRI-EFM1 in-pile experiments were selected. The SIMMER-III calculations were in good agreement with fuel penetration lengths measured in a series of the GEYSER experiments. The fuel freezing behavior in the CABRI-EFM1 experiment was also reasonably simulated by SIMMER-III. The boiling pool consisting principally of molten fuel/steel mixtures is characterized by the heat transfer between fuel and steel. The CABRI-TPA2 experiment has suggested low transient heat flux from fuel to steel due to a steel vapor blanketing around a steel droplet. SIMMER-III well simulated the steel boiling behavior observed in the CABRI-TPA2 experiment by applying reduced heat transfer coefficients between fuel and steel. These experimental analyses by SIMMER-III have also identified key processes to be clarified by mesoscopic simulations using the COMPASS code.
Hybrid Methods Approach for Solving Problems in Transport Theory
Authors: Panta Pazos Rube´n
Published: July 2009

In this work the hybrid methods approach is introduced in order to solve some problems in Transport Theory for different geometries. The transport equation is written as: ∂ψ∂t(x,v,t)+v·∇ψ(x,v,t)+h(x,μ)ψ(x,v,t)==∫Vk(x,v,v′)ψ(x,v′,t)dv′+q(x,v,t),inΩTψ(x,v,0)=φ0(x,v),in∂Ω×Vψ(x,v,t)=φ(x,v,t),in∂Ω×V×R(1) where x represents the spatial variable in a domain D, v an element of a compact set V, ψ is the angular flux, h(x, v) the collision frequency, k(x, v, v’) the scattering kernel function and q(x, v) the source function. If ψ does not depend on the time, it is said that the problem (1) is a steady transport problem. Once the problem is defined, including the boundary conditions, it is disposed a set of chained methods in order to solve the problem. Between the different alternatives, an optimal scheme for the resolution is chosen. Two illustrations are given. For two-dimensional geometries it is employed a hybrid analytical and numerical method, for transport problems: conformal mapping first, then the solution in a proper geometry (rectangular for example). Each of the following two techniques is then applied, Krylov subspaces method or spectral-LTSN method. For three-dimensional problems also it is used a hybrid analytical and numerical method, for problems with more complex geometries: a homotopy between the original boundaries (piecewise surfaces) and another (a parallelepiped for example). Then each of two techniques are applied, Krylov subspaces method or nodal-LTSN method. In this case, the design of new geometries for reactors is a straightforward task. En each case, the domain consist of three regions, one of the source, other is the void region and the third one is a shield domain. The results are obtained both with an algebraic computer system and with a language of high level. An important extension is the study and treatment of transport problems for domains with irregular geometries, between them Lipschitzian domains. One remarkable fact of this work is the combination of different modeling and resolution techniques to solve some transport problems.
Recirculation Pump Speed Control System Upgrades in Nuclear Boiling Water Reactors
Authors: Morgan James W.
Published: July 2009

The nuclear power industry is faced with determining what to do with equipment and instrumentation reaching obsolescence and selecting the appropriate approach for upgrading the affected equipment. One of the systems in a nuclear power plant that has been a source of poor reliability in terms of replacement parts and control performance is the reactor recirculation pump speed/ flow control system for boiling water reactors (BWR). All of the operating BWR-3 and BWR-4’s use motor-generator sets, with a fluid coupled speed changer, to control the speed of the recirculation water pumps over the entire speed range of the pumps. These systems historically have had high maintenance costs, relative low efficiency, and relatively inaccurate speed control creating unwanted unit de-rates. BWR-5 and BWR-6 recirculation flow control schemes, which use flow control valves in conjunction with two-speed pumps, are also subject to upgrades for improved performance and reliability. These systems can be improved by installing solid-state adjustable speed drives (ASD), also known as variable frequency drives (VFD), in place of the motor-generator sets and the flow control valves. Several system configurations and ASD designs have been considered for optimal reliability and return on investment. This paper will discuss a highly reliable system and ASD design that is being developed for nuclear power plant reactor recirculation water pump controls. Design considerations discussed include ASD topology, controls architecture, accident, transient and hydraulic analyses, potential reactor internals modifications, installation, demolition and economic benefits.
Residual Stress Fields Under Different Clamping Conditions in Circumferentially Welded Thin-Walled Cylinders
Authors: Malik Afzaal M. , Qureshi Ejaz M. , Dar Naeem Ullah , Khan Iqbal
Published: July 2009

Arc welding is a reliable joining method widely utilized in nuclear, pressure vessels, aerospace and aeronautical structures to ensure the intended in service behaviour during the thermal and/or pressure loadings. Weld induced deformations and high residual stresses often occur during the course of welding. These cause significant threats for the structural integrity of the nuclear power plant components, particularly in stress corrosion inhibited environments owing to the risk of stress corrosion cracking (SCC). In this research, the consequences of five different structural boundary conditions on the evolution of residual stress fields after the welding are investigated. Both experimental and numerical simulations based on finite element modeling are employed during the course of investigation. Full three-dimensional FE models for the circumferentially, arc welded thin-walled cylinders are developed in ANSYS®. The complex coupled, thermo-mechanical phenomenon during the welding is simulated by sequentially coupled approach enhanced by user written APDL subroutines. The role of welding restraints in minimizing / optimizing the residual stresses is presented and discussed in detail. The result reveals that residual stresses show weak dependence on the degree of the restraints. Although the stress levels slightly varies in magnitude, but similar trend is observed for all the structural clamping conditions under study. Simulation results validated through full-scale experiments with high-tech reliably instrumented welding and measuring equipments shows promising features of the developed modelling and simulation strategy for use in shop floor applications.
Recent Developments on Seismic PRA
Authors: Coman Ovidiu , Onisoru Justin
Published: July 2009

Recent developments of the Probabilistic Risk Assessment (PRA) applications have tendency to use one integrated model for Level 1 both internal and external events. In this approach external events logic models are mapped on internal model. Some relevant aspects related to selection of seismic initiating events, seismic sequences, seismic basic events, fragility analysis, development of seismic logic model and seismic quantification are discussed in the paper. Development of seismic model mapped on internal PRA model may require to define seismic basic events (and associated switches) that may replace a group of internal basic events (at the level of system function or top event). Seismic fragility task is very resource intensive especially for high quality SPRA. When more than one seismic scenario is used significant effort can be saved by scaling fragility parameters developed for one earthquake scenario to correspond to the other earthquake scenario. The paper presents some tools developed by the authors capable to perform damage state fragility calculations and fragility scaling.
CFD Simulations of Boron Mixing in ESBWR Core Bypass Regions During an Anticipated Transient Without Scram (ATWS) Event
Authors: Tallman James A. , Marquino Wayne , Alamgir M. D.
Published: July 2009

CFD was used to simulate boron mixing in the ESBWR core during a limiting ATWS event. During the ESBWR design process, questions were raised regarding the Standby Liquid Control System (SLCS), which adds a sodium penta-borate solution to the core during ATWS events for the purpose of halting reactivity and core shutdown. The questions involved the design prediction of boron transport and mixing in the core bypass spaces, and how that influenced the core shutdown time prediction associated with the SLCS. In response, a CFD simulation was performed to study the mixing of the sodium penta-borate solution in bypass spaces of the ESBWR core. The CFD solution was a transient, Reynolds-Averaged Navier-Stokes (RANS) solution using two-equation turbulence modeling and 25 million computational nodes. Compared with the design predictions, the CFD results showed a much faster transport of boron toward the center of the core and a much greater quantity of boron is ingested into the fuel bundles. These findings demonstrate an adequate level of conservativism in the core shutdown time predictions, and confirm the effectiveness of the ESBWR SLCS system.
Towards Leap-Ahead Advances in Radiation Detection
Authors: Lapinskas J. R. , Archambault B. C. , Wang J. , Webster J. A. , Zielinski S.
Published: July 2009

Tension metastable fluid states offer unique potential for leap-ahead advancements in radiation detection. Such metastable fluid states can be attained using tailored resonant acoustics to result in acoustic tension metastable fluid detection (ATMFD) systems. ATMFD systems are under development at Purdue University. Radiation detection in ATMFD systems is based on the principle that incident nuclear particles interact with the dynamically tensioned fluid wherein the intermolecular bonds are sufficiently weakened such that even fundamental particles can be detected over eight orders of magnitude in energy with intrinsic efficiencies far above conventional detection systems. In the case of neutron-nuclei interactions the ionized recoil nucleus ejected from the target atom locally deposits its energy, effectively seeding the formation of vapor nuclei that grow from the sub-nano scale to visible scales such that it becomes possible to record the rate and timing of incoming radiation (neutrons, alphas, and photons). Nuclei form preferentially in the direction of incoming radiation. Imploding nuclei then result in shock waves that are readily possible to not only directly hear but also to monitor electronically at various points of the detector using time difference of arrival (TDOA) methods. In conjunction with hyperbolic positioning, the convolution of the resulting spatio-temporal information provides not just the evidence of rate of incident neutron radiation but also on directionality — a unique development in the field of radiation detection. The development of superior intrinsic-efficiency, low-cost, and rugged, ATMFD systems is being accomplished using a judicious combination of experimentation-cum-theoretical modeling. Modeling methodologies include Monte-Carlo based nuclear particle transport using MCNP5, and also complex multi-dimensional electromagneticcum-fluid-structural assessments with COMSOL’s Multi-physics simulation platform. Benchmarking and qualification studies have been conducted with Pu-based neutron-gamma sources with encouraging results. This paper summarizes the modeling-cum-experimental framework along with experimental evidence for the leap-ahead potential of the ATMFD system for transformation impact on the world of radiation detection.