bluuebunny/crossref_metadata_2025 huggingface dataset:
Simulation of Convective Flows in Sinusoidal Partitioned Enclosure
Authors: Anilkumar S. H. , Kuzhiveli Biju T.
Published: February 2010

A numerical study is carried out for natural convective flow and heat transfer in a two-dimensional enclosure with centrally located sinusoidal thin partition for a range of Rayleigh numbers, partition heights and aspect ratios. The partition is oriented parallel to the two vertical isothermal walls and the other surfaces are insulated. The flow and temperature distributions are taken to be two-dimensional. Transport equations are modeled by a stream function-vorticity formulation and are solved numerically by finite-difference approach. Comparisons with previously published numerical and experimental works are done and found to be in excellent agreement. The Rayleigh number varies from 103 to 106 and aspect ratio from 0.5 to 5. The results are presented for different fluids in the form of streamlines, vectors and isotherm plots. The variation of local Nusselt number over the sinusoidal partition and isothermal wall provide valuable insight into the physical processes.
An Angular Finite Element Method for the Solution of the Even-Parity Equation
Authors: Becker R. , Koch R. , Modest M. F. , Bauer H.-J.
Published: February 2010

The present article introduces a new method to solve the radiative transfer equation (RTE). First, a finite element discretization of the solid angle dependence is derived, wherein the coefficients of the finite element approximation are functions of the spatial coordinates. The angular basis functions are defined according to finite element principles on subdivisions of the octahedron. In a second step, these spatially dependent coefficients are discretized by spatial finite elements. This approach is very attractive, since it provides a concise derivation for approximations of the angular dependence with an arbitrary number of angular nodes. In addition, the usage of high-order angular basis functions is straightforward. In the current paper the governing equations are first derived independently of the actual angular approximation. Then, the design principles for the angular mesh are discussed and the parameterization of the piecewise angular basis functions is derived. In the following, the method is applied to two-dimensional test cases which are commonly used for the validation of approximation methods of the RTE. The results reveal that the proposed method is a promising alternative to the well-established practices like the Discrete Ordinates Method (DOM) and provides highly accurate approximations. A test case known to exhibit the ray effect in the DOM verifies the ability of the new method to avoid ray effects.
An Opto-Thermal Micro-Disk Resonator for Temperature Detection
Authors: Rahman Anisur , Kumar Sunil
Published: February 2010

In this paper, a temperature sensor has demonstrated using the opto-thermal disk-resonator (OTDR). The sensor is designed based on Whispering Gallery Mode (WGM) principle. The OTDR is a novel technique to integrate thermal science and optical technology. Optical resonances of Transverse Electric (TE) wave propagations in dielectric micro-disk resonator have been developed based on asymptotic approach. The developed expressions for size parameter for TE wave is very simple and can be used to characterize the resonances in dielectric micro-disk resonator. Asymptotic expression has applied to predict the resonance frequency at different temperatures. The sensor performance has been studied by using finite element simulation and found encouraging result for temperature sensing. Theoretical analysis shows that present study will lead to development of novel OTDR sensors for MEMS and nano applications.
Material Parameters Influencing the Radiative Properties of Heterogeneous Optically Thick Oxide Ceramics
Authors: Rousseau Benoit , Gomart Hector , De Sousa Meneses Domingos , Echegut Patrick
Published: February 2010

In this work is described a hybrid numerical code aiming to retrieve the thermal radiative properties of opaque heterogeneous coatings. The code is based on the combination of a Monte Carlo Ray Tracing (MCRT) program and of the use of the Effective Medium Approximation (EMA). This hybrid code is able to takes account for the surface roughness, the internal porosity and the chemical composition of a given layer. The solid part of a coating is composed of a conducting oxide which contents excess oxygen (δ ∼ 0.1–0.2). Due to δ and to their thicknesses (> 2 μm), the coatings are opaque for the spectral range going from the far infrared range to the visible range. For each layers, their normal spectral emissivity have been measured by infrared spectrometry (T ∼1000K). The comparison of the numerical spectra to the experimental ones permits the discussion of the effects of both the surface roughness, and the internal porosity on the emissivity spectra.
Laser Flash Measurement of Samples With a Transparent Reference Layer
Authors: Ban Heng , Hua Zilong
Published: February 2010

The laser flash method is a standard method for thermal diffusivity measurement. This paper reports the development of a method and theory that extends the standard laser flash method to measure thermal conductivity and thermal diffusivity simultaneously. By attaching a transparent reference layer with known thermal properties on the back of a sample, the thermal conductivity and thermal diffusivity of the sample can be extracted from the temperature response of the interface between the sample and the reference layer to a heating pulse on the front surface. The theory can be applied for sample and reference layer with different thermal properties and thickness, and the original analysis of the laser flash method becomes a limiting case of the current theory with an infinitely small thickness of the reference layer. The uncertainty analysis was performed and results indicated that the laser flash method can be used to extract the thermal conductivity and diffusivity of the sample. The results can be applied to, for instance, opaque liquid in a quartz dish with silicon infrared detector measuring the temperature of liquid-quartz interface through the quartz.
Investigating the Effect of Fouling on TEMA-Type Heat Exchangers’ Selection
Authors: Al-Otaibi Abdullah M.
Published: February 2010

Fouling has been a very serious problem in heat exchangers ever since their invention. It would require increasing heat exchanger nominal size by up to 30% to compensate for the losses in heat transfer due to fouling accumulation on the heat transfer surface. The large energy losses lead to more research in this field, and coming up with new technologies that target the reduction of these losses. This paper investigates the intent of having two similar TEMA-type shell-and-tube heat exchangers, S- and T-type configuration, utilized in industrial operations. Since the main reason for having these two types are the need to facilitate maintenance activities of fouled shell-side’s shell and tube heat exchangers, this paper assesses achieving this challenge when using the proposed heat exchanger type. This paper also presents a comparison of using the two heat exchanger types regarding different factors. The main factors considered are the cost (initial, operation, time and repair), and the utilization of new technologies appearing in the field of shell-and-tube heat exchangers to mitigate fouling. This paper also employs some actual heat exchanger problems at oil and gas operations as an example of fouling mitigation at S- and T-type heat exchangers.
Exact Solution of Two-Dimensional Hyperbolic Heat Conduction Equation With Combined Boundary Conditions and Arbitrary Initial Conditions
Authors: Mahmoudi Seyed Reza , Toljic Nikola
Published: February 2010

Analytical solution of hyperbolic heat conduction equation has so far been limited only to one-dimensional frameworks. With the expanding of the application range for the fast heat sources in microelectronic industries, the two-dimensional solution of non-Fourier heat conduction becomes increasingly important. This paper presents an exact solution of hyperbolic heat conduction equation for a finite plane with sides subjected to combined boundary conditions. In the mathematical model, the heating on boundaries is treated as an apparent heat source whereas sides of the plane with the second kind boundaries are assumed to be insulated. The important characteristic of the proposed solution is its simplicity.
Analysis of FIV Characteristics on a Coaxial Double-Tube Type Hot Gas Duct for the VHTR
Authors: Song Kee-nam , Kim Yong-wan , Park S.-C.
Published: February 2010

The Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source of the order of 950°C for nuclear hydrogen generation, which can produce hydrogen from water or natural gas. A primary hot gas duct (HGD) as a coaxial double-tube type cross vessel is a key component connecting the reactor pressure vessel and the intermediate heat exchanger in a VHTR. In this study, a structural sizing methodology for the primary HGD of a VHTR is suggested in order to modulate a flow-induced vibration (FIV). And as an example, a structural sizing of a horizontal HGD with a coaxial double-tube structure was carried out using the suggested method. These activities include a decision of the geometric dimensions, a selection of the material, and a evaluation of the strength of the coaxial double-tube type cross vessel components. Also in order to compare the FIV characteristics of the proposed design cases, a fluid-structure interaction (FSI) analysis on a quarter part of the HGD was carried out using the ADINA code.
Engineering, Design, and Fabrication of Flanges and Reinforced Elbows for an Essential Service Water (ESW) Polyethylene Pipeline
Authors: Svetlik Harvey
Published: February 2010

30 years ago, steel pipe was the standard in the ASME code for raw water supply and raw water handling at Nuclear Power Plants. At some power plants, despite best efforts, that steel pipe has suffered intermittent leakage from external and severe MIC internal corrosion. Additionally, internal tuberculation and mineral build-up has severely constricted flow in other pipelines. Advanced, engineering pipe-grade polyethylene pipe has been extruded and used in some nuclear power plants as the effective method to eliminate corrosion and tuberculation of raw water system pipelines. Implementing the change to earth-quake tolerant polyethylene pipelines has resulted in decreased maintenance, increased system reliability, and improved plant longevity. The expectation is that the advanced polyethylene will provide continuous service up to 100 years from initial installation, with a very low statistical probability of any pressure rupture during its service life. Herein discussed is the engineering grade of polyethylene material, its design basis, the conversion of extruded heavy-wall pipe (Picture 1) into fabricated components, and the final production of fully pressure-rated, fabricated pipe fittings with wall thickness of up to 5-inches. Fabrication pictures 1 to 12 are included.
Flow Visualization of Acoustic Resonance for Safety Relief Valves in Power Uprated BWRs
Authors: Okuyama Keita , Takahashi Shiro , Tamura Akinori , Nishida Koji , Ohtsuka Masaya , Tsubaki Masaaki
Published: February 2010

The boiling water reactor (BWR-3) in the Quad Cities Unit 2 Nuclear Power Plant (QC-2) experienced a significant increase in steam moisture under extended power uprate conditions (EPU). Inspection of the steam dryer showed it was likely to have been damaged by high cycle fatigue carry over, while the cause of the dryer failure was considered as flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines. Acoustic resonance was considered to be generated by interaction between the sound field and unstable shear layers across the closed side branches of the SRV stub pipes. From the dryer failure experience in QC-2, it became apparent that quantitative evaluation of the dryer loading was important for BWRs which were undergoing EPU planning. Acoustic resonance frequency can be evaluated using the geometry of the SRV stub pipes and the speed of sound. To predict the fluctuation pressure, it is necessary to consider a feedback mechanism between unsteady vorticity fluctuation and aerodynamic sound. Although theoretical approaches to sound propagation have been reported, experiments focused on vorticity are quite limited. In this study, in order to clarify this mechanism, a particle image velocimetry system (PIV) is used to visualize the flow field first, and then its effect on the wall pressure fluctuation is estimated by measuring the pressure fluctuation at the top of the stub pipe with a pressure sensor. Vortex growth from the leading edge of the cavity is confirmed and its shedding frequency is close to the frequency of the fluctuating pressure at the top of the stub pipe. Additionally, the pattern of maximum vorticity versus Strouhal number corresponds to the fluctuating pressure at the top of the stub duct. Thus, detected vorticity near the trailing edge of the cavity is the aerodynamic sound source.
Calculation and Measurement of Kinetics Parameters of Tehran Research Reactor
Authors: Hosseini Seyed Abolfazl , Vosoughi Naser , Gharib Mortaza , Ghofrani Mohammad Bagher
Published: February 2010

Effective delayed neutron fraction βeff and neutron generation time Λ are important factors in reactor physics calculation and transient analysis. In first stage of this research, these kinetics parameters have been calculated for two states of Tehran research reactor (TRR), i.e. cold (fuel, clad and coolant temperature 20°C) and hot (fuel, clad and coolant temperature 65, 49 and 44°C, respectively) using MTR_PC code. In second stage, these parameters have been measured with experimental method based on Inhour equation. For cold state, calculated βeff and Λ by MTR_PC are 0.008315 and 30.190 μsec, respectively. Same parameters in hot state are 0.008303 and 33.828 μsec, respectively. The measured βeff and Λ for cold state (reactor power is range of 100–200 Watt) are 0.008088 and 32.001 μsec, respectively. The calculated and measured values are in good agreement. Relative errors are % 2.8 for βeff and % 5.6 for Λ which are smaller than the other reported results.
On Enhancing Water Heat Transfer by Nanofluids
Authors: Shimkevich Alexander L.
Published: February 2010

Conventional coolants such as water have, in general, poorer heat-transfer properties than most liquid metals. A clear need exists today to develop new strategies in improving the effective heat-transfer behavior of the water coolant. Recent development in technology has created a new class of liquids called “nanofluids” which are two-phase mixtures of liquid matrix and dispersed systems consisting of nanoparticles usually less than 50 nm in size. A material of these particles can be a metal, oxide, or another compound. The subject of this research is a microstructure modification of fine suspension in liquids. Such the studying can be useful for motivated choosing a composition of coolant suspension, developing technology for the effective nanofluid operation in the first circuit of pressure water reactors (PWR). These investigations are important for purposeful modifying the water composition, its deterministic operation under non-isothermal conditions. In this paper, the author has attempted to estimate an effect of a fractal structure of nanoparticles for enhancing thermal conductivity of water.
Development of Observation Techniques in Reactor Vessel of Experimental Fast Reactor Joyo
Authors: Takamatsu Misao , Imaizumi Kazuyuki , Nagai Akinori , Sekine Takashi , Maeda Yukimoto
Published: February 2010

During the investigation of an incident that occurred with the experimental fast reactor Joyo, In-Vessel Observations (IVO) using a standard Video Camera (VC) and a Radiation-Resistant Fiberscope (RRF) took place at (1) the top of the Sub-Assemblies (S/As) and the In-Vessel Storage rack (IVS), (2) the bottom face of the Upper Core Structure (UCS) under the condition with the level of sodium at −50 mm below the top of the S/As. A simple 6 m overhead view of each S/A, through the fuel handling or inspection holes etc, was photographed using a VC fixed to the Rotating-Plug (R/P) with the acrylic panel for making observations of the top of S/As and IVS. About 650 photographs were required to create a composite photograph of the top of the entire S/As and IVS, and a resolution was estimated to be approximately 1mm. In order to observe the bottom face of the UCS, a Remote Handling Device (RHD) equipped with RRFs (approximately 13 m long) was specifically developed for Joyo with a tip that could be bent into an L-shape and inserted into the 70 mm gap between the top of the S/As and the bottom of the UCS. A total of about 35,000 photographs were needed for the full investigation. Regarding the resolution, the sodium flow regulating grid of 0.8mm in thickness could be discriminated, and the base of thermocouple sleeves 6 mm in diameter located 450 mm above the top of S/As were also clearly observed. In both types of observations, it was confirmed that lighting adjustments play a critical role. Particularly in narrow space observations, scattered lighting with automatic dimming controlled light source was available for achieving close observations of the in-vessel structures. In addition to the successful result of the incident investigation, these experiments provided valuable insights for use in further improving and verifying in-vessel observation techniques in Sodium cooled Fast Reactors (SFRs).
Development of Control Rod Position Indicator Using Seismic-Resistance Reed Switches
Authors: Yu Je-Yong , Kim Ji-Ho , Huh Hyung , Choi Myong-Hwan , Sohn Dong-Seong
Published: February 2010

The system-integrated reactor currently under development at the Korea Atomic Energy Research Institute is being designed with a soluble boron free operation and the use of a nuclear heating for the reactor start-up. These design features require a Control Element Drive Mechanism (CEDM) for the integral reactor to have a fine-step movement capability as well as a high reliability for a fine reactivity control. Also the reliability and accuracy of the information for the control rod position is very important to the reactor safety as well as the design of the core protection system. The position indicator is classified as a Class 1E component because the rod position signal of the position indicator is used in the safety related systems. Therefore it will be separated from the control systems to the extent that a failure of any single control system component of a channel and will have sufficient independence, redundancy, and testability to perform its safety functions assuming a single failure. The position indicator is composed of a permanent magnet, reed switches and a voltage divider. Four independent position indicators around the upper pressure housing provide an indication of the position of a control rod comprising of a permanent magnet with a magnetic field concentrator which moves with the extension shaft connected to the control rod. The arranged reed switches are positioned along a line parallel to the path of the movement of the permanent magnet and it is activated selectively when the permanent magnet passes by. The traditional reed switch is sealed in a glass tube filled with an inert gas. So it is easily damaged during an assembly and test. Therefore, this research adopts seismic-resistance reed switches to the position indicator in order to reduce the damages or impacts during the handling of the position indicator and earthquake. The control rod position indicator having a seismic-resistance characteristic for the integral reactor was developed and tested. The performance test of a position indicator with a compact arrangement of the seismic-resistance reed switches for a high resolution was successfully carried out. The results of this study have shown that a position indicator including seismic-resistance reed switches and a compact arrangement can be applied effectively to the integral reactor CEDM after the environmental tests.
Development of a Remote Welding Machine for a DUPIC Fuel Bundle Fabrication
Authors: Kim Soo-sung , Park Geun-il , Lee Jung-won , Koh Jin-hyun
Published: February 2010

A remote welding machine for a DUPIC (Direct Use of spent PWR fuel In CANDU reactors) bundle fabrication was designed to establish the optimum welding processes in a hot cell environment. An initial investigation for hands-on fabrication outside the hot cell was performed in the consideration of the constraints of welding machine in the hot cell conditions. Gas tungsten arc welding (GTAW), laser beam welding (LBW), friction welding (FW), and resistance welding (RW) process were assessed as candidates for this application. A preliminary welding investigation to improve the RW process was also performed. The RW process was determined to be the most suitable process in a hot cell environment for joining an endplate to an endcap. An advantage of RW would be a qualified process for overlapped plates welding for which there is extensive production experience were available. A preliminary investigation for a hands-on fuel fabrication outside the hot cell was conducted in the consideration of the constraints caused by a welding in a hot cell. The optimum resistance welding parameters for the endplate welding process were obtained in terms of the current, electrode pressure and welding cycle. This paper presents an outline of the developed RW machine for a DUPIC bundle fabrication and reviews the conceptual design of a remote RW welder by using a master-slave manipulator. The design of RW machine by using the 3D modeling method was also designed. Furthermore the integrity of the welds by the resistance welding was confirmed by the results of the torque test, an examination of the microstructure and the fracture surfaces of the welds.
Simulation of Containment Thermal-Hydraulic Phenomena in the Marviken Blowdown Experiment With the CONTAIN and ASTEC CPA Codes
Authors: Kljenak Ivo , Mavko Borut
Published: February 2010

The Boiling Heavy Water Reactor Blowdown 16 Experiment, which was performed in the Marviken experimental facility, was simulated with the CONTAIN and ASTEC CPA codes. The main purpose of the work was the assessment of the codes for simulating thermal-hydraulic phenomena in a BWR containment at accident conditions. Simulated pressures, atmosphere temperatures and wetwell pool characteristics are compared to experimental measurements. The results show that both codes satisfactorily reproduced the containment thermal-hydraulic behavior.
System-Core Thermal Hydraulic Interaction During a Reflood Phase of a Cold-Leg LBLOCA in the Atlas Integral Effect Test Facility
Authors: Cho Seok , Park Hyun-Sik , Choi Ki-Yong , Kang Kyong-Ho , Kim Yeon-Sik , Baek Won-Pil
Published: February 2010

A series of experimental tests to simulate a reflood phase of a cold-leg LBLOCA of the APR1400 have been performed by using the ATLAS facility. This paper describes related experimental results with respect to the thermal-hydraulic behavior in the core and the system-core interactions during the reflood phase of cold-leg LBLOCA condition. The present descriptions will be focus on the LB-CL-09, LB-CL-11, LB-CL-14, and LB-CL-15 tests performed by using the ATLAS. The LB-CL-09 is the integral effect test with conservative boundary condition, and LB-CL-11, and -14 are the integral effect tests with realistic boundary condition, and the LB-CL-15 is the separated effect test. The objectives of these tests are to investigate the thermal-hydraulic behavior during an entire reflood phase, and to provide reliable experimental data for validating the LBLOCA analysis methodology for the APR1400. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results for the LBLOCA scenario of the APR1400. The ECC water flow rate from the safety injection tanks and the decay heat were simulated from the start of the reflood phase. The present experimental data showed that the cladding temperature behavior is closely related to the collapsed water level in the core and the downcomer.
Design and Fabrication of the FBR Fuel Disassembly System
Authors: Kitagaki Toru , Tasaka Masayuki , Higuchi Hidetoshi , Koizumi Kenji , Hirano Hiroyasu , Washiya Tadahiro , Kobayashi Tsuguyuki
Published: February 2010

Japan Atomic Energy Agency (JAEA) has been developing a reliable disassembly system for FBR fuel reprocessing as a part of Fast Reactor Cycle Technology Development (FaCT). As FBR fuel pins are installed in a hexagonal shaped wrapper tube made of stainless steel, the fuel pins should be separated from the wrapper tube prior to the shearing process. JAEA has been developing the laser beam cutting method and the mechanical cutting method as the disassembly system. Although Fiber laser system showed a good cutting performance, it couldn’t completely avoid fuel pin damage and adhesion during the cutting operation. So we focused on the mechanical method to minimize such troubles. Two types of mechanical cutting modes have to be developed to realize the disassembly procedure, namely, the slit-cut for the wrapper tube and the crop-cut for the end plug region of the fuel pin bundle. To ensure disassembly technology of commercial reactor fuel assemblies, we designed and fabricated the testing machine of disassembly system having the cutting modes in engineering scale. We confirmed basic functions of this machine and improved its performance. We will soon demonstrate engineering operation by a series of disassembling and pin bundle handling procedure; separating fuel pins from wrapper tube, transferring them to the fuel magazine for shearing. Scattering of cutting dust cause machine troubles and transition of it to the dissolution process together with pins causes unknown problems. To resolve the problems, collection device of cutting dust will be tested and the cutting condition to make the disassembly easy to cut will be improved.
Corrosion Behavior of FBR Structural Materials in High Temperature Supercritical CO2
Authors: Furukawa Tomohiro , Inagaki Yoshiyuki , Aritomi Masanori
Published: February 2010

A key problem in the application of a supercritical carbon dioxide (CO2) turbine cycle to a fast breeder reactor is the corrosion of structural materials by supercritical CO2 at high temperature. In this study, corrosion tests on the candidate materials, high-chromium martensitic steel (12Cr-steel) and FBR grade type 316 stainless steel (316FR), were performed for up to approximately 2000h at 400–600°C in supercritical CO2 pressurized at 20MPa. Corrosion due to the high temperature oxidation was measured in both steels. Results showed that the behavior differed greatly. For 12Cr-specimens, weight gain showed parabolic growth as exposure time increased at each temperature. The oxidation coefficient could be estimated by the Arrhenius function. The specimens were covered by two successive layers of oxide, an Fe-Cr-O layer and an Fe-O layer. A partial thin oxide diffusion layer appeared between the base metal and the Fe-Cr-O layer. The corrosion behavior was equivalent to that in supercritical CO2 at 10MPa, and no effects of CO2 pressure were observed in this study. For 316FR specimens, the weight gain was significantly lower than that of 12Cr-specimen, and good resistance against corrosion was observed. No dependency of temperature or immersed time on weight gain was observed under the test conditions, and the value of all tested specimens was within 2g/m2. Some nodule shape oxide was observed on the surface of the 316FR specimen. Carburizing, known as a factor in the occurrence of breakaway corrosion and/or the degradation of ductility, was observed on the surface of both steels.
Development of Safety Assessment Code for Decommissioning of Nuclear Facilities (DecDose)
Authors: Shimada Taro , Ohshima Soichiro , Sukegawa Takenori
Published: February 2010

A safety assessment code, DecDose, for decommissioning of nuclear facilities has been developed, based on the experiences of the decommissioning project of Japan Power Demonstration Reactor (JPDR) at Japan Atomic Energy Research Institute (now Japan Atomic Energy Agency). DecDose evaluates the annual exposure dose of the public and workers according to the progress of decommissioning of the plant, and also evaluates the public dose at accidental situations including fire and explosion. The public dose at normal situations during decommissioning is evaluated from the amount of radionuclides discharged from the plant to the atmosphere and the ocean. The amounts of radionuclides discharged depend on which and how activated and/or contaminated components and structures are dismantled. The amount is predicted by using the radioactive inventory given by the plant. The filtration efficiency of the ventilation system and decontamination factors of the liquid waste treatment system of the plant are also considered. Both of the internal dose caused by inhalation and ingestion of agricultural crops and seafood, and the external dose by radioactive aerosols airborne and radioactive deposition at soil surfaces are calculated for all of possible pathways. Also included in the external dose are direct radiation and skyshine radiation from waste containers which are packed and temporarily stored in the in-site building. For external dose of workers, the radiation dose rate from dismantling contaminated components and structures is calculated using the dose rate library which was previously evaluated by a point kernel shielding code. In this condition, radiation sources are regarded to be consisted of two parts; one is a dismantling object of interest, and the other is the sum of surrounding objects. Difference in job type or position is taken into account; workers for cutting are situated closer to a dismantling component, other workers help them at some distance, and the supervisor watches their activities from away. For worker’s internal dose, the radionuclide concentrations in air for individual radionuclides are calculated from a dismantling condition, e.g. cutting speed, cutting length of the dismantling component and exhaust velocity. A calculation model for working time on dismantling was developed using more segmented WBS (work breakdown structure). DecDose was partially verified by comparison with measured the external dose of workers during JPDR Decommissioning Project. The DecDose is expected to contribute to utilities in formulating rational dismantling plans and to the safety authority in estimating conservativeness in safety assessment of licensing application or risk-based regulatory criteria.
Ageing Management Status of HANARO Research Reactor
Authors: Kim Hyungkyoo , Jung Hoansung , Lim Incheol , Cho Yeonggarp , Ryu Jeongsoo , Park Youngchul
Published: February 2010

Ageing means time-dependent changes of the quality and design characteristics having existed at the time of an approval, and ageing may occur during the operating time of a facility. The management of equipment begins with an awareness that ageing is not about how old the equipment is, but is about what is known about its condition, and the factors that influence the onset, evolution and mitigation of its degradation. HANARO, the open-tank-in-pool type research reactor with 30MW thermal power, achieved its first criticality in 1995. The reactor power has been gradually increased to the full power reached in June 2004. Many activities for an ageing management including improvement of the reactor systems were conducted during 13 years. The in-service inspection plans, for the reactor structure and the safety related piping designed as ASME safety category III, were established and carried out. The preventive maintenance for the primary cooling system including the pumps and heat exchangers, the electric system was also performed. The safety diagnosis and reinforcement for an effective life management and insurance of the safety of the building structures such as the reactor building, stack, secondary cooling system equipment room and cooling tower in HANARO were performed. The upgrade of the instrumentation and control system has been carried out gradually since 2001. This paper describes the conducted ageing management activities and long term ageing management plan of HANARO.
Hydraulic Modeling of Air-Entraining Vortex Formation During Flow Withdrawal From Water Storage Tanks
Authors: Johansson Andrew , Padmanabhan Mahadevan , Cain Stuart , Meyer Bryan , Schowalter David
Published: February 2010

In response to recent U.S. Nuclear Regulatory Commission (NRC) Component Design Bases Inspection (CBDI) issues, many US nuclear power stations have been required to demonstrate that minimum submergence requirements were properly determined for flow withdrawal from various safety related storage tanks. In many cases, the licensees failed to consider a vortex allowance, or applied an inappropriate vortex methodology. For Duke Energy’s McGuire Station, a Refueling Water Storage Tank (RWST) model was constructed using a geometric scale of 1:4.073. Testing included transient water level conditions simulating the field for selected flows (corresponding to prototype flows of 1,600 to 19,700 gpm) and water levels giving submergences of 1 to 5 ft above the suction nozzle in the model (prototype submergences of 4 to 20.3 ft). Results showed that with no return flow, the submergence at the onset of air entrainment ranged from 0.049 to 0.705 ft prototype for flows ranging from 1,600 to 19,700 gpm prototype, respectively. Based on the test results, it was determined that a vortex suppression device was not required for the McGuire RWST, as the expected water levels during operation would be higher than those indicated for onset of air entrainment for a given flow. The scale model testing showed that the critical submergences for initiation of air-entraining vortices were much lower than those predicted by Hydraulic Institute guidelines.
Assessment of Post-CHF Heat Transfer Models for the SPACE Code
Authors: Choi K. Y. , Yun B. J. , Park H. S. , Moon S. K. , Kim K. D.
Published: February 2010

The SPACE (Safety and Performance Analysis Code) which is based on a multi-dimensional two-fluid, three-field model is under development for a licensing purpose of pressurized water reactors in Korea. A total of 12 wall-to-fluid heat transfer modes were defined and a heat transfer mode selection logic was developed according to the noncondensable gas quality, the void fraction, the degree of subcooling and the wall temperature. Among the constitutive equations of the SPACE code, post-CHF heat transfer models are considered to have the most significant uncertainties because their physical phenomena are not fully understood yet. Though a variety of models and correlations for the post-CHF heat transfer are available, there is no reliable model to reproduce the heat transfer rate as realistically as possible. Several post-CHF heat transfer models are implemented into the SPACE code; critical heat, transition and film boiling models. The present paper describes the model assessment progress which was done for the transition and film boiling models of the SPACE code. A heat flux partition into the continuous liquid, entrained droplet and vapor fields should be taken into account in order to be in line with physical phenomena as the SPACE code has three-field equations in its hydraulic solver. Existing energy partitioning methods were peer-reviewed in order to determine the best model which can be applicable to the SPACE code. The present work will help to consolidate the developed wall-to-fluid heat transfer package of the SPACE code.
Transient Flow Field on Containment Floor Predicted by Two-Dimensional Shallow Equation Solver
Authors: Bang Young Seok , Lee Gil-Soo , Huh Byung-Gil , Oh Deog-Yeon , Woo Sweng-Woong
Published: February 2010

For the analysis of debris transport on containment floor, a model to predict the flow field should have a fast-running capability and high accuracy. A model is developed to calculate the transient flow field on the containment floor involving a complex geometry in the advanced pressurized water reactor (PWR) such as Advanced Power Reactor (APR)-1400, which does not have a switchover from injection to recirculation following a loss-of-coolant accident (LOCA). Two-dimensional shallow water equation (SWE) is solved using the finite volume method (FVM). Unstructured triangular meshes are used to simulate the complex structures on the containment floor. Harten-Lax-van Leer (HLL) scheme, one of the approximate Riemann solver, is adopted to capture the dry-wet interface and to determine the momentum flux at the interface. An experiment of a sudden dam break having water reservoir and L-shape open channel is simulated and compared with the calculated result, which supports the validity of the present model. The model is also applied to calculation of the flow field of APR-1400. The calculated flow field can be characterized by the propagation of waves generated by surface level difference and by the reflection of waves from solid wall. The transient flow rates entering to the Holdup Volume Tank (HVT) can be predicted within a practical limit of computational resource.
CFD Application of the Diffusion-Inertia Model to Bubble Flows and Boiling Water Problems
Authors: Filippov Alexander S. , Alipchenkov Vladimir M. , Drobyshevsky Nickolay I. , Mukin Roman V. , Strizhov Valeri Th. , Zaichik Leonid I.
Published: February 2010

The paper is aimed at the application of a model for simulating the dispersed turbulent flows. The model presented proceeds from a kinetic equation for the probability density function of the particle velocity distribution in turbulent flow. This approach is called the diffusion-inertia model (DIM). Applications of the model to droplet and bubble flows are presented. In the case of vaporized liquid, the interphase heat and mass transfer is introduced by adding the corresponding governing equations. This extended version of the DIM was applied to simulating the boiling water flow in a heated pipe.
Nondestructive Evaluation of Neutron Irradiation Damage on Austenitic Stainless Steels by Measurement of Magnetic Flux Density
Authors: Takaya Shigeru , Nagae Yuji , Aoto Kazumi , Yamagata Ichiro , Ichikawa Shoichi , Konno Shotaro , Ogawa Ryuichiro , Wakai Eiichi
Published: February 2010

Magnetic flux densities for neutron irradiated specimens of austenitic stainless steels, SUS304 and Fast Breeder Reactor grade type 316 (316FR), were measured by using a flux gate (FG) sensor to investigate the nondestructive evaluation method of irradiation damage parameters, dose and He content. Specimens were irradiated in each one of the experimental fast reactor JOYO, the Japan Materials Testing Reactor, and the Japan Research Reacter-3M (JRR-3M), or in both of JRR-3M and JOYO (coupling irradiation). Irradiation in various reactors and the coupling irradiation provided irradiation conditions which could be hardly obtained by irradiation in a single reactor. The range of dose, He content and irradiation temperature of the neutron irradiated samples studied in this paper were 0.01–30 displacement per atom (dpa), 1.0–17 appm and 470–560 °C, respectively. Magnetic flux density increased with dose although there may be a threshold dose for magnetic property to change between 2 and 5 dpa for 316FR. This result shows the possibility of nondestructive evaluation of dose by measuring magnetic flux density by an FG sensor. On the other hand, magnetic flux density did not depend on He content.
Frequency Domain Linear Stability Analysis of Simplified Supercritical Water-Cooled Reactor
Authors: Liu Pengfei , Yang Yanhua , Xue Aijun , Cheng Xu
Published: February 2010

This paper presents the stability analysis of a supercritical water-cooled system. A model of simplified supercritical water-cooled system is introduced and then its thermal-hydraulic equations, initial conditions and boundary conditions are given. Based on perturbation linearization and Laplace transformation, the transfer function between inlet mass flow rate oscillation and pressure drop oscillation of the simplified system model is established, and the characteristic equation of the simplified system model is derived. Applying control theory to solve the characteristic equation, the stability boundary points are found by judging whether the real parts of all roots solved by characteristic equation are greater than zero. A stability map which consists of these stability boundary points is generated by using both dimensionless sub-pseudo-critical number and trans-pseudo-critical number. An unstable region nearby the pseudo-critical point is determined. The effects of some important parameters on the stability map are investigated also by using decay ratio. The sensitivity analysis shows that the system is stabilized with a higher hydraulic system resistance, fluid inlet velocity or system pressure. It also shows that a longer heating zone, a harder pump characteristics or a larger gravitational acceleration (orientation angle) leads to a larger stability margin of the system. The stability map is also found to be not sensitive to a higher friction resistance or system pressure.
A Review of Neural Network Application for Fault Diagnosis of NPP
Authors: Wang He
Published: February 2010

Artificial Neural Network (ANN) with its self-learning capabilities, nonlinear mapping ability and generalization ability, has been widely applied for fault diagnosis of complex system like Nuclear Power Plant (NPP). In this paper, an overview of the application of supervised multi-layer feed-forward neural network for fault diagnosis of NPP is presented, including the following aspects: the acquisition of the training sample data, the determination of appropriate input and output data, the choice of hidden layer structure and the evaluation of network model performance. Finally, a number of key issues about the engineering application of neural network fault diagnosis in practice were discussed.
Non-Linear Characteristics of Saturated and Subcooled Pool Boiling CHF for Various Test Heater Configuration
Authors: Fukuda Katsuya , Liu Qiusheng
Published: February 2010

Many aspects of saturated and subcooled pool boiling CHF in liquids have been investigated by many researchers for pressure, subcooling, test heater configurations, thermal properties, etc., assuming a CHF model mainly based on a kind of hydrodynamic instability at CHF with a linear relation of CHF for subcooling at a pressure for about a half century until now. Recently it was clarified that the CHFs measured were mainly divided into two groups for lower and higher subcooling at a pressure as a typical case. It was assumed that the former and latter CHFs occur due to hydrodynamic instability (HI) and explosive-like heterogeneous spontaneous nucleation (HSN) respectively. It was attempted to clarify in the present paper the effect of test heater configurations on the saturated and subcooled pool boiling CHFs resulting from the HI and the HSN using existing data for horizontal test cylinders with the diameters ranging from 0.2 to 5 mm, a vertical cylinder, a vertically oriented rectangular ribbon and a horizontally suspended vertically oriented rectangular ribbon in liquids by the authors and other researchers. The CHF correlations representing the CHFs resulting from the HI and the empirical correlation resulting with HSN including test heater configuration effects respectively were suggested.
Modernization Project of Ikata Unit 1/2 Main Control Room Using Full Digital Control Board
Authors: Sakuma Tomohide , Kitamura Kazuo , Okamoto Hiroki , Nakao Tadashi , Watanabe Hiroshi
Published: February 2010

The Mitsubishi Group and Japanese PWR utilities have been engaged in a continual process of main control board design improvement. Advanced main control boards with fully digital instrumentation and touch operation technology have been adopted in newly constructed plants. Moreover, instrumentation and control systems have been fully digitalized and applied not only to non-safety systems but to safety systems as well. Such a system in general is referred to as a fully digital system. From this time forward, the application of this fully digital system will become mainstream with the modernization of existing plants. Yet, unlike newly constructed plants, restrictions unique to existing plants will need to be taken into consideration. In the construction of Ikata Units 1 and 2 of Shikoku electric power co., measures such as those described below were implemented so as to achieve standardization with future construction in mind: (1) A thorough interface check was conducted to ensure compatibility with sections outside the scope of modernization. Additionally, to minimize the affect modernization had on the operating system, the system was verified by operators and verification results were reflected in the board layout, system specifications, and the like. (2) Due to limited space for board installation, a plan was devised for housing a processor and determining the remote input/output board location. (3) Onsite tests was minimized in order to complete construction within a short period of time, i.e., within the regularly scheduled inspection period, and the degree of completion of the product was enhanced prior to shipment by a full combination test conducted at the plant. In addition, the local construction period was shortened by devising a plan for installing the board and conducting interface tests during the previous regularly scheduled inspection and the current regularly scheduled inspection.
An Integrated Approach to Human System Interface Design in New Nuclear Power Plants
Authors: Viguera Jose´ Manuel , Jime´nez Alfonso , Burillo Juan Antonio
Published: February 2010

Nuclear power plants design is moving toward a wider use of digital computers, especially microprocessors, in information and control systems. The amount of automation and the role of the operator are under discussion in many countries. The view of the operator’s role presently varies. The main opinions can be summarized as follows: 1. Move toward a high degree of automation, fostering the machine role. 2. Use of computer-generated procedures providing information to skilled operators for them to make the final decision. 3. Use of digital systems to help the operator identify problems, decide on the appropriate corrective actions and aid in the execution of those actions. Tecnatom, S.A. has developed an integrated Human Factor Engineering (HFE) methodology, based on international regulations and experience obtained from several national and international projects, combining technology, organization and human elements to generate a Human-Centered Design. Human Factor Engineering (HFE) is the application of the knowledge of human capabilities and characteristics to develop equipment, facilities and systems. With the application of this knowledge, human performance, and therefore system performance, can be dramatically improved. Man/machine systems designed with the human as a key element are inherently safer and more reliable than those that are not. Until recently, design of these human-equipment interfaces has been secondary to “pure hardware” design; that is, equipment and facilities were designed without formal consideration of the implications for operators. Our approach is to systematically apply an HFE methodology that will produce: a) Human-System Interfaces that are easy, friendly to and consistent for the operators. b) Simulator-Assisted Engineering platforms for validation activities in the logic, control and human-system interface areas. c) Training Programs based on the systematic analysis of job and task requirements. d) Procedures derived from the same design process and analyses as the Human-System Interface and Training. Application of good HFE methodology during system development, implementation and operation is, from our point of view, vital for optimal system performance regarding operation activities. “A disciplined approach to HFE helps ensure that humans are considered integral system components, requiring careful consideration of how they will interact with their equipment.”