bluuebunny/crossref_metadata_2025 huggingface dataset:
Measurement and Calculation of Gamma Radiation from HWZPR Reactor
Authors: Jalali Majid
Published: September 2008

HWZPR is a research reactor with natural uranium fuel, D2O moderator and graphite reflector with maximum power of 100 W. It is a suitable means for theoretical research and heavy water reactor experiments. Neutrons from the core participate in different nuclear reactions by interactions with fuel, moderator, graphite and the concrete around the reactor. The results of these interactions are the production of prompt gammas in the environment. Useful information is gained by the reactor gamma spectrum measurement from point of view of relative quantity and energy distribution of direct and scattered radiations. Reactor gamma ray spectrum has been gathered in different places around the reactor by HPGe detector. In analysis of these spectra, 1H(n,γ)2H, 16O(n,n′γ)16O, 2H(n,γ)3H and 238U(n,γ)239U reactions occurring in reactor moderator and fuel, are important. The measured spectrum has been primarily estimated by the MCNP code. There is agreement between the code and the experiments in some points. The scattered gamma rays from 27Al (n,γ)28Al reaction in the reactor tank, are the most among the gammas scattered in the reactor environment. Also the dose calculations by MCNP code show that 72% of gamma dose belongs to the energy range 3–11 MeV from reactor gamma spectrum and the danger of exposure from the reactor high-energy photons is serious.
Optimal Preventive Maintenance Under Decision Dependent Uncertainty
Authors: Galenko Alexander , Popova Elmira , Kee Ernie , Grantom Rick
Published: September 2008

We analyze a system of N components with dependent failure times. The goal is to obtain the optimal block replacement interval (different for each component) over a finite horizon that minimizes the expected total maintenance cost. In addition, we allow each preventive maintenance action to change the future joint failure time distribution. We illustrate our methodology with an example from South Texas Project Nuclear Operating Company.
Devices to Mitigate Gas Entrainment in the Surge Tank of PFBR
Authors: Ramdasu D. , Shivakumar N. S. , Padmakumar G. , Anand Babu C. , Vaidyanathan G.
Published: September 2008

Surge tank is one of the important components in the secondary circuit of a typical sodium cooled fast breeder reactor, provided to take care of pressure surges in case of a sodium water reaction in Steam Generators (SG). The blanket of argon cover gas at the top of the tank acts as a cushion for the surges. The argon gas above the free surface of sodium in the tank is a source of entrainment into the sodium which is undesirable from the consideration of effective heat transfer in Inter mediate Heat Exchanger (IHX) and SG cavitation in pumps and operational problems of continuous feed and bleed of cover gas, thus leading to unfavourable reactor operating conditions. To investigate the phenomenon of gas entrainment in surge tank, hydraulic experiments were conducted in water using 1/12 scaled model. The minimum height of liquid column in the tank when gas entrainment is completely avoided was established. Different methods to mitigate gas entrainment were tested in the model and a combination of porous plate and stiffener ring was found to be optimum in reducing the liquid column required to mitigate gas entrainment in surge tank.
WWER Steam Generators Tubing Performance and Aging Management
Authors: Trunov Nikolay B. , Davidenko Stanislav E. , Grigoriev Vladimir A. , Popadchuk Valery S. , Brykov Sergey I. , Karzov Georgy P.
Published: September 2008

At WWER NPPs the horizontal steam generators (SGs), are used that differ in design concept from vertical SGs mostly used at western NPPs. Reliable operation of SG heat-exchanging tubes is the crucial worldwide problem for NPP of various types. According to the operation feedback the water chemistry is the governing factor affecting operability of SG tubing. The secondary side corrosion is considered to be the main mechanism of SG heat-exchanging tubes damage at WWER plants. To make the assessment of the tubing integrity the combination of pressure tests and eddy-current tests is used. Assessment of the tubing performance is an important part of SG life extension practice. The given paper deals with the description of the tube testing strategy and the approach to tube integrity assessment based on deterministic and probabilistic methods of fracture mechanics. Requirements for eddy-current test are given as well. Practice of condition monitoring and implementing the database on steam generators operation are presented. The approach to tubes plugging criteria is described. The research activities on corrosion mechanism studies and residual lifetime evaluation are mentioned.
Metastable Regimes: A Parametric Study in Reference to Single-Phase Parallel Channel Natural Circulation Systems
Authors: Gartia M. R. , Pilkhwal D. S. , Vijayan P. K. , Saha D.
Published: September 2008

The objective of the present study is to predict the existence of metastable regime (multiple steady states) and effects of various parameters on it, in a parallel channel system under natural circulation. To examine the basic behavior of such a system, a three parallel channel system with non-uniform heat inputs and a common downcomer, was investigated in detail both analytically and using RELAP5/Mod 3.2 computer code. The results showed the existence of a metastable regime for a given set of heat input ratio. A study was also performed to investigate the effect of diameter, channel height and orificing at the inlet of heater, on the metastable regime. A Non-Dimensional Heat Flux Ratio (NH) has been proposed to predict the flow reversal. The heat flux level at which flow reversal occurs has been predicted for different rates of power rise and power reduction (with different input heat flux). It was found that the power level at which the flow reversal occurs depends upon the operating procedures (i.e. power rise and power reduction). The flow direction in different channels will depend on the prevailing density difference among channels and downcomer. It was found that, the flow direction in a channel, where power variation is taking place, follows the maximum buoyancy hydraulic path.
Validation of a CFD Condensation Model Based on Heat and Mass Transfer Analogy by TOSQAN Facility ISP47 Test
Authors: Bucci Matteo , Ambrosini Walter , Forgione Nicola , Oriolo Francesco , Paci Sandro
Published: September 2008

The content of this paper is focused on a computational fluid dynamics analysis of the test performed within the facility TOSQAN as a part of the International Standard Problem 47 (ISP 47). The aim of the study is to contribute to the understanding of the heat and mass transfer mechanisms and to check the possibility to use a commercial CFD code for simulating the mass transfer phenomena of interest in nuclear reactor containment design and safety analysis. In this aim, the FLUENT 6.2 code has been used. The effect of the condensation rate onto the vessel walls was simulated by appropriate source terms introduced by user-defined subroutines into the mass, energy and momentum balance equations. In the paper the time trends of the average temperature and pressure of the atmosphere inside the TOSQAN vessel have been compared with the available experimental data, obtaining a good agreement. Spatial profiles have been also analysed and compared with the experimental ones for the main physical variables in the first, second and fourth steady-state phases which the test consists of.
Benchmarking and Qualification of PAC-Femlab for Resonant Acoustic Chamber Design
Authors: Butt A. , Xu Y. , Webster J. , Shaw P. , Taleyarkhan R. P.
Published: September 2008

The design of high-powered resonant acoustic systems capable of inducing large pressure oscillations in the 105 Hz to 106 Hz range requires a validated simulation platform, one that includes complexities of multi-dimensional fluid-structure interactions. Past efforts at designing such systems have relied mainly on time-consuming, trial-error based heuristic approaches (West et al., 1967; Taleyarkhan et al., 2002;2004). A robust design-cum-simulation platform is required to enable rapid strides and motivated this study for which the PAC-Femlab model was developed and successfully qualified against detailed experiment data as well as against data from a second independent experiment conducted elsewhere (Cancelos et al., 2004).
Low Burnup Inert Matrix Fuels Performance: TRANSURANUS Analysis of the Halden IFA-652 First Irradiation Cycle
Authors: Calabrese R. , Vettraino F. , Tverberg T.
Published: September 2008

The inert matrix fuels are a promising option to reduce-eliminate worldwide plutonium stockpiles by burning it in LWRs. These fuels, where plutonium is hosted in a U-free inert matrix phase, may reach high burning efficiency while preventing new plutonium build-up under irradiation. A specific investigation on CSZ and thoria inert matrices has been developed by ENEA since several years. In-pile testing on the ENEA-conceived innovative fuels is ongoing in the OECD Halden HBWR since June 2000 (IFA-652 experiment). The registered burnup at the end of 2005 is about 38 MWd·kgUeq−1 vs. 45 MWd·kgUeq−1 (40 MWd·kgUOXeq−1) target. Fuel pins are equipped with fuel temperature thermocouples, internal pressure transducers and fuel stack elongation sensors, with the task of studying thermal conductivity and its degradation with burnup, densification-swelling behaviour and the FGR. In this paper, the response at low burnup (< 7 MWd·kgUeq−1) of CSZ-based fuels loaded in IFA-652, is analysed by means of the TRANSURANUS code. To this purpose, a comprehensive modelling of the above mentioned un-irradiated fuels, mainly relying on the thermophysical characterisation performed at the JRC/ITU-Karlsruhe, has been implemented in a custom TRANSURANUS version (TU-IMF). A comparison of the code predictions vs. the experimental data, aimed at evaluating the early-stage under irradiation phenomena, particularly densification and relocation, has been performed.
Analysis of Flow Instabilities in Forced-Convection Steam Generator
Authors: Wu Ge Ping , Qiu Sui Zheng , Su Guang Hui , Jia Dou Nan
Published: September 2008

Because of the practical importance of two-phase instabilities, substantial efforts have been made to date to understand the physical phenomena governing such instabilities and to develop computational tools to model the dynamics. The purpose of this paper is to present a numerical model for the analysis of flow-induced instabilities in forced-convection steam generator. The model is based on the assumption of homogenous two-phase flow and thermodynamic equilibrium of the phases. The thermal capacity of the heater wall has been included in the analysis. The model is used to analysis of flow instabilities in steam generator and to study the effects of system pressure, mass flux, inlet temperature and inlet/outlet restriction, gap size, the ratio of do/di and the ratio of qi/qo on system behavior.
Experimental Research on Dryout Point of Flow Boiling in Narrow Annuli
Authors: Wu Ge Ping , Qiu Sui Zheng , Su Guang Hui , Jia Dou Nan
Published: September 2008

An experimental research on the dryout point of flow boiling in narrow annuli is conducted under low mass flux with 1.5mm and 1.0mm gap, respectively. Distilled water is used as work fluid, the range of pressure is from 2.0 to 4.0MPa and that of mass flux is 26.0∼69.0 kg/(m2 · s). The relation of CHF and critical qualities with mass flux and pressure are revealed. It is found that the critical qualities decrease with the mass flux and increase with the inlet qualities in externally heated annuli. Under the same conditions critical qualities in outer tube are always larger than that in inner tube. Кутателадзе’s correlations is cited and modified to predict the location of dryout and proved to be not a proper one. Considering in detail the effects of the geometry of annuli and heat flux on dryout, an empirical correction is finally developed to predict dryout point in narrow annuli under low mass flux condition which has a good agreement with experimental data.
Digital Instrumentation and Control Failure Events Derivation and Analysis by Frame-Based Technique
Authors: Huang Hui-Wen , Shih Chunkuan , Yih Swu , Tzeng Yen-Chang , Chen Ming-Huei
Published: September 2008

A frame-based technique, including physical frame, logical frame, and cognitive frame, was adopted to perform digital I&C failure events derivation and analysis for generic ABWR. The physical frame was structured with a modified PCTran-ABWR plant simulation code, which was extended and enhanced on the feedwater system, recirculation system, and steam line system. The logical model is structured with MATLAB, which was incorporated into PCTran-ABWR to improve the pressure control system, feedwater control system, recirculation control system, and automated power regulation control system. As a result, the software failure of these digital control systems can be properly simulated and analyzed. The cognitive frame was simulated by the operator awareness status in the scenarios. Moreover, via an internal characteristics tuning technique, the modified PCTran-ABWR can precisely reflect the characteristics of the power-core flow. Hence, in addition to the transient plots, the analysis results can then be demonstrated on the power-core flow map. A number of postulated I&C system software failure events were derived to achieve the dynamic analyses. The basis for event derivation includes the published classification for software anomalies, the digital I&C design data for ABWR, chapter 15 accident analysis of generic SAR, and the reported NPP I&C software failure events. The case study of this research includes (1) the software CMF analysis for the major digital control systems; and (2) postulated ABWR digital I&C software failure events derivation from the actual happening of non-ABWR digital I&C software failure events, which were reported to LER of USNRC or IRS of IAEA. These events were analyzed by PCTran-ABWR. Conflicts among plant status, computer status, and human cognitive status are successfully identified. The operator might not easily recognize the abnormal condition, because the computer status seems to progress normally. However, a well trained operator can become aware of the abnormal condition with the inconsistent physical parameters; and then can take early corrective actions to avoid the system hazard. This paper also discusses the advantage of Simulation-based method, which can investigate more in-depth dynamic behavior of digital I&C system than other approaches. Some unanticipated interactions can be observed by this method.
Performance Assessment of Sodium to Air Finned Heat Exchanger for FBR
Authors: Vinod V. , Suresh Kumar V. A. , Noushad I. B. , Ellappan T. R. , Rajan K. K. , Rajan M. , Vaidyanathan G.
Published: September 2008

In pool type Fast Breeder Reactors (FBR) a passive Safety Grade Decay Heat Removal (SGDHR) system removes decay heat produced in the core when normal heat removal path through steam water system is not available. This is essential to maintain the core temperatures within limits. A Decay Heat Exchanger (DHX) picks the heat from the pool and transfers the heat to atmosphere through sodium to Air Heat Exchanger (AHX) situated at high elevation. Due to the temperature differences existent in the system density differences are generated causing a buoyant convective heat transfer. The system is completely passive as primary sodium, secondary sodium and air flows under natural convection. DHX is a sodium to sodium counter flow heat exchanger with primary sodium on shell side and secondary sodium on tube side. AHX is a cross flow heat exchanger with sodium on tube side and air flows in cross flow across the finned tubes. Capacity of a single loop of SGDHR is 8MW. Four such loops are available for the decay heat removal. It has been seen that the decay heat removal to a large extent depends on the AHX performance. AHX tested have shown reduced heat removal capacity much as 30 to 40%, essentially due to the bypassing of the finned tubes by the air. It was felt that a geometrically similar AHX be tested in sodium. Towards this a 2MW Sodium to air heat exchanger (AHX) was tested in the Steam Generator Test Facility (SGTF) constructed at Indira Gandhi Center for Atomic Research (IGCAR), Kalpakkam. The casing arrangement of the AHX was designed to minimise bypassing of air.
Learning From the Best with Process Benchmarking in Nuclear Power Plants
Authors: Jansky Andy
Published: September 2008

Process benchmarking between power plants of different utilities has been a tedious task due to the difficulty to obtain significant and suitable data. Only independent consulting agencies with unrestricted access to sensitive operational data gained through projects in various comparable power plants can perform such a task. In order to conduct a holistic, integral benchmark analysis, the basis of such an analysis — the measurement data and key performance indicators — need to be determined. In a second step the verified data needs to be standardized to ensure comparability before the organizational structure and technical processes can be adjusted to the best-practice.
Identification of Error Sources in Fatigue Analyses for Thermal Loadings
Authors: Binder Franz , Gantz Dieter
Published: September 2008

To identify thermal loadings (thermal shocks and thermal stratification), in German NPPs special fatigue monitoring systems have been installed. The detailed temperature measurement uses sheathed thermocouples, which are located on the external component surface. Tightening straps are used for the widespread method of locking the thermocouples into position. The basic objective of the paper is to identify several sources of error in fatigue analyses and to evaluate their specific importance. In fatigue analyses based on the outer-surface temperature, the thermal situation at the inner-surface has to be determined. This leading analysis step is not regulated in the ASME III code. In the paper, the variance in the calculated fatigue is quantified for different inner surface temperature profiles with a common good compliance to the outer surface temperature. Another focus of the paper is the presentation of findings concerning the exactness of fatigue analyses based on FEA. The effect of geometrical truncation (mesh density) and the time-increments on the calculated fatigue level is shown. Furthermore, regarding discontinuities with thermal shock loading, the differences in the calculated fatigue between applying ASME NB-3200 on the one hand and NB-3600 on the other are quantified. Finally, the uncertainty in the calculated fatigue associated with the known variance of the physical material property values (E-Modulus, thermal conductivity, thermal expansion) is highlighted.
An Evaluation of the Hydrogen Flammability and the Hydrogen Concentration in the Reactor Building for Wolsong 2
Authors: Kim See Darl , Jin Youngho , Park Soo Yong , Song Yong Man
Published: September 2008

Following a core uncovery, the core will heat up and the fuel pin cladding will oxidize in the presence of steam. The reaction follows a parabolic rate law, and a significant cladding oxidation does not occur until the cladding temperature reaches approximately 1800° F. One of the products of the cladding oxidation reaction is hydrogen, which can accumulate in the PHTS if the PHTS is intact or in the reactor building if a pathway exists from the PHTS. If the hydrogen concentration reaches 4 to 6% in the reactor building (which is very possible following a significant core damage), the hydrogen can ignite and cause a spike in the reactor building pressure and temperature. If enough hydrogen is allowed to build up in the reactor building, the ignition of the hydrogen can lead to a pressure spike that exceeds the reactor building ultimate design pressure, thereby jeopardizing the reactor buildings integrity. Even if reactor building integrity is not jeopardized by a global hydrogen burn, the fact that a global hydrogen burn could occur will prevent the TSC from exiting the SAMG. Hydrogen generation and flammability is a concern throughout most of the guidelines in the Wolsong 2 (CANDU 6 type reactor) SAMG. The purpose of this paper is to define whether the hydrogen in the reactor building atmosphere is flammable, and to estimate the hydrogen concentration in the reactor building atmosphere based on an estimated oxidation percentage. For the Wolsong 2 this CA, four sets of hydrogen figures were developed for the case of no venting (no core/concrete interaction), a 15% venting (no core/concrete interaction), a 30% venting (no core/concrete interaction), and no venting (core/concrete interaction for 24 hours resulting in a 114°F superheat and a addition of non-condensable gases).
Estimation of Critical Flow Velocity for Collapse of Gas Test Loop Booster Fuel Assembly
Authors: Guillen Donna Post , Russell Mark J.
Published: September 2008

This paper presents calculations performed to determine the critical flow velocity for plate collapse due to static instability for the Gas Test Loop booster fuel assembly. Long, slender plates arranged in a parallel configuration can experience static divergence and collapse at sufficiently high coolant flow rates. Such collapse was exhibited by the Oak Ridge High Flux Reactor in the 1940s and the Engineering Test Reactor at the Idaho National Laboratory in the 1950s. Theoretical formulas outlined by Miller, based upon wide-beam theory and Bernoulli’s equation, were used for the analysis. Calculations based upon Miller’s theory show that the actual coolant flow velocity is only 6% of the predicted critical flow velocity. Since there is a considerable margin between the theoretically predicted plate collapse velocity and the design velocity, the phenomena of plate collapse due to static instability is unlikely.
Potential Advantages of Underground Nuclear Parks
Authors: Myers Carl W. , Elkins Ned Z. , Kunze Jay F. , Mahar James M.
Published: September 2008

In this paper we argue that an underground nuclear park (UNP) could potentially lead to lower capital and operating cost for the reactors installed in the UNP compared to the traditional approach, which would be to site the reactors at the earth’s surface at distributed locations. The UNP approach could also lead to lower waste management cost. A secondary benefit would be the increased margins of safety and security that would be realized simply as a consequence of siting the reactors underground. Lowered capital and operating cost for a UNP relative to traditional reactor siting is possible through the aggregate effect of the elimination of containment structures, in-place decommissioning, reduced physical security costs, reduced weather-related costs, reduced cost of liability insurance and reduced unit-cost for the nth reactor made possible through the continuous construction of multiple reactors at the same underground location. Other cost reductions might be possible through the transfer of the capital cost for part of the underground construction from the reactor owners to the owners of the UNP. Lower waste management cost is possible by siting the UNP at a location where there are geological and hydrological conditions suitable for hosting both the reactors and the repository for the waste from those reactors. After adequate storage and cooling, and assuming direct disposal, this would enable the spent fuel from the reactors to be transported directly to the repository and remain entirely underground during the transport process. Community concerns and transportation costs would be significantly reduced relative to current situations where the reactors are separated from the repository by long distances and populated areas. The concept for a UNP in bedded salt is used to develop a rough order of magnitude cost estimate for excavation of the reactor array portion of a UNP. Excavation costs appear to be only a small fraction of the overall power plant costs for an UNP in salt. Many engineering, safety, environmental, regulatory and cost-benefit and technical issues related to the UNP concept need to be evaluated.
POLCA-T: Validation of Transient Bundle Flow Predictions During an All Recirculation Pumptrip in a BWR
Authors: Bredolt Ulf
Published: September 2008

In this paper predicted results obtained with the POLCA-T code are compared to measurement data from an all recirculation pump trip test made at the Nuclear Power Plant Olkiluoto 1 during its commissioning in 1978. Olkiluoto 1 is one of three nuclear power plants owned by the electrical power company Teollisuuden Voima OY in Finland. The prediction is based on a best estimate reactor analysis code that incorporates a full 3D model of a reactor core into a system model. The use of a coupled code provides a means to simulate interactions between the reactor core and the plant dynamics to increase the understanding of different events and its impact on the core behaviour. The following conclusions are made based on the simulations with POLCA-T of the Olkiluoto 1 commissioning test of tripping all recirculation pumps: • POLCA-T predicts the event accurately, the flow coast down and the power decay is in good agreement with measured data. • 3D effects in channel flow can not be observed. The observed effects support the current use of a 1D approach since this is conservative with respect to power and the evaluation of Operating Limit Critical Power Ratio, (OLMCPR). This conclusion is also supported by measurement data.
Instantaneous Leakage Evaluation of Metal Cask at Drop Impact
Authors: Takeda Hirofumi , Kageyama Norio , Wataru Masumi , Sonobe Ryoji , Shirai Koji , Saegusa Toshiari
Published: September 2008

There have been a lot of tests and analyses reported for evaluation of drop tests of metal casks. However, no quantitative measurement has ever been made for any instantaneous leakage through metal gaskets during the drop tests due to loosening of the bolts in the containments and lateral sliding of the lids. In order to determine a source term for radiation exposure dose assessment, it is necessary to obtain fundamental data of instantaneous leakage. In this study, leak tests were performed by using scale models of the lid structure and a full scale cask without impact limiters simulating drop accidents in a storage facility, with aim of measuring and evaluating any instantaneous leakage at drop impact. Prior to drop tests of a full scale metal cask, a series of leakage tests using scale models were carried out to establish the measurement method and to examine a relationship between the amount of the lateral sliding of the lid and the leak rate. It was determined that the leak rate did not depend on the lateral sliding speeds. Drop tests of a full scale metal cask without impact limiters were carried out by simulating drop accidents during handling in a storage facility. The target was designed to simulate a reinforced concrete floor in the facility. The first test was a horizontal drop from a height of 1 m. The second test simulated a rotational impact around an axis of a lower trunnion of the cask from the horizontal status at a height of 1 m. In the horizontal drop test, the amount of helium gas leakage was calculated by integrating the leak rate with time. The total amount of helium gas leakage from the primary and secondary lids was 1.99×10−6Pa · m3. This value is 9.61×10−9% of the initially installed helium gas. The amount of leakage was insignificant. In the rotational drop test, the total amount of leakage from the primary and secondary lids was 1.74×10−5Pa·m3. This value is 8.45×10−8% of the initially installed helium gas. This value was larger than that of the horizontal drop test. Nevertheless, the amount of leakage was also insignificant. The relationship between the maximum sliding displacement of the lid and the leak rate coincided between the tests of a scale model and a full scale metal cask.
Study on Pool-Nucleate Boiling Heat Transfer Characteristics by Using Artificial Cavities
Authors: Hateruma Ryo , Sato Takato , Koizumi Yasuo , Ohtake Hiroyasu
Published: September 2008

Pool boiling heat transfer experiments were performed by using the well-controlled/defined heat transfer surface for water. Uni-size and -shape artificial cavities were created on the mirror-finished silicon plate by utilizing the MEMS technology. Experimental results agreed well with what were predicted by the traditional boiling theory. The mirror-finished surface showed only the tendency of natural circulation heat transfer. The artificial-cavity heat transfer surface followed the pool-nucleate boiling trend. The onset of the pool-nucleate boiling was well predicted by the traditional pool-nucleate boiling theory. These results indicated that the artificial cavities behave just like natural cavities. The results indicated the artificial cavities are quite useful and promising to examine the true features of complicated boiling that have been overshadowed by complicatedness. From recorded high speed video pictures, the coalescence of bubbles that were growing on the cavities were classified into four categories; the normal lift (no coalescence), the vertical coalescence, the declining coalescence and the horizontal coalescence. As the cavity interval was increased, the horizontal coalescence decreases to zero, the vertical coalescence also decreases, and on the contrary to these, vertical coalescence and normal lift increase. The cavity interval 3 mm (S/Lc ≈ 1.2) seemed to be the border whether the horizontal coalescence occurs or not.
Diagnosis of Thermal Efficiency of Nuclear Power Plants Using Optical Torque Sensors
Authors: Umezawa Shuichi , Adachi Jun
Published: September 2008

A new optical torque measuring method was applied to diagnosis of thermal efficiency of nuclear power plants. The sensor allows torque deformation of the rotor caused by power transmission to be measured without contact. Semiconductor laser beams and small pieces of stainless reflector that have bar-code patterns are employed. The intensity of the reflected laser beam is measured and then input into a computer through an APD and an A/D converter having high frequency sampling rates. The correlation analysis technique can translate these data into the torque deformation angle. This angle allows us to obtain the turbine output along with the torsional rigidity and the rotating speed of the rotor. The sensor was applied to a nuclear plant of Tokyo Electric Power Company, TEPCO, following its application success to the early combined cycle plants and the advanced combined cycle plants of TEPCO. As the turbine rotor of the nuclear power plant is less exposed than that of the combined cycle plants, the measurement position is confined to a narrow gap. In order to overcome the difficulty in installation, the shape of the sensor is modified to be long and thin. Sensor performance of the nuclear power plant was inspected over a year. The value of the torsional rigidity was analyzed by the finite element method at first. Accuracy was improved by correcting the torsional rigidity so that the value was consistent with the generator output. As a result, it is considered that the sensor performance has reached a practical use level.
Molecular Dynamic Simulation of Sodium in 7-Pin LMFBR Bundle Under Hypothetical Accident Conditions
Authors: Bottoni Maurizio , Bottoni Claudio , Scanu John
Published: September 2008

In the frame of safety analysis of liquid metal fast breeder reactors (LMFBRs) under hypothetical Unprotected Loss of Flow (ULOF) conditions two-phase flow of sodium is simulated in a 7-pin bundle, with hexagonal lattice. Molecular dynamics, with the application of the Direct Simulation Monte Carlo (DSMC) method, and a macroscopic model describing rewetting sequences due to the flow of a sodium liquid film along the pin surfaces, are applied to simulate the coolant in the bundle. The pin surfaces and the inner surface of the hexagonal canning are treated in the Monte Carlo simulation as diffusively reflecting surfaces. Collisions of sodium molecules are computed with the “hard-sphere” model. With respect to previous work the following improvements of the computational code were made: i) The full bundle is simulated, thus allowing for asymmetries, like a skewed power distribution, to be accounted for; ii) A pin model calculates detailed temperature distributions in the pins, so that temperature boundary conditions are computed and not imposed; iii) Post processing visualisation of computed results was developed. An out of pile sodium boiling experiment run at the Nuclear Research Center of Karlsruhe, Germany, is simulated and conclusions are drawn about the applicability of the methodology in computer codes dedicated to breeder reactors safety analysis.
Qualifying Elbow Meters for High Pressure Flow Measurements in an Operating Nuclear Power Plant
Authors: Chan A. M. , Maynard K. J. , Ramundi J. , Wiklund E.
Published: September 2008

To support the installation and use of elbow meters to measure the high pressure emergency coolant injection flow in an operating nuclear station, a test program was performed to qualify: (i) the “hot” tapping procedure for field application and (ii) the use of elbow meters for accurate flow measurements over the full range of station ECI flow conditions. This paper describes the design conditions and major components of a flow loop used for the elbow meter calibrations. Typical test results are presented and discussed.
Industrial Maintenance Strategies
Authors: Akbar Sajjad
Published: September 2008

Industrial plants have become more complex due to technological advancement. This has made the task of maintenance more difficult. The maintenance costs in terms of resources and downtime loss are so high that maintenance function has become a critical factor in a plant’s profitability. Industry should devote as much forethought to the management of maintenance function as to production. Maintenance has grown from an art to a precise, technical engineering science. Planning, organizing scheduling and control of maintenance using modern techniques pays dividends in the form of reduced costs and increased reliability. The magnitude and the dimension of maintenance have multiplied due to development in the engineering technologies. Production cost and capacities are directly affected by the breakdown time. Total operating cost including the maintenance cost plays an important role in replacement dimension. The integrated system approach would bring forth the desired results of high maintenance standards. The standards once achieved and sustained, would add to the reliability of the plan and relieve heavy stresses and strains on the engineering logistic support.
Wavelet Spatial Energy Spectrums Studies on Drag Reduction by Microbubble Injection
Authors: Zhen Ling , Hassan Yassin A.
Published: September 2008

In this study, continuous wavelet transforms and spatial correlation techniques are employed to determine the space-localized wavenumber energy spectrum of the velocity signals in turbulent channel flow. The flow conditions correspond to single phase flow and microbubbles injected two phase flow. The wavelet energy spectrums demonstrate that the wavenumber (eddy size) content of the velocity signals is not only space-dependent but also microbubbles can impact the eddy size content. Visual observations of the wavelet energy spectrum spatial distribution was realized by using Particle Image Velocimetry (PIV) measurement technique. The two phase flow condition corresponds to a drag reduction of 38.4% with void fraction of 4.9%. The present results provide evidence that microbubbles in the boundary layer of a turbulent channel flow can help adjust the eddy size distributions near the wall. This can assist in explaining that microbubbles are performing as buffers to keep the energy of fluid particles going in streamwise direction and reducing the energy of fluid particles going in normal direction.
The Results of 10 Years Periodic In-Service Inspections in HANARO
Authors: Park Yong-Chul , Lee Yong-Sup , Kim Bong-Soo
Published: September 2008

HANARO, an open-tank-in-pool type research reactor of 30 MWth power in Korea, has been operating normally since its initial criticality in February, 1995. For the last ten years, HANARO has carried out ten years periodic in-service inspections (ISI as below) in accordance with Article IWD in ASME SEC. XI to verify the mechanical and structural integrities of the pressure retaining components of the safety related systems and the integral attachments of the supports and restraints of the components which are NPS 4 and above, to be within a specified boundary. This paper describes the results of the ISI including a system pressure test and a VT-3 visual inspection. From the results, it was confirmed through the ISI that the pressure retaining components and parts were stable to within the specified boundaries for their mechanical and structural integrities.
Test Design, Results, and Archived Database for the OECD Lower Head Failure Program Integral Experiments
Authors: Humphries Larry L. , Chu Tze Yao , Bentz John H.
Published: September 2008

In the event of a severe core meltdown accident, core material can relocate to the lower head of a pressurized water reactor (PWR) vessel resulting in significant thermal and pressure loads to the vessel. The potential for failure of the pressure vessel makes possible the release of core material to the containment. The objective of this experimental/analytical program is to characterize the mode, timing, and size of lower head failure (LHF) under severe accident conditions. The OECD Lower Head Failure (OLHF) project investigates lower head failure for conditions of low reactor coolant system (RCS) pressure (2–5 MPa) and prototypic through-wall temperature differential (ΔTW >200K). Low RCS pressure is motivated by the desire to use the data to develop models for assessing accident management strategies involving reactor pressure vessel (RPV) depressurization. Pressure transient is useful in assessing the effect of water injection as part of accident management strategy. Prototypic through-wall temperature differential, ΔTW, is of importance because of the need to provide data where stress redistribution in the vessel wall occurs (as a result of decreasing material strength with temperature). Test design and results for the four OLHF integral tests are reported and summarized in this paper. A short description of the test conduct and heating history is followed by a description of the vessel failure site, the vessel deformation, temperature profiles, stress state, and rupture dynamics for each test. Key observations and conclusions are summarized for each test. The ∼1/5 scale tests are extensively instrumented to provide temperature, pressure, and displacement data. The vessel surfaces are mapped both before and after the test to provide measurements of pre-test thickness, post-test thickness, and cumulative vessel deformation. Data has been assessed and qualified in data reports for each test. The data has been preserved in MSEXCEL™ spreadsheets with macro utilities to facilitate access and analysis of the data. As a result, there exists a well-archived, well-qualified database for model development and validation.
Influence of Test Tube Material on Subcooled Flow Boiling Critical Heat Flux in Short Vertical Tube
Authors: Hata Koichi , Shiotsu Masahiro , Noda Nobuaki
Published: September 2008

The steady state subcooled flow boiling critical heat flux (CHF) for the flow velocities (u = 4.0 to 13.3 m/s), the inlet subcoolings (ΔTsub,in = 48.6 to 154.7 K), the inlet pressure (Pin = 735.2 to 969.0 kPa) and the increasing heat input (Q0exp(t/τ), τ = 10, 20 and 33.3 s) are systematically measured with the experimental water loop. The 304 Stainless Steel (SUS304) test tubes of inner diameters (d = 6 mm), heated lengths (L = 66 mm) and L/d = 11 with the inner surface of rough finished (Surface roughness, Ra = 3.18 μm), the Cupro Nickel (Cu-Ni 30%) test tubes of d = 6 mm, L = 60 mm and L/d = 10 with Ra = 0.18 μm and the Platinum (Pt) test tubes of d = 3 and 6 mm, L = 66.5 and 69.6 mm, and L/d = 22.2 and 11.6 respectively with Ra = 0.45 μm are used in this work. The CHF data for the SUS304, Cu-Ni 30% and Pt test tubes were compared with SUS304 ones for the wide ranges of d and L/d previously obtained and the values calculated by the authors’ published steady state CHF correlations against outlet and inlet subcoolings. The influence of the test tube material on CHF is investigated into details and the dominant mechanism of subcooled flow boiling critical heat flux is discussed.
Economic Evaluation of Nuclear Reactor Coupling with Seawater Desalination Facility
Authors: Song Danrong , Liu Tong , Zhao Hua
Published: September 2008

Nowadays only 1/2000 of overall fresh water resource can be used in the world, 1.1 billion people in the world are short of safety drinkable water, so the supply of freshwater is one of the major problems facing to the world in 21st century. China is one of the thirteen water-poorest countries and fresh water has become a choke point of the economic development in those water-poor regions. The desalination of sea water is a very important way to find new fresh water resource and sea water is inexhaustible.
Evaluation of the Sub-Channel Code COBRA-TF for Prediction of BWR Fuel Assembly Void Fraction Distribution
Authors: Aydogan Fatih , Hochreiter Lawrence E. , Ivanov Kostadin
Published: September 2008

Good quality experimental data is needed to refine the thermal hydraulic models for the prediction of rod bundle void distribution and critical heat flux (CHF) or dry-out. The Nuclear Power Engineering Corporation (NUPEC) has provided a valuable database to evaluate the thermal hydraulic codes [1]. Part of this database was selected for the NUPEC BWR Full-size Fine-Mesh Bundle Tests (BFBT) benchmark sponsored by US NRC, METI-Japan, NEA/OECD and Nuclear Engineering Program of the Pennsylvania State University (PSU). Twenty-five organizations from ten countries have confirmed their intention to participate and will provide code predictions to be compared to the measured data for a series of defined exercises within the framework of the BFBT benchmark. This benchmark data includes both the fine-mesh high quality sub-channel void fraction and critical power data. Using a full BWR rod bundle test facility, the void distribution was measured at mesh sizes smaller than the sub-channel by using a state-of-the-art computer tomography (CT) technology [1]. Experiments were performed for different pressures, flow rates, exit qualities, inlet sub-cooling, power distributions, spacer types and assembly designs. There are microscopic and sub-channel averaged void fraction data from the CT scanner at the bundle exit as well as X-ray densitometer void distribution data at different elevation levels in the rod bundle. Each sub-channel’s loss coefficient was calculated with using the Rehme method [2,3], and a COBRA-TF sub-channel model was developed for the NUPEC facility. The BWR assembly that was modeled with COBRA-TF includes two water rods at the center. The predicted sub-channel void fraction values from COBRA-TF are compared with the bundle exit void fraction values measured using the CT-scanner void fraction from the BFBT benchmark data. Different plots are used to examine the code prediction of the void distribution at a sub-channel level for the different sub-channels within the bundle.
The OECD Pipe Failure Data Exchange Project: Validation of Canadian Data
Authors: Viglaski Tom , Blahoianu Andrei , Lydell Bengt , Riznic Jovica
Published: September 2008

Structural integrity of piping systems is important to plant safety and operability. Information on degradation and failure of piping components and systems is collected and evaluated by regulatory agencies, international organizations (e.g., OECD/NEA and IAEA) and industry organizations worldwide to establish systematic feedback to reactor regulation and research and development programs associated with non-destructive examination (NDE) technology, in-service inspection (ISI) programs, leak-before-break evaluations, risk-informed ISI, and probabilistic safety assessment (PSA) applications involving passive component reliability. In 2002, the Nuclear Energy Agency (NEA) of the Organization for Economic Co-Operation and Development (OECD) has initiated an international pipe failure data collection and exchange project. The OECD Pipe Failure Data Exchange (OPDE) Project has been established to encourage multilateral co-operation in the collection and analysis of data relating to pipe failure events in commercial nuclear power plants. At present, the database contains 3644 records to which twelve participating countries contributed. This paper presents a brief description of the ODPE project objectives and work scope, as well as the Canadian contribution on data validation with respect to development and application of the pipe failure data collection on which OPDE is based. It gives a number of tables and figures that can be obtained from these records, with selected data ranging from a very broad (i.e. level of participation in the database from each member country), to very specific (i.e. plant operational state at time of pipe failure discovery for CANDU reactors).
Plant Outage Time Savings Provided by Subcritical Physics Testing at Vogtle Unit 2
Authors: Cupp Philip , Heibel M. D.
Published: September 2008

The most recent core reload design verification physics testing done at Southern Nuclear Company’s (SNC) Vogtle Unit 2, performed prior to initial power operations in operating cycle 12, was successfully completed while the reactor was at least 1% ΔK/K subcritical. The testing program used was the first application of the Subcritical Physics Testing (SPT) program developed by the Westinghouse Electric Company LLC. The SPT program centers on the application of the Westinghouse Subcritical Rod Worth Measurement (SRWM) methodology that was developed in cooperation with the Vogtle Reactor Engineering staff. The SRWM methodology received U. S. Nuclear Regulatory Commission (NRC) approval in August of 2005. The first application of the SPT program occurred at Vogtle Unit 2 in October of 2005. The results of the core design verification measurements obtained during the SPT program demonstrated excellent agreement with prediction, demonstrating that the predicted core characteristics were in excellent agreement with the actual operating characteristics of the core. This paper presents an overview of the SPT Program used at Vogtle Unit 2 during operating cycle 12, and a discussion of the critical path outage time savings the SPT program is capable of providing.