bluuebunny/crossref_metadata_2025 huggingface dataset:
Theoretical Design of Thermosyphon for Process Heat Transfer From NGNP to Hydrogen Plant
Authors: Sabharwall Piyush , Patterson Mike , Gunnerson Fred
Published: July 2009

The Next Generation Nuclear Plant (NGNP) will most likely produce electricity and process heat, with both being considered for hydrogen production. To capture nuclear process heat, and transport it to a distant industrial facility requires a high temperature system of heat exchangers, pumps and/or compressors. The heat transfer system is particularly challenging not only due to the elevated temperatures (up to ∼ 1300K) and industrial scale power transport (≥50 MW), but also due to a potentially large separation distance between the nuclear and industrial plants (100+m) dictated by safety and licensing mandates. The work reported here is the preliminary analysis of two-phase thermosyphon heat transfer performance with alkali metals. A thermosyphon is a device for transporting heat from one point to another with quite extraordinary properties. In contrast to single-phased forced convective heat transfer via ‘pumping a fluid’, a thermosyphon (also called a wickless heat pipe) transfers heat through the vaporization / condensing process. The condensate is further returned to the hot source by gravity, i.e. without any requirement of pumps or compressors. With this mode of heat transfer, the thermosyphon has the capability to transport heat at high rates over appreciable distances, virtually isothermally and without any requirement for external pumping devices. Two-phase heat transfer by a thermosyphon has the advantage of high enthalpy transport that includes the sensible heat of the liquid, the latent heat of vaporization, and vapor superheat. In contrast, single-phase forced convection transports only the sensible heat of the fluid. Additionally, vapor-phase velocities within a thermosyphon are much greater than single-phase liquid velocities within a forced convective loop. Thermosyphon performance can be limited by the sonic limit (choking) of vapor flow and/or by condensate entrainment. Proper thermosyphon requires analysis of both.
Benchmark Efforts to Support Studies of Advanced VHTRs
Authors: Ames David E. , Tsvetkov Pavel V.
Published: July 2009

The Very High Temperature Reactor (VHTR) is the leading candidate for the reactor component of the Next Generation Nuclear Plant (NGNP). This is because the VHTR demonstrates great potential in improving safety characteristics, being economically competitive, providing a high degree of proliferation resistance, and producing high outlet temperatures for efficient electricity generation and/or other high temperature applications, most notably hydrogen production. In addition, different fuel types can be utilized by VHTRs, depending on operational goals. In this case, the recovery and utilization of the valuable energy left in LWR fuel in order to create ultra long life single batch cores by taking advantage of the properties of TRU fuels. This paper documents the initial process in the study of TRU fueled VHTRs, which concentrates on the verification and validation of the developed whole-core 3D VHTR models. Many of the codes used for VHTR analysis were developed without a full appreciation of the importance of randomness in particle distribution. With this in mind, the SCALE code system was chosen as the computational tool for the study. It provides the opportunity of utilizing SCALE versions 5.0 and 5.1, making it possible to compare and analyze different techniques accounting for the double heterogeneity effects associated with VHTRs. Startup physics results for Japan’s High Temperature Test Reactor (HTTR) were used for experiment-to-code benchmarking. MCNP calculations were employed for code-to-code benchmarking. Results and analysis are included in this paper.
Evaluation of the Phase Composition, Crystallinity and Trace Isotope Variation of SiC in Experimental TRISO Coated Particles
Authors: de Villiers Johan P. R. , Ngoepe Noko , Roberts James , Tuling Alison S.
Published: July 2009

The silicon carbide layers in experimental TRISO coated particles with zirconia kernels were evaluated for their phase composition, their impurity levels and the crystal perfection and twinning of the crystallites in the layers. This evaluation was necessary to compare the different SiC layers and to relate these properties to various quality tests and ultimately to manufacturing parameters in the CVD coater. Identification of the various polytypes was done using electron diffraction methods. This is the only method for the unequivocal identification of the different polytypes. The 3C, and 6H polytypes were positively identified. A feature of the SiC in some samples is the disordered nature of the phase. The disorder is characterised by planar defects, of different width and periodicity, giving rise to streaking in the diffraction pattern along the [111] direction of the 3C polytype. Polarised light microscopy in transmission is a useful tool to easily distinguish between the cubic (beta) and non-cubic (alpha) SiC in the layers. It also provides valuable information about the distribution of these phases in the layers. Raman spectroscopy was used to examine the distribution of Si in the SiC layers of the different samples. Two samples contain elevated levels of Si, of the order of 50%, with the highest levels on the inside of the layers. The elevated Si levels also occur in most of the other samples, albeit at lower Si levels. This was also confirmed by use of SEM electron backscatter analysis. Rietveld analysis using X-ray diffraction is presently the only reliable method to quantify the polytypes in the SiC layer. It was found that the SiC layer consists predominantly (82% to 94%) of the 3C polytype, with minor amounts of the 6H and 8H polytypes. Impurities in the SiC and PyC could be measured with sufficient sensitivity using laser ablation inductively coupled mass spectrometry (LA-ICP-MS). The SiC and PyC layers are easily located from the intensity of the C13 and Si29 signals. In most cases the absolute values are of less concern than the variation of impurities in the samples. Elevated levels of the transition elements Cu, Ni, Co, Cr and Zn are present erratically in some samples. These elements, together with Ag107 and Ag109, correlate positively, indicating impurities, even metallic particles. Elevated levels of these transition elements are also present at the SiC/OPyC (Outer Pyrolytic Carbon) interface. The reasons for this are unknown at this stage. NIST standards were used to calibrate the impurity levels in the coated particles. These average from 1 to 18 ppm for some isotopes.
Evolution of Near Term PBMR Steam and Cogeneration Applications
Authors: Kuhr Reiner W. , Hannink Ryan , Paul Keith , Kriel Willem , Greyvenstein Renee , Young Roger
Published: July 2009

US and international applications for large onsite cogeneration (steam and power) systems are emerging as a near term market for the PBMR. The South African PBMR demonstration project applies a high temperature (900°C) Brayton cycle for high efficiency power generation. In addition, a number of new applications are being investigated using an intermediate temperature range (700–750°C) with a simplified heat supply system design. This intermediate helium delivery temperature supports conventional steam Rankine cycle designs at higher efficiencies than obtained from water type reactor systems. These designs can be adapted for cogeneration of steam, similar to the design of gas turbine cogeneration plants that supply steam and power at many industrial sites. This temperature range allows use of conventional or readily qualifiable materials and equipment, avoiding some cost premiums associated with more difficult operating conditions. As gas prices and CO2 values increase, the potential value of a small nuclear reactor with advanced safety characteristics increases dramatically. Because of its smaller scale, the 400–500MWt PBMR offers the economic advantages of onsite thermal integration (steam, hot water and desalination coproduction) and of providing onsite power at cost versus at retail industrial rates avoiding transmission and distribution costs. Advanced safety characteristics of the PBMR support the location of plants adjacent to steam users, district energy systems, desalination plants, and other large commercial and industrial facilities. Additional benefits include price stability, long term security of energy supply and substantial CO2 reductions. Target markets include existing sites using gas fired boilers and cogeneration units, new projects such as refinery and petrochemical expansions, and coal-to-liquids projects where steam and power represent major burdens on fuel use and CO2 emissions. Lead times associated with the nuclear licensing process may support early applications in the 2018–2020 timeframe. This paper summarizes the design options likely to shape these emerging steam and cogeneration applications.
Effect of Helium Injection on Natural Circulation Onset Time in a Simulated Pebble Bed Reactor
Authors: Yurko Joseph P. , Sorensen Katrina M. , Kadak Andrew , Yan Xing L.
Published: July 2009

This paper describes the experimental validation of a proposed method that uses a small amount of helium injection to prevent the onset of natural circulation in high temperature gas reactors (HTGR) following a depressurized loss of coolant accident. If this technique can be shown to work, air ingress accidents can be mitigated. A study by Dr. Xing L. Yan et al. (2008) developed an analytical estimate for the minimum injection rate (MIR) of helium required to prevent natural circulation. Yan’s study used a benchmarked CFD model of a prismatic core reactor to show that this method of helium injection would impede natural circulation. The current study involved the design and construction of an experimental apparatus in conjunction with a CFD model to validate Yan’s method. Based on the computational model, a physical experimental model was built and tested to simulate the main coolant pipe rupture of a Pebble Bed Reactor (PBR), a specific type of HTGR. The experimental apparatus consisted of a five foot tall, 2 inch diameter, copper U-tube placed atop a 55-gallon barrel to reduce sensor noise from outside air movement. Hot and cold legs were simulated to reflect the typical natural circulation conditions expected in reactor systems. FLUENT was used to predict the diffusion and circulation phases. Several experimental trials were run with and without helium injection. Results showed that with minimal helium injection, the onset of natural circulation was prevented which suggests that such a method may be useful in the design of high temperature gas reactors to mitigate air ingress accidents.
MELCOR Code Validation on HE-FUS3 Loop
Authors: Sallus Laurent , Van Hove Walter
Published: July 2009

Since the last decade, Tractebel Engineering has been involved in several consecutive projects in the field of High Temperature Gas Reactor (HTGR). The objectives of the present project called RAPHAEL (www.raphael-project.org) is to provide R&D results in order to consolidate available data on generic V/HTR technologies and to develop innovative solutions to further contribute to the improvement of HTR performances. One of the objectives of the RAPHAEL Sub-project Safety is to qualify tools for performing safety analyses and supporting the safety approach and demonstration. One of the work packages concerns the validation of the existing thermal-hydraulic system codes capabilities needed to perform transient analysis in V/HTR. This validation is carried out by benchmarking against experimental data and by comparing simulation results given by several codes. The current paper presents the work performed at Tractebel Engineering on the simulation of the HE-FUS3 experimental loop — ENEA facility, Brasimone (Italy) — with the MELCOR v.1.8.6 code. The HE-FUS3 loop contains a wide range of components characteristic of a V/HTR like compressor, pipes, diffusers, valves, heaters and heat exchangers. Even if the loop characteristics/configuration is not prototypical of a V/HTR design, the loop is useful to assess the objectives identified by the Project, i.e. helium operating fluid, design pressure and temperature set respectively at 10.5 MPa and 530 °C. The experimental data of the HE-FUS3 loop made available for the benchmark are a set of steady state tests for the thermal-hydraulic characterization of the loop and two transient tests — Loss Of Flow Accidents (LOFA). Moreover, to assess the characteristics of the compressor, data have also been provided from a compressor test campaign. From the code-to-experiment comparison the ability of MELCOR v.1.8.6 to reproduce the experimental results is judged.
Evaluation of a Vessel Cooling System for the VHTR
Authors: Kim Min-Hwan , Lee Won-Jae , Richards Matt
Published: July 2009

The Very High Temperature Reactor (VHTR) has been selected by the U.S. as the Generation IV technology for the Next Generation Nuclear Plant (NGNP), and also by the Republic of Korea for the Nuclear Hydrogen Development and Demonstration (NHDD) project. One of the key long-lead items for the VHTR is the Reactor Pressure Vessel (RPV). In the absence of active vessel cooling, the RPV temperature during normal operation is determined by the design point selected for the primary coolant inlet temperature and the design of the reactor internal components, including the physical location of riser channels that route the coolant flow to the plenum above the reactor core. For the VHTR, the primary coolant (helium) inlet temperature is expected to be in the range 490°C to 590°C and the outlet temperature is expected to be in the range 850°C to 950°C. For the RPV, both SA-508/533 steel and higher alloy steels with higher temperature capability (e.g., 9Cr-1Mo-V steel) are being considered. Because of its extensive experience base as an ASME Section III code-approved material for Light Water Reactor (LWR) pressure vessels, SA-508/533 steel is emerging as a strong candidate for the VHTR RPV. However, in order to use this material, the RPV temperature must be maintained below ASME code limits, which are 371°C during normal operation and 538°C for up to 1000 h during accident conditions.
Performance Comparison of VHTR Plants With Direct and Indirect Energy Conversion Cycles
Authors: El-Genk Mohamed S. , Tournier Jean-Michel
Published: July 2009

This paper compared the performance of very high temperature reactor (VHTR) plants with direct and indirect closed Brayton Cycles (CBCs) and investigated the effect of the molecular weight of the CBC working fluid on the number of stages in and the size of the single shaft turbomachines. The CBC working fluids considered are helium (4 g/mole) and He-Xe and He-N2 binary mixtures (15 g/mole). Also investigated are the effects of using LPC and HPC with inter-cooling, cooling the reactor pressure vessel with He bled off at the exit of the compressor, and changing the reactor exit temperature from 700°C to 950°C on the plant thermal efficiency, CBC pressure ratio and the number of stages in and size of the turbo-machines. Analyses are performed for reactor thermal power of 600 MW, shaft rotation speed of 3000 rpm, and IHX temperature pinch of 50 °C.
Thermo-Hydraulics Characteristic Study of Vapor Explosion
Authors: Li Tianshu , Yang Yanhua , Yuan Minghao , Hu Zhihua
Published: July 2009

The fuel/coolant interaction (FCI) vapor explosion generates more heat than the loss. Low temperature molten materials jet and single droplet fuel/coolant interaction has been investigated by using an observable experimental apparatus. The experimental results showed that many factors affect the vapor explosion, such as the temperature conductivity coefficient of simulant molten materials; the triggering; the Taylor instability; penetrating manner; molt metal temperature and coolant temperature etc. For the molten metal jet vapor explosion, the higher temperature conductivity coefficient is, the easier the vapor explosion is, and the higher temperature the explosion needs. The Rayleigh-Taylor instability and the triggering promote the vapor explosion. When the molten tin alloy temperature is about 600°C to 700°C, the pressure reached maximum value. The increasing coolant temperature would decrease the pressure of the vapor explosion.
Integrated Software Safety Analysis Method for Digital I&C Systems
Authors: Huang Hui-Wen , Hung Hung-Chih , Shih Chunkuan , Yih Swu , Chen Ming-Huei , Yu Yuan-Chang , Chen Shu-Chuan , Tu Wan-Tsz , Yang Wei-Yi
Published: July 2009

The digitalized Instrumentation and Control (I&C) system of Nuclear power plants can provide more powerful overall operation capability, and user friendly man-machine interface. The operator can obtain more information through digital I&C system. However, while I&C system being digitalized, three issues are encountered: 1) software common-cause failure, 2) the interaction failure between operator and digital instrumentation and control system interface, and 3) the non-detectability of software failure. These failures might defeat defense echelons, and make the Diversity and Defense-in-Depth (D3) analysis be more difficult. This work developed an integrated methodology to evaluate nuclear power plant safety effect by interactions between operator and digital I&C system, and then propose improvement recommendations. This integrated methodology includes component-level software fault tree, system-level sequence-tree method and nuclear power plant computer simulation analysis. Software fault tree can clarify the software failure structure in digital I&C systems. Sequence-tree method can identify the interaction process and relationship among operator and I&C systems in each D3 echelon in a design basis event. Nuclear power plant computer simulation analysis method can further analyze the available backup facilities and allowable manual action duration for the operator when the digital I&C fail to function. Applying this methodology to evaluate the performance of digital nuclear power plant D3 design, could promote the nuclear power plant operation safety. The operator can then trust the nuclear power plant than before, when operating the highly automatic digital I&C facilities.
Lingao II Source Term Analysis of a Mitigated Severe Accident Scenario
Authors: Chen Yaodong , Wu Yongzhong , Zhang Hui
Published: July 2009

A typical scenario, Station-Black-Out, which makes up an important part of contribution to core degradation frequency of three loop PWR Lingao phase II according to the PSA level 1 studies, was selected in this paper for Severe Accident Analysis with the application of combination of MELCOR and COCOSYS. Of the scenario, concurrent with postulated power recovery before degradation proceeded to failure of lower head, the evolution of in_vessel phase was studied; a detailed phenomena covering heat-up and dry-out of primary system, collapse and melt down of fuel assemblies, thermal hydraulic response inside the containment compartments, release and accumulation of H2 and fission product (FP) source terms, their leakage into environment were simulated and investigated. The effectiveness of preventive and mitigative measures, such as dedicated core depressurization system (DCDS), passive autocatalytic recombiner (PAR) system, are somehow validated through the calculation; the results also indicated that with the power recovered at 6 hour to start up ECCS, the corium could be retained above support plate.
Swirling Annular Flow in a Steam Separator
Authors: Kataoka Hironobu , Sou Akira , Hosokawa Shigeo , Tomiyama Akio
Published: July 2009

Effects of pick-off ring configuration on the separation performance of a downscaled model of a steam separator for a boiling water nuclear reactor are examined using various types of pick-off rings. The experiments are conducted using air and water. Pressure drops in a barrel and a diffuser, and diameters and velocities of droplets at the exit of the barrel are measured using differential pressure transducers and particle doppler anemometry, respectively. As a result, the following conclusions are obtained: (1) the separation performance does not depend on the shape of pick-off ring, but strongly depends on the width of the gap between the pick-off ring and the barrel wall, (2) the pressure drop in the barrel is well evaluated using the interfacial friction factor for unstable film flows, and (3) the radial distribution of droplet velocity at the exit of the separator is of use for the evaluation of carry-under.
Characteristic of Natural Circulation Temperature Fluctuation Under Rolling Motion Condition
Authors: Tan Sichao , Gao Puzhen , Su Guanghui
Published: July 2009

The characteristics of natural circulation temperature fluctuation are studied experimentally and theoretically. Experimental results show that rolling motion may cause the coolant fluctuation which results in the periodical oscillation of wall and fluid temperatures. The amplitude of wall temperature fluctuation is larger than that of fluid temperature oscillation, and amplitude of wall temperature fluctuation increases with rolling period and amplitude increasing. Because of the wall heat capacity, there is a phase difference between the fluctuation curves of fluid and outlet fluid temperatures. A mathematical model is also conducted to simulate the temperature fluctuation under coolant flow fluctuation condition. The theoretical results agree well with experimental data.
Study on an Optimization Algorithm of Generating Test Vectors for Digital Reactor Protection System Testing
Authors: Sheng Xin , Huang Xiaojin , An Zhencai , Guo Yin
Published: July 2009

As a safety-critical system for the NPP, the digital Reactor Protection System (RPS) has replaced the traditional analog Reactor Protection System in the most newly-built NPPs. A new type digital RPS developed by INET, Tsinghua University, must pass the hardware qualification and software Verification & Validation (V&V) to satisfy the requirements of quality criterion and safety laws. The stimulation/response testing method is always used in the integration testing phase of software V&V. The test vectors group would be very large if the digital RPS has many input variables. Therefore, In order to find out all of the failure of software, the less testing vectors would be benefit to limit the testing time and cost. A black box model is always be used for those systems with few known information for the Conner. All testing vectors would be generated by nature sequence. The black box model has good features. It does not rely on any prior knowledge about the objective system. However, the black box model may increase the average number of test vectors and average time to find out all of the failure. If a grey box model can be adopted in the testing process, a lot of known information of the objective system can be used and the test time would be saved prominently. As independent developed digital RPS by INET, there is enough information of the testing objects, which can be used to apply the grey box model on the digital RPS testing procedure and to generate the test vectors. An optimization algorithm of test vectors generating is as follows: a) Firstly, a different weight factors would be set to different combination of input variables by expert knowledge and logic design rules; b) Secondly, a particle movement algorithm is used to optimize, compare and select random test vectors by weight factors. The primary simulation results indicate that the average testing time and the number of test vectors are both less than the normal test strategy which based on the black box model. The optimization algorithm of test vectors generating based on the particle movement may be more efficient to find out all of the failure. Therefore, the testing cost and time would be saved in consequence.
An Aging Simulation Model for the Emergency Diesel Generators of Angra 2 by Supplementary Variables
Authors: Santos de Oliveira Iona´ Maghali , Frutuoso e Melo Paulo Fernando Ferreira , de Pinho Marcos Oliveira
Published: July 2009

The extension of the useful lifetime of nuclear power plants is an issue of great importance and concern. From the reliability point of view, this problem requires the consideration of time-dependent failure rates and possible failure dependencies. This analysis has been typically performed through a Markovian approach. To illustrate this point, we have developed a computerized reliability analysis of the emergency diesel generators (EDGs) of a four “loop” PWR plant, considering the hypothesis of aging and perfect repair by using Supplementary Variables to cast the initially Nonmarkovian model into a Markovian one. In order to perform such analysis and to simulate aging effects, a nuclear plant has been taken for reference, which has been commercially operating for only six years. Failure rates were obtained from similar EDGs of another plant, already under aging, while repair data were taken from its technical specifications. Discontinuous repair rates were considered in order to improve maintenance strategies. Several ages were attributed to these equipments, allowing the calculus of the failure probability as well as their availability according to each regarded age. In this sense, the EDGs behavior as to aging can be obtained in detail and decisions concerning maintenance and useful lifetime extension can be made on a stronger basis. To get the desired results in terms of reliability figures and due to the discontinuous repair rates that had to be taken into account, a new numerical method that uses a part of the analytical solution, called Euler Iterative + Characteristic, has been developed in order to solve the differential equations systems, making the solving of the system faster and more efficient.
Inert Matrix Fuels Analysis by Means of the TRANSURANUS Code: The Halden IFA-652 In-Pile Test
Authors: Calabrese R. , Vettraino F. , Tverberg T.
Published: July 2009

Inert matrix fuels are a possible option to reduce separated plutonium stockpiles by burning it in LWR fleet. A high burning efficiency targeted by preventing new plutonium build-up under irradiation (U-free fuel), a proved high radiation damage and leaching resistance are fundamental requirements when a once-through fuel cycle strategy is planned. Among other options, both calcia-stabilised zirconia (csz) and thoria fulfill these criteria standing as the most promising matrices to host plutonium. While several in-pile tests concerning thoria fuels are found, calcia-stabilised zirconia under-irradiation performance is still to be fully assessed, with this regard the thermal conductivity, markedly lower than UOX and MOX cases, plays a fundamental role. For this reason, ENEA has conceived a comparative in-pile testing of three different U-free inert matrix fuel concepts, that have been performed in the OECD Halden HBWR (IFA-652 experiment). The discharge burnup accomplished about 90–97% of the 45 MWd/kgUeq target under typical LWR irradiation conditions. The test-rig is a six-rod bundle loaded with IM, IMT and T innovative fuels. IM and T fuels have, respectively, csz and thoria as matrix, the fissile phase being HEU oxide (UO2 93% 235U enriched). IMT is a ternary fuel composed by csz+thoria matrix and HEU oxide as fissile phase. Thoria is added in IMT fuel to improve the low IM reactivity feedback coefficients. Pins are instrumented providing fuel centerline temperature, pin inner pressure and fuel stack elongation measurements. Our purpose is to investigate the key processes of IMF under-irradiation behaviour by means of the TRANSURANUS code. Thermal conductivity and its degradation with burnup, densification-swelling response and FGR are tentatively modelled in the burnup domain of IFA-652. In particular it is pointed out the effects of pellet geometry and fuel microstructures in the IM and IMT cases. The consistency of our results is discussed aiming at understanding the in-pile response, as a fundamental step, in the perspective of future deployment of the nuclear fuels we are dealing with Notwithstanding this ambitious objective, it is clear, however, that these results rely on a limited data set and that, as TRANSURANUS is a semi-empirical code mostly tailored for commercial fuels, the modelling of the IMF is still a work in progress.
Conceptual Thermal Hydraulic Design of a 20MW Multipurpose Research Reactor
Authors: Chae Heetaek , Seo Chulgyo , Park Jonghark , Park Cheol , Vinh Vinh L.
Published: July 2009

The conceptual thermal hydraulic design analyses for a 20 MW reference AHR core have been jointly performed by KAERI and DNRI (VAEC). The preliminary core thermal hydraulic characteristics and the safety margins for the AHR core were studied for various core flow rates and fuel assembly powers. Statistical method and the MATRA_h subchannel code have been applied to evaluate the thermal hydraulic performances of the AHR core under the forced convection cooling mode during a nominal power operation and the natural circulation mode during a reactor shutdown condition. In addition, the safety margin evaluations were carried out for 2 typical accidents, a loss of flow accident by a primary pump seizure and a reactivity induced accident by a CAR rod withdrawal during a normal full power operation. Major design parameters for these analyses are a minimum critical heat flux ratio (MCHFR), an onset of a nucleate boiling (ONB) margin and a maximum fuel temperature. Results of the thermal hydraulic analyses show that the normal full power operation of the AHR could be ensured with a sufficient thermal margin for the onset of a nucleate boiling for a coolant velocity larger than 7.3 m/s. The AHR is also thought to have a sufficient natural circulation cooling capability up to at least 1.8MW to cool the core without the onset of a nucleate boiling in a channel after a normal reactor shutdown and during anticipated transients. It was also confirmed that the AHR core was sufficiently protected from the loss of a flow by a primary cooling pump seizure and the overpower transients by a CAR withdrawal from the MCHFR and fuel temperature points of view.
Preliminary Research of Auxiliary Bearing in HTR-10GT Project
Authors: Qin Qingquan , Yang Guojun , Shi Zhengang , Yu Suyuan
Published: July 2009

Active Magnetic Bearing (AMB) was used in the project of 10MW high temperature gas-cooled reactor (HTR-10GT) for the advantages over conventional mechanical bearings: without any mechanical friction and lubrication, etc. Auxiliary Bearings (ABs) is one of the most important parts in the AMB system, and its main function is to support the rotor at rest and provide protection for the rotor system during an overload or magnetic bearings failure situation. This paper introduced auxiliary bearings used in the HTR-10GT project and compared its advantages and disadvantages with other types of auxiliary bearings. The dynamic behaviors and temperature variation are the most important factors that may affect the performance of auxiliary bearings in a rotor drop event, this paper also analyzed the touching down course and dynamics in detail, divided the drop down process into four distinct stages of motion: free fall, impact, sliding-whirling and rolling. Finally, a test rig built up for the following rotor drop test is presented in the article. Test results at lower drop down speed were discussed. The result of the theory and experiment research has important reference value for the auxiliary bearings design of HTR-10GT.
Investigations to Arrive at a Stable Start-Up Procedure for AHWR
Authors: Bagul R. K. , Pilkhwal D. S. , Vijayan P. K. , Saha D.
Published: July 2009

Natural circulation is being adopted as a mode of core heat removal in several nuclear reactors that are under development. This is due to the passive nature of natural circulation that enhances the system safety and reliability. However, major concern in the design of natural circulation based reactor systems is to avoid the flow instabilities that may occur under certain operating conditions, i.e. unstable operational regime. Therefore various reactor operational transients such as start-up, power raising, setback and also the steady state operating points must fall within the stable regime. The choice of operating procedures needs to be made judiciously and which also needs to be validated and supported by experiments. Advanced Heavy Water Reactor (AHWR) being developed in India is a pressure tube type natural circulation boiling water-cooled reactor, wherein major part of the power is generated by thorium. Experiments and analytical studies have been performed to arrive at a rational start-up procedure for AHWR. Experimental results obtained in simple rectangular natural circulation loops as well as in a scaled down facility have revealed the importance of external pressurization to avoid the flashing and Type-I instabilities that occur at low pressure during start-up.
Experiments on Eurofer Steel Corrosion by Pb-16Li
Authors: Gessi Alessandro , Aiello Antonio , Benamati Gianluca
Published: July 2009

Corrosion samples of the EUROFER 97 steel were exposed to Pb-16Li in the LIFUS II loop at 480°C and 550°C, with a liquid metal velocity of about 0,01 m/s in agreement with the foreseen operative conditions of the HCLL (Helium Cooled Lithium Lead) blanket concept. The specimens were extracted after 1500, 3000 and 4500 hours exposure at the lower temperature, while a further time step of 6000 hours was added at the higher temperature. After extraction, weight change measurements and metallurgical analysis were performed on the corrosion specimens. The experimental results demonstrated a linear corrosion mechanism, together with no preferential elemental depletion of the steel in contact with the liquid alloy. In this paper the results are reported and discussed with reference to previous similar works.
Non-Boussinesq Integral Model for Horizontal Turbulent Strongly Buoyant Plane Jets
Authors: Xiao Jianjun , Travis John R. , Breitung Wolfgang
Published: July 2009

Horizontal buoyant jets are fundamental flow regimes for hydrogen safety analyses in the nuclear power plants. Integral model is an efficient, fast running engineering tool that can be used to obtain the jet trajectory, centerline dilution and other properties of the flow. In the published literature, most of the integral models that are used to predict the horizontal buoyant jet behavior employ the Boussinesq approximation, which limits the density range between the jets and the ambient. CorJet, a long researched, developed, and established commercial model, is such a Boussinesq model, and has proved to be accurate and reliable to predict the certain buoyant jet physics. In this study, Boussinesq and non-Boussinesq integral models with modified entrainment hypothesis were developed for modeling horizontal turbulent strongly buoyant plane jets. All the results and data where the Boussinesq model is valid will collapse to CorJet when they are properly normalized, which implies that the calculation is not sensitive to density variations in Boussinesq model. However, non-Boussinesq results will never collapse to CorJet analyses using the same normalized scaling, and the results are dependent on the density variation. The reason is that CorJet employs the Boussinesq approximation in which density variations are only important in the buoyancy term. For hydrogen safety analyses, the large density variation between hydrogen and the ambient, which is normally the mixture of air and steam, will make the Boussinesq approximation invalid, and the effect of the density variation on the inertial mass of the fluid can not neglected. This study highlights the assumption of the Boussinesq approximation as a limiting, simplified theory for strongly buoyant jets. A generalized scaling theory for horizontal strongly buoyant jet seems not to exist when the Boussinesq approximation is not applicable. This study also reveals that the density variation between jets and the ambient should be less than 10% to accurately model horizontal buoyant jets when the Boussinesq approximation is applied. Verification of this integral model is established with available data and comparisons over a large range of density variations with the CFD codes GASFLOW and Fluent. The model has proved to be an efficient engineering tool for predicting horizontal strongly buoyant plane jets.
Risks for Flow Induced Vibrations (FIV) at EPU
Authors: Nilsson Per , Lillberg Eric
Published: July 2009

This work deals with risk areas for flow induced vibration at extended power uprate, EPU. The focus is on the mechanisms of excitation in one phase relevant for Swedish BWRs and PWRs. FIV-events that have occurred in nuclear power plants over the world have been collected and categorized. The most relevant events for EPU are summarized to: vibrations in steam systems due to turbulence or vortex shedding and resonance, vibrations of internal parts and also thermal mixers and sleeves or in valves and vibrations of tube banks in partial or full cross flow. Based on the collected events and some semi-empirical methods, a simple search list for FIV by power uprate has been developed. In principle these changes lead to increased risks: changed flow velocity, decreased water temperature and increased steam temperature and decreased structural damping, mass or stiffness. In addition to that, the typical collected events should be regarded.
Pipe Diameter Effect on the Two-Phase Flow Separator for Removal of Liquid in Horizontal Pipes
Authors: Chang J. S. , Looy P. C. , Harvel G. D.
Published: July 2009

A pipe diameter effect on the two-phase flow liquid separation phenomenon by an annular type liquid separator is experimentally investigated, where this type of separator does not have any intrusive section inside the pipe. Past work demonstrated that a 2 cm diameter flow separator could effectively separate liquid from an annular flow regime without the use of intrusive devices in the main pipe. The current study examines the results of a 10 cm diameter pipe and compares to the 2 cm results previously observed. The separator consists of segments of Lucite tubing containing multiple holes in the separation sections. An experiment was conducted for a 10 cm inner diameter flow pipe and inlet two-phase flow regimes of stratified smooth, stratified wavy and annular mist are studied for superficial gas velocities from 0 to 8 m/s and superficial liquid velocities from 0 to 0.005 m/s. The number of holes, the pitch, and the shape of the holes in the wall of the separator were varied. The results show that the separation efficiencies were observed in the range of 80% to 100%, where the separation efficiency decreases with increasing superficial liquid velocity. Separation efficiencies of 100% were observed for stratified flow conditions. Separation efficiencies greater than 90% were usually observed for annular mist conditions. For stratified wavy conditions, separation efficiencies ranged from 80% to 100%. The separation efficiency does increase with the number of holes as expected. In comparison to the smaller 12 cm diameter test section, the results indicate the same trends, where the superficial velocities can be used as a first order scaling parameter for the separator.
Thermal Cycle Evaluation for Feedwater Heater Out of Service Condition
Authors: Dziuba Lindsey L. , Stakenborghs Robert J.
Published: July 2009

Large power plant secondary side thermal cycles typically include multiple parallel trains of condensate and feedwater heating to improve cycle efficiency. These heaters are typically arranged in multiple identical trains and can contain by-pass lines to accommodate removal from service for on-line maintenance. In a closed system, the heaters are typically horizontal u-tube heat exchangers with various combinations of internal and external drains cooling. Heating fluid is provided from steam extraction and drains flow. Drains flow can be piped either in a forwards (pumped forward) or backwards (cascading drains) direction. All of these possible equipment combinations result in a complex system where the impact of taking single or multiple heaters out of service for on-line maintenance can be difficult to properly determine. This paper presents a description of the various impacts of removing feedwater heaters from service during operation. The impacts can include increased steam flow through the downstream turbine stages, increased extraction flows and drains flows to the remaining heaters, and increased heater tube side flows that can all potentially cause equipment damage. Proper precautions should be taken to protect plant equipment, including reducing plant thermal power output when necessary. This paper provides specific plant examples to illustrate the complexity of the problem and its proper solution. A simplified evaluation methodology is also provided to allow the user to determine the quantitative impact of removing equipment from service. For each particular feedwater heater out of service arrangement, the method uses mass and energy balances around the remaining in-service feedwater heaters to determine the magnitude of the changes in turbine stage, extraction and drain flowrates. The appropriate power level for that condition is determined by applying specific acceptance criteria for the impacted equipment.
CEA Preliminary Studies for Industrial MOx Reprocessing: From Laboratory Results to Successful First Industrial Campaign on UP2-800 Areva Plant
Authors: Reynier Tronche N. , Eysseric C. , Dancausse JP.
Published: July 2009

In November/December 2004, AREVA succeeded in reprocessing for the first time MOX spent fuels with a continuous dissolution process on the UP2-800 plant in La Hague. It was the industrial conclusion of preliminary and extensive studies realized by the CEA teams at the laboratory scale in Atalante facilities. The experiments were realized using MOx fuel obtained by MIMAS process and irradiated in Saint-Laurent des Eaux power reactor. The average burn-ups of the pin samples used for the studies were comprised between 28 GWd/T and 43 GWd/T. As the MOx fuel present a specific behavior versus dissolution due to the presence of PuO2 clusters, the solubility tests were mainly focused on the Pu behavior as a function of the temperature of dissolution (from 88 °C to boiling point) and the fuel burn-up. The fixed dissolution parameters were the residence time (8 hours), the initial acidity (near 8 M) and the sample size (35 mm). Available results concern dissolution kinetics, plutonium solubility and its distribution in the different flux (insoluble solids, solution).
Role of Initial Melt Inventory on Debris Generation With Woods Metal Water Interaction Under Drop Mode
Authors: Narayanan K. S. , Das S. K. , Rao E. H. V. M. , Lydia G. , Murthy S. S. , Kumaresan M. , Harvey J. , Kasinathan N. , Rajan M.
Published: July 2009

Thermal interaction of a hot liquid with a cold volatile liquid was investigated with woods metal melt water system at first contact temperature just above the spontaneous nucleation temperature of water. Drop mode experiments were carried out up to 1000 g of melt and resulting debris sizes and shapes were analyzed. Experiment was carried out with melt temperature of 425 °C and water pool at 30 °C. One of the 100 g run resulted in an energetic interaction with a mass median diameter 2.5 mm ± 1.58. The MMD is seen to change to higher values for non-energetic cases and higher melt inventories. Dynamic pressure signals are also recorded. Relative contribution of thermal and hydrodynamic fragmentation is assessed in pool and narrow geometries. Dependence of mass median diameter on initial melt inventory has been brought out.
Dimensionless Analysis of Heat Transfer and Flow Resistance on Laminar Flow Recuperator
Authors: Yin Chunyu , Yang Xiaoyong , Wang Jie
Published: July 2009

Recuperator is one of the key components in the helium-turbine cycle coupled with High Temperature Gas cooled Reactor (HTGR). Synthetically considering the heat transfer coefficients, the pressure drop and installation space of recuperator, it is obviously a trend to use compact heat exchanger as recuperator in nuclear power plant. Recuperator recovers heat from the turbine exhaust gas. It promotes the cycle efficiency over entire power range and in all typical modes including start up and shut down modes. The recuperator’s heat transfer coefficients, height, pressure drop have effect on the recuperator’s effectiveness. The main purpose of this paper is to present the law of heat transfer and flow resistance in laminar flow compact exchanger. Based on the similarity theory, the dimensionless parameters of the plate-fin heat exchanger is given in this paper; and then the the dimensionless analysis of the over-all heat transfer coefficient, recuperator’s effectiveness and flow resistance is presented. Furthermore, relationship between the pressure drop and length is also developed.
Primary Study on Carbon Dioxide Power Cycle of HTGR
Authors: Duan Chengjie , Yang Xiaoyong , Wang Jie , Yu Suyuan
Published: July 2009

At present, power cycles used in HTGR are indirect steam Rankine cycle and helium Brayton cycle. Using water or helium as working fluid which transform thermal energy into mechanical energy for HTGR power cycle has many disadvantages. Steam cycle could choose steam system which is similar to conventional coal-fired power plant, but because of the limit of material and equipments, there is big temperature difference between the steam and the helium, that makes big loss of thermal power and lowers the cycle efficiency. Helium can reach a high temperature in HTGR Brayton cycle and it has good stability, but because of helium has big isentropic exponent and low density, it is difficult to compress and makes helium turbine has shorter blades and more stages than normal gas turbine. Carbon dioxide has good thermal stability and physical properties. To avoid the reaction of CO2 with graphite and canning of fuel element at high temperature, it should be used in an indirect cycle as second loop working fluid. CO2 has appropriate critical pressure and temperature (7.38MPa, 304.19K) and can choose three types of cycle: supercritical cycle, subcritical-pressure cycle and trans-critical-pressure cycle (CO2 sometimes works under supercritical pressure, some times under subcritical-pressure). Carbon dioxide cycle works in a high pressure, so it makes pressure loss lower. When CO2 works close to its critical point, its density become larger than other conditions, and not change very much, this permits to reduce compress work. The thermal physical properties of carbon dioxide are totally different from helium due to CO2 works as real gas in the cycle. That causes the calculation of CO2 thermal physical properties, heat transfer and power cycle efficiency become difficult and need to be iterated. A systematic comparison between helium and carbon dioxide as working fluid for HTGR has been carried out. An empirical equation had been selected to estimate the thermal physical properties of carbon dioxide. Three types of carbon dioxide power cycle have been analyzed and the thermal efficiency has been calculated. A detailed introduction to the basic calculation process of the CO2 cycle thermal efficiency had been presented in the paper.