bluuebunny/crossref_metadata_2025 huggingface dataset:
Numerical Evaluation of Wall Temperature Measurement Methods Developed to Investigate Thermal Fatigue at T-Junction Pipe
Authors: Miyoshi Koji , Nakamura Akira , Takenaka Nobuyuki
Published: October 2013

Thermal fatigue cracking may initiate at a T-junction pipe where high and low temperature fluids flow in from different directions and mix. Thermal stress is caused by a temperature gradient in a structure and its variation. The accurate simulation of the temperature distribution in structures is, therefore, important for estimating thermal fatigue. In this study, an experimental method using a T-junction pipe with thermocouples was developed. Wall temperatures in the experiment have to be obtained at the inner surface of the pipe to validate numerical simulation results. The difference of position between the inner surface and the measurement point where using thermocouples may, however, have an effect on temperature data. The numerical simulation results showed that the amplitude of temperature fluctuations was reduced to 54% and the difference of phase was 0.91 radians when the sinusoidal temperature fluctuation of 5Hz was applied from the inner surface to the thermocouple measurement points. These results showed that wall temperatures at the inner surface should be estimated from measured data obtained by the thermocouples. A transfer function was, therefore, obtained to calculate wall temperatures at the inner surface from measured data. In addition, the numerical simulation results showed that the amplitude and the phase of temperature fluctuations differed depending on existence of voids around thermocouples. Respective differences of it in the amplitude and the phase were about 5% and 3% when the sinusoidal temperature fluctuation of 5Hz was applied at the inner surface. These results showed that thermocouples should be installed in pipes without voids to measure accurate temperature fluctuations. A mock-up test showed that voids stayed behind the thermocouples when the thermocouples were brazed into the pipe wall at atmospheric pressure, but the voids disappeared when thermocouples were brazed in a vacuum atmosphere into the inner surface of the pipe.
The Variation in Effectiveness of Low-Finned Tubes Within a Shell-and-Tube Heat Exchanger for Supercritical CO2
Authors: Fourspring Patrick M. , Nehrbauer Joseph P.
Published: October 2013

Low-finned tubes can be effective in baffled flow heat exchangers, if the heat transfer coefficients on either side of the heat exchanger differ greatly and therefore limit the thermal conductance of the heat exchanger. Low-finned tubes can increase thermal conductance by providing additional heat transfer area on the limiting side.

The height and the spacing of the low-fins must be greater than the thickness of the thermal boundary layer on the low-finned side of the heat exchanger. Otherwise, the effectiveness of the additional area that the low-finned tubes provide will be reduced. The boundary layer thickness is dependent on the velocity and the thermophysical properties of the fluids. Therefore, in a standard shell-and-tube heat exchanger, the number of heat exchanger shell-side baffles needs to be properly considered to provide the correct shellside velocity without introducing too much pressure drop.

Testing of a shell-and-tube heat exchanger containing low-finned tubes varied the flow rate and pressure of the supercritical CO2 on the shell side as water provided the cooling on the tube side. The testing maintained the temperature and pressure of the CO2 above the critical point in order to determine the changes in the effectiveness of the low-finned tubes and thus the heat transfer rate of the heat exchanger.

The results show that the additional heat transfer area provided by the low-finned tubes will remain fully effective, even as the supercritical fluid nears its critical point or a pseudo-critical temperature. This result also supports (but is not sufficient to prove) the guidance to limit the estimated thickness of the thermal boundary layer to the fin height and twice the fin spacing to ensure the additional heat transfer area provided by the low-finned tubes remain effective.
Comparison of Containment Response for a CANDU 6 Station Blackout Sequence Using MAAP4-CANDU (v4.0.7) and GOTHIC (v7.2b)
Authors: Petoukhov S. M. , Nguyen Tuan H. , Awadh B. , Mathew P. M.
Published: October 2013

The objective of this study is to compare containment simulation results predicted by MAAP4-CANDU v4.0.7 and GOTHIC v7.2b for a Station Blackout severe core damage accident, when the containment dousing spray system is not credited.

The input parameters (mass and energy source terms into containment) for GOTHIC were generated by MAAP4-CANDU for the Station Blackout sequence. The GOTHIC (lumped parameter model) and MAAP4-CANDU containment nodalization models were very similar, thus facilitating direct comparison of the results.

Analysis of the results demonstrated that the simulation results obtained with the two codes are similar. Good agreement was obtained for the containment pressure, gas temperature and containment atmosphere composition predictions.
Experimental Study on Condensate Inundation for Condensation of Horizontally Flowing Steam in a Large Condenser Tube Bundle
Authors: Fujita Issaku , Inoue Koichi , Machii Kotaro , Hiraoka Satoshi , Ishihara Manabu
Published: October 2013

While 1700MW class Pressurized Water Reactor (PWR) projects like US-APWR are conducted by Mitsubishi Heavy Industry (MHI) now, the size of condenser tube bundle becomes the largest ever constructed by MHI to be accordance with the increased heat rejection rate from the turbine system. Where, it becomes difficult to ignore the condensate sub-cooling and heat transfer deterioration by condensate inundation. Therefore, we have been working on research about the practical application of large-sized condenser since 2008, and tube arrangement optimization is one of the important subjects.

As suitable tube arrangement for a large-sized condenser, the proven streamlined shape tube bundle was proposed, which had been adopted in the fossil fuel power units. In the streamlined shape tube bundle, most steam condenses on tubes under horizontal or downward steam flow conditions. In large tube bundle, the heat transfer deterioration due to inundation apparently occurs. On the other hand, condensate is blown away by the horizontal steam flow, and the effect of inundation would be reduced. For the tube arrangement optimization, the accurate knowledge on heat transfer including the behavior of such condensate under the horizontal steam flow condition is essential. However, most research on steam condensation on tube bundle has been conducted for the downward steam flow but the research for horizontal steam flow condensation is very limited so far (1–5).

In the present research, condensation experiments were conducted by using a horizontal tube bank containing 36 cooling tubes with 12 condensate supply tubes. Steam was horizontally supplied to the tube bank at velocities 15–27m/s at pressures of 8.8kPa. Cooling tubes were made of copper and have an outer diameter of 19.1mm and condensing length of 150mm. In order to clarify the effect of condensate inundation on condensation heat transfer in detail, local heat transfer coefficients and the condensate flow in the tube bank were measured. In this paper, we describe findings on the condensate behavior and heat transfer in horizontal flow obtained by the experiment.
AP1000 Plant CILC/CIPS Risk Assessment Using Advanced TH Methodology
Authors: Wang G. , Sapienza P. , Fetterman R. J. , Young M. Y. , Secker J. R. , Nelson S. W. , Oelrich R. L. , Sung Y.
Published: October 2013

Similar to many existing Pressurized Water Reactors (PWR), the AP1000® cores will undergo sub-cooled nucleate boiling in the upper grid spans of some fuel assemblies at normal operating conditions. Sub-cooled nucleate boiling may increase crud deposits on the fuel cladding surface which may increase the risk of Crud Induced Power Shift (CIPS) and/or Crud Induced Localized Corrosion (CILC). A CIPS/CILC risk assessment has been performed to support the AP1000 fuel assembly design finalization.

In this paper, the advanced thermal-hydraulic (TH) methodology used in the AP1000 plant CIPS/CILC risk assessments are summarized and discussed, and the relationship between the CIPS/CILC mechanisms, fuel reliability, and plant operating conditions is also presented. Finally, acceptable AP1000 core CIPS/CILC risk assessment results are summarized and suggestions that specifically target reducing CIPS/CILC risks for AP1000 plants are described.
Risk Analysis of the Residual Heat Removal System at South Texas Project With a Special Focus on Breach of Containment
Authors: Solom Matt , Chance Christopher , Pannier Christopher , Seager Robert , Lee Alan , Green Jordan , Duong Tam , Alicea Paola
Published: October 2013

A unique feature in the design of the reactors at South Texas Project (STP) is that each unit’s Residual Heat Removal System (RHRS) is located within containment. The aim of this work is to identify the potential failure modes of the Residual Heat Removal System that could lead to a breach of containment during reactor operation and thereby may increase Core Damage Frequency (CDF). The analysis began with a Failure Modes and Effects Analysis (FMEA) of the RHRS based on a Piping and Instrumentation Diagram. The two motor operated valves that isolate the RHRS from the Reactor Coolant System (RCS) were assumed to fail with an internal leak, exposing downstream components to reactor coolant. Pathways for coolant to exit containment were identified and analyzed for severity, occurrence, and detectability of the failure modes. The analyses of these factors lead to the determination of a criticality rating, which assisted in the ultimate findings. The results of the FMEA were used to construct an event tree of the failure modes of interest and the composite probability of each failure. The highest probability failure mode of interest was a breach of containment by a tube of the heat exchanger leaking into the Component Cooling Water (CCW) with a failure probability of 2.5E−10 per reactor year. The insights gained in this analysis will be used by the South Texas Project for future risk analysis and decision-making regarding the RHRS.
Basic Experiment on Lithium Removal Technique
Authors: Furukawa Tomohiro , Hirakawa Yasushi
Published: October 2013

As a preliminary investigation into the establishment of a lithium removal technique for the components used at the International Fusion Materials Irradiation Facility (IFMIF), experiments were performed on the dissolution of lithium in three solvents: ethanol, pure water, and ethanol–water. In these experiments, hemispherical lithium was immersed in the solvents at constant temperatures, and the degree of dissolution was measured continuously from the height of the sample. From the obtained data, the average dissolution rate in the solvents at each testing temperature (10–90 °C) and the amount of hydrogen generated by the chemical reaction were calculated. The average dissolution rates in ethanol, pure water, and ethanol–water at 30 °C were 0.01, 1.6, and 0.43 mm/min, respectively. Although the average dissolution rate increased with the testing temperature in the low-temperature range (10–50 °C) for all solvents, this increase was saturated in the high-temperature range (50–90 °C) in experiments with pure water and ethanol–water as solvents. The volume of gas collected during each experiment was in good agreement with the volume of hydrogen assumed to be generated from the chemical reaction of lithium with the solvents.
Investigation of PAR Behavior in the REKO-4 Test Facility
Authors: Simon B. , Reinecke E.-A. , Klauck M. , Heidelberg D. , Allelein H.-J.
Published: October 2013

Passive auto-catalytic recombiners (PARs) play a key role in the hydrogen mitigation strategy of European LWRs. In order to avoid possible threats related to hydrogen combustion, PARs are installed to remove hydrogen released during a loss-of-coolant accident. The possible impact of hydrogen explosions became evident during the reactor accident in Fukushima (Japan) in March 2011, where leaked hydrogen ignited and largely destroyed the upper part of the reactor building. The mitigation strategy is based and verified by computational accident assessments. Code validation against experimental data is vital in order to achieve reliable results.
Study of Aerosol Scavenging by Rain in Case of a Radioactive Contamination of the Atmosphere
Authors: Quérel Arnaud , Lemaitre Pascal , Monier Marie , Porcheron Emmanuel , Flossmann Andrea
Published: October 2013

The analysis of the radioactive aerosol scavenging by rain after the Chernobyl accident highlights some differences between the modelling studies and the environmental measurements. Part of this gap is due to the uncertainties on the scavenging efficiency of aerosol particles by raindrops, in particular for drops with a diameter larger than one millimeter. The IRSN (Institut de Radioprotection et de Sûreté Nucléaire) has decided to launch an experimental study to measure with a better accuracy the scavenging efficiency of large raindrops.

The scavenging efficiency of aerosol has been determined by measuring precisely the mass of aerosol particles collected by a single drop after its path through an atmosphere loaded with particles.

The collection efficiencies for drop diameters of 2 mm and 2.6 mm (previously unknown for atmospheric aerosols) are measured. The impact of these new data on modeling of the washout of the atmosphere by the rain is noticed.
Validation of the LP Code MELCOR 1.8.6 for Containment Analyses With Hydrogen Recombiners Using Results From the OECD/NEA Experimental Project THAI
Authors: Kotouč Miroslav
Published: October 2013

Two experiments with different types of PAR unit (from the vendors Areva and NIS, respectively) from the international experimental project THAI (organized by OECD/NEA) were simulated with the lumped parameter code MELCOR 1.8.6. A simple model of PAR units, based on identification of sinks and sources of mass and energy in each time step, was implemented in the code, taking advantage of FORTRAN functions and vendors-furnished correlations for the recombination rate. The two experiments, HR-12 and HR-14, were characterised by elevated pressure and temperature in steam-saturated atmospheres with lack of oxygen (i.e. oxygen starvation conditions). The MELCOR code with the simple PAR model succeeded to accurately predict the recombination rate, including the efficiency drop due to oxygen starvation. The precision achieved in calculation of other quantities such as pressure and gas concentrations was less accurate, reflecting the possibilities of a lumped parameter code.
Improving Air-Cooled Condenser Performance and Availability Through Innovative Cleaning, Inspection and In-Situ Sleeving
Authors: Fischer Gary , Ripley Craig
Published: October 2013

Whether a power plant chooses an air-cooled condenser (ACC) because of siting issues, or because of changing environmental laws, the utilization of this dry cooling option has seen a dramatic level of growth over the past decade around the world. Like the steam surface condenser, the ACC is prone to tube fouling, albeit to the external finned tube surfaces instead of internal surfaces. This tube fouling can dramatically impact the performance of the ACC causing plant heat rate to suffer, increasing the consumption of fuel, increasing CO2 emissions and reducing megawatt output. In addition to fouling, ACC units are exposed to dramatic swings in ambient temperature. Units exposed to multiple freeze-thaw cycles each year often develop numerous points of inleakage into the vacuum boundary, also causing unit efficiency to suffer. Unlike the steam surface condenser, the ACC can be elevated at 60–75 feet in the air which poses considerable challenges to locating and repairing the sources of leaks.

This paper will examine the performance improvements realized by the Yellowstone Power Plant, located in Billings, Montana on their ACC unit after utilizing an innovative cleaning technology as well as the use of a new finned tube “sleeve” to repair leaking tubes.
Scaling Analysis of the Pressurizer Surge Line in Nuclear Plant Integral Test Facilities
Authors: Deng Cheng-Cheng , Chang Hua-Jian , Qin Ben-Ke , Wang Han , Chen Lian
Published: October 2013

During small break loss of coolant accident (SBLOCA) of AP1000 nuclear plant, the behavior of pressurizer surge line has an important effect on the operation of ADS valves and the initial injection of IRWST, which may happen at a time when the reactor core reaches its minimum inventory. Therefore, scaling analysis of the PRZ surge line in nuclear plant integral test facilities is important. Four scaling criteria of surge line are developed, which respectively focus on two-phase flow pattern transitions, counter-current flow limitation (CCFL) behavior, periodic draining and filling and maintaining system inventory. The relationship between the four scaling criteria is discussed and comparative analysis of various scaling results is performed for different length scale ratios of test facilities. The results show that CCFL phenomenon and periodic draining and filling behavior are relatively more important processes and the surge line diameter ratios obtained by the two processes’ scaling criteria are close to each other. Thus, an optimal scaling analysis considering both CCFL phenomenon and periodic draining and filling process of PRZ surge line is given, which is used to provide a practical reference to choose appropriate scale of the surge line for the ACME (Advanced Core-cooling Mechanism Experiment) test facility now being built in China.
Coupled Code SOCRAT-BN Development for Safety Analysis of Sodium-Cooled Fast Reactors
Authors: Usov Eduard V. , Kudashov Ivan G. , Zhigach Sergey A. , Butov Anton A. , Pribaturin Nikolay A. , Lezhnin Sergey I. , Chalyy Ruslan V. , Yakush Sergey E. , Vinogradova Uliya
Published: October 2013

SOCRAT-BN is a software package to simulate design and severe accidents of sodium-cooled fast reactors. The package consists of modules for calculating damage to the reactor’s core, thermohydraulic processes and neutron physics.

The thermohydraulic module has been developed to calculate one- and two-phase flows in channels with different geometry and bundles. The module is based on a two-fluid model for equal pressures of phases.

In this paper we present an explanation of the deciding constitutive models for equations used in the system. Validation of the module was performed on the experimental data for one- and two-fluid flows in complex geometry channels and on calculation of running a first loop of the reactor BN-600 in nominal mode.
Dynamic Simulation Research on Primary Coolant System of Sodium Cooled Fast Reactor
Authors: Yu Yang , Guo Yun
Published: October 2013

The Chinese Experimental Fast Reactor (CEFR) is a 65MWt/20MWe sodium cooled fast reactor. It is a pool-type reactor where the reactor and other internals such as pumps and intermediate heat exchangers (IHX) are immersed in a sodium pool. In this paper a one-dimensional dynamic code was developed to model the primary sodium circuit which included the reactor core, IHX, pumps, hot and cold pool etc. Moreover, the model of the property of sodium flow and heat transfer correlations was collected and compiled. This paper also discusses the mathematical models of various components of the primary sodium circuit, the numerical techniques to solve the models, the thermal-hydraulic studies of some design basis events such as the loss of primary pump or secondary pump accident etc, the comparison of the results of the code with that of the safety analysis report.

Studies were conducted simulating both full and low power operating conditions. The dynamic code has been validated, and the results show that it has a benign response to some typical accidents. Finally, the model and code derived in this paper could be used in the safety analysis of pool-type sodium cooled fast reactor, and adopted in the development of CEFR simulation platform.
Research on Enhancement of Natural Circulation Capability in Lead-Bismuth Alloy Cooled Reactor by Using Gas-Lift Pump
Authors: Zuo Juanli , Tian Wenxi , Qiu Suizheng , Su Guanghui
Published: October 2013

The gas-lift pump has been adopted to enhance the natural circulation capability in the type of lead-bismuth alloy cooled reactors such as ADS and LMFR. The natural circulation ability and the system safety are obviously influenced by the two phase flow characteristics of liquid metal-inert gas. In this study, the numerical research has been performed to evaluate the natural circulation capability of lead-bismuth cooled ADS with gas-lift pump. The drift flow theory, void fraction prediction model and friction pressure drop prediction model have been adopted in this numerical simulation. The effects of the gas flow rate, the bubble diameter and the height of ascension pipe on natural circulation capability of gas-lift pump have been analyzed. The results showed that in bubbly flow pattern, for a fixed value of gas volume flow rate, the natural circulation capability increased with the decrease of the bubble diameter. In the bubbly flow, slug flow, churn flow and annular flow pattern, with the gas volume flow rate increasing, the natural circulation capability initially increased and then declined. As the height of ascension pipe increased, the natural circulation flow rate went up. Besides, the flow parameters influenced the thermal hydraulic characteristics of the reactor core obviously. Therefore, in the practical engineering application, the suitable gas volume flow rate, bubble diameter and the height of ascension pipe are important parameters to the capability of natural circulation in lead bismuth alloy loop of this research. The present work is helpful for revealing the law of enhancing the natural circulation capability by gas-lift pump, and providing theoretical basis of the optimization design of cooling and system safety.
Research on Low Cycle Fatigue in 26m2 Low Pressure Inner-Casing of Nuclear Power Plant
Authors: Hu Yifeng , Chen Gang , Kang Ming
Published: October 2013

High reliability and long life endurance are very important factors in the operation of large steam turbines for nuclear power plant (NPP). The low cycle fatigue (LCF) is necessary take into account problem in the design of turbine component. Thus the calculation method and assessment criterion of LCF are important for the design and development of the low pressure (LP) inner-casing.

The main activity of this study was focus on the research of the low cycle fatigue in 26m2 LP inner-casing of steam turbines for NPP, which is manufactured by Shanghai Electric Power Generation Equipment Co., Ltd: STP. And detailed calculation method and assessment criterion combined with FEA analysis, fatigue curve and linear cumulative damage theory are developed to judge the life of the LP inner-casing, The approach is applied in the process of the design work of large steam turbine inner-casing.
Control Strategies for VHTR Gas-Turbine System With Dry Cooling
Authors: Sato Hiroyuki , Yan Xing L. , Ohashi Hirofumi , Tachibana Yukio , Kunitomi Kazuhiko
Published: October 2013

An original control strategy for very high temperature reactor (VHTR) gas-turbine system with dry cooling against ambient air temperature fluctuation was established in order to enable the freedom of site selection wherever desired without significant drawbacks on the performance. First, the operability of power conversion system and degradation of power generation efficiency were examined considering not only the thermodynamics but also the mechanical efficiency of compressor based on detailed performance map derived from experimental data. Second, control simulations for large ambient temperature fluctuations were conducted by system analysis code with the built-in control strategy. In addition, the sensitivity of power generation efficiency for typical steam cycle with dry cooling to ambient air temperature changes was assessed for the comparison. It was shown that the design goal can be effectively met simply by monitoring and controlling a few of key operating parameters such as reactor outlet temperature, primary coolant pressure. Furthermore, distinctive advantages of the VHTR gas-turbine system over nuclear power plant employing Rankine cycle was demonstrated when installing in inland area.
Application of Fine Depletion Calculation for LLFPs In-Core Transmutation Evaluation in PWRs
Authors: Liu Kun , Wu Hongchun , Cao Liangzhi , Zheng Youqi , Wang Changhui
Published: October 2013

An in-core transmutation analysis and evaluation code, named CATE, considering in-core fine flux calculation and fine depletion process, is verified and validated in the present paper. Verification and Validation of implementations for the OECD/NEA PWR cell benchmark for actinides transmutation, IAEA PWR benchmark and infinite homogenized plate problem to confirm reliability and numerical accuracy for the code have been performed in presented paper. The numerical performance of the code system is demonstrated in the analyses of the in-core fuel management calculation. It is found that the present code system gives stability in prediction of critical concentration of boric solution and radial power distribution. Based on the verifications and validations, a preliminary LLFP transmutation pattern is calculated. Numerical results indicate that CATE can be used not only for the fuel management calculation, but also for in-core transmutation evaluation of PWR.
Visualization on Inert Gas Jets Impinging to a Glass Tube Submerged in Liquid Sodium
Authors: Kudoh Hideyuki , Sugiyama Ken-ichiro , Narabayashi Tadashi , Ohshima Hiroyuki , Kurihara Akikazu
Published: October 2013

For the sodium-water reaction accident, it is important to grasp the structure of gas jets submerged in liquid sodium and associated droplet sizes. In this study, we successfully obtained visualized images of inert gas jets injected into liquid sodium. Formation processes of liquid sodium droplets entrained into the gas jets and drop-size distributions are discussed.

Visualization was conducted on upward argon gas jets impinging to a glass tube in stationary liquid sodium pool. The gas jet behavior was observed by using a bore scope inserted inside of the glass tube. The gas jet velocities at the nozzle exit were 13 < u0 < 125 m/s.

For u0 > 50 m/s, liquid droplets began to be formed. The formation processes of liquid droplets were classified into (1) spray type; a liquid film formed on the nozzle is blown and atomized by the gas jets and (2) entrainment type; attributed to the instability of a gas cavity-liquid phase interface. But the difference in the drop-size distribution of the both formation processes was not significant. Mean droplet diameter of liquid sodium was confirmed to be agreed well with an empirical equation suggested from a model experiment using water.
Estimation of Fragmentation on Jet Breakup in Coolant
Authors: Kuroda Taihei , Abe Yutaka , Kaneko Akiko , Yuzuru Iwasawa , Nariai Hideki , Sakaba Hiroshi , Koyama Kazuya , Matsuo Eiji
Published: October 2013

Fast Breeder Reactor (FBR) is designed with safety in mind. However, there is billion to one possibility that a hypothetical Core Disruptive Accident (CDA) occurs. When CDA occurs, the Post Accident Heat Removal (PAHR) must be achieved. In the PAHR, the molten material is required to be fragmented and solidified in sodium coolant. In order to estimate whether the molten material jet is completely solidified in sodium coolant or not, it is significant to estimate jet breakup length. Although, the jet breakup length is influenced with fragmentation behavior, the correlation between them is not clear yet. Therefore, it is strongly required to clarify the mechanism of the fragmentation behavior on the jet surface. The objective of the present study is to estimate fragmentation on jet breakup in coolant experimentally. Tap water and Fluorinert™ (FC-3283) are used as simulated coolant and molten material, respectively. Flourinert is transparent and colorless liquid and its density is higher than water, therefore we can observe internal flow structure of Fluorinert. Fluorinert injected into water, and the jet breakup behavior and the fragmentation behavior of the jet are observed by using high speed video camera.

In order to estimate fragmentation on liquid jet, we identified the position of the interface with back lighting technique and also, we conducted velocity measurement with Particle Image Velocimetry (PIV) technique simultaneously. It is observed that interfacial waves of the jet are generated. Waves are pulled with surrounding liquid and grown up. Finally, a fragment is separated as a droplet from front edge of the wave. Also, the vorticity is evaluated from the velocity data in order to investigate influence of the flow field in detail. From the result of calculating vorticity, the high value was estimated when jet was fragmented. It is suggested that fragmentation behavior correlates with the surrounding flow field. And the energy ratio contributing to fragmentation is calculated from velocity field. The energy ratio is important to investigate the amount of the fragmentation on liquid jet. Fragmentation on jet breakup in coolant is estimated.
Missile Safety Analysis of Nuclear Steam Turbine
Authors: Chen Rong , Liang Feng , Hua Wen Xiang , Jiang Pu Ning , He Xiao Zhong
Published: October 2013

This paper introduced the basic turbine missile safety analysis process for the Shanghai Turbine Plant’s nuclear power station. The turbine-generator unit is not considered as part of a nuclear power plant’s safety related equipment, but there is still the possibility that a high-energy missile created by failure of the low-pressure turbine disks could penetrate the stationary barriers of the turbine (especially the last stage of the low-pressure turbine), and damage plant safety-related equipment. The failure mechanisms of the turbine failure that could hypothetically generate missiles were described. Rotor damage due to stress corrosion cracking with brittle fracture and ductile fracture and destructive over-speed protection control system failure were discussed as the main elements to evaluate the safety of the turbine unit. Rotor stress analysis using FEA was used to evaluate the strength of the turbine rotor. Key factors that influence fracture mechanisms were discussed. Crack initiation and growth to the critical size was estimated. The strike probability of the turbine missile was accessed to assure the safety of the nuclear safety related equipment from a steam turbine unit generated missile.
Prediction of Hydrogen Distribution Under Severe Accident Conditions With CFD and Lumped Parameter Codes
Authors: Szabó Tobias , Benz Stefan , Kretzschmar Frank , Royl Peter , Jordan Thomas
Published: October 2013

In case of a severe accident, the containment is the ultimate barrier to the environment. Therefore, reliable simulations tools for containment thermal hydraulics, including hydrogen distribution are indispensable. We simulated the behavior of the containment atmosphere under severe accident conditions with a postulated source term of water, steam and hydrogen. We used a detailed 3D CFD code (GASFLOW) and a lumped parameter code (MELCOR) in order to compare and assess their modeling capabilities. A simplified generic containment including all important components was used as a test bed. We analyzed the calculated pressure histories, mass and energy balances, convective flow as well as steam and hydrogen distributions. Integral values were modeled in good agreement by both codes. The overall flow was reasonably predicted. However we observed discrepancies in the calculated steam and hydrogen concentrations.
Validation Experiment for the Cooling Capability of Passive Auxiliary Feedwater System (PAFS) With PASCAL Facility
Authors: Bae Byoung-Uhn , Kim Seok , Park Yu-Sun , Kim Bok-Deuk , Kang Kyoung-Ho , Yun Byong-Jo
Published: October 2013

The Passive Auxiliary Feedwater System (PAFS) is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor Plus) which is intended to completely replace the conventional active auxiliary feedwater system. It removes the decay heat by cooling down the secondary system of the SG using condensation heat exchanger installed in the Passive Condensation Cooling Tank (PCCT). With an aim of validating the cooling and operational performance of the PAFS, PASCAL (PAFS Condensing Heat Removal Assessment Loop), was constructed to experimentally investigate the condensation heat transfer and natural convection phenomena in the PAFS. It simulates a single tube of the passive condensation heat exchangers, a steam-supply line, a return-water line, and a PCCT with a reduced area, which is equivalent to 1/240 of the prototype according to a volumetric scaling methodology with a full height. The objective of the experiment is to investigate the cooling performance and natural circulation characteristics of the PAFS by simulating a steady state condition of the thermal power. From the experiment, two-phase flow phenomena in the horizontal heat exchanger and PCCT were investigated and the cooling capability of the condensation heat exchanger was validated. Test results showed that the design of the condensation heat exchanger in PAFS could satisfy the requirement for heat removal rate of 540 kW per a single tube and the prevention of water hammer phenomenon inside the tube. It also proved that the operation of PAFS played an important role in cooling down the decay heat by natural convection without any active system. The present experimental results will contribute to improve the model of the condensation and boiling heat transfer, and also to provide the benchmark data for validating the calculation performance of a thermal hydraulic system analysis code with respect to the PAFS.
Development of a Heat Transfer Correlation for Supercritical CO2 Based on Multiple Data Sets
Authors: Preda Tiberiu , Saltanov Eugene , Pioro Igor , Gabriel Kamiel S.
Published: October 2013

Currently, increase in thermodynamic efficiency of water-cooled Nuclear Power Plants (NPPs) can only be achieved by raising the coolant’s operating conditions above the supercritical point. The critical point of water is 22.06 MPa and 373.95°C, making supercritical water research very power-intensive and expensive. CO2 behaves in a similar manner once in the supercritical state, but at significantly lower pressure and temperature, since critical point of CO2 is 7.37 MPa and 30.98°C. The applications of supercritical CO2 research range from using it as a modelling fluid, to supercritical turbine applications in Liquid Metal Fast Breeder Reactors (LMFBRs), and use in a supercritical Brayton cycle. Therefore, it is of prime importance to model its behaviour as accurately as possible. For this purpose, experimental data of Koppel (1960), He (2005), Kim (2005) and Bae (2007) for CO2 were analyzed, and a new correlation was developed. The dataset consists of 1409 wall temperature points with pressures ranging from 7.58 to 9.58 MPa, mass fluxes from 419 to 1200 kg/m2s, and heat fluxes from 20 to 130 kW/m2. All runs take place in bare tubes of inner diameters from 0.948 to 9.00 mm in both vertical and horizontal configurations. The proposed correlation takes a wall-temperature approach to predicting the Nusselt number. This paper compares the new correlation with other work which has been done at the University of Ontario Institute of Technology by Mokry et al. (2009), as well as with correlations by Swenson et al. (1965) and Dittus-Boelter (1930). It was found that the new correlation has an overall RMS error of 13% for Heat Transfer Coefficient (HTC) values and 5% for calculated wall temperature values. The correlation can be used as a conservative approach to predict wall temperature values in Supercritical Water Reactor (SCWR) preliminary calculations, to predict heat transfer in secondary-loop turbine/ heat exchanger applications, as with the LMFBR, and to help validate scaling parameters used for water and other coolants.
Development of a Core Management Tool for MYRRHA: Work in Progress
Authors: Jaluvka David , Van den Eynde Gert , Vandewalle Stefan
Published: October 2013

Currently, the advanced nuclear irradiation facility MYRRHA is under development at the Belgian Nuclear Research Centre SCK•CEN in Mol, Belgium. MYRRHA is intended as a multi-purpose machine for material and fuel irradiations, radioisotopes production, silicon doping, transmutation research, education, etc. A special effort is being made within the project to develop an in-core fuel management tool for MYRRHA.

Various requirements have been imposed on the core management tool; e.g., the tool should allow a user to specify a variety of optimization objectives and constraints. Based on that, a modular design of the tool was developed that makes it independent of the particular metaheuristic method used for the optimization as well as of the reactor analysis code used for an evaluation of candidate solutions during the search.

In the paper, the current version of the core management tool for MYRRHA is presented together with some results on an illustrative in-core fuel management optimization problem.
Assessment of Different Turbulence Models in Helically Coiled Pipes Through Comparison With Experimental Data
Authors: Colombo Marco , Cammi Antonio , Ricotti Marco E.
Published: October 2013

This paper deals with a comprehensive study of fully developed single-phase turbulent flow and pressure drops in helically coiled channels. To the aim, experimental pressure drops were measured in an experimental campaign conducted at SIET labs, in Piacenza, Italy, in a test facility simulating the Steam Generator (SG) of a Generation III+ integral reactor. Very good agreement is found between data and some of the most common correlations available in literature. Also more data available in literature are considered for comparison. Experimental results are used to assess the results of Computational Fluid Dynamics (CFD) simulations. By means of the commercial CFD package FLUENT, different turbulence models are tested, in particular the Standard, RNG and realizable k-ε models, Shear Stress Transport (SST) k-ω model and second order Reynolds Stress Model (RSM). Moreover, particular attention is placed on the different types of wall functions utilized through the simulations, since they seem to have a great influence on the calculated results. The results aim to be a contribution to the assessment of the capability of turbulence models to simulate fully developed turbulent flow and pressure drops in helical geometry.
Analysis of the Threshold Peculiarities of Density Wave Instabilities in a Parallel Vertical Twin Channels With Risers: IPPE Database
Authors: Kornienko Yuri
Published: October 2013

This paper presents practically unknown outside Russia experimental data about density wave instability (DWI) regions and its thresholds, as well as their primary analysis. This database has been produced by IPPE installation [1–3] with two vertical boiling channels with risers. Regime conditions: 6÷16 MPa pressures; −0.3÷0.75 MW/m2 heat fluxes; low mass velocities 120 ÷ 600 kg/m2s, with different flow additional throttling at the inlet and exit of the channels. Input coolant temperatures served as a DWI threshold parameter and varied in these experiments between 20 °C and Ts − (5÷50) °C. The bottom and top instability thresholds were determined, accordingly, by growth or attenuation of mass flow regular out-of-phase oscillations in both channels. The basic experimental tendencies and peculiarities of DWI instability thresholds behavior typical to all researched pressures are: 1) presence of two instability regions (with stability zone between them) — high subcooled boiling DWI-1 and low subcooled boiling DWI-2, shaping as a descending ‘wedge’ of DWI region widening with growing Zu number; 2) stability zone diminishes with increasing Zu numbers and reduces right up to merging of DWI-1 and DWI-2 regions; 3) increasing additional inlet throttling diminish area of DWI-1 and 2 regions and move it to more higher Zu number; 4) the instability thresholds lines curvature grows with Zu number; 5) the threshold lines behavior directly points at the existence of asymptotic subcooling values restricting the beginning and end of density wave’s oscillation regions and intervals with possible linear approximation.
AREVA’s Advanced Thermal-Hydraulic Codes and Methods for Levels III and IV Crud Risk Assessment
Authors: Jones John H. , Anghelescu Mihnea S. , Bradbury Michael S. , Martin Mathieu G. , Galimov Azat Y. , Hatman Anca , Harne Richard L. , Keheley Thomas H. , Wende Uta , Bigot Jérôme
Published: October 2013

From a crud deposition perspective the achievement of zero fuel failures requires the integration of core neutronics, core thermal-hydraulics, and plant chemistry disciplines. The various level of detail required in the assessment of crud risk is based on guidelines and a checklist developed and published by EPRI. In the guidelines EPRI defined two levels of assessment where core neutronics and thermal-hydraulics are coupled with plant chemistry. These two levels are called Levels III and IV in the EPRI guidelines.

AREVA developed a process using the standard licensing tools and a number of specialized application tools and interfaces that allow Level III and IV calculations. The Level III calculations are performed using a typical reload licensing subchannel node scale (scale of several centimeters) whereas the Level IV calculations are performed on fractions of a millimeter scale.

This paper provides an overview of the neutronics and thermal-hydraulic process used to perform Levels III and IV assessments and provides some results for a typical B&W-designed 177-fuel assembly operating plant. This same process could be applied to other plant types. The details of the plant chemistry process are not covered in this paper, as they are covered in other publications.
Operation and Maintenance Solutions for Generator Water Cooling
Authors: Svoboda Matthias , Svoboda Robert
Published: October 2013

Plugging of copper hollow conductors in water cooled generators is an issue occasionally encountered in large nuclear, fossil and hydro powered power plants, causing load limitations or even unplanned shut-down. A response is the proper choice of stator water chemistry and of the purification systems. The common water chemistry regimes are with low or high oxygen content, and neutral or alkaline pH in the stator water. All these regimes have proven to work in the field, but none is without problems. This can be due to various causes. There is one common trait however: problems arise when the system is being exposed for any length of time to a regime it was not designed for. Stainless steel hollow conductors have only simple requirements on water chemistry. Hollow conductor plugging can lead to overheating and possibly to catastrophic failure, so swift action to remedy the situation is recommended. There are several options. Hot reverse flush can remove debris and temporarily open up the plugs, and mechanical cleaning can open totally blocked hollow conductors. There are several methods for chemical cleaning, the most common being different forms of acid cleaning and EDTA cleaning. The former has the disadvantage of dissolving base metal and brazing as well as the oxides, while the latter method manipulates less aggressive substances that only dissolve the copper oxides in the system.
High-Speed Steam Turbine Systems for Small-Scale Power Generation Applications
Authors: Petrov Miroslav P. , Fridh Jens , Göransson Ake , Fransson Torsten H.
Published: October 2013

Energy utilization from low-grade fuels of either fossil or renewable origin, or from medium-temperature heat sources such as solar, industrial waste heat, or small nuclear reactors, for small-scale power generation via steam cycles, can be reasonably enhanced by a simple technology shift. This study evaluates the technical feasibility of a compact power generation package comprising a steam turbine directly coupled to a high-speed alternator delivering around 8–12 MW of electrical power. Commercial or research-phase high-speed electrical generators at MW-scale are reviewed, and a basic thermodynamic design and flow-path analysis of a steam turbine able to drive such a generator is attempted.

High-speed direct drives are winning new grounds due to their abilities to be speed-controlled and to avoid the gearbox otherwise typical for small system drivetrains. These two features may offer a reasonable advantage to conventional drives in terms of higher reliability and better economy.

High-speed alternators with related power electronics are nowadays becoming increasingly available for the MW-size market. A generic 8 to 12 MW synchronous alternator running respectively at 15,000 to 10,000 rpm, have been used as a reference for evaluating the fundamental design of a directly coupled steam turbine prime mover.

The moderate steam parameter concept suits well for converting mid-temperature thermal energy into electrical power with the help of low-tech steam cycles, allowing for distributed electricity production at reasonable costs and efficiency. Steam superheat temperatures below 350°C (660°F) at pressures of maximum 20 bar would keep the steam volumetric flow sufficiently high in order to restrain the turbine losses typical for small-scale turbines, while helping also with simpler certification and safety procedures and using primarily established technology and standard components. The proposed steam turbines designs and their characteristics thereof have been evaluated by computer simulations using the in-house code ProSteam and its sub-procedures AXIAL and VaxCalc, by courtesy of Siemens Industrial Turbomachinery and its steam turbine division located in Finspong, Sweden.

The first results from this study show that high-speed steam turbines of the proposed size and type are possible to design and manufacture based on conventional components, and can be expected to deliver a very satisfactory performance at variable power output.