bluuebunny/crossref_metadata_2025 huggingface dataset:
The Perspective of Creating a Closed Fuel Cycle Based on the Fast Reactors in Ukraine
Authors: Tereshchenko I. A. , Ustimenko S. O.
Published: October 2013

It is known that the IAEA considers options for a fuel cycle with light water reactors of new generation (LWR); however, the uranium reserves will not last forever, so the input of fast reactors in order to “close” the fuel cycle is currently the best option. Therefore, the preliminary calculations of the approximate cycle took place, and a comparative analysis of cycles using only LWR and the cycle with a gradual replacement of LWR by fast breeder reactors was carried out as well. It is appropriate, because there is a sufficiently large number of spent nuclear fuel now is accumulated in temporary storage, but soon it has to be either converted into fresh fuel or disposed (that is unacceptable in all respects). The main problem of the closed fuel cycle is usage of MOX-fuel, but this type of fuel is elaborated now and will be improved upon.
Station Blackout Analysis of Lungmen ABWR Using TRACE
Authors: Wang Jong-Rong , Lin Hao-Tzu , Chen Hsiung-Chih , Wang Wei-Chen , Shih Chunkuan
Published: October 2013

The Lungmen NPP is the first ABWR (Advanced Boiling Water Reactor) nuclear power plant in Taiwan, consisting of two identical units with 3,926 MWt rated thermal power each and 52.2×106 kg/h rated core flow. The core of Lungmen NPP has 872 bundles of GE14 fuel. There are 10 reactor internal pumps (RIP) in the reactor vessel, providing 111% rated core flow at the nominal operating speed of 151.84 rad/sec. A station blackout (SBO) is defined as the loss of offsite electrical power concurrent with turbine trip and unavailability of the onsite emergency AC power. These result in the loss of core cooling and heat removal systems that rely on the above AC power for their operation. In this research, the TRACE SBO model of Lungmen ABWR has been developed in order for the analysis of SBO transient. The initial condition of SBO transient is 100% rated power/100% rated core flow. The TRACE’s results show that the reactor fuel temperature has been reached 1088.71 K (the zirconium-water reaction may generate) at about 3200 sec. It indicates that the fuels might be damaged after 3200 sec if the RCIC and ACIWA failed to activate in this transient.
Calculation Study on Onset of Nucleate Boiling of Natural Circulation Based on Density Gradient
Authors: Cheng Sheng , Zhou Tao , Zhang Lei , Huang Yanping
Published: October 2013

The flow characteristics of natural circulation are relying on the heated fluid density difference as a driving force. Based on density gradient theory, a characteristic factor of natural circulation was proposed. The forced circulation models of Onset of Nucleate Boiling (ONB) of Bergles&Rohsenow and SU Shunyu were modified through the characteristic factor. The calculation results showed that the modified forced circulation models could calculate the occurrence of ONB in the condition of natural circulation better. Meanwhile, the calculating results and differences of heat flux at ONB between the models of natural and forced circulation were given by changing the thermal parameters. The results revealed that, most of the calculated values of natural circulation model were lower than those of forced circulation models when inlet subcooling and pressure varied, which meant the occurrence of ONB in natural circulation preceded that in forced circulation. The calculated results of forced circulation models could close to those of the natural circulation model through modification, which indicated the characteristic factor of natural circulation proposed in this paper was one of the respects which could reflect the inherent mechanistic difference between natural and forced circulation.
Development for CANDU-6 Moderator Temperature Prediction by Using Porous Media Approach
Authors: Lee Jae Ryong , Yoon Han Young , Kim Hyoung Tae , Jeong Jae Jun
Published: October 2013

In this study, a thermal hydraulic behavior of the moderator in the CANDU reactor was numerically investigated by using CUPID code. KAERI has been developing a component-scale thermal hydraulics code, CUPID. The aim of the code is multi-dimensional, multi-physics and multi-scale thermal hydraulics analysis. This code adopts a three-dimensional, transient, two-phase and three-field model, and includes physical models and correlations of the interfacial mass, momentum, and energy transfer for the closure. To avoid the complexity to generate computational geometry around the matrix of 440 Calandria tubes, a porous media approach was applied. Flow resistance inside the porous media zone was derived from the empirical correlation of the frictional pressure loss. In order to consider the turbulent jet inflows from the inlet nozzles, the standard k-ε turbulence model was applied. For the grid dependency test, three different grid systems were tested. The moderator test vessel at Stern Laboratories Inc. (SLI) for the validation is a cylinder with a diameter of 2m and a length of 0.2m (a thin “slice” of CANDU-6 Calandria vessel). Since the axial flow is assumed to be invariant, two-dimensional calculation was performed. Vertical profile of the liquid temperature was compared with other calculation results as well as experimental data.
The Thermal-Hydraulic Analyses of Transients in PBWFR
Authors: Tain Yonghong , Su Guanghui , Wang Jun , Tian Wenxi , Qiu Suizheng
Published: October 2013

Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR) can produce steam from direct contact of feedwater and lead bismuth eutectic (LBE) in the chimney of 3m height, which could eliminate the steam generators. Meanwhile it circulates LBE by means of buoyancy of steam bubbles and the primary pump is not necessary. The concept makes the reactor simple and compact. The thermal-hydraulic behavior and safety performance were studied by a computer program in the paper. The reactor power was simulated by the point reactor kinetics equations. The drift flux model was adopted to calculate the fluid density and pressure drops in the chimney. The rate of heat exchange of the feed water and LBE in the chimney was supposed to be constant. The results showed that PBWFR had very good inherent safety. The designs could ensure the integrity of the claddings. Meanwhile the impacts of axial conduction were also researched. It showed that even in the PBWFR, in which the temperature difference between inlet and outlet of core was 150°C, the impacts could also be ignored.
Study on the Minimum Drag Coefficient Phenomenon of Supercritical Pressure Water in the Pseudocritical Region
Authors: Lei Xianliang , Li Huixiong , Yu Shuiqing , Zhang Yifan , Chen Tingkuan
Published: October 2013

With the development of supercritical (and even ultra-supercritical) pressure boilers (SCBs) with high capacities, and at the same time, with the consideration of supercritical pressure water-cooled reactors (SCWRs) as one of the six most promising reactor concepts accepted in the Generation IV International Forum (GIF), flow and heat transfer of supercritical water becomes more and more important for both the design and operation safety of the related facilities. Thermo-hydraulic characteristics are among the issues, which are of special significance for the SCBs and SCWRs. It has been found that at supercritical pressures, the hydraulic resistance of water exhibits special characteristics in regions near its pseudo-critical point, which is hereafter called the minimum drag coefficient phenomenon. Experimental investigation was carried out in the present study to investigate further the characteristics of drag coefficient of supercritical pressure water under different conditions. The total pressure drop characteristic of water flowing in smooth tube and internally ribbed tube under the supercritical pressures was measured in experiments with a wide range of operational parameters, such as the system pressures ranging from 23 to 28 MPa, the average heat fluxes varied from 100 kW/m2 to 500kW/m2, and the mass fluxes of water in a range of 600 ∼ 1050 kg/m2s. The experimental data were compared with prediction results calculated by existing common correlations for single phase pressure drops, and large discrepancies were observed between the experimental data and the prediction results. Furthermore, the pressure drops characteristics of supercritical pressure water in cases with different tube arrangement and test conditions were compared with each other, such as that in horizontal tubes and vertical tubes, and that in isothermal flows and in non-isothermal flows. Additionally, this phenomenon observed in the present studies was also analyzed by using computational fluid dynamics technology, and the mechanism of pressure drop variation was reasonably explained. It was found that the deviation appeared between the previously proposed drag coefficient correlations and the present experimental data was mainly owning to ignoring the variation of an existence of the minimum drag coefficient in the pseudo critical region in previous studies, and based on the data obtained in this study, a new correlation for drag coefficient for supercritical pressure water was presented.
Development of Modeling Transport of Graphite Dust in HTGR During Normal Operation Condition
Authors: Chen Zhipeng , Xie Fei , Zheng Yanhua , Shi Lei , Li Fu
Published: October 2013

High temperature gas-cooled reactor (HTGR), especially the pebble-bed core type reactor, will inevitably cause the wear the graphite components and generate graphite dust in the core. The graphite dust is taken away by helium coolant and deposited on the surface of the primary circuit, and the fission products may be absorbed on the dust. Since it is possible that the fission products are released with dust under the accident conditions such as depressurization events, they have a potential hazard of radiation exposure to the environment. The objective of this paper is to develop a code for calculating the behaviour of graphite dust in the primary circuit of HTGR. The paper is focused on development of models for predicting the deposition rates of the dust. The purpose of the work is to estimate the amount and distribution of deposited dust during plant life time, which was assumed to be 40 full-power years. The result will lay the foundation for further studies of fission products releasing and interaction with dust under accident conditions.
Study on Mixing Process of Two Component Gases in a Vertical Fluid Layer: Effectiveness on One-Dimensional Natural Circulation
Authors: Mizuno Hiroki , Takeda Tetsuaki , Funatani Shumpei
Published: October 2013

This study is to investigate an effect of natural convection or natural circulation on a transport process by molecular diffusion in a stratified fluid layer consisting of two component gases. There are many experiment and analysis regarding natural convection or natural circulation in the vertical slot. However, there are few studies on natural convection or circulation and molecular diffusion in the stratified fluid layer consisting of two component gases. It was confirmed that these phenomena appear when the depressurization accident occurs in the very high temperature reactor (VHTR). Therefore it is important to evaluate the transport and mixing processes during the depressurization accident of the VHTR.

The experiment has been performed regarding the combined phenomena of molecular diffusion and natural convection or circulation in a two parallel vertical slots filled with two component gases. The vertical slot consists of the one side heated wall and the other side cooled wall. The other slot consists of the two cooled walls. The dimension of heated wall is 500mm×200mm and thickness is 3mm. The width of the slot is 20mm and the aspect ratio is 25. Combination of nitrogen(N2)/argon(Ar), neon(Ne)/argon(Ar), helium(He)/nitrogen(N2) and helum(He)/argon(N2) was used as the first step of the experiment of the two component gases. The density change of the gas mixture and the gas temperature distribution in the slots were obtained. The mixing process of the heavier gas from the bottom side of the slot filled with the lighter gas was discussed in this paper.

The experimental results showed that the transport phenomena by the molecular diffusion were influenced by the localized natural convection or circulation of the gas mixture in the stratified fluid layer. From the experimental results, it was found that the mixing process by molecular diffusion was affected significantly by the natural convection or circulation induced by the slight temperature difference between both vertical walls.
Insulation Material Deposition and Distribution in a PWR Fuel Assembly Cluster
Authors: Renger Stefan , Alt Sören , Kästner Wolfgang , Seeliger André , Zacharias Frank
Published: October 2013

Background of experimental and methodical work is the loss of coolant accident (LOCA) with release of fibrous pipe insulation material. Latest investigations were focused on material deposition and distribution (cross mixing) in the reactor core. Therefore, a 2×2 PWR fuel assembly (FA) cluster was constructed. Four shortened PWR-FA-dummies are provided with separated in- and outlets. Every 16×16 fuel rod dummy consists of 20 control rod simulators, two spacers, FA-head and FA-bottom with a 3.5×3.5 mm integrated debris-screen filter (IDF). The cluster is encased in an acrylic housing for visual observation. It is connected with the test facility “Zittau Flow Tray” (ZFT), a simplified sump model, which allows inclusion and investigation of complex phenomena like material sedimentation in the sump and strainer blockages. A well mixing of air in the fluid was also considered by free jet expansions and flows through full cone-nozzles as well as marginal air entrainments.

This Paper includes descriptions of applied measuring techniques (digital image processing, thrubeam laser sensors etc.) and an overview of all considered boundary conditions. Experimental results, aiming at the development, implementation and verification of multiphase flow and strainer models, are presented.
Measurements of RF-Induced SOL Modifications in Tore Supra Tokamak
Authors: Kubič Martin , Gunn James P. , Colas Laurent , Heuraux Stéphane , Faudot Eric
Published: October 2013

Since spring 2011, one of the three ion cyclotron reconance heating (ICRH) antennas in the Tore Supra (TS) tokamak is equipped with a new type of Faraday screen (FS). Results from Radio Frequency (RF) simulations of the new Faraday screen suggest the innovative structure with cantilevered bars and ‘shark tooth’ openings significantly changes the current flow pattern on the front of the antenna which in turn reduces the RF potential and RF electrical field in particular parallel to the magnetic field lines which contributes to generating RF sheaths. Effects of the new FS operation on RF-induced scrape-off layer (SOL) modifications are studied for different plasma and antenna configurations — scans of strap power ratio imbalance, phasing, injected power and SOL density.
Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration — (4) Measurement of Bubble Deformation Near Wall Under Flow Rate Fluctuation
Authors: Okachi Satoshi , Seto Masaki , Monji Hideaki , Kaneko Akiko , Abe Yutaka , Yoshida Hiroyuki , Takase Kazuyuki
Published: October 2013

In order to clear the two-phase flow behavior under earthquake, a systematic study is done experimentally and numerically. The present study is one on the series of the study on two-phase flow under earthquake, and focuses on the flow rate fluctuation. The flow rate fluctuation was added to bubbly or plug flow in a horizontal pipe, and flow behavior was measured by PIV and image processing. The bubble deformation near the pipe wall was observed and the velocity field around the bubble was shown. The bubble coalescence was also observed under the flow rate fluctuation condition.
Stress Test Evaluation of Refueling and Storage Pools for Loviisa NPP
Authors: Varpasuo Pentti , Kähkönen Jukka
Published: October 2013

The paper will describe the following analysis of Loviisa plant’s spent fuel pools. As a consequence of stress tests for the existing NPP’s in Finland after the experiences gathered from the Tohoku -Taheiyou-Oki event in Japan in March of this year Fortum Power and Heat Oy (Fortum) has initiated the following analyses of the Loviisa power plant’s refueling pools and spent fuel intermediate storage pools for the combined cooling loss and the earthquake loads. The following loads will be analyzed: 1) The spent fuel pool and refueling pools water temperature is 100 degrees Celsius. The heat load duration is undetermined; 2) The earthquake ground motion applied simultaneously with the thermal load is defined as follows: (ground motion response spectrum is defined in Guide YVL 2.6), the maximum horizontal acceleration is assumed to be 0.1g, 0.2g, 0.3g and 0.4g, respectively; 3) Own weight; 4) The pool of water, hydrostatic pressure load plus the sloshing load because of earthquake motion. Te analysis is aimed to demonstrate the structural integrity and leak-tightness of the pools under the effect of above loads. The analysis is nonlinear taking into account the cracking of the concrete. As a result of the analysis the maximum strains will be determined in the pool stainless steel liner as well in the pool concrete walls.
A Preliminary Study of the MSFR Dynamics
Authors: Guerrieri Claudia , Aufiero Manuele , Cammi Antonio , Fiorina Carlo , Luzzi Lelio
Published: October 2013

In the last years, increasing interest has been focused on an innovative concept of Molten Salt Reactor (MSR) characterized by a fast neutron spectrum that combines the favorable characteristics of MSRs adopting molten salt fluorides both as fuel and coolant with those ones of fast neutron reactors. As a matter of fact, the Molten Salt Fast Reactor (MSFR) has been recognized as a long term alternative to solid-fuelled fast neutron systems and has been identified as reference Gen-IV configuration. Although considerable studies have been carried out for the analysis of the graphite-moderated MSR dynamics, the adoption of a fast spectrum configuration without graphite in the core is expected to notably modify the dynamic behavior of the system, thus requiring further investigation.

In this paper, a preliminary analysis of the MSFR dynamics is performed allowing for the evaluation of the impact of some safety parameters (e.g., feedback coefficients and delayed neutron fraction) on the system behavior for different fuel cycle strategies. For this purpose, a simplified non-linear one-dimensional model of the primary circuit has been developed and the dynamic response of the system has been investigated with reference to different significant transient initiators, namely: unprotected transient overpower, unprotected loss of flow, and unprotected loss of heat sink.

These analyses are thought to give a basic understanding of the MSFR dynamics, as well as significant indications in terms of the system safety parameters.
Experimental Study of Heat Transfer in a 7-Element Bundle Cooled With Supercritical Freon-12
Authors: Richards G. , Samuel J. , Shelegov A. S. , Kirillov P. L. , Pioro I. L. , Harvel G.
Published: October 2013

Experimental data on SuperCritical-Water (SCW) cooled bundles are very limited. Major problems with performing such experiments are technical difficulties in testing and experimental costs at high pressures, temperatures and heat fluxes. Also, there are only a few SCW experimental setups currently in the world capable of providing data. SuperCritical Water-cooled nuclear Reactors (SCWRs), as one of the six concepts of Generation IV reactors, cannot be designed without such data. Therefore, a preliminary approach uses modeling fluids such as carbon dioxide and refrigerants instead of water are practical. In particularly, experiments in supercritical refrigerant-cooled bundles can be used. One of the SC modeling fluids typically used is Freon-12 (R-12) with the critical pressure of 4.136 MPa and the critical temperature of 111.97°C.

A set of experimental data obtained at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia) in a vertically-oriented bundle cooled with supercritical R-12 was analyzed. This dataset consisted of 20 runs. The test section was 7-element bundle installed in a hexagonal flow channel with 3 grid spacers. Data was collected at pressures of approximately 4.65 MPa for several different combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature. The values of mass flux were ranged from 400 to 1320 kg/m2s and inlet temperatures ranged from 72 to 120°C. The test section consisted of fuel-element simulators that were 9.5 mm in OD with the total heated length of about 1 m. Bulk-fluid and wall temperature profiles were recorded using a combination of 8 thermocouples.

Analysis of the data has confirmed that there are three distinct heat-transfer regimes for forced convention in supercritical fluids: 1) Normal heat transfer; 2) Deteriorated heat transfer characterized with higher than expected wall temperatures; and 3) Enhanced heat transfer characterized with lower than expected wall temperatures. It was also confirmed that the effects of spacers are evident which was previously observed in sub-critical experimental data. Further analysis needs to be conducted on the deteriorated heat transfer phenomena for low mass flux cases which represent accident scenarios. This can be done by designing a natural circulation experimental test loop.
Identification of Non-Safety Systems to Withstand Beyond Design Basis Events
Authors: Nygaard James , Wan Ping , Chan Desmond , Barrientos Sara
Published: October 2013

As an aftermath of the natural disasters affecting the Fukushima Daiichi nuclear power plants in Japan, there has been great attention to provide assurance of safety of nuclear power plants around the world. Accordingly, many countries are requiring “stress tests” for their plants to assess the ability to withstand disaster scenarios for which they were not originally designed. Additional efforts are underway to capture lessons learned related to the operation of critical or major systems. Each operator and each country’s regulatory authority may be at different levels of completion for these activities. However, effects on non-safety related or peripheral systems have not been specifically addressed as standalone items or in an integrated systems approach.

This paper seeks to produce an initial assessment of vulnerable systems, structures or components of non-safety related areas that may become critical to the safe operation of a nuclear plant or to the first steps to maintain stability of the plant during a postulated beyond design basis event. The same assessment is valid for events of significant magnitude, or for events affecting the entire site or region, even if a plant’s design basis is not exceeded.

The initial assessment is based on widespread events, such as at the Fukushima Daiichi station, with focus on large nuclear power reactors. Certain peripheral plant systems support plant operators and staff or emergency responders such as by affording communication or physical access to plant areas. Other peripheral systems support plant operation or recovery, for example provision of diverse power supply or cooling means. Passive components common to multiple systems such as cables and piping are also assessed. Once vulnerable systems, structures or components are identified, various modifications or mitigation approaches will be discussed.
Void Fraction Covariance in Two-Phase Flows
Authors: Brooks Caleb S. , Liu Yang , Hibiki Takashi , Ishii Mamoru
Published: October 2013

A void fraction covariance in the expression for area-averaged local relative velocity has been assumed to be one by current and past researchers. Similarly, in a multi-bubble group approach, void covariances for each bubble group appear in the group area-averaged relative velocity expressions. The covariance terms have been analyzed with a substantial database from literature including upward flow in pipe diameters of 1.27 cm to 15.2 cm, downward flow in pipe diameters of 2.54 cm and 5.08 cm, and upward flow in an annulus (Dh = 1.9cm) under adiabatic, boiling, and condensing conditions. Simple relations are proposed to specify the covariance in order to improve the prediction of area-averaged local relative velocity. The correlations were found to agree well with the experimental data for the flow configurations and conditions analyzed.
Vibration Analysis of Coolant Pump With Two Unbalanced Disks Based on the State-Space Newmark Method
Authors: Wang YuJin , Wang DeZhong , Hu YaoYu
Published: October 2013

This paper discusses the vibration of the latest vertical coolant pump with two disks and nonlinear bearing oil film force, namely rotor-bearing-disk system. To analyze the vibration of the rotor, the finite element method is adopted with the Timoshenko beam, and the dynamical equations are solved by using state-space Newmark method. Results of this study reveals the different mass eccentricity and phase angle of two disks are the important variables to the rotor with two disks under bearing nonlinear oil film force and periodic force, and affect the critical speed obviously.
A Strain Based Failure Criterion for Stainless Steel 316L
Authors: Akinci Necip Onder , Jaquay Kenneth
Published: October 2013

This study presents a proposed failure criterion for evaluation of stainless steel 316L plates subjected to multiaxial stresses. Effects of membrane and bending stresses are investigated. The experimental data used in the development of the failure criterion is based on 316L forming limit diagram. Strains obtained from the forming limit diagram are used for calculation of stresses and triaxiality factors. The proposed failure criterion is compared with those available in the literature. Effects of bending stresses on the ductility of 316L plates are discussed.
Experimental Investigation of Friction Reduction in Transport of Fluids Using Surfactants
Authors: Sanchez-Silva Florencio , Carvajal-Mariscal Ignacio , Quinto-Diez Pedro , Barbosa-Saldana Juan Gabriel
Published: October 2013

This paper presents the results of the influence of surfactants in reducing friction while driving fluid in pipes. Experimental research was conducted with water-surfactant mixtures which were tested three types of these additives: anionic, cationic and nonionic surfactants. Was designed and built an experimental facility in which the test area was acrylic pipe with an inner diameter of D = 19 mm and a length of 300 D. The concentrations of surfactants in the mixtures were 150, 250, 500, 750 and 1000 ppm, added according to a pilot program that took into account the amount and type of additive added in different liquid mass fluxes. Pressure losses were compared against those obtained when flow is the same water flows through the installation. The results obtained show a reduction of up to 43.9% of the friction which is achieved with a Re = 11243 and surfactant concentrations of 250 ppm (cetyl trimethyl chloride ammonium), to which was added as a stabilizer for the micro structure of the surfactant, sodium salicylate, which applies only to the cationic type surfactants. The results are promising but left to study such issues as: the injection and recovery of surfactant, more efficient mixing, the mechanisms that lead to a reduction of friction and the effect of temperature among others.
Reactor Kinetics in a Loss-of-Forced-Cooling (LOFC) Test of HTGRs
Authors: Takamatsu Kuniyoshi , Sawa Kazuhiro
Published: October 2013

The High-Temperature Engineering Test Reactor (HTTR) is the first High-Temperature Gas-cooled Reactor (HTGR) with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of JAEA. At present, test studies are being conducted using the HTTR to improve HTGR technologies in collaboration with domestic industries that also contribute to foreign projects for the acceleration of HTGR development worldwide. To improve HTGR technologies, advanced analysis techniques are currently under development using data obtained with the HTTR, which include reactor kinetics, thermal hydraulics, safety evaluation, and fuel performance evaluation data (including the behavior of fission products). In this study, a three gas circulator trip test and a vessel cooling system (VCS) stop test were performed as a loss of forced cooling (LOFC) test to demonstrate the inherent safety features of HTGR. The VCS stop test involved stopping the VCS located outside the reactor pressure vessel to remove the residual heat of the reactor core as soon as the three gas circulators are tripped. All three gas circulators were tripped at 9, 24 and 30 MW. The primary coolant flow rate was reduced from the rated 45 t/h to 0 t/h. Control rods (CRs) were not inserted into the core and the reactor power control system was not operational. In fact, the three gas circulator tripping test at 9 MW has already been performed in a previous study. However, the results cannot be disclosed to the public because of a confidentiality agreement. Therefore, we cannot refer to the difference between the analytical and test results.

We determined that the reactor power immediately decreases to the decay heat level owing to the negative reactivity feedback effect of the core, although the reactor shutdown system was not operational. Moreover, the temperature distribution in the core changes slowly because of the high heat capacity due to the large amount of core graphite. Core dynamics analysis of the LOFC test for the HTTR was performed. The relationship among the reactivities (namely, Doppler, moderator temperature, and xenon reactivities) affecting recriticality time and reactor peak power level as well as total reactivity was addressed. Furthermore, the analytical results for a reactor transient of hundred hours are presented. Based on the results, emergency operating procedures can be developed for the case of a loss of coolant accident in HTGR when the CRs are not inserted into the core and the reactor power control system is not operational. The analytical results will be used in the design and construction of the Kazakhstan High-Temperature Reactor and the realization of commercial Very High-Temperature Reactor systems.
Research on the Loading Pattern of Reactor Core With MOX Fuel in the First Cycle
Authors: Zhang Tianqi , Yu Shihe , Cao Xinrong
Published: October 2013

In order to research the performance of Pressurized Water Reactor (PWR) with 1/3 MOX fuel in the initial cycle, this paper serves Qinshan II reactor core as the reference core to design suitable MOX assemblies and study relevant core properties. The analyses documented within use assembly cross section calculation code CASMO-4 and core calculation code SIMULATE-3 studied by Studsvik. The purpose of this paper is to demonstrate that the Qinshan II reactor is capable of complying with the requirement for MOX fuel utilization without significant changes to the design of the plant. Several impacts on key physics parameters and safety analysis assumptions, introduced by MOX, are discussing in the paper.
Supercritical-Water Experimental Setup for In-Pile Operation
Authors: Miletić M. , Růžičková M. , Fukač R. , Pioro I. , Peiman W.
Published: October 2013

The main goal of the Generation-IV nuclear-energy systems is to address the fundamental research and development issues necessary for establishing the viability of next-generation reactor concepts to meet future needs for clean and reliable energy production. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable.

One of the six Generation-IV concepts is a SuperCritical Water-cooled Reactor (SCWR), which continues the utilization of well-known light-water-reactor technologies. Research Centre Rez Ltd. has taken part in a large European joint-research project dedicated to Generation-IV light-water reactors with objectives to contribute to the fundamental research and development of the SCWRs by designing and building a test facility called “SuperCritical Water Loop (SCWL)”.

The main objective of this loop is to serve as an experimental facility for in-core and out-of-core corrosion studies of structural materials, testing and optimization of suitable water chemistry for future SCWRs, studies of water radiolysis at supercritical conditions and nuclear fuels. This paper summarizes the concept of the SCWL, its design, utilization and first results obtained from non-active tests already performed within the supercritical-water conditions.
Analysis of Graphite Dust Deposition in Steam Generator of HTR
Authors: Peng Wei , Yang Xiao-yong , Yu Su-yuan , Wang Jie
Published: October 2013

The behavior of the graphite dust in the steam generator is important to the safety of High Temperature gas-cooled Reactor (HTR). The present study numerically investigates the effect of steam generator structure on deposition and diffusion of graphite dust particles in HTR, then the effect of helium velocity, tube spacing and graphite dust particle diameter on the deposition and diffusion characteristics of graphite dust are analyzed. The results showed that the flow field around the heat transfer tube is complex and high turbulence intensity; the deposition efficiency of the graphite dust will increase as the tube spacing increases, while the deposition efficiency of the graphite dust will decrease with the inlet velocity increases.
How the Use of Limited Plant Data Can Support Limited Improvements in Combined Cycle Operation
Authors: Perez James M. , Toburen Tina L.
Published: October 2013

Process improvement and plant optimization are common initiatives among Utilities and Independent Power Producers (IPP). This paper intends to outline the value and challenges with obtaining data to support these initiatives and provide for good business decisions.

Typical information, such as performance acceptance tests and heat balance design data is essential when evaluating plant performance. Normal, on-going operating data at full load conditions is also needed when making decisions regarding plant improvements. However, obtaining all the necessary data is not always possible and with the current market conditions, conducting performance tests may not be economical. When faced with a known performance deficiency, determining the root cause and options for performance recovery can be difficult in any situation. When design data is hard to come by and operating data difficult to obtain, making an informed decision can be more of an art than a science.

An example of the decision process used to upgrade the capacity for a gas turbine combined cycle facility is explained in detail in this paper. The results of the process show that even with limited available data for the facility, improvements in performance can be achieved.
In Vessel Loss of Coolant Accident Analysis for the JET Tokamak in Support of the Safety Case Review
Authors: Vallory Joelle , Boyer Helen , Moreno Fabien , Kadri Dominique
Published: October 2013

The Joint European Torus is currently the largest operating tokamak in the world. JET was established with a long-term objective to contribute to the creation of safe, environmentally sound prototype fusion reactors. To meet this objective, JET operates using deuterium as a fuel but also in D-T mode where the fusion reaction is fuelled by deuterium and tritium. To justify the safety of D-T experiments, safety reports have been produced at stages to obtain approval for the past experimental campaigns. The Safety Case is currently under review in preparation of the next tritium campaign in 2014/ 2015.

In this prospect, the hazard assessments have been reviewed and when possible, the calculations in support of these assessments have been updated. For the in vessel loss of coolant accident, the state of the art CEA developed thermal hydraulics code CATHARE 2 V2.5_2, has been adapted down to pressure relevant to fusion applications. The complexity of the heat transfer mechanisms involved in the vacuum vessel has been represented with different elements available in the code to improve the 0D model. Modeling parameters have been studied and any compromise in modeling was done by increasing conservatism of the peak pressure in the VV. Some physical parametric studies have also been run especially on the failure of safety equipment designed to minimize the in vessel consequences on the LOCA. One major result of the analysis is that the initial calculations were conservative in their assumptions and that some break size initially considered as beyond design could lead to pressure peak in the vessel which could be limited by the safety systems below the bursting disk pressure.
Performance Evaluation of Evaporative Compressor Inlet Air Cooling System in a Gas Turbine-Based Cogeneration Plant
Authors: Zabihian Farshid , Fung Alan S. , Schuler Fabio
Published: October 2013

Gas turbine-based power plants are very sensitive to ambient conditions and their output power and efficiency can be decreased significantly with increase in the ambient temperature. Various compressor inlet air cooling systems have been proposed and utilized to reduce inlet air temperature to the system, including evaporative systems e.g. media and fogging, and mechanical cooling systems. In this work, different techniques for compressor inlet air cooling are briefly reviewed. Then, the fogging system employed in the Whitby cogeneration power plant is explained with particular attention to the location of the system installation. A model of the gas turbine-based cogeneration plant is also developed to simulate the Whitby cogeneration power plant. The effects of fogging compressor inlet air cooling system on the performance of the plant are investigated. The results indicate that at an ambient temperature of 30°C and relative humidity of 40% the inlet cooling of as high as 8.4°C is possible which can increase output power to more than 50 MW. Also, it is found that the model can predict the gas turbine exhaust temperature and the plant’s power production with the error level of lower than 0.5% and 3%, respectively.
Performance Experienced From Switching Copper Condenser Tubing to a High Performance Stainless
Authors: Goebel Summer , Criminger Colin R.
Published: October 2013

Over a period of five years, the condensers of two sister 500 MW Units, in the Midwest, were retubed using a SEA-CURE® S44660 material. S44660 is a high alloy, super ferritic stainless steel. The S44660 material replaced the original Admiralty and copper/nickel condenser tubing. Several factors contributed to the decision to use S44660 as the condenser tube material. The critical factor in the decision was a pending lower copper discharge limit in the plant’s National Pollutant Discharge Elimination System (NPDES). The plant was unable to meet the lower copper limits with copper condenser tube material.

In service since the early 1970’s, Unit A and B condensers were subjected to conditions, including silting and operational upsets, which negatively affected condenser performance. A third 500 MW sister Unit, Unit C, operational in the mid 80’s, experienced similar conditions. All three units are based load coal units and use cooling towers to cool the circulating water. The cooling tower make-up comes from the Arkansas River, which also receives the effluent from the plant.

During the summer of 2011, shortly after the Unit B condenser retube, both A & B condensers showed degradation in performance. Through a series test and trials, the causes of degradation were identified and corrected. The cause was a combination of the tube material, lower river levels, higher than normal temperatures, macro fouling, and air in leakage.
Simulation of Mixing Induced by a Hot PAR Exhaust Plume
Authors: Andreani Michele , Kelm Stephan
Published: October 2013

Passive Autocatalytic Recombiners (PARs) are installed in various reactor containment designs to mitigate the hydrogen risk. For the evaluation of the effectiveness of these devices, validated computational tools are needed. To build confidence in the codes, their capability must also be assessed against separate effect tests addressing specific phenomena. Within the OECD SETH 2 project three experiments have been performed in the large-scale PANDA facility, where the thermal effect of a PAR was simulated by means of a heater and the plume generated by the heat source interacted with an initially stratified ambient. In these tests, helium was used instead of hydrogen. The position of the heater and the presence of simultaneous injection of steam were varied in these tests. These experiments have been analyzed with the GOTHIC and the ANSYS CFX codes. This paper reports only the results obtained with the GOTHIC code. In general, the GOTHIC code in conjunction with a coarse mesh could predict the mixing process reasonably well. The only substantial discrepancy with the experiments was the overprediction of the velocity at the inlet of the heater case, but this had little effect on the simulation of the overall mixing.
Development of a Safety Analysis Code SCTRAN for SCWR
Authors: Gou Junli , Wu Pan , Shan Jianqiang , Zhang Bo , Yang Jue
Published: October 2013

Safety analysis is one of the chief difficulties during the research and design of SCWRs. Currently, the development of the SCWR safety analysis code is still in its infancy all around the world, and very few computer codes could carry out the trans-critical calculations where significant changes in water properties would take place. In this paper, a safety analysis code SCTRAN for SCWRs, has been developed based on RETRAN-02, the best estimate code used for safety analysis of Light Water Reactors. The ability of SCTRAN to simulate transients where both supercritical and subcritical regimes are encountered has been verified by comparing with the APROS and RELAP5-3D codes, respectively. The results show that the SCTRAN code developed in this study is capable of performing safety analysis for SCWRs, and the results are reliable. SCTRAN developed in this study is of great value and engineering significance, and could be utilized in the Chinese SCWR research and development.
Capability of the Integral Code ASTEC V2 to Simulate a Blow Down in Plant Scale Tests
Authors: Brähler Thimo , Risken Tobias , Koch Marco K.
Published: October 2013

In this paper the Accident Source Term Evaluation Code (ASTEC) is validated against the blowdown experiments Marviken M19 and M24. These tests mainly differed by the mass flow of the released steam from the pressure vessel and the configuration of the vent pipes in the pressure suppression chamber — while in M19 all vent pipes were arranged to one pool, in M24 they were split up by 27 pipes in one and one pipe in another pool.

For the simulation of both tests, an existing model of the facility for another lumped parameter code COCOSYS was transferred to ASTEC. In this data set a simple zone and a flow connection was used to model the pressure suppression chamber. Further simulations were performed with another approach for the pressure suppression chamber, so called “DRASYS”-zones. Using the DRASYS model, the user has to specify more inputs for the geometry of the pressure suppression chamber. The results of the simulations are in good agreement to the measured pressure and temperature of both tests. By using the DRASYS model in ASTEC, the results were improved slightly for M19 compared to the simple pressure suppression zone model. In opposite, the results of the simple model are in better agreement to M24. Overall the conclusion is that ASTEC is able to simulate a Blow Down in plant scale with both models.
Plant Dynamics Evaluation of a Monju Ex-Vessel Fuel Storage System During a Station Blackout
Authors: Mori Takero , Sotsu Masutake , Honda Kei , Suzuki Satoshi , Ohira Hiroaki
Published: October 2013

The prototype fast breeder reactor “Monju” has an ex-vessel fuel storage system (EVSS) which consists mainly of an ex-vessel fuel storage tank (EVST) and an EVST sodium cooling system. EVST uses natural circulation of sodium for decay heat removal. Natural circulation in the EVST is generated by the decay heat from the spent fuel assemblies and the cooling of the cooling coils installed in the EVST. The EVST sodium cooling system consists of three independent loops. In each loop, sodium is circulated by electromagnetic pumps and the heat is removed by an air cooler with blowers. This system has the ability to remove the maximum decay heat using two loops, and thus, it uses two of the three loops for normal operation. During a station blackout (SBO), the pumps and blowers are stopped. However, the three air coolers are installed about 13.5 m higher than the cooling coils, and therefore, the EVST sodium cooling system potentially retains some cooling ability because of natural circulation. In this study, an analysis and evaluation of the plant dynamics for the spent fuel and the EVSS structural integrity during an SBO were performed.

The ultimate heat sink for the EVST sodium cooling system is the atmosphere, and the air coolers have an exhaust stack for efficient natural circulation caused by the chimney effect. However, the EVST sodium cooling system loses pressure and the heat transfer characteristics change if the flow rate is low. It was, therefore, necessary to confirm the temperature and flow rate behavior of EVSS in this analysis.

In the present calculations, the plant dynamics analysis program “Super-COPD” was used. The factors affecting the cooling ability were investigated and analytical cases were determined. In one case, the two operated loops were switched to natural circulation after an SBO. The number of cooling loops was then changed from two to three by having an operator open the vane and dampers of the standby loop. In this case, sodium temperature in the EVST increased to approximately 320°C. When the number of cooling loops was not changed and natural circulation occurred in only two loops, the sodium temperature in the EVST increased to approximately 450°C. In both cases, however, the structural integrity of the EVSS was maintained. These analytical results, therefore, help clarify the number of necessary cooling loops for efficient decay heat removal and sodium temperature behavior in an SBO.
Seismic Fragility Analysis of a Condensate Storage Tank With Multiple Uncertain Degradation Scenarios
Authors: Nie Jinsuo R. , Braverman Joseph I. , Hofmayer Charles H. , Choun Young-Sun , Kim Min Kyu , Choi In-Kil
Published: October 2013

The Korea Atomic Energy Research Institute (KAERI) and Brookhaven National Laboratory (BNL) are continuing a collaborative effort to achieve a better understanding of the effects of aging on the performance of structures and passive components (SPCs) in nuclear power plants (NPPs). This paper presents a seismic fragility analysis of a condensate storage tank (CST) with multiple degradation scenarios that are treated in a non-perfectly correlated manner. The analysis utilizes a set of optimum Latin Hypercube samples to characterize the deterioration behavior of the fragility capacity as a function of age-related degradations. This study is an addition to the previous study summarized in an ICONE19 paper entitled “Seismic Fragility Analysis of a Degraded Condensate Storage Tank” [1], which considered individual degradation scenarios and multiple degradations occurring in a perfectly correlated manner.