bluuebunny/crossref_metadata_2025 huggingface dataset:
Analysis of Bypass Effect on a 250 MW High Temperature Gas-Cooled Reactor
Authors: Zheng Yanhua , Chen Fubing , Shi Lei
Published: November 2014

Pebble bed modular high temperature gas-cooled reactors (HTR), due to their characteristics of low power density, slender structure, large thermal inertia of fuel elements and reactor component materials (graphite), have good inherent safety features. However, the reflectors consisting of large piles of graphite blocks will form huge numbers of certain bypass gaps in the radial, axial and circumferential directions, thus affecting the effective cooling flow into the reactor core, which is one of the concerned issues of HTRs.

According to the preliminary design of the Chinese high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the thermal-hydraulic calculation model is established in this paper. Based on this model, considering different bypass flow, that is to say, different core cooling flow, fuel element temperature, outlet helium temperature and the core pressure drop in the normal operation, as well as the maximal fuel temperature during the depressurized loss of forced cooling (DLOFC) accident are analyzed. This study on bypass effects on the steady-state and transient phases can further demonstrate the HTR safety features.
233U Evaluation Comparison Study
Authors: Alrwashdeh Mohammad , Abdalla Aniseh , Wang Kan
Published: November 2014

The aim of this study is to investigate the availability and accuracy of the cross section data for 233U to perform the calculations of the critical system. Two evaluated data libraries are available, U.S. data bank (ENDF) and the Japanese data bank (JENDL), by using BAYES method for resonance parameters available in SAMMY code and weighted least square method with nonlinear regression by using FITWR computer code. Evaluation of the 233U has been investigated by using of SAMMY code, in order to generate a useful data base for critical calculations, the computer code FITWR for experimental - experimental data fitting show same results obtained from Bayes method included within SAMMY code, with a slight deference in the results at the evaluated cross sections due to different mathematical methods have different results.
Application of Dual-Adaptive Niched Genetic Algorithm in Optimal Design of Nuclear Power Components
Authors: Wang Cheng , Yan Chang-qi , Wang Jian-jun , Chen Lei , Li Gui-jing
Published: November 2014

Genetic algorithm (GA) has been widely applied in optimal design of nuclear power components. Simple genetic algorithm (SGA) has the defects of poor convergence accuracy and easily falling into the local optimum when dealing with nonlinear constraint optimization problem. To overcome these defects, an improved genetic algorithm named dual-adaptive niched genetic algorithm (DANGA) is designed in this work. The new algorithm adopts niche technique to enhance global search ability, which utilizes a sharing function to maintain population diversity. Dual-adaptation technique is developed to improve the global and local search capability at the same time. Furthermore, a new reconstitution operator is applied to the DANGA to handle the constraint conditions, which can avoid the difficulty of selecting punishment parameter when using the penalty function method. The performance of new algorithm is evaluated by optimizing the benchmark function. The volume optimization of the Qinshan I steam generator and the weight optimization of Qinshan I condenser, taking thermal-hydraulic and geometric constraints into consideration, is carried out by adopting the DANGA. The result of benchmark function test shows that the new algorithm is more effective than some traditional genetic algorithms. The optimization design shows obvious validity and can provide guidance for real engineering design.
Safety Analysis of a High Breeding Fast Reactor Cooled by Supercritical-Pressure Light Water: Accidents and Abnormal Transients at Supercritical Pressure
Authors: Guo Rui , Oka Yoshiaki
Published: November 2014

The high breeding core of a supercritical water cooled fast reactor has been designed with tightly packed fuel rod assemblies to obtain a high breeding ratio and negative reactivity coefficients. A high breeding capability of less than 50 years of compound system doubling time was reported by the conceptual design study of the reactor at ICONE21. The present paper describes the safety analysis of the reactor for accidents and abnormal transients at supercritical pressure. The safety principle, safety system configuration and types of the abnormal transients and accidents are the same as those of the Super FR [1]. Safety criteria such as cladding temperature, pressure and fuel enthalpy are similar to those of Super FR. Results indicate that all safety criteria are satisfied.
Large Electro-Magnetic Pump Design for Application in the ASTRID Sodium-Cooled Fast Reactor
Authors: Aizawa Rie , Suzuki Tetsu , Laffont Guy , Rey Frédéric
Published: November 2014

Within the framework of the French Sodium Fast Reactor (SFR) prototype called ASTRID (Advance Sodium Technological Reactor for Industrial Demonstration), an application of Large capacity Electro-Magnetic Pumps (LEMP) is considered as a main concept of the circulating pump on intermediate sodium circuits. The use of LEMP has several merits in the design of reactor, operation, and maintenance. Furthermore, high efficiency is acquired when heat-resistant coil insulation is used for this LEMP. Nevertheless, some theoretical and technological developments have to be carried out in order to validate the design tools by taking into account Magneto Hydro Dynamic (MHD) phenomena and the applicability of the LEMP to ASTRID steady state and transient operating conditions. In this aim, a collaboration agreement between the CEA and TOSHIBA Corporation came into force in April 2012 to carry out a joint work program on the ASTRID EMP design and development. This paper describes the dedicated design studies and experimental activities for the LEMP development within the framework of the CEA-TOSHIBA collaboration.
A New Paradigm: Cycling Operations at Nuclear Power Plants in the United States
Authors: Hilleman Douglas , Kumar Nikhil , Lefton Steven
Published: November 2014

Nuclear power plants are no longer immune to cycling operation. While certain nuclear power plants in Europe have been performing load following operation, this type of operation has largely been avoided in the United States. Due to increasing contribution of nuclear generation in the mix, European operators were forced to make modifications to increase the maneuverability of their nuclear generation assets. However, in the United States, nuclear generation is still a relatively smaller contributor (19%). Still, with rapid increase in renewable generation, some nuclear plants are being asked to operate at reduced power and cycle to lower power levels. With most future renewable integration studies advocating for increased flexibility on the grid, nuclear generation maneuverability will allow system operators with another resource to mitigate system costs.

This paper presents the results of a detailed study of a 1,150 MW boiling water reactor nuclear plant when cycled to low loads. The authors present the relative damage of cycling to various reduced power levels 80% to 15% power levels compared to a cold startup and shutdown of a nuclear plant. An assessment was made of the systems that had fatigue damage and costs. We also discuss some of the limitations of cycling that a nuclear plant has and present and discuss recommendations to reduce damage and costs.
The Code Development for the Thermo-Hydrodynamic Characteristics Study of Reflood Phase
Authors: Niu Shipeng , Zhang Yapei , Tian Wenxi , Qiu Suizheng , Su Guanghui
Published: November 2014

An thermo-hydraulic analysis model for the reflood phase was established in the paper, based on the flow and heat transfer characteristics of the reflood. A code based on the model was developed for the thermo-hydraulic analysis of reflood. By comparing and analyzing the calculation results with the experimental results, the influences of the system pressure and the subcooling of coolant on reflood phase were studied. The study also provided theoretical basis for safety analysis of fuel element of reflood phase.
Thermal Modeling of the HTR-10 Using the RELAP5-3D Code
Authors: Scari Maria Elizabeth , Costa Antonella Lombardi , Pereira Claubia , da Silva Clarysson Alberto Mello , Veloso Maria Auxiliadora Fortini
Published: November 2014

Several efforts have been considered in the development of the modular High Temperature Gas cooled Reactor (HTGR) planned to be a safe and efficient nuclear energy source for the production of electricity and industrial applications. In this work, the RELAP5-3D thermal hydraulic code was used to simulate the steady state behavior of the 10 MW pebble bed high temperature gas cooled reactor (HTR-10), designed, constructed and operated by the Institute of Nuclear and New Energy Technology (INET), in China. The reactor core is cooled by helium gas. In the simulation, results of temperature distribution within the pebble bed, inlet and outlet coolant temperatures, coolant mass flow, and others parameters have been compared with the data available in a benchmark document published by the International Atomic Energy Agency (IAEA) in 2013. This initial study demonstrates that the RELAP5-3D model is capable to reproduce the thermal behavior of the HTR-10.
Adsorption Properties of Sr(II) on Zeolite Type Adsorbents and Their Irradiation Stabilities
Authors: Mimura Hitoshi , Susa Shunsuke , Ito Yoshiyuki , Saito Yasuo , Matsukura Minoru
Published: November 2014

Radioactive waste treatment is planned in LWTF (Low-level radioactive Waste Treatment Facility, JAEA) for LLW generated from the Tokai-reprocessing facility. The target LLW consists of highly concentrated sodium nitrate (5 M NaNO3) containing low-level 90Sr. In this study, selective adsorption properties of Sr2+ for highly functional A type zeolites (A51-JHP, A51-J (Union Showa) and A-4, X type zeolite (F-9) and Titanic acid-PAN (polyacrylamide) were clarified by batch and column adsorption methods. The irradiation stabilities of these adsorbents were also evaluated.

The distribution properties of Sr2+ on different adsorbents were compared in simulated waste solution (5 M NaNO3, 0.1 ppm Sr2+, 85Sr as tracer). The order of distribution coefficients (Kd,Sr) was Titanic acid-PAN > A51-JHP > A51-J > A-4 > F-9. The largest value of Kd,Sr for titanic acid-PAN was estimated to be 218 cm3/g, while the saturated capacity (Qmax) was very small. Titanic acid-PAN had also the largest uptake rate of Sr2+ ions and the uptake attained equilibrium within 8 h. On the other hand, A51-JHP had a relatively large Kd,Sr value above 100 cm3/g and a Qmax value of 0.65 mmol/g.

The breakthrough properties of Sr2+ were examined by varying cations present (single and mixed solutions) and flow rate (0.08 and 0.17 cm3/min). The components for the single solution were 400 g/L NaNO3, 100 ppm Sr2+, 85Sr as tracer, and the mixed solution contains 200 ppm Cs+, 100 ppm Ca2+, 50 ppm Mg2+, 50 ppm RuNO3+ in addition to the single solution components. The breakthrough curve for Titanic acid-PAN column using single solution had an S-shaped profile, while the “concentration phenomenon” exceeding C/C0 (breakthrough ratio) = 1 was observed in the case of mixed solution. As for the A51-JHP column, the breakthrough curve for single solution was similar to that for mixed solution and the 5% breakpoint was enhanced by decreasing the flow rate.

The A51-JHP was stable under 60Co-irradiation up to 2.54 MGy; Kd,Sr and Qmax values were almost constant. In contrast, Titanic acid-PAN was affected above 0.28 MGy, due to the radiolysis of PAN matrix, and this surface alteration led to the release of active component of titanic acid.

The novel A type zeolite (A51-JHP) is thus expected for the selective removal of Sr2+ in LWTF. The optimization of particle size and flow rate should be examined before practical use.
Research on Performance of Gas Methyl Iodide Removal
Authors: Zhou Yanmin , Sun Zhongning , Gu Haifeng , Wang Junlong , Miao Zhuang
Published: November 2014

The characteristics of gas methyl iodide removed by containment filter venting system is similar with bubble column reactor, this study is focused on the absorption performance of methyl iodide in a bubble column reactor with experiment and mathematical calculations, the alkalescent sodium thiosulphate solution in a scale-up facility is used as absorber in the research. The results show that the gas methyl iodide removal efficiency is mainly influenced by the temperature of solution, system pressure, gas flow rate and liquid level respectively. In the range of 0∼80°C, the removal efficiency improve obviously with increasing temperature, while the chemical reaction process is a major factor that limiting the removal efficiency, when the temperature is higher than 80°C, the efficiency is no longer sensitive to the variation of temperature and the mass transfer process become the main limiting factors. The increase of system pressure and height of solution can enhance gas absorption process significantly, the removal efficiency improve linearly with the two parameters. However, the gas volume flow rate plays an opposite role on absorption process mainly because of the reducing on contacting time. In addition, the variation of entrance concentration has a little impact on removal efficiency. The mathematical calculations of removal efficiency fit well with experimental results in the bubble column reactor.
Application of Hybrid Subgroup Decomposition Method to Gas Cooled Thermal Reactor Problem
Authors: Yasseri Saam , Rahnema Farzad
Published: November 2014

In this paper, a newly developed hybrid subgroup decomposition method is tested in a 1D problem characteristic of gas cooled thermal reactors (GCR). The new method couples an efficient coarse-group eigenvalue calculation with a set of fine-group transport source iterations to unfold the fine-group flux. It is shown that the new method reproduces the fine-group transport solution by iteratively solving the coarse-group quasi transport equation. The numerical results demonstrate that the new method applied to 1D GCR problem is capable of achieving high accuracy while gaining computational efficiency up to 5 times compared to direct fine-group transport calculations.
Analysis of Some Accidents With Deborated Slug Passing Through the WWER Core
Authors: Ponomarenko Grigory L.
Published: November 2014

Deborated slug may be formed by the different ways. Stagnant areas may be arise in circulation loops when Reactor Coolant Pumps (RCP) do not work and decay heat removing system works. First RCP start-up in the loop with formed deborated slug is the most dangerous situation. Accelerated slug enters into the core and quickly (during 1 s) may bring reactor from deep subcriticality into above-critical state and generate dangerous neutron splash. Discharged energy may damage or distract fuel rods. However in reality, at the core enter there is not a distillate with zero concentration of boric acid (CB=0 g/kg). There is a coolant with lowered 0 < CB < 16 as compared to shut-down value (16 g/kg), as a result of coolant mixing and washing-out of deborated slug.

Analysis of this beyond design basis accident was performed with use of certified system coupled code KORSAR/GP [1–3] with spatial neutron kinetics and account of incomplete coolant mixing in reactor.

Mixing process was modeled by KORSAR/GP on the base of experimental results obtained on the experimental facility scaled WWER-1000 as 1:5. As a result of the carried out work the technique of the analysis is developed. Essential features of mode with deborated slug are revealed. Necessary provisions are determined, and recommendations for safety maintenance and improve are formulated. Safe volumes of deborated slugs and the time intervals of inadvertent deboration were determined. These not dangerous volumes and corresponding time intervals appeared big enough to provide high probability not to exceed them during start-up of modernized WWER-1000 power units. Influence of increase in CR quantity from 61 up to 121 pcs. on safety parameters in the given mode is evaluated inter alia.
Lab-Scale Design, Demonstration and Safety Assessment of an FPGA-Based Post Accident Monitoring System for Westinghouse AP1000 Nuclear Power Plants
Authors: McNelles Phillip , Lu Lixuan
Published: November 2014

A Field Programmable Gate Array (FPGA) is a type of integrated circuit (IC), which is programmed after it is manufactured. FPGAs are referred to as a form of programmable hardware, as there is typically no software or operating system running on the FPGA itself. A significant amount of design work has been performed regarding the application of FPGAs in the nuclear field in recent years, with much of that work centered around safety related Instrumentation and Control (I&C) systems and safety systems. These new FPGA based systems are considered to be viable alternatives to replace many old I&C systems that are commonly used in Nuclear Power Plants (NPPs). Many of these older analog and digital systems are obsolete, and it has become increasingly difficult to maintain and repair them. FPGAs possess certain advantages over traditional analog circuits, PLCs and microprocessors, when considering nuclear I&C and safety system applications. This paper describes how FPGA technology has been used to construct a lab-scale implementation of a Post-Accident Monitoring System (PAMS), for a Westinghouse AP1000 Nuclear Power Plant, using a National Instruments “cRIO” chassis and I/O modules. This system will perform the major functions of the existing PAMS, including monitoring the vital values such as temperature, water level, pressure, flow rate, radiation levels and neutron flux in the event of a serious reactor accident. These values are required in standards such as United States Nuclear Regulatory Commission (NRC), Canadian Nuclear Safety Commission (CNSC), International Electrotechnical Commission (IEC), and Institute of Electrical and Electronics Engineers (IEEE). All of the input signals are read and processed using the FPGA, which includes alarms if the values go beyond the specified range, or if the values change rapidly. The values were then output to the computer through the FPGA interface to provide information to the operator, as well as being sent through analog and digital output modules for further processing. The system was tested using both simulated and real inputs from sensors. Furthermore, the reliability of the new system has also been analyzed, using the Dynamic Flowgraph Methodology (DFM). DFM has been successfully applied in both the nuclear and aerospace fields, and has been described as one of the best methodologies for modeling software/hardware interactions, by the scientific literature as well as in NRC reports. DFM was applied to fine-tune the design parameters by determining the potential causes of faults in the design, as well as to highlight the effectiveness of DFM in nuclear and in FPGA applications.
Structure Design of the Pressured Vessel in Experimental Test on Steam Condensation on the Cold Surface for CAP1400
Authors: Yang Lin , Han Liyong
Published: November 2014

The advanced pressurized water reactor (APWR) uses a passive safety system relying on heat removal by condensation to maintain the containment within the design limits of pressure and temperature. The passive containment cooling system (PCCS) includes many natural phenomena mechanisms. Steam condensation is one of the most important phenomena. It is very significant for engineering designing and code developing to study the mechanism of steam condensation on cold surface. In this paper, the test pressurized vessel in the experimental test on steam condensation on the cold surface for CAP1400 is designed, and the structure pressure is calculated.
The Treatment of Low-Level Radioactive Wastewater by the Nuclear Weapons Using Ultrasonic Standing Wave Methods
Authors: Hou Suxia , Luo Jijun , He Bin , Xu Jun , Li Rusong
Published: November 2014

The radiation hazards of radionuclide arising from the storage of nuclear weapons can not be ignored to the operators. Ultrasonic standing wave methods can be considered as the green cleaning separation techniques with high efficiency, and it allows the elimination of many chemical flocculants to reduce radioactive contamination. The application of ultrasonic standing wave methods for liquid radioactive wastes treatment requires solving many problems connected with the proper selection of the frequency and power of ultrasonic transducers, and the processing time, etc. Based on the model of one single suspended radioactive particle subjected to in the field of ultrasonic standing wave, the principle of the treatment of low-level radioactive wastewater by ultrasound was analyzed. The mathematic equation representing the change of the number of the suspended particles was built and calculated. The theoretical results show that under the action of ultrasonic standing wave, the particle will move toward the wave node plane, and the time of particle reaching the plane become shorter when the radius particle and the frequency and power of ultrasound was enlarged. The experimental apparatus for liquid low-level radioactive wastes treatment was designed according to the results of theoretical analysis. The experimental results show that the radioactive concentration of wastewater could be reduced from 400Bq · L−1 to 9.3Bq · L−1 and the decontamination efficiency was 97.68%. The decontamination efficiency could not be obviously improved by further increasing the treating time.
Development and Application of a UTSG Thermal-Hydraulic Analysis Code
Authors: Cong Tenglong , Su Guanghui , Tian Wenxi , Qiu Suizheng
Published: November 2014

Structural integrity of steam generator should be maintained during operation, since it performs as the pressure and heat transfer boundary of primary side coolant. Localized thermal-hydraulic parameters of secondary side are essential for the analysis of tube wastage, fatigue and failure. In this paper, a three-dimensional thermohydraulics analysis code, named STAF, is developed based on FLUENT. With STAF code, three-dimensional thermohydraulics of secondary side of AP1000 steam generator are generated. This code is developed based on the porous media theory. In this code, the drift flux two-phase model coupled with a simplified flow boiling model is utilized to present two-phase flow among the U-tube bundle. Downcomer, tube bundle, support plates and primary separators in steam generator are considered in STAF code. The calculated results are compared with a general steam generator thermohydraulic analysis code ATHOS, which is developed by EPRI steam generator group. The comparison indicates that STAF code performs well in evaluating thermal-hydraulic parameters in steam generator. The results show that the flow field varies significantly at different position in AP1000 steam generator. Flow vapor quality at the inlet of primary separators varies significantly, which is a severe challenge to the capacity design of separators.
Feasibility of a Nuclear Reactor Dedicated to Desalination
Authors: Jung Yong Hun , Jeong Yong Hoon , Moon Jang Sik
Published: November 2014

The development of a small-sized nuclear heat-only plant with maximized safety features dedicated to seawater thermal desalination was proposed to address both a serious water crisis and nuclear safety issues, which continue to be perennial problems. In this study, the feasibility of a dedicated nuclear heat-only desalination system for a target country was evaluated in comparison with a target nuclear thermal desalination system. First, the target country was selected, and its current energy and desalination status was investigated. The suitable nuclear desalination options for the target country were then selected. Finally, using corresponding analysis tools, performance and economic analyses were conducted for a dedicated nuclear heat-only desalination system and the target nuclear thermal desalination system. The results of the analyses indicate that operating the small-sized nuclear heat-only plant at low pressures coupled with a seawater thermal desalination plant will considerably improve both the safety and economy without a significant loss in desalination performance. In conclusion, the proposed dedicated nuclear heat-only desalination system is expected to have high potential for solving both problems.
Thermal Hydraulic Analysis of a Water Tank With Large Aspect Ratio
Authors: Min Qi , Zhang Li , Wang Hongtao , Zhai Junpeng
Published: November 2014

A special-shaped water tank with large aspect ratio and limited volume for cooling was investigated using computational fluid dynamics. The influence of a separator on the heat transfer ability in the water tank is analyzed. When there is no separator, the arrangement of cooling pipes is very important to the heat transfer and temperature field in the water tank. The total heat flux of the pipe bundle and the temperature field will become bad if the pipe bundle is arranged not uniform in the water tank. Adding a separator can greatly enhance the integral natural convection of cold and hot water in the water tank and a uniform temperature field and regular velocity field could be got. The heat transfer ability for the structure with a separator is better than the structure without a separator, and is not sensible to the arrangement of the pipe bundle. The heat transfer ability also did not change when the position of separator and pipe bundle exchanged, and is not a strong function of the distance between separator and the pipe bundle or the wall of the water tank. Finally, the inclination of the water tank is discussed.
Enhanced Thermal Conductivity UO2-BeO Fuels Fabrication Methods and Their Thermal Performance in Light Water Reactors
Authors: Zhou Wenzhong , Liu Rong
Published: November 2014

An enhanced thermal conductivity UO2-BeO composite nuclear fuel was studied. A methodology to generate ANSYS (an engineering simulation software) FEM (Finite Element Method) thermal models of enhanced thermal conductivity oxide nuclear fuels was developed. Two fabrication methods to produce high thermal conductivity UO2-BeO oxide nuclear fuels were summarized. These two processing routes generated pellets with two different microstructures. The characteristics and microstructures of the fuel are determined for use in FEM thermal models, and the relevant thermal properties for UO2-BeO fuels by two different fabrication methods were determined. The results showed significant increase in the fuel thermal conductivities. The reactor performance analysis showed that the decrease in centerline temperature was 250–350K for the UO2-BeO composite fuel, and thus we can improve nuclear reactors’ performance and safety, and high-level radioactive waste generation.
Cosmic-Ray Muon Imaging of Fukushima Daiichi
Authors: Miyadera Haruo , Morris Christopher , Bacon Jeffery D. , Borozdin Konstantin N. , Fujita Kyouichi , Kume Naoto , Mizokami Shinya , Nakayama Kohichi , Otsuka Yasuyuki , Perry John O. , Sano Yuji , Sugita Tsukasa , Takakura Kei , Yamada Daichi , Yoshioka Kenichi
Published: November 2014

Reactor imaging using scattering of cosmic-ray muon is proposed to assess the damages to the reactors at Fukushima Daiichi. Simulation studies showed feasibility of the reactor imaging with muons, and the technique has been demonstrated at a research reactor, Toshiba Nuclear Critical Assembly, where the reactor core was imaged with spatial resolution of 3 cm after 1 month of exposure time.
Development of Glass Melter Technology for HLLW Vitrification in Japan
Authors: Sakai Akira , Koikegami Hajime , Miura Nobuyuki , Ochi Eiji
Published: November 2014

This paper describes the development of glass melter technology, primarily the liquid fed joule-heated ceramic melter process (LFCM) for the vitrificaton of high-level radioactive liquid waste (HLLW) since 1977 in Japan. In 2013 the active test at the vitrification facility (K-facility) in Rokkasho commercial reprocessing plant was successfully completed for the final acceptance test.

During this period many activities on LFCM process development have been carried out in the engineering scale or the full-scale inactive cold tests including the radioactive laboratory scale hot tests. In particular, the design of melter bottom structure and the operating method should be optimized in order to avoid the operational problems caused by accumulation of noble metals (Ru, Rh, Pd), electro-conducive deposits on the melter bottom.

Through the operation of inactive and active test facilities in Tokai, the design basis for the Tokai Vitrification Facility (TVF) has been provided. The hot operation of the TVF was started in 1995 to demonstrate the LFCM process including the performance of the melter off-gas clean-up system etc. The TVF has provided the basis of the process design and the operation method for the K-facility melter in Rokkasho.

In case of commercial scale vitrification, the glass production rate of the melter should be several times larger than that of the TVF. The K-facility full-scale inactive mock-up melter (KMOC) has been planned to confirm the influence of scale-up factors and the difference between Tokai and Rokkasho wastes. Through the testing operation of the KMOC, which was initially started in 2000, it has been found that the stable formation of a cold cap on a molten glass surface is fundamentally important to avoid the excessive precipitation of noble metals and the yellow phase formation.

The active test of the K-facility has been proceeding under the same conditions as the KMOC, and was successfully completed in May, 2013.

The advanced glass melter development programs have also commenced from 2009 to ensure a more robust and noble metals are compatible with the LFCM system and also to provide a higher processing rate. The second K-facility full-scale inactive mock-up melter (K2MOC) has been installed in the vitrification technology development facility (X-14) at Rokkasho. Its testing operation has commenced from November, 2013.
Fundamental Study on Application of Ultrasonic Velocimetry to Molten Glass
Authors: Ihara Tomonori , Tsuzuki Nobuyoshi , Kikura Hiroshige
Published: November 2014

This paper describes fundamental study intended to apply ultrasonic velocimetry to molten glass. Buffer rod is employed to transmit ultrasound into molten glass. Several ceramics are tested for capability, which regards transmission characteristic and corrosion resistance. Among tested materials, mullite provides perfect wetting with a borosilicate glass and highest transmission ratio. Trailing echo level, which is the level of spurious signal arouse inside buffer rod, is evaluated both numerically and experimentally. Trailing echo level can be improved by using higher frequency of ultrasound. Acoustic field after buffer rod is investigated. Measurement near the buffer rod surface should be avoided due to ununiformed pressure distribution as a normal planner ultrasonic transducer has. Ultrasonic beam become sharp due to absorption on the side wall when buffer rod is equipped. Sound speed in the molten glass is obtained from 1000 °C to 1200 °C. Obtained values are almost constant within this temperature range although small negative slope is observed against temperature. Attenuation coefficient of ultrasound in the molten glass is also measured. The coefficient is well modelled to temperature dependent exponential function. Finally, application of velocimetry via buffer rod is demonstrated. From velocity profiles, maximum measurable range is investigated. Experimental results suggest that velocimetry measurement can be done unless ultrasound energy is attenuated to trailing echo level.
Development of Evaluation Method for Impact of Aircraft Fuel Explosion and Fire on BWR Reactor Building
Authors: Sakai Norio , Shoji Yoshihiro , Jimbo Masakazu , Tanaka Nobuhiko
Published: November 2014

A method was studied to evaluate the influences of aircraft fuel explosion and consecutive fire that would be expected to occur after a postulated aircraft crash on a BWR reactor building. In addition to the classical semi-empirical method for vapor cloud explosion (VCE) to evaluate overpressure, The Fire Dynamics Simulator version 6 (FDS6) was applied to analyze the fireball evolution by combustion of the dispersed fuel droplets, and the consecutive pool fire. The FDS spray model was used to simulate the fuel droplet dispersal. Input parameters such as droplet initial velocity and initial dispersal direction were presumed on the basis of the observations from the IMPACT water missile experiment. Our calculation result was in good agreement with the experimental data in terms of the droplet ‘cloud’ expanding speed by approximately 20%. This modeling manner for the droplet dispersal was also applied to the analysis of explosion in a middle-size commercial aircraft crash to the sidewall of a building that assumed a BWR reactor building. The analysis predicted that approximately 60% of aircraft fuel would be consumed in the initial fireball and most of the residual fuel would attach to the ground area close to the wall. This result can be used for the realistic determination of the initial mass and position of the residual aircraft fuel to evaluate wall integrity against thermal attack and transport of the combustion products from the secondary fire.
Electromagnetic Pump Flow Simulation by 3D MHD Code
Authors: Asada Takatoshi , Hirata Yosuke , Aizawa Rie , Suzuki Tetsu , Fujishima Yasushi
Published: November 2014

Electromagnetic (EM) pumps, which drive liquid metal by electromagnetic force in Fast Reactors, have been developed [1] [2]. Annular Linear Induction Pump (ALIP) is one of EM pumps. There have been some reports in which developed pressure’s drop and fluctuation were described near the top of the relation curve between developed pressure and flow rate (P-Q curve). These phenomena were reported by Gailitis, Kirillov and Araseki. Araseki simulated this phenomenon by two-dimensional (2D) code and reported that it is characterized by vortexes, which were initiated by azimuthal non-uniformity of sodium flow velocity and/or magnetic field, in the sodium flow.

We calculated the developed pressure of an EM pump by 2D magnetohydrodynamics (MHD) code and designed the EM pump. The code simulated the developed pressure with high accuracy in normal operations. When the flow rate was lower than the one at the top of P-Q curve, the developed pressure’s drop and fluctuation occurred. The fluctuation could disturb the stable operation of the pump. For fear of this phenomenon, the design of EM pump is sometimes too conservative. To simulate this phenomenon and examine its occurring conditions, we have developed a new three-dimensional (3D) MHD code. Clarification of these conditions for the occurrence of this phenomenon will enable to design a new structure or determine its operation conditions reasonably.

This paper describes the simulation results focusing on the occurrence of developed pressure’s drop and fluctuation. We use initial conditions which have azimuthal non-uniformity of sodium velocity to simulate the initiation of vortexes in sodium. The 3D MHD code simulates the developed pressure’s drop and fluctuation by vortexes and provides more clear illustration of their occurrence.
Thermal Performance Monitoring and Optimization in Dukovany NPP
Authors: Pliska Jiri , Machat Zdenek , Veznik Libor , Smisek Jiri
Published: November 2014

Competition in the electricity market forces producers to achieve — in compliance with safety — efficiency of production as high as possible. This efficiency and heat rate is an important indicator of both the condition of the power plant equipment and the quality of power plant operation. To cope with these challenges, powerful methods are process data reconciliation, statistical data processing of large data sets and process simulation. These functions and methods can be used to obtain useful information about process quality and equipment and sensor health.

The paper discusses practical experience from six years of using a thermal performance monitoring and optimization system in the Dukovany nuclear power plant. The system is integrated into the overall nuclear power plant process information system and data warehouse. The system provides information in near real time.

The major benefit of the system lies in a deep view into equipment behaviour and process which ensures timely detection and identification of functionality degradation of process and equipment or sensor faults. The system also helps to find and use margins of equipment operation and the overall thermal cycle.

Selected practical examples are used to demonstrate specific benefits of the system for operation and maintenance of the Dukovany nuclear power plant. There are examples of equipment fault detection and sensor degradation detection. The optimization function is explained with an example of cooling circuit optimization aimed to increase the delivery of electrical power into the grid. A detailed description of behaviour of the main components can be used for their performance evaluation and their repair planning. The benefit of more accurate determination of parameter values is reflected in more accurate determination of reactor thermal output.

The conclusion of the paper provides an overall evaluation of system benefits for operation and maintenance of a nuclear power plant.
Methodology to Evaluate Nuclear Power Plant Vulnerability to Events in the Transmission Grid
Authors: Del Rosso Alberto , Roy Jean-François , Rahn Frank , Capara Alejandro
Published: November 2014

This paper presents a general approach to evaluate the risk of trip or Loss of Off-site Power (LOOP) events in nuclear power plants due to contingencies in the power grid. The proposed methodology is based on the Zone of Vulnerability concept for nuclear plants introduced by EPRI in previous work. The proposed methodology is intended to be part of an integrated probabilistic risk assessment tool that is being developed under ongoing EPRI R&D programs. A detailed analysis of many events occurred in actual nuclear plants has been performed in order to identify, classify and characterize the various vulnerability and type of failures that may affect a nuclear plant. Based the outcome of that analysis, a methodology for evaluating the impact of off-site transmission system events on nuclear plants has been outlined. It includes description of the type of contingencies and conditions that need to be included in the analysis, as well as provisions regarding the simulation tools and models that should be used in each case. The methodology is illustrated in a simplified representation of the Western Electricity Coordinating Council (WECC) system in the U.S.
The Role of CHECWORKS™ in an Effective FAC Program
Authors: Smith David , Horowitz Jeffrey
Published: November 2014

Flow-accelerated corrosion (FAC) is a degradation mechanism that attacks carbon steels under conditions often found in nuclear and fossil power plants. FAC has been responsible for a number of significant accidents in nuclear power plants. The most recent noteworthy accident was at the Mihama Unit 3 (Japan) where the catastrophic failure of a pipe in the condensate system resulted in five fatalities.

FAC can affect virtually all of the carbon steel piping and components in the power cycle of nuclear reactors. The presence of wall thinning caused by FAC is determined through the use of non destructive examination (NDE) techniques. For large-bore piping components the most commonly used approach is ultrasonic technique (UT) using an inspection grid applied to the piping components.

As FAC is a reasonably well-understood degradation mechanism, a number of computer programs have been developed to help utility engineers determine the inspection locations and managing the data. CHECWORKS™ — the most commonly used computer program for this purpose — is the program discussed in this paper.

The use of CHECWORKS™ in utilities’ FAC programs will be described. Particular emphasis will be placed on: inspection planning, handling power uprates and other changes to operating conditions.

Inspection planning is one of the most common uses of the CHECWORKS™ software. As there are typically around 5,000 FAC-susceptible components in a reactor system, utility engineers must select the components with the highest FAC-rates for inspection. CHECWORKS™ uses its predictions combined with plant inspection data to provide a best estimate of FAC rates. From these FAC rate predictions and knowledge of the piping schedule and allowable wall thicknesses, inspection locations are determined.

Both the water chemistry used and the local operating conditions strongly influence the rate of FAC. CHECWORKS™ is used to study potential changes to water chemistry, system operation or power level to determine the impact of such changes on FAC rates and hence inspection locations. Of particular interest is the use of CHECWORKS™ to determine the impacts of power uprates. Because of the complicated parametric behavior of FAC rates, changing the power level will likely increase the FAC rates in some areas of the plant while other areas will likely see a decrease in FAC rates. This fact requires a pre-uprate analysis to determine how an inspection program will need to be modified.

This paper provides a description of how CHECWORKS™ is used in the above applications as well as showing typical examples of its usefulness in these analyses.
The Load Rejection Transient Analysis of Chinshan NPP (BWR/4) Using TRACE/PARCS
Authors: Huang Jhih-Jhong , Chen Hsiung-Chih , Wang Jong-Rong , Liao Lih-Yih , Shih Chunkuan , Lin Hao-Tzu
Published: November 2014

Chinshan Nuclear Power Plant (NPP) is the first boiling water reactor (BWR) NPP in Taiwan. It has two units of BWR/4 reactor made by GE Company and each rated thermal power was 1775 MW without power uprate (now its rated thermal power is 1805 MW after power uprate). This research focuses on the development of the Chinshan NPP TRACE (TRAC/RELAP Advanced Computational Engine)/PARCS (Purdue Advanced Reactor Core Simulator) model. The model is done in two steps: The first step is the development of a TRACES/PARCS model of Chinshan NPP which includes the vessel, fuel assemblies, the main steam lines and important control systems (such as the feedwater control system, recirculation flow control system, etc.). Key parameters (such as power, feedwater flow rate, reactor dome temperature, etc.) were identified to refine the model further in the frame of a steady state analysis. The second step is development of TRACE/PARCS model for the load rejection transient. In order to check the system response of the Chinshan NPP TRACE/PARCS model, this study uses the load rejection transient results of startup tests to benchmark the analysis results of Chinshan NPP TRACE/PARCS model. The trends of TRACE/PARCS analysis results were consistent with the startup test data. It indicates that there is a respectable accuracy in the Chinshan NPP TRACE/PARCS model for the load rejection transient.
Design and Development of Failed Fuel Location Module for Prototype Fast Breeder Reactor
Authors: Raju Vijayashree , Vinayagamoorthy Rajan Babu , Uppala Parthasarathy , Ramachandran Sureshkumar , Sundararajan Raghupathy , Perumal Chellapandi
Published: November 2014

Prototype Fast Breeder Reactor (PFBR), India’s first commercial fast breeder reactor employing fast fission is a challenging project from technological point of view to meet the energy security of the country. PFBR is a sodium cooled fast reactor. There are 198 fuel sub assemblies with mixed oxide fuel in the reactor. The fuel is provided with a leak tight metal cladding for containment of the fission products. There are risks of sodium circuit contamination and the fission products blocking the coolant flow to the fuel sub assemblies in case of clad rupture and release of solid fission products into the coolant. Hence PFBR is equipped with an elaborate failed fuel detection and location system. Failed Fuel Location Module is one of the subsystems used to identify the sub-assembly having fuel pins with clad failure. This paper discusses about the conceptual design, design specifications, detailed design, manufacture, assembly and some of the results of functional testing of failed fuel location module of PFBR.
Application of a CFD-Based Forced Cross Flow Model to a Sub-Channel Analysis Tool
Authors: Zimmermann Markus , Cheng Xu , Otic Ivan , Alleborn Norbert , Sieber Galina
Published: November 2014

A detailed CFD analysis was performed to investigate the effect of mixing vanes on single-phase flow distribution in a 5×5 rod bundle simulating PWR operating conditions. The simulation model was 1 m long and contains one split-type mixing vane grid. For the inlet, a fully developed velocity profile from a previous rod-bundle simulation was used. The simulation case was isothermal to avoid a superposition with wall boiling effects. For the simulation of turbulence effects a non-linear k-ε model was used. This model is able to reproduce the swirling flow in the wake of the grid spacer as well as secondary flow patterns that typically occur in sub-channel flows and needs less computational time compared to Reynolds Stress models. The results show that two large vortices are formed at the mixing vane tips in the center of an interior sub-channel. These vortices are co-rotating within the sub-channel and counter-rotating compared to the adjacent sub-channels. The presence of these vortices influence the radial pressure distribution and the cross-flows significantly. A new model approach was developed, which accounts for the effects of mixing vanes and the vortex structures. The new model requires additional input parameters that can be derived from the CFD results. The proposed model was implemented into the sub-channel analysis tool COBRA-FLX™. A comparison between the old and the new approaches for modeling the forced cross-flow shows a significant improvement of prediction of the mass flux distribution.