bluuebunny/crossref_metadata_2025_split huggingface dataset:
Study on Primary and Secondary Heat-Transport Systems for Sodium-Cooled Fast Reactor
Authors: Dragunov Alexey , Saltanov Eugene , Pioro Igor , Harvel Glenn , Ikeda Brian 
Published: February 2014

One of the current engineering challenges is to design next generation (Generation IV) Nuclear Power Plants (NPPs) with significantly higher thermal efficiencies (43–55%) compared to those of current NPPs to match or at least to be close to the thermal efficiencies reached at fossil-fired power plants (55–62%). The Sodium-cooled Fast Reactor (SFR) is one of the six concepts considered under the Generation IV International Forum (GIF) initiative.
The BN-600 reactor is a sodium-cooled fast-breeder reactor built at the Beloyarsk NPP in Russia. This concept is the only one from the Generation IV nuclear-power reactors, which is actually in operation (since 1980’s). At the secondary side, it uses a subcritical-pressure Rankine-steam cycle with heat regeneration. The reactor generates electrical power in the amount of 600 MWel. The reactor core dimensions are 0.75 m (height) by 2.06 m (diameter). The UO2 fuel enriched to 17–26% is utilized in the core.
There are 2 loops (circuits) for sodium flow. For safety reasons, sodium is used both in the primary and the intermediate circuits. Therefore, a sodium-to-sodium heat exchanger is used to transfer heat from the primary loop to the intermediate one. In this work major parameters of the reactor are listed. The actual scheme of the power-conversion heat-transport system is presented; and the results of the calculation of thermal efficiency of this scheme are analyzed. Details of the heat-transport system, including parameters of the sodium-to-sodium heat exchanger and main coolant pump, are presented.
In this paper two possibilities for the SFR in terms of the power-conversion cycle are investigated: 1. a subcritical-pressure Rankine-steam cycle through a heat exchanger (current approach in Russian and Japanese power reactors); 2. a supercritical-pressure CO2 Brayton gas-turbine cycle through a heat exchanger (US approach).
With the advent of modern super-alloys, the Rankine-steam cycle has progressed into the supercritical region of the coolant and is generating thermal efficiencies into the mid 50% range. Therefore, the thermal efficiency of a supercritical Rankine-steam cycle is also briefly discussed in this paper.
According to GIF, the Brayton gas-turbine cycle is under consideration for future nuclear power reactors. The supercritical-CO2 cycle is a new approach in the Brayton gas-turbine cycle. Therefore, dependence of the thermal efficiency of this SC CO2 cycle on inlet parameters of the gas turbine is also investigated.
Study on Mixing Process of Two Component Gases in a Stable Stratified Fluid Layer: Effectiveness on Natural Convection
Authors: Takeda Tetsuaki , Funatani Shumpei 
Published: February 2014

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)/Ar, helium(He)/N2 and He/Ar 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.
Structure and Erosion Characteristics of Underexpanded Inert-Gas Jets Impinging on a Single Cylinder Submerged in Liquid Sodium
Authors: Kudoh Hideyuki , Sugiyama Ken-ichiro , Narabayashi Tadashi , Ohshima Hiroyuki , Kurihara Akikazu 
Published: February 2014

When a heat transfer tube wall in a steam generator of a sodium-cooled fast reactor fails, high-pressure steam leaks into low-pressure liquid sodium side. Then the chemical reaction between steam and liquid sodium causes the erosive and corrosive reaction jet, and the secondary failure of a neighboring heat transfer tube is expected. The objectives of the present experimental study are to understand the structure of an underexpanded impinging gas jet which is injected into liquid sodium and to grasp a relationship between the gas jet velocity and the erosion characteristics of liquid sodium droplets which are entrained into the gas jet.
In the present study, the structure of an underexpanded inert gas jet impinging on a single cylinder within liquid sodium was clarified using the visualization method, which has been constructed by the authors. Then the erosion tests were carried out at the most erosive geometry, which was confirmed by the visualization experiments.
From the visualization, the underexpanded impinging gas jet in liquid sodium could be classified into four regions: core region, droplet flow region, liquid film flow region, and separation region. The erosion phenomenon on the cylinder surface was observed only when the droplet flow region was impinged. In this region, the liquid droplet impingement type erosion was confirmed. The erosion rate, the ratio of the volume loss of specimens to the duration time, was correlated as the function of the gas jet velocity at the end of the external expansion region.
Preliminary Study on Thermal-Load Fluctuation Test Using HTTR
Authors: Tochio Daisuke , Shimazaki Yosuke , Iigaki Kazuhiko , Nakagawa Shigeaki , Takada Shoji , Sakaba Nariaki , Sawa Kazuhiro 
Published: February 2014

JAEA has designed and developed a commercial very high temperature reactor hydrogen cogeneration system named GTHTR300C. JAEA has investigated that a hydrogen production facility is treated as not a reactor grade facility but a general industrial plant grade facility. The reactor plant system should be designed absorbing the disturbance under AOO and becoming to steady state in the case that thermal-load of the facility is fluctuated or lost. Also, high-accuracy plant dynamics code should be developed to design the GTHTR300C.
In order to demonstrate continuing the reactor operation in if the AOO is occurred in the hydrogen production facility and to obtain the validation data for high-accuracy plant dynamics code, the thermal-load fluctuation tests were planned using the HTTR in JAEA. In this study, it is concluded that the thermal-load fluctuation tests can be carried out without reactor scram and without modification of the HTTR facility and that the test data for validation of the code can be obtained sufficiently.
Study of Pulsating Laminar Flow Field and Resistance Characteristics in Rectangular Channel
Authors: Liu Yusheng , Zhang Chunming , Tan Sichao , Wen Lijing , Ma Shuai 
Published: February 2014

Pulsating laminar flow field and resistance characteristics in rectangular channel are investigated analytically and experimentally. Mathematical model is established. The influences of flow fluctuation period on the flow field in rectangular channel are analyzed. The velocity gradient in the pulsating laminar flow is compared with that in the steady laminar. Analytical results are verified by experiment.
The results show that flow pulsation causes phase difference between the velocity and pressure drop. Compared with steady flow filed, the flow field changes a lot under pulsating condition due to the oscillating velocity component. The velocity gradient in the pulsating laminar flow is changed periodically by the oscillating velocity component, which leads some change in friction coefficient.
3D Calculations of PWR Vessels Neutron Fluence With EFLUVE 3D Code
Authors: Gosmain Cécile-Aline , Rollet Sylvain , Schmitt Damien 
Published: February 2014

In the framework of surveillance program dosimetry, the main parameter in the determination of the fracture toughness and the integrity of the reactor pressure vessel (RPV) is the fast neutron fluence on pressure vessel. Its calculated value is extrapolated using neutron transport codes from measured reaction rate value on dosimeters located on the core barrel.
EDF R&D has developed a new 3D tool called EFLUVE3D based on the adjoint flux theory. This tool is able to reproduce on a given configuration the neutron flux, fast neutron fluence and reaction rate or dpa results of an exact Monte Carlo calculation with nearly the same accuracy. These EFLUVE3D calculations does the Source*Importance product which allows the calculation of the flux, the neutronic fluence (flux over 1MeV integrated on time) received at any point of the interface between the skin and the pressure vessel but also at the capsules of the pressurized water reactor vessels surveillance program and the dpa and reaction rates at different axial positions and different azimuthal positions of the vessel as well as at the surveillance capsules. Moreover, these calculations can be carried out monthly for each of the 58 reactors of the French current fleet in challenging time (less than 10mn for the total fluence and reaction rates calculations considering 14 different neutron sources of a classical power plant unit compared to more than 2 days for a classic Monte Carlo flux calculation at a given neutron source).
The code needs as input:
- for each reaction rate, the geometric importance matrix produced for a 3D pin by pin mesh on the basis of Green’s functions calculated by the Monte Carlo code TRIPOLI;
- the neutron sources calculated on assemblies data (enrichment, position, fission fraction as a function of evolution), pin by pin power and irradiation. These last terms are based on local in-core activities measurements extrapolated to the whole core by use of the EDF core calculation scheme and a pin by pin power reconstruction methodology.
This paper presents the fundamental principles of the code and its validation comparing its results to the direct Monte Carlo TRIPOLI results. Theses comparisons show a discrepancy of less than 0,5% between the two codes equivalent to the order of magnitude of the stochastic convergence of Monte Carlo results.
Evaluation of Damaging Effects on a NPP Integrity Subjected to Beyond Design Earthquake
Authors: Lo Frano Rosa , Forasassi Giuseppe 
Published: February 2014

The aim of this study is to evaluate the influence of the damaging or failure effects induced by seismic loadings on the loading bearing capability of the overall nuclear structures, systems and components (SSCs). In particular it was analyzed the behaviour of the SSCs in terms of their required safety functions and capacity assuming the occurrence of an earthquake having magnitude beyond the design basis value (in agreement with the “stress tests” suggestions, presently foreseen by the European and International Associations).
To the purpose, quite refined FEM models were set up and implemented considering suitable materials behaviour and constitutive laws for the reactor materials, particularly for the concrete material behaviour (which could suffer failure and damage mechanisms during the ground shaking). Several type of damaging mechanics were considered.
The obtained results seemed to confirm the overall containment reliability even if with minor upgrading actions in operating procedures and design characteristics. Moreover they allowed to appropriately check the NPP containment strength reserve.
Experimental Study of Feed Water Level Decreasing Effect on VVER Steam Generator Model Operation in Condensation Mode
Authors: Morozov A. V. , Remizov O. V. , Soshkina A. S. 
Published: February 2014

The essential technological distinctions of the AES-2006 project with VVER-1200 reactor, equipped with passive safety systems, determine some specifics in the character of accidents with the coolant leaks from the reactor primary circuit and the refusal of active part of the emergency core cooling system.
In the case of an accident with depressurization of the reactor primary circuit, the system of passive heat removal (PHRS) ensures the transition of steam generators (SG) into the mode of steam condensation. As a result, a condensate comes to the core, providing its additional cooling. But in the case of beyond design basis accident the rupture of second circuit pipelines or PHRS steam-condensation path is possible. At such type of accidents the feed water level in SG vessel will be decrease, what can results in the deterioration of its condensation power.
For experimental investigation of the condensation mode of VVER steam generator, a large scale HA2M-SG test facility was constructed in IPPE. The test facility incorporates VVER reactor SG model with volumetric-power scale of piping is 1:46, PHRS heat exchanger imitator, cooling by process water and buffer tank, equipped by steam supply system from the IPPE heat power plant. The facility main equipment connected by pipelines and equipped by valves. The elevations of the main equipment correspond to those of reactor project.
Experiments at the HA2M-SG test facility have been performed at the pressure 0.36 MPa, correspond to VVER reactor pressure at the last stage of the beyond design basis accident. On the base of the results of these experiments the correlation of condensation power from feed water level in SG model was obtained. The results of carried out tests make it possible to draw a conclusion about sufficiently large stability of VVER steam generator working in condensation mode to feed water level decreasing.
Scientific Bases of Water Chemistry for Corrosion Control of NPPs by Integration of Radiation- and Electro-Chemistry
Authors: Saji Genn 
Published: February 2014

In this paper, the author continues his investigation on the scientific basis of water chemistry specifications by applying his recent theory, which integrates the elemental radiation- and electro-chemistry reactions in the “Butlar-Volmer equation.” The B-V equation is well established as the basic material balance equation in corrosion science.
The author’s new approach has been compared with the published in-pile test results of the electrochemical potential differences between the in-flux and out-flux regions for both the PWR- and BWR water chemistry environment. Although the theoretical estimation generally reproduced the experimental results, there remains significant deviation from the experimental results at the very low DH region (<10cc-STP/kg-water) in PWRs as well as the low DO region (<10ppb) in BWRs. Although these regions are outside of the water chemistry specifications of general interest, the scientific causes of the deviation must be clarified.
In this paper, the author found that the deviations are due to the dominant radiation-chemical reactions involving hydrogen ions and hydrogen peroxide at the lower ends. Although the radiation- and electrochemical reaction was further exploited with respect to the potential differences induced by the hydrogen peroxide, the effects were disappointingly small, when estimated in terms of a mixed potential of the electrode reactions. This leads the author to suspect that hydrogen-ion-radical reactions should be the main causes. Currently further analyses are in progress.
A Post Accident Safety Analysis Report of the Fukushima Accident — Future Direction of Evacuation: Lessons Learned
Authors: Saji Genn 
Published: February 2014

In nuclear safety, the source term is introduced to provide adequate isolation of the nuclear hazards from the public, by establishing a concept of ‘effective distance.’ This combines a geographical distance to the site boundary and an effective ‘distance’ with the use of engineered safety features (i.e., a containment system and its cooling system), combined with evacuation procedures to prevent radiation injury. Severe accidents occur when these safety systems failed to function. This basic safety approach was once again jeopardized by the Fukushima accident, which followed the Chernobyl accident. The factors that mitigated the effluent releases however depend greatly on the intrinsic safety features combined with the accident management. The multi-layered retention/decontamination factors that a nuclear power plant possesses should be incorporated in specifying the environmental source term. The Fukushima accident provides a reasonable upper bound with respect to environmental releases due to a LUHS (loss of ultimate heat sink), which triggered a prolonged SBO (station blackout).
Due to the anticipated radiological consequences, the Japanese Government issued a series of evacuation orders, resulting in the evacuation of approximately 160,000 people from the Fukushima area. The prolonged evacuation is believed to be the cause of over one thousand “disaster-related (pre-mature) deaths (DRDs)” which have been reported among the evacuees due to psychosomatic effects (48%) and the disruption of medical and social welfare facilities (18%). In the future these types of deaths should be avoided.
A Numerical Simulation on Effects of Distance Between Bubble and the Wall on Behavior of Rising Bubble
Authors: Huang Ying , Gao Puzhen 
Published: February 2014

A numerical investigation of two-dimensional air bubble behaviors under the effect of gravity in still water based on the VOF (Volume-Of-Fluid) method is carried out.
Initially, the surface tension effects on the behavior of the bubble is analyzed, which contains the simulation of the ascending motion of a single air bubble in liquid and the study of the interaction between bubbles in terms of coalescence.
Additionally, the differences of single bubble’s rising motion in an infinite surroundings and in a vertical narrow channel are analyzed. The coalescence of bubbles is also studied. The motion of bubbles with different diameters in a vertical channel is simulated. It is found that the bubbles’ behavior depends on the distance between the bubble and the wall.
Finally, numerical simulation of the motion of several bubbles of the same size, at the same initial horizontal position and with uniform distribution is carried out. The result reveals that the bubbles at different distances from the wall have different velocities, after a while, the bubbles distribution presents as “U”.
Model Passivated Carbon Electrodes for Fluorine Generation in MSRs and the Nuclear Fuel Cycle
Authors: Seto Kelvin S.-H. , Ikeda Brian M. 
Published: February 2014

Elemental fluorine, F2, is used in the nuclear fuel cycle for the isotopic separation of uranium-235 and 238, as well as for the purification of LiF-BeF2 in molten salt reactors. F2 is generated on an industrial scale by an electrochemical process using carbon electrodes in a KF-2HF molten salt. Carbon electrodes are used for industrial F2 generation due to its chemical stability, high conductivity, and relatively low cost. One of the main issues faced when using carbon electrodes in this chemical system is passivation through the formation of C-F compounds on the surface of the electrode. This results in a loss of anode wettability to the electrolyte and diminished charge transfer rate. The voltage needed for the fluorine evolution reaction increases which negatively impacts the safety of the system, increases the operating costs, and leads to faster degradation of the electrode.
The degradation of electrical properties during passivation is progressive, eventually leading to electrode deactivation. The process of deactivation begins with a passivating C-F layer at potentials above the equilibrium potential (2.92 V). The layer is both non-wetting to the KF-2HF media and insulating. Deactivation begins with inhibited F2 bubble detachment, formation of a persistent gas layer, and finally deactivation as the electrode surface is completely covered by a thick, insulating C-F layer causing charge transfer to cease. Only a small current is able to flow, even at high potentials (up to 9 V), indicating F2 generation is completely inhibited. The purpose of this study is to manufacture and test model carbon electrodes and, to examine the non-wetting properties of a partially fluorinated surface. The electrodes will be prepared by mixing PTFE-particles with Vulcan carbon powder and then pressing to form pellets. These electrodes should have a reproducible surface for electrochemical performance studies that will lead to a better understanding of the surface chemistry. The research will develop novel electrodes with a goal to minimize the voltage required for F2 production. This will enhance the efficiency in the overall process and lower the manufacturing costs for F2.
Carbon electrodes with different PTFE-content (20 w.% and 35 w.%) were synthesized. Electrochemical fluorination was then carried out at different potentials in the F2 generation region (4 to 8 V) in molten KF·2HF electrolyte at ∼90 °C. The electrochemical behaviour of the carbon-PTFE electrodes was examined and compared for both fluorine passivated and non-passivated graphite, amorphous carbon, and vitreous carbon electrodes. The growth of the electrical double-layer capacitance between the carbon electrodes and the KF·2HF molten salt was studied. The effects of composition of fluorinated and non-fluorinated carbon on electrode performance are presented.
Validation of the System Code APROS for Fast Transients
Authors: Raqué Manuel , Herbell Heiko , Schulenberg Thomas 
Published: February 2014

The thermal-hydraulic system code APROS Version 5.09 [1] is being applied in the European project SCWR-FQT to evaluate the performance of the safety systems for a nuclear test facility operated with supercritical water. In order to validate the commercial code for predictions of transient phenomena, two adequate hydraulic experiments from literature have been simulated.
The experiment of Fujii and Akagawa [2] investigated hydraulic shocks as they will occur in case a pipe, which is stationary passed through by water, is abruptly closed. A simple numerical model with adapted time and space nodalization was able to reproduce the observed physical phenomena, such as the magnitude of the initial pressure wave and reflection time, in detail.
In a second experiment, a test series was performed by Becker et al. [3] and Mathisen [4] in order to examine the natural circulation in a closed loop for different heating rates and system pressures. For a stepwise power increase, the typical mass flow characteristic for boiling channels was recorded until the onset of flow oscillations. In further runs, the effect of different initial conditions on the flow stability was analyzed. This paper compares the numerical predictions with both experimental results. The numerical models could describe the physical phenomena with appropriate accuracy.