bluuebunny/crossref_metadata_2025 huggingface dataset:
Numerical Analysis for a Molten Salt Reactor in the Presence of Fissile Lump
Authors: Xiao Yao , Zhang Dalin , Guo Zhangpeng , Qiu Suizheng
Published: October 2013

Molten salt reactors (MSRs) have seen a marked resurgence of interest over the past few decades, highlighted by their inclusion as one of the six Generation IV reactor types. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this paper, the attention is focused on the behaviors of a MSR in the presence of localized perturbations caused by fissile precipitates. A neutron kinetic model considering the fuel salt flow is established based on the neutron diffusion theory, which consists of two-group neutron diffusion equations for the fast and thermal neutron fluxes and six-group balance equations for delayed neutron precursors, and the group constants dependent on the temperature are calculated by the code DRAGON. In addition, the k-epsilon turbulent model is adopted to establish the flow and heat transfer. The thermo-hydraulic and neutronic models which are coupled through the temperature, heat source and velocity are coded in a program. The effects of the localized perturbation on the distributions of power, temperature, neutron fluxes and delayed neutron precursors are obtained and discussed in detail. The results provide some valuable information for the research and design of this new generation reactor.
Dynamic Behavior Analysis of a Supercritical Water Cooled System Using Modified ATHLET
Authors: Zhou Chong , Cheng Xu , Yang Yanhua
Published: October 2013

This paper presents the application of the modified ATHLET code to the dynamic behavior analysis of a supercritical water cooled system. A small scale fuel assembly under the “evaporator” condition of the HPLWR core design concept, with a supercritical water cooled experiment loop, is about to be tested in the LVR-15 research reactor of the Nuclear Research Institute in Rez in Czech Republic. With the purpose to analyze the safety characteristic and prepare for the licensing of the test loop, some design based accidents like loss of coolant due to break of coolant supply lines and trip of the primary pump are simulated in this paper. All the results achieved indicate that the modified ATHLET code has good feasibility in simulation of supercritical water cooled systems.
Two-Phase Flow in an Annular Channel With an Obstacle
Authors: Kashinsky O. N. , Lobanov P. D. , Kurdyumov A. S. , Pribaturin N. A.
Published: October 2013

Experimental study of gas-liquid two-phase flow in an annular channel is performed. The channel consisted of two coaxial tubes with the diameters of 42 and 20 mm. An obstacle covering a quarter of the channel section was placed in the channel to produce a strong three-dimensional disturbance of the flow. Gas-liquid flow was produced by injecting air bubbles at the channel entrance through a special mixer. Measurements of local wall shear stress are performed using an electrochemical technique. Measurements of time-averaged and fluctuational wall shear stress are performed at various points relative to the obstacle, this allowed to study the field of the hydrodynamic parameters of the flow. Local void fraction is measured using a conductivity probe which traversed across the channel. The distribution of local void fraction in the region downstream the obstacle is obtained. Increased values of local void fraction in the region close to the obstacle are detected.

The experimental data obtained can be used for validation of existing and developing computer codes accounting for a 3-D structure of two-phase flows.
Investigation of a Heavy Liquid Metal Free Surface Target Experiment
Authors: Litfin Karsten , Fellmoser Frank , Batta Abdalla , Class Andreas G. , Wetzel Thomas
Published: October 2013

Several distinct reactor strategies are proposed within the context of the IP Eurotrans framework programme for the transmutation of nuclear waste. A pool type reactor filled with liquid heavy metal and containing a subcritical core is one of the promising designs. Additional neutrons required for the nuclear reaction are generated by a spallation reaction inside the core. A high power proton beam is guided through a vacuum tube from an accelerator into the liquid heavy metal pool i.e. into the reactor core. At the point where the beam hits the metal surface special construction effort is indispensable to handle the high heat production.

A specific target design is used to ensure a high fluid velocity and a stable surface at the beam entry. This design employs a concentric vertical feeder establishing a free conical surface with velocity up to 2.5 m/s ensuring stable surface flow and appropriate heat removal. The proposed target geometry has been constructed using underlying rules developed by the MYRRHA design group for the free surface target in the MYRRHA research reactor.

A full scale model of this design using lead bismuth eutectic (LBE) has been set up and experimentally investigated at the of the Karlsruhe Institute of Technology (KIT). Measurements taken by high speed digital imaging visualize both conical inner and outer jet free surface. They show a stable surface in a wide range of operating conditions starting from 35% of the nominal flow rate and agree well with numerical investigations using commercial CFD code Star-CD and Star-CCM+. Previous concerns related to splashing or cavitation during the start-up or shut-down procedure proofed unjustified.
Assessing Sensitivity of Observations in Source Term Estimation for Nuclear Accidents
Authors: Ma Yuanwei , Wang Dezhong , Tan Wenji , Ji Zhilong , Zhang Kuo
Published: October 2013

In the Fukushima nuclear accident, due to the lack of field observations and the complexity of source terms, researchers failed to estimate the source term accurately immediately. Data assimilation methods to estimate source terms have many good features: they works well with highly nonlinear dynamic models, no linearization in the evolution of error statistics, etc. This study built a data assimilation system using the ensemble Kalman Filter for real-time estimates of source parameters. The assimilation system uses a Gaussian puff model as the atmospheric dispersion model, assimilating forward with the observation data.

Considering measurement error, numerical experiments were carried on to verify the stability and accuracy of the scheme. Then the sensitivity of observation configration is tested by the twin experiments.

First, the single parameter release rate of the source term is estimated by different sensor grid configurations. In a sparse sensors grid, the error of estimation is about 10%, and in a 11*11 grid configuration, the error is less than 1%.

Under the analysis of the Fukushima nuclear accident, ahead for the actual situation, four parameters are estimated at the same time, by 2*2 to 11*11 grid configurations. The studies showed that the radionuclides plume should cover as many sensors as possible, which will lead a to successful estimation.
A Study of Air-Coupled Ultrasonic Flowmeter
Authors: Tsukada Keisuke , Ihara Tomonori , Kikura Hiroshige
Published: October 2013

A new non-contact flow metering system is developed using air-coupled ultrasonic technique. The principle of the present system is based on the time-of-flight (TOF) ultrasonic flowmeter. The TOF ultrasonic flowmeter has two transducers, a transmitter and a receiver, which are strapped on the outside wall of a pipe. However, it is difficult to set the transducers on very high temperature pipe walls, since the transducers loose piezoelectricity over Curie point. Therefore, we focused on air-coupled ultrasonic waves for flow field measurement, and developed an air-coupled ultrasonic transducers and a signal processing system.
Sensitivity Analysis of Fuel Centerline Temperature in SCWRs
Authors: Abdalla Ayman , Peiman Wargha , Pioro Igor , Gabriel Kamiel
Published: October 2013

The Generation IV International Forum (GIF) is intended to encourage the world’s leading nuclear countries to develop nuclear energy systems that can supply future energy demands. There are six nuclear reactor concepts under research and development as part of the GIF. The SuperCritical Water-cooled Reactor (SCWR) is one of these six nuclear-reactor concepts. The proposed SCWRs operate at high temperatures and pressures at around 625°C and 25 MPa, respectively. These high operating parameters are essential in order to achieve a thermal efficiency of around 45–50%, which is significantly higher than those of the current conventional nuclear power plant (NPPs) which operate at a thermal efficiency in the range of 30–35%.

The SCWRs high operating temperatures and pressures impose many challenges. One of these challenges is the heating of the fuel to temperatures that can cause fuel melting. The main objective of this paper is to conduct a sensitivity analysis in order to determine the factors mostly affecting the fuel centerline temperature. In this process, different thermal conductivity fuels such as Mixed Oxide Fuel (MOX), Uranium Oxide + Beryllium Oxide (UO2+BeO), and Uranium Carbide (UC) will be examined enclosed in a 54-element fuel bundle. Other factors such as the sheath material and the Heat Transfer Coefficient (HTC) might also affect the fuel centerline temperature. The HTC will be increased by a multiple of two and the fuel centerline temperature will be calculated. Therefore, in this paper the HTC, bulk-fluid, sheath and fuel centerline temperature will be calculated along the heated length of a generic SCWR fuel channel at an average channel thermal power of 8.5 MWth.
Experiment of Adiabatic Two-Phase Flow in an Annulus Under Low-Frequency Vibration
Authors: Chen Shao-Wen , Brooks Caleb S. , Macke Chris , Hibiki Takashi , Ishii Mamoru , Mori Michitsugu
Published: October 2013

In order to investigate the possible effect of seismic vibration on two-phase flow dynamics and thermal-hydraulics of a nuclear reactor, experimental tests of adiabatic air-water two-phase flow under low-frequency vibration were carried out in this study. An eccentric cam vibration module operated at low motor speed (up to 390rpm) was attached to an annulus test section which was scaled down from a prototypic BWR fuel assembly sub-channel. The inner and outer diameters of the annulus are 19.1mm and 38.1mm, respectively. The two-phase flow operating conditions cover the ranges of 0.03≤ ≤1.46m/s and 0.25≤≤1.00m/s and the vibration displacement ranges from ±0.8mm to ±22.2mm. Steady-state area-averaged instantaneous and time-averaged void fraction was recorded and analyzed in stationary and vibration experiments. A neural network flow regime identification technique and fast Fourier transformation (FFT) analysis were introduced to analyze the flow regimes and void signals under stationary and vibration conditions. Experimental results reveal possible changes in flow regimes under specific flow and vibration conditions. In addition, the instantaneous void fraction signals were affected and shown by FFT analysis. Possible reasons for the changes include the applied high acceleration and/or induced resonance at certain ports under the specific flow and vibration conditions.
Comparison of Existing Supercritical Carbon Dioxide Heat Transfer Correlations for Horizontal and Vertical Bare Tubes
Authors: Surendran Prabu , Gupta Sahil , Preda Tiberiu , Pioro Igor
Published: October 2013

This paper presents a thorough analysis of ability of various heat transfer correlations to predict wall temperatures and Heat Transfer Coefficients (HTCs) against experiments on internal forced-convective heat transfer to supercritical carbon dioxide conducted by Koppel [1], He [2], Kim [3] and Bae [4]. It should be noted the Koppel dataset was taken from a paper which used the Koppel data but was not written by Koppel. All experiments were completed in bare tubes with diameters from 0.948 mm to 9 mm for horizontal and vertical configurations. The datasets contain a total of 1573 wall temperature points with pressures ranging from 7.58 to 9.59 MPa, mass fluxes of 400 to 1641 kg/m2s and heat fluxes from 20 to 225 kW/m2.

The main objective of the study was to compare several correlations and select the best of them in predicting HTC and wall temperature values for supercritical carbon dioxide. This study will be beneficial for analyzing heat exchangers involving supercritical carbon dioxide, and for verifying scaling parameters between CO2 and other fluids. In addition, supercritical carbon dioxide’s use as a modeling fluid is necessary as the costs of experiments are lower than supercritical water.

The datasets were compiled and calculations were performed to find HTCs and wall and bulk-fluid temperatures using existing correlations. Calculated results were compared with the experimental ones. The correlations used were Mokry et al. [5], Swenson et al. [6] and a set of new correlations presented in Gutpa et al. [7]. Statistical error calculations were performed are presented in the paper.
Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration (6) Numerical Simulation of Bubble Deformation Near Wall Under Accelerating Condition
Authors: Yoshida Hiroyuki , Nagatake Taku , Takase Kazuyuki , Kaneko Akiko , Monji Hideaki , Abe Yutaka
Published: October 2013

Earthquake is one of the most serious phenomena for safety of a nuclear reactor in Japan. Therefore, structural safety of nuclear reactors has been studied and nuclear reactors were contracted with structural safety for a big earthquake. However, it is not enough for safety operation of nuclear reactors because thermal-fluid safety is not confirmed under the earthquake. For instance, behavior of gas-liquid two-phase flow is unknown under the earthquake conditions. Especially, fluctuation of void faction is an important factor for the safety operation of the nuclear reactor. In the previous work, fluctuation of void faction in bubbly flow was studied experimentally and theoretically to investigate the stability of the bubbly flow. In such studies, flow rate or void fraction fluctuations were given to the steady bubbly flow. In case of the earthquake, the fluctuation is not only the flow rate, but also body force on the two-phase flow and shear force through a pipe wall. Interactions of gas and liquid through their interface also act on the behavior of the two-phase flow. The fluctuation of the void fraction is not clear for such complicated situation under the earthquake. Therefore, the behavior of gas-liquid two-phase flow is investigated experimentally and numerically in a series of study. In this study, to develop the prediction technology of two-phase flow dynamics under earthquake acceleration, a detailed two-phase flow simulation code with an advanced interface tracking method TPFIT was expanded to two-phase flow simulation under earthquake conditions. In this paper, outline of expansion of the TPFIT to simulate detailed two-phase flow behavior under the earthquake condition was shown. And the bubbly flow in a horizontal pipe excited by oscillation acceleration and under the fluctuation of the liquid flow was simulated by using expanded TPFIT. Predicted deformation of bubbles near wall was compared with measured results under flow rate fluctuation and structural vibration.
Numerical Simulation of Graphite Oxidation in HTGR
Authors: Wang Yao , Wang Peng , Yu Suyuan
Published: October 2013

The oxidation of the bottom reflector graphite in HTR-10 of Tsinghua University during air ingress accident were studied numerically based on a theoretical calculation model, which considered the changes of chemical reaction rate and mass transfer speed with weight loss in oxidation process. The results showed that oxidation degree of the bottom reflector increased nearly exponentially with temperature; the oxidation corrosion quantity increased significantly over time accumulation; weight loss calculated by this model is larger than previous models, so reactor graphite structure design should be paid more attention.
Neutronics Performance of a Fusion-Fission Hybrid Reactor With Pressure Tube Assemblies
Authors: Zu Tiejun , Wu Hongchun , Zheng Youqi , Cao Liangzhi
Published: October 2013

A fusion-fission hybrid reactor with pressure tube assemblies is proposed for energy production in this paper. In the light water cooled blanket, spent nuclear fuel from 33 GWD/t LWR and natural uranium oxide are taken as driver fuel for neutron/energy multiplication aiming at increasing uranium utilization efficiency. Tritium self-sufficiency is obtained by Li2O. The home developed NAHR code is employed to simulate the neutronics behavior in the blanket. The plasma conditions and configuration of ITER are used to the neutronics calculations. The results show that the blanket can be operated for at least 5 years with good performance of acceptable energy multiplication and tritium breeding. The first neutron wall loading is lower than 0.57 MW/m2.
Best Estimate Analysis of PHEBUS FPT1 Test Using RELAP/SCDAPSIM Code
Authors: Kaliatka Tadas , Ušpuras Eugenijus , Vileiniškis Virginijus
Published: October 2013

The PHEBUS-FP program is an outstanding example of an international cooperative research program that is yielding valuable data for validating severe accident analysis computer codes. The main objective of the PHEBUS FPT1 experiment was to study the processes in the overheated reactor core, release of fission products and their subsequent transport and deposition under conditions representative of a severe accident of a Pressurised Water Reactor. The FPT1 test could be divided in the bundle degradation, aerosol, washing and chemistry phases. The objective of this article is the best estimate analysis of the bundle degradation phase. GRS (Germany) best estimate method with the statistic tool SUSA used for uncertainty and sensitivity analysis of calculation results and RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during severe accident conditions, was used for the simulation of this test. The RELAP/SCDAPSIM calculation results were compared with the experimental measurements and calculations results, received by employing ICARE module of ASTEC V2 code. The performed analysis demonstrated, that the best estimate method, employing RELAP/SCDAPSIM and SUSA codes, is capable to model main severe accidents phenomena in the fuel bundle during the overheating and melting of reactor core.
Methodology for the Analysis of Fuel Behavior During LOCA and RIA for Licensing Purposes
Authors: Adorni Martina , Del Nevo Alessandro , D’Auria Francesco
Published: October 2013

Licensing requirements vary by country in terms of their scope, range of applicability and numerical values and may imply the use of system thermal hydraulic computer codes. Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes, as for burst temperature, burst strain and flow blockage calculations. This may imply the use of a dedicated fuel rod thermo-mechanical computer code, which can be coupled with thermal-hydraulic system and neutron kinetic codes to be used for the safety analysis.

This paper describes the development and the application of a methodology for the analysis of the Large Break Loss of Coolant Accident (LB-LOCA) scenario in Atucha-2 Nuclear Power Plant (NPP), focusing on the procedure adopted for the use of the fuel rod thermo-mechanical code and its application for the safety analysis (Chapter 15 Final Safety Analysis Report, FSAR). The methodology implies the application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient.

A strong effort has been performed in order to enhance the fuel behaviour code capabilities and to improve the reliability of the code results.
LIFUS5/Mod2: The Experimental Facility for HLM/Water Interaction Investigation
Authors: Del Nevo Alessandro , Ciampichetti Andrea , Forgione Nicola , Mannori Simone
Published: October 2013

The interaction between heavy liquid metal (HLM) and water is a safety concern for the preliminary designs of lead fast reactor and of subcritical transmutation system prototypes. Current pool-type configurations have the steam generators inside the reactor vessel. This implies that the primary to secondary leak (e.g. steam generator tube rupture) shall be considered as a safety issue in the design, but also in the preliminary safety analysis, of this reactor types. This requires availability of qualified experiments carried out with initial and boundary conditions representative of the reactor prototype. The objective is to support the development and to demonstrate the reliability of computer codes in simulating the phenomena of interest. LIFUS5/Mod2 is a separate effect test facility, designed for investigating the interaction between heavy liquid metal (HLM) and water, installed at ENEA CR Brasimone. LIFUS5 test facility is refurbished based on the operating experience acquired during the previous experimental campaigns and the support of numerical tools (i.e. SIMMER-III, -IV and, with some extent, RELAP5). The new design involves a new facility configuration, test section, instrumentation and control systems. This paper presents the new LIFUS5/Mod2 facility, the planned experimental campaign and code calculations devoted to support the design modifications, the instrumentation and the final test matrix. The activities are performed in the framework of the EC funded THINS Project and thanks the support of the Italian Ministry for the Economic Development.
The Analysis of Instability in a Two-Phase Natural Circulation Loop Under Rolling Motion
Authors: Zhang Wenchao , Tan Sichao , Gao Puzhen
Published: October 2013

An analytic model was developed to simulate the flow oscillation in a natural circulation system under rolling motion. This model is a nodal formulation that uses one-dimensional homogeneous equilibrium assumptions for diabatic two-phase flow. This model indicates that there is single-phase flow fluctuation when heating power I s small. However, if the heating power increases sufficiently, vapor bubbles generate immediately and the period of vapor bubble generation is the same as the rolling period. The calculated results are in good agreement with the experimental data.
Measurement of Gas Entrainment Rate From Free Surface by Vortex
Authors: Koizumi Yasuo , Ohte Naosuke , Hideki Kamide , Ohno Shuji , Ito Kei
Published: October 2013

A sodium-cooled fast breeder reactor is now at the developing stage in Japan. One concern for safety is cover gas entrainment into the sodium coolant. The gas entrainment rate into liquid by the vortex formed on the free surface was examined experimentally. Liquid flowed into a cylindrical vessel from a wall tangentially. Swirl flow was formed in the vessel, and then liquid drained from the bottom outlet of the vessel. A hollow vortex was formed on the free surface in the test vessel. Air was entrained under the free surface of the vortex and carried away from the bottom of the vessel. The flow state of the gas entrainment was visually observed by using a high speed video camera. The gas entrainment rate into liquid was measured. In the present experiments, test fluid was changed from water in the previous experiments to 20 cSt silicone oil. The liquid level in the test vessel was 25 mm in the present experiments. Only the vortex-type gas-entrainment was observed as in the previous experiments since the liquid level was low. The flow state observed at the flow visualization section of the outlet pipe was only a semi-annular flow. The initiation of the gas entrainment was delayed in the case of silicone oil compared with the case of water. The increasing rate of the gas entrainment to the liquid velocity is milder in the case of silicone oil than in the case of water.
Effect of Pressure on Underdeveloped Nucleate Boiling on the Single Horizontal Tube Heated by Condensing Steam
Authors: Morozov A. V. , Remizov O. V. , Kalyakin D. S.
Published: October 2013

In new Russian NPP with VVER reactor in the event of LOCA, provision is made for the use of passive heat removal system for necessary core cooling. In the case of leakage in the primary circuit this system assures the transition of steam generators to operation in the mode of condensation of the primary circuit steam coming to steam generator piping from the reactor. As a result, the condensate from the steam generators arrives to the core providing its additional cooling.

The steam generator off-design condensation mode has the following features: undeveloped nucleate boiling on the horizontal tubes heated by condensing steam; natural circulation processes in the both steam generator circuits; low heat fluxes and temperature differences. The experimental study of undeveloped nucleate boiling on the single horizontal tube heated by condensing steam has been carried out in the Institute for Physics and Power Engineering.

The experiments have been carried out on the GROT test facility. The heart of the test facility is the single VVER steam generator tube (length l = 10.2 m, outer diameter D = 16 mm, wall thickness d = 1.5 mm). The tube is fabricated of the original stainless steel 08Cr18Ni10Ti. The length and geometry of the test tube corresponded to that of real steam generator. The test facility was equipped with thermocouples enabling the temperatures of primary and secondary facility circuits to be controlled.

The experiments were carried out at three heating steam pressures Ps1: 0.21, 0.35, 0.55 MPa. The main task of the research was to study the pressure effect on the process of undeveloped nucleate boiling on the single horizontal tube.

On the base of the results of these experiments the empirical correlations for prediction of heat transfer coefficient and heat flux were obtained. The generalizing empirical correlations obtained can be used for the substantiation of work heat-exchanging equipment of NPP with VVER reactor in the condensation mode, and also can be applied to the verification of computer codes.
Uncertainty Evaluation of the Rod Withdrawal at Power Accident Analysis Including 3D Neutron Kinetics
Authors: Bajs Tomislav , Petruzzi Alessandro , Grgić Davor
Published: October 2013

A continuous uncontrolled Rod Cluster Control Assembly (RCCA) bank withdrawal at power belongs to group of Reactivity Initiated Accidents (RIA). It will cause an increase in core heat flux and a reactor coolant temperature rise. Unless terminated by manual or automatic action, the power mismatch and resultant coolant temperature rise could eventually result in departure from nucleate boiling (DNB) and/or fuel centreline melt. The accident can be DNBR or overpower limiting accident depending on initial power level and rate and amount of reactivity addition.

The Rod Withdrawal At Power (RWAP) accident was analyzed for NPP Krško to evaluate possible Resistance Temperature Detectors (RTD) bypass removal and introduction of thermowell for the average temperature measurement. The influence of different coolant temperature measurement delays to related protection system response and limiting system variables was studied first using point kinetics model as implemented in RELAP5 code. The selected scenario (maximum insertion rate with rods in manual mode) has been re-calculated using RELAP5/PARCS coupled code. Core wide departure from nucleate boiling ratio (DNBR) calculation has been performed at the end of the coupled code calculation using COBRA based model to determine minimum DNBR for hot channel. In order to assess available safety margins following such accident CIAU methodology has been applied to evaluate the uncertainty of RELAP5 analysis and modified CIAU/TN methodology to evaluate uncertainty of the three-dimensional neutronics/thermal-hydraulics calculations. Differences between system and coupled code results and uncertainties is discussed.
The Safety Analysis of the 600 MWe Sodium-Cooled Fast Reactor With MARS-LMR
Authors: Lee Kwi Lim , Ha Kwi Seok , Jeong Hae Yong , Chang Won Pyo
Published: October 2013

Korea Atomic Energy Research Institute (KAERI) has been developing a conceptual design of the demonstration fast reactor (DFR), which is the pool type sodium cooled fast reactor with the thermal power of 1548.2 MW and the core loaded with metal fuel.

The DFR is composed of a Primary Heat Transport System (PHTS), an Intermediate Heat Transport System (IHTS), a Steam Generating System (SGs) and a decay heat removal system (DHRS). The DHRS is composed of 2 units of Passive Decay-heat Removal Circuits (PDRC) and 2 units of Active Decay-heat Removal Circuits (ADRC). The PDRC consists of two independent loops with sodium-to-sodium Decay Heat eXchanger (DHX) and natural-draft sodium-to-Air Heat eXchanger (AHX). The ADRC consists of two independent loops with sodium-to-sodium DHX and Forced-Draft sodium-to-air Heat eXchanger (FDHX) located in the upper region of the reactor building. The PDRC is very different from that of KALIMER-600 on the points of the submerged location and the heat transfer mechanism.

For the identification of safety characteristics, 5 DBE’s (Design Bases Events) are analyzed using the MARS-LMR code. The representative DBE’s are TOP (Transient of Over Power), LOF (Loss Of Flow), LOHS (Loss Of Heat Sink), Reactor Vessel Leak and Pipe Break. As a result, it is identified that the DFR were appropriately performed as designed and the temperatures of the fuel and the structure were evaluated to satisfy the criteria.
Development of New Simulation Software System to Evaluate Radiation Dose Rates in a Wide Radioactively Contaminated Area
Authors: Hashimoto Tomoharu , Kondo Masahiro , Tayama Ryuichi , Gamo Hideho
Published: October 2013

The Japanese government plans to conduct decontamination tasks in radioactively contaminated areas. For such a situation, we developed a system that evaluates radiation dose rates in a wide radioactively contaminated area by utilizing our radiation dose evaluation technology. This system can not only generate present maps of radiation dose rate in the air based on the dose rate measured at the surface of the contaminated areas, but can also quickly calculate the reduction effect of dose rate due to decontamination tasks by entering decontamination factors. The system can then formulate decontamination plans and make it possible to plan measures to reduce radiation exposure for workers and local residents.

Radioactive nuclides that contribute to gamma-ray dose rate are mainly Cs-134 and Cs-137 in soil, on trees, buildings, and elsewhere. Shapes of such radiation sources are assumed to be 10m square or 100m square. If it is unsuitable that the radiation sources assume to squares, the radiation sources can assume to point. The relation between distance from the surface or point source and the radiation dose rate is calculated using MCNP5 code (A General Monte Carlo N-Particle Transport Code - Version 5), and approximated using four-parameter empirical formula proposed by Harima et al. In addition, the system can consider shielding such as soil, concrete, and iron. When setting such shielding, the skyshine dose rate is taken into account in dose rate calculation.
Stability Analysis of a Supercritical Water Natural Circulation Loop
Authors: Li ZhongChun , Yu JiYang , Song XiaoMing
Published: October 2013

As a part of “supercritical water reactor basic research”, the stability of the natural circulation research plays an important role on the feasibility of supercritical water reactor and experiment research. In this paper, the stability of a supercritical water natural circulation loop built by Department of Engineering Physics, Tsinghua University was studied by numerical method. It was confirmed that the static or Ledinegg instability doesn’t occur in HACA system, and there are no instabilities existing when the inlet enthalpy is larger than critical enthalpy. Instability was observed by numerical way, which is similar to DWOs and PDOs in two phase natural circulation loop. The system parameters’ influence on the instability of supercritical natural circulation loop was studied.
Smart Lube Systems for Gas Turbine Engines
Authors: Kenney Thomas B.
Published: October 2013

Presently, the typical gas turbine based power generator relies on a large, expensive skid system to provide lubrication to the bearings. These skids consist of a large number of expensive components, many of which require maintenance that drives operating cost and creates environmental hazards. A smart lube system based on recently developed technology enables dramatic simplification and cost reduction of the skid system and significantly reduces oil consumption along with the frequency and cost of maintenance. Such a system would take advantage of the availability of advanced pump system controls and fluid nozzle technology to drastically reduce the quantity of oil required, and the resultant heat rejection. As a result, the sizes of reservoir, cooler, filter and piping are greatly reduced. The cost of components, support equipment and maintenance activities will also be reduced to a fraction of their present values.
Condenser Refurbishment (Retube / Rebundle): Lost Opportunity
Authors: Tagliamonte Mark
Published: October 2013

Condenser Refurbishment Projects (Re-tube / Re-bundle) focus on a tube for tube replacement with or without tube sheet replacements or a modular tube bundle replacement. These projects are large capital investments for the owner and most of the time the justification and planning efforts are focused on extending the operating life of the power plant with the goal of a 3 to 5 year return on investment. The project planning focus is typically on addressing the immediate needs of rapid return of the unit to service, copper removal, condenser performance improvement (new material selection), and or adding additional steam condensing capacity to support a plant uprate.

This paper addresses issues that are typically overlooked in the initial planning and budgeting for condenser refurbishment projects that can impact condenser performance, reliability, component maintainability, and schedule.

As with all aging equipment a detailed assessment of a condenser’s condition should be performed to evaluate not only the tubes and tube sheets, but also the general conditions of support plates, shell, penetrations, spargers (Figure12), dump valves, high energy drains, thermal sleeves (Figure 1 and Figure 2), extraction steam bellows, impingement shields, and the non condensable off gas system to name a few. Also discussed are examples of problems encountered during some refurbishment projects and the implemented solutions.
A Design of Simulation Experiment System for the Helium Turbine
Authors: Zhu Liang , Shi Zhengang
Published: October 2013

In this paper a new design of simulation experiment system for the helium turbine is described. This experiment system is for to study the rotor dynamic performance of the helium turbine in a 10 MW high temperature reactor-based helium turbine (HTR-10GT) power generation project. The system is a vertical and biaxial arrangement. The motor rotor is on one axis and the turbine-compressor rotor is another axis. The two axes are connected by a gear box having a gear ratio of 1:5 and two couplers. The motor has a rate speed of 3,000rpm, and the turbine-compressor rotor 15,000 rpm. The turbine-compressor rotor, which is simulated, has a mass of 646 kg, a length of 3,568mm, and is supported by two radial magnetic bearings along with an axial magnetic bearing. The design load carrying capacity of the axial magnetic bearing is 9.8 kN with a clearance of 0.7mm and the radial magnetic bearing is 1.96 kN with a clearance of 0.6mm. For simulating the aerodynamic force and the unbalanced force on the turbine-compressor rotor when it was turning, we designed a magnetic load simulator which can exert axial and radial load to the rotor directly. The design axial load of the magnetic load simulator is 10kN and the radial load is 715N both with a clearance of 1mm. The whole simulation experiment system includes the ventilation and cooling system and monitoring and control system. It is the first time that the simulation test has been run for the vertical helium turbine with magnetic bearings, running at full speed. The turbine-compressor rotor will pass through the critical speeds for the first and second order of bending, before reaching the rated speed. The magnetic load simulator can exert load quickly and directly, and the load form can be changed online, which is a new application way for the magnetic bearing.
Highlights of Preliminary Level 1 PSA Results for a Research Reactor
Authors: Lee Yoon-Hwan , Jung Won-Dea , Park Su-Ki , Lee Jin-Hong
Published: October 2013

This preliminary PSA (Probabilistic Safety Assessment) was undertaken to assess the level of safety for the design of a research reactor and to evaluate whether it is probabilistically safe to operate and reliable to use. The scope of the PSA reported here is a Level 1 PSA that addresses the risks associated with core damage. It includes an evaluation of the types of accidents that could lead to core damage, and an assessment of their frequencies. After reviewing the documents and the conceptual design, 8 typical initiating events are selected regarding internal events during the normal operation of the reactor.

Simple fault tree models for the PSA are developed instead of the detailed model at this conceptual design stage. The failures of the major components and dependencies between systems have been considered for the fault tree analysis. Normal operating trains were assumed to have a pump, a check valve and a manual valve. The failures of pumps and supporting systems such as the electrical power are modeled, and the failure of the check valve or manual valve is also modeled for the train. Of course, the Common Cause Failure (CCF) and operator error are modeled.

The criterion for inclusion was all sequences with a point estimate frequency greater than a truncation value of 1.0E−13/yr. LOCA-I is the dominant contribution to the total CDF by a single initiating event. The CDF from the internal events of the research reactor is estimated to be 7.38E−07/year. The CDF for the representative initiating events is less than 1.0E−6/year even though conservative assumptions are used in the reliability data. The conceptual design of the research reactor is designed to be sufficiently safe from the viewpoint of safety.
Science Translational Medicine
Podcast: 30 October 2013
Authors: Mirkin Chad A. , Frisk Megan L.
Published: October 2013

In this podcast, Chad Mirkin describes how spherical nucleic acid (SNA) nanoparticles can be used for RNAi in glioblastoma.
One Device to Find Them All
Authors: Chow Edward Kai-Hua
Published: October 2013

Graphene oxide nanosheets in a microfluidic device integrate both isolation and analysis steps for tracking circulating tumor cells.
Microbiome Assembly across Multiple Body Sites in Low-Birthweight Infants
Authors: Costello Elizabeth K. , Carlisle Erica M. , Bik Elisabeth M. , Morowitz Michael J. , Relman David A.
Published: October 2013

The purpose of this study was to evaluate the composition and richness of bacterial communities associated with low-birthweight (LBW) infants in relation to host body site, individual, and age. Bacterial 16S rRNA genes from saliva samples, skin swabs, and stool samples collected on postnatal days 8, 10, 12, 15, 18, and 21 from six LBW (five premature) infants were amplified, pyrosequenced, and analyzed within a comparative framework that included analogous data from normal-birthweight (NBW) infants and healthy adults. We found that body site was the primary determinant of bacterial community composition in the LBW infants. However, site specificity depended on postnatal age: saliva and stool compositions diverged over time but were not significantly different until the babies were 15 days old. This divergence was primarily driven by progressive temporal turnover in the distal gut, which proceeded at a rate similar to that of age-matched NBW infants. Neonatal skin was the most adult-like in microbiota composition, while saliva and stool remained the least so. Compositional variation among infants was marked and depended on body site and age. Only the smallest, most premature infant received antibiotics during the study period; this heralded a coexpansion of
Pseudomonas aeruginosa
and a novel
Mycoplasma
sp. in the oral cavity of this vaginally delivered, intubated patient. We conclude that concurrent molecular surveillance of multiple body sites in LBW neonates reveals a delayed compositional differentiation of the oral cavity and distal gut microbiota and, in the case of one infant, an abundant, uncultivated oral
Mycoplasma
sp., recently detected in human vaginal samples.

IMPORTANCE
Complications of premature birth are the most common cause of neonatal mortality. Colonization by the indigenous microbiota, which begins at delivery, may predispose some high-risk newborns to invasive infection or necrotizing enterocolitis (NEC), and protect others, yet neonatal microbiome dynamics are poorly understood. Here, we present the first cultivation-independent time series tracking microbiota assembly across multiple body sites in a synchronous cohort of hospitalized low-birthweight (LBW) neonates. We take advantage of archived samples and publically available sequence data and compare our LBW infant findings to those from normal-birthweight (NBW) infants and healthy adults. Our results suggest potential windows of opportunity for the dispersal of microbes within and between hosts and support recent findings of substantial baseline spatiotemporal variation in microbiota composition among high-risk newborns.
Randomised controlled trial of vitamin D supplementation in sarcoidosis
Authors: Bolland Mark J , Wilsher Margaret L , Grey Andrew , Horne Anne M , Fenwick Sheryl , Gamble Greg D , Reid Ian R
Published: October 2013

Objectives

The role vitamin D intake/production plays in sarcoidosis-associated hypercalcaemia is uncertain. However, authoritative reviews have recommended avoiding sunlight exposure and vitamin D supplements, which might lead to adverse skeletal outcomes from vitamin D insufficiency. We investigated the effects of vitamin D supplementation on surrogate measures of skeletal health in patients with sarcoidosis and vitamin D insufficiency.

Design

Randomised, placebo-controlled trial.

Setting

Clinical research centre.

Participants

27 normocalcaemic patients with sarcoidosis and 25-hydroxyvitamin D (25OHD) <50 nmol/L.

Intervention

50 000 IU weekly cholecalciferol for 4 weeks, then 50 000 IU monthly for 11 months or placebo.

Primary and secondary outcome measures

The primary endpoint was the change in serum calcium over 12 months, and secondary endpoints included measurements of calcitropic hormones, bone turnover markers and bone mineral density (BMD).

Results

The mean age of participants was 57 years and 70% were women. The mean (SD) screening 25OHD was 35 (12) and 38 (9) nmol/L in the treatment and control groups, respectively. Vitamin D supplementation increased 25OHD to 94 nmol/L after 4 weeks, 84 nmol/L at 6 months and 78 nmol/L at 12 months, while levels remained stable in the control group. 1,25-Dihydroxy vitamin D levels were significantly different between the groups at 4 weeks, but not at 6 or 12 months. There were no between-groups differences in albumin-adjusted serum calcium, 24 h urine calcium, markers of bone turnover, parathyroid hormone or BMD over the trial. One participant developed significant hypercalcaemia after 6 weeks (total cholecalciferol dose 250 000 IU).

Conclusions

In patients with sarcoidosis and 25OHD <50 nmol/L, vitamin D supplements did not alter average serum calcium or urine calcium, but had no benefit on surrogate markers of skeletal health and caused one case of significant hypercalcaemia.

Trial registration

This trial is registered at the Australian New Zealand Clinical Trials Registry (http://www.anzctr.org.au). The registration number is ACTRN12607000364471, date of registration 5/7/2007.
PIASγ expression in relation to clinicopathological, tumour factors and survival in indigenous black breast cancer women
Authors: Agboola Ayodeji , Musa Adewale , Banjo Adekumbiola , Ayoade Babatunde , Deji-Agboola Mopelola , Nolan Christopher , Rakha Emad , Ellis Ian , Green Andrew
Published: October 2013

Aim

Indigenous black women with breast cancer (BC) show a high frequency of triple negative breast cancer (TNBC) comprising ER-, PR- and HER2- phenotypes and BRCA1 deficiency together with a high mortality rate, prompting speculation that risk factors could be genetic and the molecular portrait of these tumours may be different to those of Western women. Protein inhibitor of activated signal transducer (PIAS) γ implicated in the BRCA1 deficiency and triple negative BC was investigated to establish the relationship among the small ubiquitin-like modifier marker, pathological features, biomarkers expression and clinical outcome in the black women.

Materials and methods

This study investigated the immunoprofiles of PIASγ in 231 Nigerian BC prepared as tissue microarrays and correlated their protein expression with clinical outcome, pathological responses and the expression of 14 other relevant biomarkers.

Results

PIASγ protein expression showed a significant correlation with higher histological grade, basal-like biomarkers expression (CK14, CK5/6 and EGFR), BRCA1 regulator (MTA1), p53, PI3KCA, basal-like phenotype and TNBC. Also, an inverse correlation with steroid hormones (ER and PgR), p27, MDM4, mucin 1 and BRCA1 was observed with PIASγ expression. Univariate and multivariate survival analyses showed PIASγ expression was a predictor of poor outcome independent of tumour histological grade and ER expression.

Conclusions

PIASγ appears to be important in breast cancer behaviour arising from Nigerian women. PIASγ may therefore be useful for the screening of basal-like and TNBC. Also, development of novel therapies towards targeting PIASγ functional pathways may enhance the BC management among this ethnic nationality.
Full blood count normal reference values for adults in France
Authors: Troussard Xavier , Vol Sylviane , Cornet Edouard , Bardet Valérie , Couaillac Jean-Paul , Fossat Chantal , Luce Jean-Charles , Maldonado Eric , Siguret Virginie , Tichet Jean , Lantieri Olivier , Corberand Joël ,   for the French-Speaking Cellular Hematology Group (Groupe Francophone d'Hématologie Cellulaire, GFHC)
Published: October 2013

Aims

To determine full blood count (FBC) normal reference values for adults.

Methods

FBC normal values for healthy adults were defined, after establishing preanalytical conditions, in a population of 33 258 subjects, 19 612 men and 13 646 women. The values were established after excluding from this population all people having conditions liable to modify, directly or indirectly, FBC parameters.

Results

Results for values of standard parameters are provided in detail for each parameter, by sex and by age group from 16 to 69 years of age. In addition, we present FBC values from a population of 339 subjects aged over 69 years with no comorbidities.

Conclusions

These normal values are proposed for use in everyday practice. They make it possible to distinguish, without ambiguity, a normal situation from a pathological situation. Moreover, they might be used over all mainland France.