bluuebunny/crossref_metadata_2025 huggingface dataset:
Preliminary Thermal-Hydraulic Analyses for Designing an Experimental Model of a Molten Salt Reactor Concept
Authors: Yamaji Bogdán , Aszódi Attila
Published: February 2014

Based on the MSFR (Molten Salt Fast Reactor) reactor concept presented within the framework of the EVOL (Evaluation and Viability of Liquid Fuel Fast Reactor System, EU FP7) international research project preliminary three-dimensional thermal-hydraulic analyses and the discussion of scaled experimental modelling will be presented.

The MSFR concept is a single region, homogeneous liquid fuelled fast reactor. The reactor concept uses fluoride-based molten salts with fissile uranium and/or thorium and other heavy nuclei content with the purpose of applying the thorium cycle and the burn-up of transuranic elements. The concept has a single region cylindrical core with sixteen radial inlet and outlet nozzles located at the bottom and top of the core. The external circuit (internal heat exchanger, pump, pipes) is broken up into sixteen identical modules distributed around the core.

Purpose of the three-dimensional computational fluid dynamics (CFD) calculations is to study the possibility of experimental investigation of the fluid flow in the core of the proposed MSFR concept using a scaled model and Particle Image Velocimetry (PIV) flow measurement technique.

First the main properties of the proposed MSFR concept are introduced, and the information on other experimental thermal-hydraulic modelling of different reactors, including MSRE (Molten Salt Reactor Experiment) are summarised.

With a scaled plexiglas MSFR model it would be possible to carry out flow field measurements under laboratory conditions using PIV method. Possible way of scaling are presented and a series of preliminary CFD calculations are discussed. Possibilities and limitations of such scaling and segmenting of a model and the use of water as substitute fluid for the experimental mock-up will be discussed.

Objectives of such a measurement series would be validation, benchmarking of CFD calculations and codes, application of CFD modelling experience in the detailed thermal-hydraulic design of the MSFR concept, possible measurements for the study of specific problems or phenomena, for example refinement of inlet geometry, effects of internal structures, coolant mixing.
Development of Numerical Simulation Method for Relocation Behavior of Molten Debris in Nuclear Reactors: 1 — Preliminary Analysis of Relocation of Molten Debris to Lower Plenum
Authors: Yamashita Susumu , Yoshida Hiroyuki , Takase Kazuyuki
Published: February 2014

We developed the numerical simulation method for predicting the melting core behavior including solidification and relocation based on the three-dimensional multi-phase thermal-hydraulic simulation models. In this code, each of gas, liquid and solid phase are treated individually, and interface between two phases are simulated directly. In this paper, the developed code was applied to numerical simulations of the melting behavior of the simulated fuel assemblies and reactor structures. In the simulation, complicated structures in the BWR lower plenum was simply modeled. A decay heat in molten or solidified debris was also considered. Moreover, several different initial conditions were used to check performance of this code and to evaluate adequacy of the present numerical method. From the present numerical results, it was confirmed that relocation of molten debris in the BWR lower plenum can be simulated by the currently developed code including effects of melting and solidification of debris.
The Investigation of Minor Losses in T-Pipe Flow
Authors: Sun Dajie , Zhang Donghui , Hu Wenjun , Ren Lixia
Published: February 2014

It is now known explicitly how to calculate the pressure losses due to a developed flow in a pipe. As for pipe system, for simplicity, the minor losses caused by valves, elbows, enlargements, inlets, outlets, and other fittings was considered to be ignorable when compared with the frictional losses. However, actually, this is not always the case. In the work designing the safety rods for next generation of fast reactor, the inner structure which makes a large part of contribution to the pressure drop when the liquid sodium flow through them, should be considered carefully. In this case, the pressure losses due to T-joint need to be calculated. And unfortunately, empirical formulas couldn’t be found through the combination of survey and the huge quantity of reference data to solve the problem perfectly. In this paper, several simplified models are proposed to calculate to minor loss in T-pipe, which could be implied to the calculation of pressure drop of safety nod, and by comparing their feather, the most reasonable model is found. It has been found that even different models give different results of minor losses, the total pressure drop of the control rod, however, deviates no more than five percent.
Numerical Computation of One-Dimensional Unsteady Two-Phase Flow Using HEM Model and IAPWS IF-97 Equations of State
Authors: Khola Nitin , Pandey Manmohan
Published: February 2014

Numerical simulation of transients in two-phase flow is crucial to simulate accident-like and aberrant conditions of nuclear reactors for safety analysis. A considerable number of such problems can be treated as one-dimensional with significant reduction in complexity without much loss in applicability. Most commercial thermal hydraulic codes are based on the two-fluid model, which solves balance equations for each phase and also accounts for thermodynamic non-equilibrium between the phases. However, the homogenous equilibrium mixture (HEM) model of two-phase flow can be employed to develop simple and efficient codes for transient simulations, using which extensive parametric studies can be carried out. In the present work, a code for numerical computation of unsteady one-dimensional two-phase flow has been developed and reactor transients have been simulated. The governing equations were obtained by the HEM model and were decoupled and approximated using the sectionalized compressible flow (SC) model and the momentum integral (MI) model. The equations of state used in the code are based on IAPWS Industrial Formulation-97. Pressure and heat flux transients for PWR and BWR were simulated with the code and compared with those reported in the literature. Further numerical simulations with the code were carried out to predict the transient response of nuclear reactors to various perturbations.
Heat Transfer Correlation for Boiling Flows in Vertical Rectangular Narrow Channel
Authors: Chen Chong , Gao Puzhen
Published: February 2014

In view of the practical significance of a correlation of heat transfer coefficient in the aspect of such applications as engineering and predictive, some efforts towards correlating flow boiling heat transfer correlation for vertical rectangular narrow channels have been made in this paper. Based on analyses of existing general correlations for flow boiling heat transfer, it is found that general correlations are not suitable to predict heat transfer coefficients in narrow channels. By considering the suppression factor S and two-phase pool boiling heat transfer coefficient hnb, introducing boiling number Bo and two-phase friction multiplier ϕtt2, the Chen correlation has been modified to be used for narrow channels. The flow boiling heat transfer was the function of boiling number Bo and two-phase friction multiplier ϕtt2, the coefficients have a regular change with the increase of boiling number Bo or two-phase friction multiplier ϕtt2. A comparison of the newly developed correlation with the existing data for narrow channels shows a satisfactory agreement. The heat transfer coefficient of the narrow channel was well predicted by the new correlation and the deviation within ±20%.
Studies on the Responses of CR-39 Detector to C3+, O3+, and Proton Irradiations
Authors: Du X. , Liu S. H. , Zhang J. H. , Chen C. Q. , Ren X. T.
Published: February 2014

In the present study, CR-39 detectors with thickness of 1.0mm are irradiated respectively by C3+(10MeV, 20MeV, 20MeV), O3+(10MeV, 20MeV, 30MeV) particles and protons (5MeV) with ion fluences ranging from 1.0×107cm−2 to 5.0×109cm−2. The etching time conditions of the latent tracks for the different irradiation particles with different fluences described above are experimental studied. The curves for track density distributions with irradiation fluences of different energy particles are simulated. The suggested irradiation fluences ranges for the three types of particles detected with CR-39 detector are offered according to the measurement results.
Numerical Investigations on Instabilities in Supercritical Water Flowing Through Heated Parallel Channels
Authors: Mangukiya Bharat , Pandey Manmohan
Published: February 2014

The supercritical water-cooled reactor (SCWR) is a concept for an advanced reactor that operates at supercritical pressure and temperature. However, near the pseudocritical region the coolant properties are very sensitive to temperature, which raises concern about flow instabilities, whose analysis is very important to ensure safety of SCWR. Existence and nature of instabilities can be investigated numerically, using commercial system codes based on one-dimensional unsteady equations, suitable equations of state for thermo-physical properties of supercritical water, and suitable correlations for hydraulic resistance and heat transfer in supercritical fluids. In the present work, numerical investigation of stability characteristics of heated parallel channel system carrying supercritical water has been carried out using the RELAP5 code. The inlet and outlet pressures, the inlet temperature, and the heat flux were specified as the boundary conditions. The numerical model was validated with the help of experimental results reported in the literature. Transient numerical simulations were carried out to study the effect of various operating parameters (e.g., the mass flow rate, the inlet temperature, and the heat flux). The trends obtained were similar to those reported in the literature, albeit the asymmetry was less pronounced due to higher system pressure.
Advanced Fuel CANDU Reactor (AFCR) Core Physics With High-Burnup Recycled Uranium and Uranium-Thorium Fuel
Authors: Boubcher Mustapha , Meng Zhiliang , Cottrell Catherine M. , Kuran Sermet
Published: February 2014

The CANDU reactor has flexibility for using alternative fuels, such as Recycled Uranium (RU), low-enriched uranium and thorium (LEU/Th) and plutonium-thorium (Pu/Th) based fuels. This capability results from a versatile pressure tube design, simple fuel bundle, on-power refuelling, and high neutron economy. The Advanced Fuel CANDU Reactor (AFCR), which retains the excellent neutron economy and fuel cycle flexibility that are inherent in the CANDU reactor design, is a design jointly undertaken by Candu Energy and its Chinese partners led by CNNC. The objective is to use RU that is planned to be available from the reprocessing of Chinese LWR fleet spent fuel and increase the overall uranium utilization at minimal cost and minimal risk at commercially proven CANDU type new build reactors. This paper is focused on various physics aspects of using RU and LEU/Th in the AFCR units in the short run. It also describes full-core management simulations for the implementation of DRU and LEU/Th fuel cycles in AFCR reactor. The CANFLEX bundle design was chosen as carrier for both DRU and LEU/Th fuel in this study.
Neutronic Analysis and Fuel Cycle Simulation of the MIT Reactor Using MCODE-FM and Experimental Validation
Authors: Sun Kaichao , Ames Michael , Newton Thomas , Hu Lin-wen
Published: February 2014

A neutronic analysis of the Massachusetts Institute of Technology Research Reactor (MITR) is performed using state-of-the-art computational tools: the continuous-energy Monte Carlo code MCNP5 and the point-depletion code ORIGEN2.2. These codes are externally coupled by the in-house code package, MCODE (MCNP-ORIGEN Coupled Depletion Program), more recently, it being extended to MCODE-FM (Fuel Management). The latter features automated input file generation, data manipulation, and post-processing of the output data for the fuel cycle analysis, so that it is used to simulate the fuel management of the MITR. MCODE-FM also has an optional criticality search algorithm to simulate control blade movement.

The code validation is carried out by comparing the calculated results to experimental data. Two sets of the comparisons are made in the present paper: 1) the Xe-135 reactivity effect during the reactor start-up and shutdown and 2) the thermal and fast neutron flux in an irradiation capsule in the reactor core. Good agreements have been found.

The validated MCODE-FM is therefore useful for neutronic analysis and the fuel cycle simulation of the MITR. The time dependent variation of the key parameters, viz. the control blades’ axial position (maintaining criticality) and the fissile inventory in the fuel, is presented.
Safety Evaluation of Nuclear Plant Fire Using CFD Analysis
Authors: Matsuda Akihiro , Ishibashi Takashi , Tsuchino Susumu
Published: February 2014

In this paper, investigations of the safety of electrical cables in nuclear plant fire were described by CFD analysis and by flammability tests of power and control cables. For CFD analysis, thermal configurations of a nuclear power plant room in a fire accident was calculated by the Fire Dynamics Simulator (FDS). For flammability tests of cables, the cone-calorimeter test machine was modified to use the cable tray type specimen. The modified cone-calorimeter test machine is able to test the flammability of cables and insulation resistance from room temperature to 800°C.

In the flammability test, electric power and control cables of the polyethylene sheathed cables which had been applied to the nuclear plants were adopt.

The cable specimens of which length were 600mm, were fixed on a cable tray specimen. The temperature of the cable specimens was raised to from 200 to 570°C using the cone-type electrical heater. During the tests, electrical resistance and temperature was measured to evaluate safety of insulation resistance of cables.

As results, the destruction of the insulation resistance of a polyethylene sheathed control cable was observed in the examinations, temperatures of which were 400°C and 500°C.

In the numerical simulation using Fire Dynamics Simulator (FDS), an electrical switchgear room in nuclear plants was supposed as an evaluation model. As the fire source, fire in motor control cabinet in the switchgear room on nearby cable tray was applied to CFD analysis. From the results of numerical simulations, the heat configuration and temperature in the switchgear room was evaluated and radiation heat from fire to the cable trays was calculated.

Considering these results given by flammability tests of electrical cables and CFD analysis, the safety of a switchgear room in nuclear plants was shown.
Numerical Simulation of Single Bubble Moving in Stagnant Solid-Liquid Mixture Pool Using Finite Volume Particle Method
Authors: Aramaki Yuki , Suzuki Takahito , Miya Ichiro , Guo Liancheng , Morita Koji
Published: February 2014

Three-phase flow formed in a disrupted core of nuclear reactors is one of the key phenomena to be simulated in reactor safety analysis. Particle-based simulation could be a powerful CFD tool to understand and clarify local thermal-hydraulic behaviors involved in such three-phase flows. In the present study, to develop a computational framework for three-phase flow simulations, a single bubble moving in a stagnant solid particle-liquid mixture pool was simulated using the finite volume particle (FVP) method. The simulations were carried out in a two dimensional system. The bubble shape change and the bubble rise velocity were compared with the newly performed experiments, which used solid particulate glasses of 0.9 mm in diameter, liquid silicone and air. The two-phase flow simulation of a single bubble rising in a stagnant liquid pool reproduced measured bubble shape and bubble rise velocity reasonably. On the other hand, the bubble rise velocity in a stagnant particle-liquid mixture pool was overestimated in comparison with the measurement. This result suggests that particle-particle and particle-fluid interactions would have dominant influence on bubble motion behavior in the particle-liquid mixture pool under the present multiphase conditions. To evaluate such interactions in the simulations, the particle-particle interactions were modeled by the distinct element method (DEM), while two models were applied to represent particle-fluid interactions. One is the theoretical model for apparent viscosity of particle-liquid mixture, which describes the viscosity increase of liquid mixed with solids based on the Frankel-Acrivos equation. The other is the drag force model for solid-fluid interactions. In the present study, we took the Gidaspow drag correlation, which is a combination of the Ergun equation and Wen-Yu equation. A comparison of both the transient bubble shape and bubble rise velocity between the results of experiment and simulation demonstrates that the present computational framework based on the FVP method and solid-phase interaction models is useful for numerical simulations of a single bubble moving in a stagnant solid particle-liquid mixture pool.
Design and Development of HTR-PM Reactor Protection System
Authors: Li Duo , Xiong Huasheng , Guo Chao
Published: February 2014

High Temperature gas-cooled Reactor-Pebble bed Module (HTR-PM) Reactor Protection System (RPS) is a dedicated system to be designed and developed according to HTR-PM Nuclear Power Plant reactor protection specifications. HTR-PM RPS has the framework of four redundant channels and has two independent and diverse subsystem x and subsystem y to perform different protection functions, which would decrease the potential common cause failure caused by software and increase the system reliability.
Development of Experimental Setup of High Temperature Helium Loop and Preparation for In-Pile Operation
Authors: Miletić Marija , Fukač Rostislav , Pioro Igor , Dragunov Alexey
Published: February 2014

Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuels supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow’s needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts which will utilize thermal neutron spectrum is a Very High Temperature Reactor (VHTR).

This reactor concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behaviour within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Centre Rez, Ltd.

The purpose of the High Temperature Helium Loop (HTHL) is to simulate technical and chemical conditions of VHTR’s coolant. The loop is intended to serve an as experimental device for fatigue and creep tests of construction metallic materials for gas-cooled reactors and it should be also employed for research in field of gaseous coolant chemistry. The loop will serve also for tests of nuclear graphite, dosing and Helium purification systems. Because the VHTR is a new reactor concept, major technical uncertainties remain relative to helium-cooled advanced reactor systems. This paper summarizes the concept of the HTHL in the Research Centre Rez Ltd., its design, utilization and future plans for experimental setup.
Hybrid Seismic Response Evaluation Method for Wall Thinning Piping System
Authors: Sakai Michiya , Matsuura Shinichi , Inada Fumio
Published: February 2014

Pipe wall thinning is a one of the major degradation mechanisms in aged nuclear power plants (NPPs). In Japan, the seismic safety of wall thinning piping system during earthquake must be evaluated in aged NPPs. Seismic safety of piping systems with wall thinning had been investigated by other researchers using shaking table tests of reduced scale and numerical analyses. However, there exist the limitations such as the scale effect of pipe model for shaking table tests and the limit of the evaluation for numerical analysis concerning the criteria of pipe integrity. By the way, elbow can be one of the most important elements to evaluate the seismic safety of piping system. So, in order to evaluate seismic safety of piping systems with wall thinning elbow, hybrid tests have been conducted, in which the seismic response of the whole piping system is treated as a numerical model, and the real piping is used only for the element on which the transformation and damage locally concentrated. The through-wall crack only occurred in the case of a uniform thinning model although cracks didn’t penetrate in the non thinning model and the local thinning model. In the experimental condition, the failure mode of wall thinning elbow under seismic loadings had been low cycle fatigue, and effectiveness of this evaluation method has been demonstrated.
CFD Investigation of Heat Transfer Deterioration in Supercritical Water Flowing Through Vertical Annular Channels
Authors: Tejaswini U. S. , Basu Dipankar N. , Pandey Manmohan
Published: February 2014

In order to enhance the efficiency of current light water reactors, the generation IV initiative has included the supercritical water reactor (SCWR) as one of the future designs. The rapid change in density in the vicinity of the pseudo-critical temperature leads to strong buoyancy effect at low flow rates and flow acceleration at high flow rates, both of which significantly influence heat transfer characteristics. Experimental investigation of such phenomena being very cumbersome and cost-intensive, numerical simulation using CFD tools is considered to be a useful option for providing better understanding of the heat transfer mechanisms in geometries and conditions typical of SCWR. The present work involves numerical analysis of the heat transfer deterioration (HTD) phenomenon in turbulent flow of supercritical water through a vertical annular channel. ANSYS-CFX 14.0 software was employed for the same. An annular fluid domain, with a heated inner wall and an insulated outer wall, was modeled and the flow was considered to be in the upward and downward directions. Grid independence study was conducted with structured mesh. The results were compared with those reported in the published literature. It is known that the HTD phenomenon causes a sudden rise in the wall temperature, and hence it is necessary to predict the effect of changes in operating and design parameters. Parametric study was done by varying pressure, inlet temperature, heat flux and mass flux. Annuli of different hydraulic diameters were also considered.
RELAP5-3D Assessment Using a Qualified Experimental Database
Authors: Fiori F. , Kovtonyuk A. , Petruzzi A. , Kovtonyuk M. , D’Auria F.
Published: February 2014

The independent code assessment constitutes an important part in the qualification of an analytical tool for performing licensing or NPP support activities. The consistent application of adopted modeling procedures together with qualified experimental database plays a key role in the code assessment process. It should be mentioned that a consistent code assessment supported by a qualified experimental database is an important step for developing a solid ground for the uncertainty evaluation in the frame of Best Estimate Plus Uncertainty (BEPU) approach. The assessment of RELAP5-3D code capabilities for reproducing various accident and transient in PWR-like experimental Integral Test Facilities (ITF) has been performed at the Nuclear Research Group of San Piero a Grado (GRNSPG) of University of Pisa. The present paper summarizes the first series of this activity, aimed for RELAP5-3D assessment at Large Break LOCA and Small Break LOCA conditions. In this framework, 2 LBLOCA and 2 SBLOCA experimental tests performed at LOBI, LOFT and SPES ITF have been simulated by RELAP5-3D. First, an experimental database has been developed for mentioned facilities and experimental tests of interest. This process has been carried out following IAEA guidelines in Safety Series Report 23 and 52 and other known best practices. The elaboration of great amount of available information and adopting quality assurance procedures resulted in the standardized-form qualified Reference Data Sets (RDS) for each ITF. The second part of the paper discusses development of RELAP5-3D models for LOBI, LOFT and SPES facilities and simulation of the 4 tests, namely LOBI A1-06 and LOFT L2-5 (2A LOCA), and LOBI A1-83 and SPES SB-04 (10% CL LOCA and 6% CL LOCA respectively). All models were developed following the same nodalization techniques adopted at GRNSPG-UNIPI. All calculations were qualified both at steady-state and transient conditions applying the same set of qualitative and quantitative criteria. The judgment of RELAP5-3D code capabilities to reproduce the relevant thermal-hydraulic phenomena has been reported.
Ultrasonic Examination of Noble Metal Precipitate Detection in Glass Melter
Authors: Koikegami Hajime , Yasunaga Shigeyoshi , Kakihara Toshiaki , Yanagihara Arisa , Kawai Hiroki , Hatanaka Hiroaki , Maruyama Tadashi
Published: February 2014

This paper describes the results of ultrasonic examination of the build-up of noble metal precipitates in the glass melter using a wave guide rod. The glass melter is used for the vitrification process of high-level radioactive liquid waste (HLLW).

A high density isotropic graphite T-6 (Ibiden Co., Ltd.) and a nickel-based corrosion and heat resistant alloy 690 were selected as candidate materials of wave guide rod. Ultrasonic propagation experiment with the frequency 0.5, 1.0 and 2.25 MHz was performed using the wave guide rods. The T-6 graphite wave guide rod in the simulant fluid (glycerin) showed a good capability of ultrasonic propagation and signal detection. However, the propagation of ultrasonic wave into molten glass was not possible due to the poor wetting of molten glass with the graphite. On the other hand, it was possible to propagate the ultrasonic wave into molten glass with the wave guide rod made of the nickel-based alloy 690. However, the intensity of reflecting echo signal was very small due to the large ultrasonic attenuation in the nickel-based alloy. The simulation study on the ultrasonic wave propagation using the three dimensional finite element method code ComWave showed good agreement with the experimental results.
Comparative Study on Self-Compacting Concrete Pouring Methods of AP1000 Structure Module
Authors: Dai Shuai , Ye Zhonghao , Li Erfei , Zheng Chuyao
Published: February 2014

Modular construction technology is one of the main technical features of AP1000. But after the construction practices of two self-reliance supporting projects, we found that concrete poured in the module is extremely difficult, the advantages of shorter construction duration in modular construction will be seriously weakened, so we designed three different pouring methods: layered pouring method, layered continuous pouring method and loop continuous pouring method. Deep comparative analysis on the principle, characteristic and implementation effect of these three pouring methods are studied in this paper.
Experimental Investigation of Boil-Off-Scenario in BWR Spent Fuel Pools
Authors: Schulz Steffen , Schuster Christoph , Hurtado Antonio
Published: February 2014

Accident scenarios in spent fuel pools (SFP) were not a main topic in nuclear safety studies so far. Heat fluxes are low and the timeframe for counteractions in case of a loss of cooling is long. However, because of their huge activity inventory safety of spent fuel pools should be analyzed with great intensity. Beside some analytical approximations from the U.S. Nuclear Regulatory Commission (NRC) there is only few experimental data about loss of coolant accidents in a SFP [1, 3, 4, 5, 6]. After the reactor accident in Fukushima Daiichi NPP this issue became omnipresent in public perception. The presented paper investigates boil-off scenario of a boiling water reactor (BWR) SFP after a station blackout with focus on experimental findings.

Since 2006 experiments for SFP boil-off scenarios were conducted at the Technische Universität Dresden [7]. For a detailed investigation under realistic geometrical fuel assembly conditions the test facility ADELA-II was built in 2011. It consists of two channels. The inner channel simulates a quarter of a BWR fuel element with a 24 rod bundle. In the outer channel 8 more heating rods were assembled to reduce heat losses from the inner channel and to simulate the surrounding. With a heated length of 3760 mm and a width of 120 mm it was designed in an axial and radial scale of 1:1. The heating profile is based on the heat flux profile of a fuel rod with low burn-up. To measure the detailed axial and radial temperature profiles 113 thermocouples are mounted. With these data conclusions for heat and mass transfer can be made with special regard to convection inside the assembly. Boil-off experiments at the ADELA-II facility with a supplied heat of 20 Watts per rod lead to rod surface peak temperatures of about 479 °C. Due to the high temperature and narrowed geometry the axial heat transport is harmed and radial heat conduction had a great influence on the test results. Accordingly, quasi adiabatic boil-off could not be verified. Thus, coolability of a fuel assembly strongly depends on the (radial) heat removal from the fuel assembly box to the surrounding. Further investigations have to be done for a better consideration of neighbored fuel elements and global thermohydraulic effects in the spent fuel pool.
Risk Insights and Use Based on Fukushima Accident
Authors: Chapman James R. , Fine K. Raymond , Grantom C. Rick
Published: February 2014

A beyond design basis event resulted in severe core damage at 3 units at Fukushima Dai-iche. In response the International community identified lessons learned, conducted initial stress tests, and requested nuclear power plant operators to conduct additional evaluations and act on the results of these evaluations. In the US actions have included development of enhanced strategies to deal with an extended loss of offsite power or the ultimate heat sink, design and procurement of portable equipment to support core cooling indefinitely, and external hazard evaluations. The key external hazard evaluations are related to external flooding and seismic hazards.

In addition to directed evaluations, NRC, ASME and NEI are pursuing refinements to the risk informed approach to decision making. The use of PRA in a risk informed decision making process, combined with defense in depth and safety margin considerations, has expanded considerably in the past two decades. So the questions addressed in this paper are: Why expand this approach further? How are PRA and traditional practices balanced? And what is the outcome if we do not expand this approach?

The insights are based on the ongoing activity by NRC to address Near Term Task Form (NTTF) Recommendation 1 [1,2] NUREG-2150 [3] which proposes a more risk informed regulatory process, ASME [4] and Industry activity lead by the Nuclear Energy Institute (NEI) [5]; and past and ongoing uses of PRA.

In summary, the paper concludes that an improved and increased use of a risk informed process is appropriate so as to better address lessons learned from the events at Fukushima, to better understand and act on limiting plant features (“vulnerabilities”), and have a framework in place to address any future beyond design basis events or new information more coherently.

A key lesson learned from Fukushima Dai-ichi is that PRA can identify vulnerabilities which can be acted on to reduce the potential for loss of another plant. But this requires the International community to embrace PRA, and to date the International community has not. This is because most in the Industry rely on checklists and are challenged to envision low frequency high consequence events.
A Research on Eliminating the Deposited Water After Water-Ingress Accident in HTR
Authors: Liu Peng , Zheng Yanhua
Published: February 2014

Water-ingress accident, caused by steam generator heating tube rupture of a high temperature gas-cooled reactor (HTGR) is an important accident to consider because it will introduce positive reactivity leading the nuclear power increase rapidly, as well as the chemical reaction of graphite fuel elements and reflector structure material with steam. Researches and simulations (Zuoyi Zhang et al. 1995; Zheng Yanhua et al. 2009) have been carried out for calculating the total amount of water ingress and to validate the safety and security of HTR. The water ingress amount, which is our mainly concerned, ranges from a few hundred kilograms to thousands of kilograms, because of the different reactors and different computing methods. The place, where the water deposits, is most likely to be the bottom of the steam generator.. Such liquid water removal, with the approach of providing a forced circulation in primary loop and accelerating the evaporation, is analyzed in this paper.

Many experimental data have been got on water evaporation rate (Dalton et al.; Willis Carrie et al. 1914; Yoshida, Hyodo et al. 1970; Sweer et al. 1976; Pauken et al. 1995). All these formulas have a common form, ṁ = hc(Pw − Pa)/hw, which shows the relationship between evaporation rate and velocity over water face, water temperature and the relative partial pressure of the water vapor. This formula has been used widely in chemical industry and other fields and shows good agreement.

FLUENT CFD code (ANSYS Fluent 14) is used for the calculation of the distribution of the flow and temperature fields. The evaporation rate is estimate combined thermal fluid data with empirical formula. As the working condition of empirical formula and that of actual reactor don’t match very well, sensitivity analysis is necessary in this report.
Issues Associated With the Development of Severe Accident Management Guidelines for NPPs in China
Authors: Yang Zhiyi , Chong Yimin , Li Chun , Deng Jian , Xu Xianhong , Chai Guohan
Published: February 2014

After Fukushima nuclear accident, the Severe Accident Management Guidelines (SAMGs) are required according to the policy of the regulatory body in China. Most nuclear power plants (NPPs) in China adopt the technical approach of generic-SAMG of the Westinghouse Owner Group, consisting of severe accident control room guideline (SACRG), diagnostic flow chart (DFC), severe accident guideline (SAG), severe challenge status tree (SCST), severe challenge guideline (SCG), technical support center (TSC) long term monitoring guideline, and SAMG termination guideline (SAEG). A number of issues have been identified during the development of the SAMGs for M310+ NPPs, which is a dominant reactor type in China. The paper discussed these issues and identified some considerations for their resolution.
Feasibility Study of Utilizing TRISO Fuel in PWR Technology
Authors: Hussain Anwar , Khan Ammar
Published: February 2014

The aim of this paper is to summarize the research that has already been conducted in the field of PWR technology which utilizes TRISO fuel (Tri-Structural Isotropic) instead of standard PWR fuel. TRISO fuel which in mainly used in HTGRs has been used in SMRs. This type of reactor can be a potential source of energy for a limited power production, heating and desalination facilities at remote location. Combination of TRISO fuel and PWR technology results in a reliable and safe NPP design. Due to enhanced and outstanding passive safety features advanced SMRs are being studied around the world and substantial amount of conceptual research has already been conducted in this field. However, further research is required to be undertaken to optimize these design concepts.

Finally, few areas have been pointed out for further research to overcome the challenges and gaps in this research field.
Ventilation Management for the AP1000 Containment Building During Construction Stage
Authors: Sun Zhiyou , Shen Hongwei
Published: February 2014

Confined space can be defined as closed or semi-closed equipments, facilities and places which potentially have poor ventilation, poisonous and hazardous gases or poor oxygen. Confined space has complex working environment and more hazardous factors which makes safety accidents occurred frequently. Modularized Construction, as the typical feature of AP1000 nuclear project construction, shortens the construction schedule but increases safety risk, since there is more confined space and cross work, especially in the Containment building after the setting of CVTH.

This thesis analyses the confined space construction feature of the containment building and emphasizes the necessity of ventilation and fume control. Some ventilation methods have been taken on site which has effectively reduced the construction risk inside containment buildings. Therefore, this thesis has practical value for the consequent AP1000 project construction.
Deposition Fractions of Fission Product and Heavy Elements on Soil in Fukushima Dai-Ichi Nuclear Power Plant Accident
Authors: Yamamoto Toru
Published: February 2014

Based on radioactivity measurement of soil samples in the site of Fukushima Dai-Ichi Nuclear Power Station, radioactivity of Sr, Nb, Mo, Tc, Ru, Ag, Te, I, Cs, Ba, La, Pu, Am, and Cm isotopes were compiled as radioactivity ratios to 137Cs. By exponentially fitting or averaging, the radioactivity ratios at the core shutdown were estimated. They were divided by those of the fuel of the core at the shutdown to obtain a deposited radioactivity fractions of the nuclides as relative values to 137Cs, which also correspond to deposition fractions of the elements as relative values to Cs. They were estimated to be orders of 10−4 to 10−3 for Sr, 10−4 for Nb, 10−2 to 10−1 for Mo, 10−1 for Ag, 10−1 to 100 for Te, 100 for I, 10−3 for Ba, 10−6 to 10−5 for Pu, 10−6 to 10−5 for Am, and 10−6 for Cm. The observed radioactivity ratios to 137Cs were compared with those obtained by severe accident analysis to assess the validation of the analysis.
Estimation of Fukushima-Daiichi NPS PCV Cooling States Based on Measurement Data After Nitrogen Injection
Authors: Ichikawa Ryoko , Masuhara Yasuhiro
Published: February 2014

The cooling states in the PCV are not clear after the TEPCO’s Fukushima Daiichi NPS accident. But from the standpoint of the long-term stability cooling, the understanding of the PCV cooling condition is important.

Therefore, the simple method was developed in which the PCV water and gas leak diameters and those locations, etc. could be estimated based on the time transient data. By applying this method to Fukushima Daiichi NPS accident after the nitrogen injection for prevention of the hydrogen explosion, the PCV leak locations, diameters and partial pressures, etc. were estimated.

It was found that the partial pressure of hydrogen and oxygen was kept low and the nitrogen injection to PCV was effective as a prevention of hydrogen explosion. The decay heat from the fuel was mainly removed by the sub-cooling water injected and steam condensation on the PCV wall. The heat release by leakage was estimated to be small.
Application of Cloud Chambers for Heuristic Comprehension of Radiation
Authors: Tawara Yuzuru , Hara Shinji , Koga Kazuo , Tsuji Kenji
Published: February 2014

In the nuclear-related public relation center, a variety of displays such as models of nuclear facility, panel presentations to explain radiation properties, radiation detectors are used for the visitors to help the understanding of nuclear power and radiation. The PR center demonstrates various aspect of the radiation such as the presence of natural background radiations around us in the daily life, shielding effect for the different kind of radiations. Cloud Chambers are often used to demonstrate the presence of natural background radiation, showing tracks caused by the ionizing effect. The shielding effect of radiation is demonstrated by inserting a shield material between a radiation source and the detector such as GM counter. It is usually illustrated in the panel that the penetration properties are different for kinds of radiation but actual demonstration is seldom used.

Then a question arises that people cannot properly understand overall characteristics of the radiation in the above described demonstrations in the PR center. So we have been trying to improve a utilization method of a cloud chamber to help the deep understanding of properties of radiation. The improved cloud chamber has the area size of 225 mm × 225 mm with 100 mm in depth and has the structure for the function of insertion and extraction of both a radiation source and a shield material, independently. When a radiation source such as Cs-137 or Fe-55 is inserted in the chamber, it is clearly shown that high energy electron created by gamma-ray or X-ray emitted from such radioisotope can make track with different length. It is also shown using lantern core with thorium-series isotopes that thick track can be made by alpha particle. Fe-55 radioisotope generates 5.9 keV X-rays, which produces about 1mm track in the chamber through photoelectron.

When a shield material of lead with a thickness of 1mm was inserted in front of the source, the fading out of the tracks is clearly observed. Thus shielding effect can be easily confirmed by eye using cloud chamber. The demonstrations of shielding effect described above can help more clear and essential understanding of the radiations. This was shown by the questionnaire survey done before and after the demonstrations for the 32 participants.

For further improvements of the cloud chamber, we are planning to enlarge the size of cloud chamber and to get much clearer track image by improving track illumination method. Finally we will re-consider more effective explanation to give correct understanding of the radiation and will verify the effectiveness of utilization method of new cloud chamber.
Helium ion microscopy
Authors: Hlawacek Gregor , Veligura Vasilisa , van Gastel Raoul , Poelsema Bene
Published: February 2014

Helium ion microcopy based on gas field ion sources represents a new ultrahigh resolution microscopy and nanofabrication technique. It is an enabling technology that not only provides imagery of conducting as well as uncoated insulating nanostructures but also allows to create these features. The latter can be achieved using resists or material removal due to sputtering. The close to free-form sculpting of structures over several length scales has been made possible by the extension of the method to other gases such as neon. A brief introduction of the underlying physics as well as a broad review of the applicability of the method is presented in this review.
Levels and Patterns of Daily Physical Activity and Sedentary Behavior Measured Objectively in Older Care Home Residents in the United Kingdom
Authors: Barber S.E. , Forster A. , Birch K.M.
Published: February 2014

Background:

Physical activity is important for maintaining independence and quality of life in older people living in care homes. Little is known about patterns of physical activity or sedentary behavior in this population.

Methods:

Thirty-three care home residents (82.6 ± 9.2 years) wore an ActiGraph GTX3 accelerometer for seven days, which provided minutes of sedentary behavior and low, light, and moderate-to-vigorous physical activity. Participants undertook the Mini-Mental State Examination and care staff reported activities of daily living (Barthel index) and functional ambulation classification (FAC) for each participant.

Results:

Participants spent on average 79% of their day sedentary, 14% in low, 6% in light, and 1% in moderate-to-vigorous physical activity. Activity levels did not significantly differ between days or hours of the day (P > .05).

Conclusion:

Levels of physical activity were very low and time being sedentary was high. This study can inform physical activity and sedentary behavior interventions for care homes’ residents.
Whole-Genome Shotgun Assembly and Analysis of the Genome of
Shiraia
sp. Strain Slf14, a Novel Endophytic Fungus Producing Huperzine A and Hypocrellin A
Authors: Yang Huilin , Wang Ya , Zhang Zhibin , Yan Riming , Zhu Du
Published: February 2014

Here, we report the draft genome sequence of
Shiraia
sp. strain Slf14 (China Center for Type Culture Collection [CCTCC] no. 209294), which is used to produce huperzine A and hypocrellin A. The genome sequence will allow for the characterization of the molecular mechanisms underlying its beneficial properties.
Genome Sequence of Legionella anisa, Isolated from a Respiratory Sample, Using an Amoebal Coculture Procedure
Authors: Pagnier Isabelle , Croce Olivier , Robert Catherine , Raoult Didier , La Scola Bernard
Published: February 2014

Legionella anisa
is a gammaproteobacterium from the class
Legionellaceae
, which is responsible for nosocomial pneumonia. We sequenced the genome from the
L. anisa
strain Linanisette, which was recovered from a clinical sample using an amoebal coculture procedure but not with standard culture methods.
Draft Genome Sequence of Clostridium ultunense Strain BS (DSMZ 10521), Recovered from a Mixed Culture
Authors: Wei Yongjun , Zhou Haokui , Zhang Lei , Zhang Jun , Wang Yuezhu , Wang Shengyue , Zhou Zhihua , Yan Xing
Published: February 2014

Clostridium ultunense
BS is the first isolated strain (type strain) of
C. ultunense
that was identified as a mesophilic syntrophic acetate-oxidizing bacterium (SAOB). Here, we report the draft genome sequence of this strain, which will help us to elucidate the mechanism of syntrophic acetate oxidization.
How well do doctors think they perform on the General Medical Council's Tests of Competence pilot examinations? A cross-sectional study
Authors: Mehdizadeh Leila , Sturrock Alison , Myers Gil , Khatib Yasmin , Dacre Jane
Published: February 2014

Objective

To investigate how accurately doctors estimated their performance on the General Medical Council's Tests of Competence pilot examinations.

Design

A cross-sectional survey design using a questionnaire method.

Setting

University College London Medical School.

Participants

524 medical doctors working in a range of clinical specialties between foundation year two and consultant level.

Main outcome measures

Estimated and actual total scores on a knowledge test and Observed Structured Clinical Examination (OSCE).

Results

The pattern of results for OSCE performance differed from the results for knowledge test performance. The majority of doctors significantly underestimated their OSCE performance. Whereas estimated knowledge test performance differed between high and low performers. Those who did particularly well significantly underestimated their knowledge test performance (t (196)=−7.70, p<0.01) and those who did less well significantly overestimated (t (172)=6.09, p<0.01). There were also significant differences between estimated and/or actual performance by gender, ethnicity and region of Primary Medical Qualification.

Conclusions

Doctors were more accurate in predicating their knowledge test performance than their OSCE performance. The association between estimated and actual knowledge test performance supports the established differences between high and low performers described in the behavioural sciences literature. This was not the case for the OSCE. The implications of the results to the revalidation process are discussed.