Science and Technology of Nuclear Installations The latest articles from Hindawi Publishing Corporation © 2015 , Hindawi Publishing Corporation . All rights reserved. Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code Wed, 25 Feb 2015 06:41:19 +0000 Cases of core coolant flow blockage transient have been simulated and analysed for the TRIGA IPR-R1 research reactor using the RELAP5-MOD3.3 code. The transients are related to partial and to total obstruction of the core coolant channels. The reactor behaviour after the loss of flow was analysed as well as the changes in the coolant and fuel temperatures. The behaviour of the thermal hydraulic parameters from the transient simulations was analysed. For a partial blockage, it was observed that the reactor reaches a new steady state operation with new values for the thermal hydraulic parameters. The total core blockage brings the reactor to an abnormal operation causing increase in core temperature. Patrícia A. L. Reis, Antonella L. Costa, Claubia Pereira, Maria Auxiliadora F. Veloso, and Amir Z. Mesquita Copyright © 2015 Patrícia A. L. Reis et al. All rights reserved. Safety Assessment of Low-Contaminated Equipment Dismantling at Nuclear Power Plants Mon, 23 Feb 2015 08:49:56 +0000 The decommissioning of nuclear facilities requires adequate planning and demonstration that dismantling and decontamination activities can be conducted safely. Existing safety standards require that an appropriate safety assessment be performed to support the decommissioning plan for each facility (International Atomic Energy Agency, 2006). This paper presents safety assessment approach used in Lithuania during the development of the first dismantling and decontamination project for Ignalina NPP. The paper will mainly focus on the identification and assessment of the hazards raised due to dismantling and decontamination activities at Ignalina Nuclear Power Plant and on the assessment of the nonradiological and radiological consequences of the indicated most dangerous initiating event. The drop of heavy item was indicated as one of most dangerous initiating events for the discussed Ignalina Nuclear Power Plant dismantling and decontamination project. For the analysis of the nonradiological impact the finite element model for the load drop force calculation was developed. The radiological impact was evaluated in those accident cases which would lead to the worst radiological consequences. The assessments results show that structural integrity of the building and supporting columns of building structures will be maintained and radiological consequences are lower than the annual regulatory operator dose limit. Egidijus Babilas, Eugenijus Ušpuras, Sigitas Rimkevičius, Gintautas Dundulis, and Mindaugas Vaišnoras Copyright © 2015 Egidijus Babilas et al. All rights reserved. AP1000 Shield Building Dynamic Response for Different Water Levels of PCCWST Subjected to Seismic Loading considering FSI Wed, 18 Feb 2015 08:29:59 +0000 Huge water storage tank on the top of many buildings may affect the safety of the structure caused by fluid-structure interaction (FSI) under the earthquake. AP1000 passive containment cooling system water storage tank (PCCWST) placed at the top of shield building is a key component to ensure the safety of nuclear facilities. Under seismic loading, water will impact the wall of PCCWST, which may pose a threat to the integrity of the shield building. In the present study, an FE model of AP1000 shield building is built for the modal and transient seismic analysis considering the FSI. Six different water levels in PCCWST were discussed by comparing the modal frequency, seismic acceleration response, and von Mises stress distribution. The results show the maximum von Mises stress emerges at the joint of shield building roof and water around the air inlet. However, the maximum von Mises stress is below the yield strength of reinforced concrete. The results may provide a reference for design of the AP1000 and CAP1400 in the future. Daogang Lu, Yu Liu, and Xiaojia Zeng Copyright © 2015 Daogang Lu et al. All rights reserved. Development of the Noncontact Temperature Sensor Using the Infrared Optical Fiber Coated with Antifog Solution Wed, 11 Feb 2015 11:24:15 +0000 This study developed a noncontact fiber-optic temperature sensor that can be installed in a spent nuclear fuel pool. This fiber-optic temperature sensor was fabricated using an infrared optical fiber to transmit the infrared light emitted from water at a certain temperature. To minimize the decrease in the detection efficiency of the fiber-optic temperature sensor due to vapor generation, its surface was coated by spraying an antifog solution and drying several times. The measurement data of the fiber-optic temperature sensor was almost linear in the range of 30~70°C. This sensor could be used as an auxiliary temperature monitoring system in a spent nuclear fuel pool. Rinah Kim, Chan Hee Park, Arim Lee, and Joo Hyun Moon Copyright © 2015 Rinah Kim et al. All rights reserved. Scientific and Engineering Literature Mini Review of Molten Salt Oxidation for Radioactive Waste Treatment and Organic Compound Gasification as well as Spent Salt Treatment Tue, 10 Feb 2015 08:35:18 +0000 Literature review was performed for the molten salt oxidation technology in order to collect all available scientific and engineering information for further use of this technology in nuclear applications. This report provides a summary of a review of scientific and engineering literature on MSO treatment of a wide variety of radioactive wastes, organic compound gasification, and related studies such as radioactive spent salt processing that was found important for further development of the MSO technology in the nuclear field for radioactive waste treatment. Miscellaneous nonnuclear uses of molten salts, such as converting carbon monoxide to carbon dioxide, are not discussed. Petr Kovařík, James D. Navratil, and Jan John Copyright © 2015 Petr Kovařík et al. All rights reserved. Study on the Use of Hydride Fuel in High-Performance Light Water Reactor Concept Sun, 01 Feb 2015 07:22:33 +0000 Hydride fuels have features which could make their use attractive in future advanced power reactors. The potential benefit of use of hydride fuel in HPLWR without introducing significant modification in the current core design concept of the high-performance light water reactor (HPLWR) has been evaluated. Neutronics and thermal hydraulic analyses were performed for a single assembly model of HPLWR with oxide and hydride fuels. The hydride assembly shows higher moderation with softer neutron spectrum and slightly more uniform axial power distribution. It achieves a cycle length of 18 months with sufficient excess reactivity. At Beginning of Cycle the fuel temperature coefficient of the hydride assembly is higher whereas the moderator and void coefficients are lower. The thermal hydraulic results show that the achievable fuel temperature in the hydride assembly is well below the design limits. The potential benefits of the use of hydride fuel in the current design of the HPLWR with the achieved improvements in the core neutronics characteristics are not sufficient to justify the replacement of the oxide fuel. Therefore for a final evaluation of the use of hydride fuels in HPLWR concepts additional studies which include modification of subassembly and core layout designs are required. Haileyesus Tsige-Tamirat and Luca Ammirabile Copyright © 2015 Haileyesus Tsige-Tamirat and Luca Ammirabile. All rights reserved. Research and Evaluation for Passive Safety System in Low Pressure Reactor Mon, 26 Jan 2015 06:42:18 +0000 Low pressure reactor is a small size advanced reactor with power of 180 MWt, which is under development at Nuclear Power Institute of China. In order to assess the ability and feasibility of passive safety system, several tests have been implemented on the passive safety system (PSS) test facility. During the LOCA and SBO accident, the adequate core cooling is provided by the performance of passive safety system. In addition the best-estimate thermal hydraulic code, CATHARE V2.1, has been assessed against cold leg LOCA test. The calculation results show that CATHARE is in a satisfactory agreement with the test for the steady state and transient test. Peng Chuanxin, Zhuo Wenbin, Chen Bingde, Nie Changhua, and Huang Yanping Copyright © 2015 Peng Chuanxin et al. All rights reserved. Single Phase Natural Circulation Behaviors of the Integral Type Marine Reactor Simulator under Rolling Motion Condition Mon, 26 Jan 2015 06:37:38 +0000 During operation in the sea the reactor natural circulation behaviors are affected by ship rolling motion. The development of an analysis code and the natural circulation behaviors of a reactor simulator under rolling motion are described in this paper. In the case of rolling motion, the primary coolant flow rates in the hot legs and heating channels oscillated periodically, and the amplitude of flow rate oscillation was in direct proportion to rolling amplitude, but in inverse proportion to rolling period. The total mass flow rate also oscillated with half the rolling period, and the average total mass flow rate was less than that in steady state. In the natural circulation under a rolling motion, the flow rate oscillations in the hot legs were controlled by the tangential force; however, the mass flow rate oscillations in the total natural circulation and the heating channels were a result of the combined action of the change of inclination angle, flow resistance, and the extra force arising from the rolling motion. The extra tangential force brought about intense flow rate oscillations in the hot legs, which resulted in increasing total flow resistance; however the extra centrifugal force played a role in increasing thermal driving head. Hou-jun Gong, Xing-tuan Yang, Yan-ping Huang, and Sheng-yao Jiang Copyright © 2015 Hou-jun Gong et al. All rights reserved. INPRO Activities on Development of Advanced Tools to Support Judgment Aggregation for Comparative Evaluation of Nuclear Energy Systems Wed, 21 Jan 2015 14:08:42 +0000 This paper presents first results of the INPRO Collaborative Project on Key Indicators for Innovative Nuclear Energy Systems, which has the objective to develop guidance and tools for comparative evaluation of the status, prospects, benefits, and risks associated with development of innovative nuclear technologies for a more distant future. Presented results illustrate expedience of application of the multicriteria decision analysis methods, which are able to provide the added value to comparative assessment of nuclear energy systems. First, the paper presents a short review of the multicriteria decision analysis methods appropriate to support judgment aggregation within comparative evaluations of nuclear energy systems based on key indicators and highlights the methodology to perform such assessments. Second, a set of key indicators elaborated in the INPRO Collaborative Project on Global Architecture of Innovative Nuclear Energy Systems Based on Thermal and Fast Reactors Including a Closed Fuel Cycle (GAINS) were evaluated for comparative evaluation of nuclear energy system evolution scenarios. Third, a numerical example is presented of application of the selected key indicators, methods, and tools for judgment aggregation in comparative assessment of the GAINS nuclear energy systems. V. Kuznetsov, G. Fesenko, A. Andrianov, and I. Kuptsov Copyright © 2015 V. Kuznetsov et al. All rights reserved. SIMMER-III Analyses of Local Fuel-Coolant Interactions in a Simulated Molten Fuel Pool: Effect of Coolant Quantity Wed, 14 Jan 2015 12:04:42 +0000 Studies on local fuel-coolant interactions (FCI) in a molten pool are important for the analyses of severe accidents that could occur for sodium-cooled fast reactors (SFRs). To clarify the mechanisms underlying this interaction, in recent years, several experimental tests, with comparatively larger difference in coolant volumes, were conducted at the Japan Atomic Energy Agency by delivering a given quantity of water into a molten pool formed with a low-melting-point alloy. In this study, to further understand this interaction, interaction characteristics including the pressure buildup as well as mechanical energy release and its conversion efficiency are investigated using the SIMMER-III, an advanced fast reactor safety analysis code. It is found that the SIMMER-III code not only reasonably simulates the transient pressure and temperature variations during local FCIs, but also supports the limited tendency of pressurization and resultant mechanical energy release as observed from experiments when the volume of water delivered into the pool increases. The performed analyses also suggest that the most probable reason leading to such limited tendency should be primarily due to an isolation effect of vapor bubbles generated at the water-melt interface. Songbai Cheng, Ken-ichi Matsuba, Mikio Isozaki, Kenji Kamiyama, Tohru Suzuki, and Yoshiharu Tobita Copyright © 2015 Songbai Cheng et al. All rights reserved. PSO Based Optimization of Testing and Maintenance Cost in NPPs Tue, 09 Dec 2014 08:59:21 +0000 Testing and maintenance activities of safety equipment have drawn much attention in Nuclear Power Plant (NPP) to risk and cost control. The testing and maintenance activities are often implemented in compliance with the technical specification and maintenance requirements. Technical specification and maintenance-related parameters, that is, allowed outage time (AOT), maintenance period and duration, and so forth, in NPP are associated with controlling risk level and operating cost which need to be minimized. The above problems can be formulated by a constrained multiobjective optimization model, which is widely used in many other engineering problems. Particle swarm optimizations (PSOs) have proved their capability to solve these kinds of problems. In this paper, we adopt PSO as an optimizer to optimize the multiobjective optimization problem by iteratively trying to improve a candidate solution with regard to a given measure of quality. Numerical results have demonstrated the efficiency of our proposed algorithm. Qiang Chou, Daochuan Ge, and Ruoxing Zhang Copyright © 2014 Qiang Chou et al. All rights reserved. A Nuclear Reactor Transient Methodology Based on Discrete Ordinates Method Thu, 04 Dec 2014 00:10:08 +0000 With the rapid development of nuclear power industry, simulating and analyzing the reactor transient are of great significance for the nuclear safety. The traditional diffusion theory is not suitable for small volume or strong absorption problem. In this paper, we have studied the application of discrete ordinates method in the numerical solution of space-time kinetics equation. The fully implicit time integration was applied and the precursor equations were solved by analytical method. In order to improve efficiency of the transport theory, we also adopted some advanced acceleration methods. Numerical results of the TWIGL benchmark problem presented demonstrate the accuracy and efficiency of this methodology. Shun Zhang, Bin Zhang, Penghe Zhang, Hui Yu, and Yixue Chen Copyright © 2014 Shun Zhang et al. All rights reserved. Modelling of QUENCH-03 and QUENCH-06 Experiments Using RELAP/SCDAPSIM and ASTEC Codes Tue, 02 Dec 2014 11:28:27 +0000 To prevent total meltdown of the uncovered and overheated core, the reflooding with water is a necessary accident management measure. Because these actions lead to the generation of hydrogen, which can cause further problems, the related phenomena are investigated performing experiments and computer simulations. In this paper, for the experiments of loss of coolant accidents, performed in Forschungszentrum Karlsruhe, QUENCH-03 and QUENCH-06 are modelled using RELAP5/SCDAPSIM and ASTEC codes. The performed benchmark allowed analysing different modelling features. The recommendations for the model development are presented. Tadas Kaliatka, Algirdas Kaliatka, Virginijus Vileiniškis, and Eugenijus Ušpuras Copyright © 2014 Tadas Kaliatka et al. All rights reserved. The Dilution Dependency of Multigroup Uncertainties Tue, 02 Dec 2014 00:10:06 +0000 The propagation of nuclear data uncertainties through reactor physics calculation has received attention through the Organization for Economic Cooperation and Development—Nuclear Energy Agency’s Uncertainty Analysis in Modelling (UAM) benchmark. A common strategy for performing lattice physics uncertainty analysis involves starting with nuclear data and covariance matrix which is typically available at infinite dilution. To describe the uncertainty of all multigroup physics parameters—including those at finite dilution—additional calculations must be performed that relate uncertainties in an infinite dilution cross-section to those at the problem dilution. Two potential methods for propagating dilution-related uncertainties were studied in this work. The first assumed a correlation between continuous-energy and multigroup cross-sectional data and uncertainties, which is convenient for direct implementation in lattice physics codes. The second is based on a more rigorous approach involving the Monte Carlo sampling of resonance parameters in evaluated nuclear data using the TALYS software. When applied to a light water fuel cell, the two approaches show significant differences, indicating that the assumption of the first method did not capture the complexity of physics parameter data uncertainties. It was found that the covariance of problem-dilution multigroup parameters for selected neutron cross-sections can vary significantly from their infinite-dilution counterparts. M. R. Ball, C. McEwan, D. R. Novog, and J. C. Luxat Copyright © 2014 M. R. Ball et al. All rights reserved. Large Scale Gas Stratification Erosion by a Vertical Helium-Air Jet Sun, 30 Nov 2014 08:39:13 +0000 Containment conditions after certain postulated severe accident scenarios in nuclear power plants might result in the accumulation of hydrogen in the vessel dome. Inspired by these accident scenarios an experiment for the OECD/NEA benchmark exercise (2014) was carried out in the large scale PANDA facility at the Paul Scherrer Institut in Switzerland. The benchmark experiment was conducted at room temperature and under conditions characterized by an initially positively buoyant jet which becomes negatively buoyant while interacting with a helium layer. The experiment addresses (i) the initial conditions especially at the tube exit and (ii) the details of the entrainment of the helium stratification into the jet and the transport of the mixture towards the lower parts of the vessel. For the tube exit velocity mean and fluctuating quantities we find a reasonable agreement with pipe flow data, but a lack of agreement between past tube exit measurements and our results. It is shown that the axial velocity of the jet experiences a strong deceleration in the vicinity of the helium-rich layer and is finally stopped. Fluid accumulates in this zone and part of this fluid is flowing back in a narrow annular region around the upward flowing jet. Consequently, part of the annular flow is reentrained into the rising jet. During the layer erosion, the flow structure changes from a more downwards oriented annular type to a more horizontally oriented mushroom type of flow. It is found that locations for which we record considerable turbulent kinetic energy extends above the region where the velocity magnitude has decayed to almost zero, indicating that the jet deceleration and redirection introduces considerable turbulence in the helium stratification. R. Kapulla, G. Mignot, S. Paranjape, L. Ryan, and D. Paladino Copyright © 2014 R. Kapulla et al. All rights reserved. Heat Transfer Analysis of Passive Residual Heat Removal Heat Exchanger under Natural Convection Condition in Tank Thu, 20 Nov 2014 00:00:00 +0000 Aiming at the heat transfer calculation of the Passive Residual Heat Removal Heat Exchanger (PRHR HX), experiments on the heat transfer of C-shaped tube immerged in a water tank were performed. Comparisons of different correlation in literatures with the experimental data were carried out. It can be concluded that the Dittus-Boelter correlation provides a best-estimate fit with the experimental results. The average error is about 0.35%. For the tube outside, the McAdams correlations for both horizontal and vertical regions are best-estimated. The average errors are about 0.55% for horizontal region and about 3.28% for vertical region. The tank mixing characteristics were also investigated in present work. It can be concluded that the tank fluid rose gradually which leads to a thermal stratification phenomenon. Qiming Men, Xuesheng Wang, Xiang Zhou, and Xiangyu Meng Copyright © 2014 Qiming Men et al. All rights reserved. The Mass Attenuation Coefficients, Electronic, Atomic, and Molecular Cross Sections, Effective Atomic Numbers, and Electron Densities for Compounds of Some Biomedically Important Elements at 59.5 keV Wed, 12 Nov 2014 00:00:00 +0000 The mass attenuation coefficients for compounds of biomedically important some elements (Na, Mg, Al, Ca, and Fe) have been measured by using an extremely narrow collimated-beam transmission method in the energy 59.5 keV. Total electronic, atomic, and molecular cross sections, effective atomic numbers, and electron densities have been obtained by using these results. Gamma-rays of 241Am passed through compounds have been detected by a high-resolution Si(Li) detector and by using energy dispersive X-ray fluorescence spectrometer (EDXRF). Obtained results have been compared with theoretically calculated values of WinXCom and FFAST. The relative difference between the experimental and theoretical values are −9.4% to +11.9% with WinXCom and −11.8% to +11.7% FFAST. Results have been presented and discussed in this paper. Burcu Akça and Salih Z. Erzeneoğlu Copyright © 2014 Burcu Akça and Salih Z. Erzeneoğlu. All rights reserved. Validation of Effective Models for Simulation of Thermal Stratification and Mixing Induced by Steam Injection into a Large Pool of Water Thu, 16 Oct 2014 13:12:30 +0000 The Effective Heat Source (EHS) and Effective Momentum Source (EMS) models have been proposed to predict the development of thermal stratification and mixing during a steam injection into a large pool of water. These effective models are implemented in GOTHIC software and validated against the POOLEX STB-20 and STB-21 tests and the PPOOLEX MIX-01 test. First, the EHS model is validated against STB-20 test which shows the development of thermal stratification. Different numerical schemes and grid resolutions have been tested. A grid with second order scheme is sufficient to capture the vertical temperature distribution in the pool. Next, the EHS and EMS models are validated against STB-21 test. Effective momentum is estimated based on the water level oscillations in the blowdown pipe. An effective momentum selected within the experimental measurement uncertainty can reproduce the mixing details. Finally, the EHS-EMS models are validated against MIX-01 test which has improved space and time resolution of temperature measurements inside the blowdown pipe. Excellent agreement in averaged pool temperature and water level in the pool between the experiment and simulation has been achieved. The development of thermal stratification in the pool is also well captured in the simulation as well as the thermal behavior of the pool during the mixing phase. Hua Li, Walter Villanueva, Markku Puustinen, Jani Laine, and Pavel Kudinov Copyright © 2014 Hua Li et al. All rights reserved. A Numerical Analysis Research on Earlier Behavior of Molten Droplet Covered with Vapor Film at the Stage of Triggering and Propagation in Steam Explosion Sun, 12 Oct 2014 00:00:00 +0000 When the molten fuel with high temperature falls into the cavity water, it will be dispersed into droplets which are covered with vapor films due to the rapid heat transfer with phase transition. This situation cannot be simply described by liquid-liquid or gas-liquid systems. And there are no sufficient experimental studies on the behavior of droplet covered with vapor film because of the rapid reaction and the difficulty in capture of the film configuration. In this paper, a multiphase code with the volume of fluid (VOF) method is used to simulate the earlier behavior of droplet when vapor film exits. The earlier behavior is defined as behavior of the droplet before its disintegration. Thermal effect and pure hydrodynamic effect are, respectively, considered. The simulation results indicate that the film thickness and material density have significant effect on the earlier behavior of droplet. The situation assumed in Ciccarelli and Frost’s model (1994) is observed in current simulation of earlier thermal droplet behavior. The effect of triggering pressure pulse on earlier hydrodynamic behavior is also discussed and it indicates that vapor film has little effect on the hydrodynamic droplet deformation when the intensity of the pressure pulse is very high. Mingjun Zhong, Yankai Li, Meng Lin, Minghao Yuan, and Yanhua Yang Copyright © 2014 Mingjun Zhong et al. All rights reserved. Uncertainty Analysis of Method-Based Operating Event Groups Ranking Thu, 18 Sep 2014 00:00:00 +0000 Safe operation and industrial improvements are coming from the technology development and operational experience (OE) feedback. A long life span for many industrial facilities makes OE very important. Proper assessment and understanding of OE remains a challenge because of organization system relations, complexity, and number of OE events acquired. One way to improve OE events understanding is to focus their investigation and analyze in detail the most important. The OE ranking method is developed to select the most important events based on the basic event parameters and the analytical hierarchy process applied at the level of event groups. This paper investigates further how uncertainty in the model affects ranking results. An analysis was performed on the set of the two databases from the 20 years of nuclear power plants in France and Germany. From all uncertainties the presented analysis selected ranking indexes as the most relevant for consideration. Here the presented analysis of uncertainty clearly shows that considering uncertainty is important for all results, especially for event groups ranked closely and next to the most important one. Together with the previously performed sensitivity analysis, uncertainty assessment provides additional insights and a better judgment of the event groups’ importance in further detailed investigation. Zdenko Šimić, Benoit Zerger, and Reni Banov Copyright © 2014 Zdenko Šimić et al. All rights reserved. Analysis on Steady-State Operation and Heat Loss of Chinese Integrated Pressurized Water Reactor Mon, 01 Sep 2014 12:03:16 +0000 Chinese integrated pressurized water reactor (CIPWR) has compact configuration and high inherent safety, which is appropriate for nuclear power plants of small and medium scale. Heat balance model has been adopted widely in thermal power calibration of PWRs because of its advantage of accuracy. In this paper, a package based on FORTRAN language is developed and added into RELAP5 to calculate the heat loss value needed in heat balance analysis. The steady-state operation of CIPWR is modelled correctly by RELAP5. The heat loss of CIPWR is calculated by the package, and the comparison of the main values of parameters needed in the heat loss calculation between RELAP5 and the package has been done. It shows that the package has high calculation accuracy and can be applied in reactor design and monitoring. Zhang Fan, Lu Dao-Gang, Sui Dan-Ting, Guo Chao, and Yuan Bo Copyright © 2014 Zhang Fan et al. All rights reserved. Modelling of the Radiological Contamination of the RBMK-1500 Reactor Water Purification and Cooling System Sun, 31 Aug 2014 08:45:56 +0000 This paper presents modelling results on the RBMK-1500 reactor water purification and cooling system (PCS) components contamination at Ignalina NPP Unit 1. The modelling was performed using a computer code LLWAA-DECOM (Tractebel Energy Engineering, Belgium), taking into consideration PCS components characteristics, parameters of the water flowing in circuits, system work regimes, and so forth. During the modelling, results on activity of PCS subsystems and components’ deposits and nuclide composition of deposits at the moment of the final shutdown of the reactor, as well as activity decay of the most contaminated PCS components’ deposits and dose rates after the final shutdown of the reactor, were obtained. Significant difference of contamination levels was revealed among PCS subsystems and subsystems components. The subsystem of nonpurified water is the most contaminated in PCS, and the activity of the least contaminated component in this subsystem is only 1.42% compared to the activity of the most contaminated component. The most contaminated and the least contaminated components of the purified water subsystem comprise 28.33% and 0.86% of activity, respectively, compared to the activity of the most contaminated PCS component. G. Poskas, R. Zujus, P. Poskas, and G. Miliauskas Copyright © 2014 G. Poskas et al. All rights reserved. A Computing Approach with the Heat-Loss Model for the Transient Analysis of Liquid Metal Natural Circulation Loop Thu, 28 Aug 2014 11:33:55 +0000 The transient behaviors of natural circulation loop (NCL) are important for the system reliability under postulated accidents. The heat loss and structure thermal inertia may influence the transient behaviors of NCL greatly, so a transient analysis model with consideration of heat loss was developed based on the MATLAB/Simulink to predict the thermal-hydraulic characteristic of liquid metal NCL. The transient processes including the start-up, the loss of pump, and the shutdown of thermal-hydraulic ADS lead bismuth loop (TALL) experimental facility were simulated by using the model. A good agreement is obtained to validate the transient model. The appended structure would provide significant thermal inertia and flatten the temperature distribution in the transients. The oscillations of temperature and flow rate are also weakened. The temperature difference between hot leg and cold leg would increase with the decrease of heat loss, so the flow rate increases as well. However, a significant increase of hot section temperature may cause a failure of facility integrity due to the decrease of heat loss. Hence, the full power of the core tank may also be limited. Daogang Lu, Xun Zhang, and Chao Guo Copyright © 2014 Daogang Lu et al. All rights reserved. Development of a Remote Handling Robot for the Maintenance of an ITER-Like D-Shaped Vessel Wed, 27 Aug 2014 11:02:08 +0000 Robotic operation is one of the major challenges in the remote maintenance of ITER vacuum vessel (VV) and future fusion reactors as inner operations of Tokamak have to be done by robots due to the internal adverse conditions. This paper introduces a novel remote handling robot (RHR) for the maintenance of ITER-like D-shaped vessel. The modular designed RHR, which is an important part of the remote handling system for ITER, consists of three parts: an omnidirectional transfer vehicle (OTV), a planar articulated arm (PAA), and an articulated teleoperated manipulator (ATM). The task of RHR is to carry processing tools, such as the viewing system, leakage detector, and electric screwdriver, to inspect and maintain the components installed inside the D-shaped vessel. The kinematics of the OTV, as well as the kinematic analyses of the PAA and ATM, is studied in this paper. Because of its special length and heavy payload, the dynamics of the PAA is also investigated through a dynamic simulation system based on robot technology middleware (RTM). The results of the path planning, workspace simulations, and dynamic simulation indicate that the RHR has good mobility together with satisfying kinematic and dynamic performances and can well accomplish its maintenance tasks in the ITER-like D-shaped vessel. Peihua Chen and Qixin Cao Copyright © 2014 Peihua Chen and Qixin Cao. All rights reserved. Investigation of TASS/SMR Capability to Predict a Natural Circulation in the Test Facility for an Integral Reactor Thu, 21 Aug 2014 11:11:53 +0000 System-integrated modular advanced reactor (SMART) is a small-sized advanced integral type pressurized water reactor (PWR) with a rated thermal power of 330 MW. It can produce 100 MW of electricity or 90 MW of electricity and 40,000 ton of desalinated water concurrently, which is sufficient for 100,000 residents. The design features contributing to safety enhancement are basically inherent safety improvement and passive safety features. TASS/SMR code was developed for an analysis of design based events and accidents in an integral type reactor reflecting the characteristics of the SMART design. The main purpose of the code is to analyze all relevant phenomena and processes. The code should be validated using experimental data in order to confirm prediction capability. TASS/SMR predicts well the overall thermal-hydraulic behavior under various natural circulation conditions at the experimental test facility for an integral reactor. A pressure loss should be provided a function of Reynolds number at low velocity conditions in order to simulate the mass flow rate well under natural circulations. Young-Jong Chung, Sung-Won Lim, and Kyoo-Hwan Bae Copyright © 2014 Young-Jong Chung et al. All rights reserved. Supercritical Water-Cooled Reactors Mon, 18 Aug 2014 07:50:06 +0000 Jiejin Cai, Claude Renault, and Junli Gou Copyright © 2014 Jiejin Cai et al. All rights reserved. Impact of Thorium Capture Cross Section Uncertainty on the Thorium Utilized ADS Reactivity Calculation Sun, 17 Aug 2014 09:38:03 +0000 Recently, the researches on fast neutron spectrum system utilized thorium fuel are widely conducted. However, the recent thorium cross section libraries are limited compared to uranium cross section libraries. The impact of thorium cross section uncertainty on thorium fuel utilized accelerator driven system (ADS) reactivity calculation is estimated in this study. The uncertainty of the caused by 232Th capture cross section of JENDL-4.0 is about 1.3%. The uncertainty of JENDL-4.0 is needed to be enhanced to provide more reliable results on reactivity calculation for the fast system. The impact of uncertainty of  232Th capture cross section of ENDF/B-VII is small (0.1%). Therefore, it will cause no significant impact of the thorium cross section library on the thorium utilized ADS design calculation. Thanh Mai Vu and Takanori Kitada Copyright © 2014 Thanh Mai Vu and Takanori Kitada. All rights reserved. Natural Circulation Characteristics of a Symmetric Loop under Inclined Conditions Thu, 14 Aug 2014 06:52:08 +0000 Natural circulation is an important process for primary loops of some marine integrated reactors. The reactor works under inclined conditions when severe accidents happen to the ship. In this paper, to investigate the characteristics of natural circulation, experiments were conducted in a symmetric loop under the inclined angle of 0~45°. A CFD model was also set up to predict the behaviors of the loop beyond the experimental scope. Total circulation flow rate decreases with the increase of inclined angle. Meanwhile one circulation is depressed while the other is enhanced, and accordingly the disparity between the branch circulations arises and increases with the increase of inclined angle. Circulation only takes place in one branch circuit at large inclined angle. Also based on the CFD model, the influences of flow resistance distribution and loop configuration on natural circulation are predicted. The numerical results show that to design the loop with the configuration of big altitude difference and small width, it is favorable to reduce the influence of inclination; however too small loop width will cause severe reduction of circulation ability at large angle inclination. Xingtuan Yang, Yanfei Sun, Zhiyong Liu, and Shengyao Jiang Copyright © 2014 Xingtuan Yang et al. All rights reserved. Compilation of Existing Neutron Screen Technology Mon, 11 Aug 2014 00:00:00 +0000 The presence of fast neutron spectra in new reactors is expected to induce a strong impact on the contained materials, including structural materials, nuclear fuels, neutron reflecting materials, and tritium breeding materials. Therefore, introduction of these reactors into operation will require extensive testing of their components, which must be performed under neutronic conditions representative of those expected to prevail inside the reactor cores when in operation. Due to limited availability of fast reactors, testing of future reactor materials will mostly take place in water cooled material test reactors (MTRs) by tailoring the neutron spectrum via neutron screens. The latter rely on the utilization of materials capable of absorbing neutrons at specific energy. A large but fragmented experience is available on that topic. In this work a comprehensive compilation of the existing neutron screen technology is attempted, focusing on neutron screens developed in order to locally enhance the fast over thermal neutron flux ratio in a reactor core. N. Chrysanthopoulou, P. Savva, M. Varvayanni, and N. Catsaros Copyright © 2014 N. Chrysanthopoulou et al. All rights reserved. An Evaluation of SMR Economic Attractiveness Tue, 05 Aug 2014 05:25:09 +0000 The nuclear “renaissance” that is taking place worldwide concerns the new build of GW size reactor plants, but smaller GenIII+ NPP (Small Modular Reactors, SMR) are on the verge to be commercially available and are raising increasing public interest. These reactor concepts rely on the pressurized water technology, capitalizing on thousands of reactor-years operations and enhancing the passive safety features, thanks to the smaller plant and equipment size. On the other hand, smaller plant size pays a loss of economy of scale, which might have a relevant impact on the generation costs of electricity, given the capital-intensive nature of nuclear power technology. The paper explores the economic advantages/disadvantages of multiple SMR compared to alternative large plants of the same technology and equivalent total power installed. The metrics used in the evaluation is twofold, as appropriate for liberalized markets of capital and electricity: investment profitability and investment risk are assessed, from the point of view of the plant owner. Results show that multiple SMR deployed on the same site may prove competitive with investment returns of larger plants, while offering, in addition, unique features that mitigate the investment risk. Sara Boarin and Marco E. Ricotti Copyright © 2014 Sara Boarin and Marco E. Ricotti. All rights reserved.