Science and Technology of Nuclear Installations The latest articles from Hindawi Publishing Corporation © 2014 , Hindawi Publishing Corporation . 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. Experimental and Numerical Study on Pressure Distribution of 90° Elbow for Flow Measurement Mon, 04 Aug 2014 11:18:04 +0000 Numerical simulation is performed to investigate the pressure distribution of helium gas under high pressure and high temperature for 10 MW High Temperature Gas-Cooled Reactor (HTGR-10). Experimental studies are first conducted on a self-built test system to investigate the static pressure distribution of a 90° elbow and validate the credibility of the computational approach. The 90° elbow is designed and manufactured geometrically the same as HTGR-10. Based on the experimental data, comparison of static pressure of inner wall and outer wall of 90° elbow with numerical results is carried out to verify the numerical approach. With high agreement between experimental results and numerical results of water flowing through 90° elbow, flow characteristics of helium gas under high pressure and high temperature are investigated on the confirmed numerical approach for flow measurement. And wall pressure distribution of eight cross sections of 90° elbow is given in detail to represent the entire region of the elbow. Beibei Feng, Shiming Wang, Shengqiang Li, Xingtuan Yang, and Shengyao Jiang Copyright © 2014 Beibei Feng et al. All rights reserved. Lessons Learned from the Fukushima Accident: An Integrated Perspective Wed, 23 Jul 2014 05:59:03 +0000 Inn Seock Kim, Akira Omoto, Enrico Zio, Joon-Eon Yang, and Yanko Yanev Copyright © 2014 Inn Seock Kim et al. All rights reserved. Subchannel Analysis, CFD Modeling and Verifications, CHF Experiments and Benchmarking Sun, 20 Jul 2014 06:52:22 +0000 Baowen Yang, Yassin A. Hassan, Jianqiang Shan, Bin Zhang, Junli Gou, and Liangzhi Cao Copyright © 2014 Baowen Yang et al. All rights reserved. Approach and Development of Effective Models for Simulation of Thermal Stratification and Mixing Induced by Steam Injection into a Large Pool of Water Thu, 17 Jul 2014 09:22:44 +0000 Steam venting and condensation in a large pool of water can lead to either thermal stratification or thermal mixing. In a pressure suppression pool (PSP) of a boiling water reactor (BWR), consistent thermal mixing maximizes the capacity of the pool while the development of thermal stratification can reduce the steam condensation capacity of the pool which in turn can lead to pressure increase in the containment and thereafter the consequences can be severe. Advanced modeling and simulation of direct contact condensation in large systems remain a challenge as evident in commercial and research codes mainly due to small time-steps necessary to resolve contact condensation in long transients. In this work, effective models, namely, the effective heat source (EHS) and effective momentum source (EMS) models, are proposed to model and simulate thermal stratification and mixing during a steam injection into a large pool of water. Specifically, the EHS/EMS models are developed for steam injection through a single vertical pipe submerged in a pool under two condensation regimes: complete condensation inside the pipe and chugging. These models are computationally efficient since small scale behaviors are not resolved but their integral effect on the large scale flow structure in the pool is taken into account. Hua Li, Walter Villanueva, and Pavel Kudinov Copyright © 2014 Hua Li et al. All rights reserved. Procedure of Active Residual Heat Removal after Emergency Shutdown of High-Temperature-Gas-Cooled Reactor Mon, 07 Jul 2014 06:57:27 +0000 After emergency shutdown of high-temperature-gas-cooled reactor, the residual heat of the reactor core should be removed. As the natural circulation process spends too long period of time to be utilized, an active residual heat removal procedure is needed, which makes use of steam generator and start-up loop. During this procedure, the structure of steam generator may suffer cold/heat shock because of the sudden load of coolant or hot helium at the first few minutes. Transient analysis was carried out based on a one-dimensional mathematical model for steam generator and steam pipe of start-up loop to achieve safety and reliability. The results show that steam generator should be discharged and precooled; otherwise, boiling will arise and introduce a cold shock to the boiling tubes and tube sheet when coolant began to circulate prior to the helium. Additionally, in avoiding heat shock caused by the sudden load of helium, the helium circulation should be restricted to start with an extreme low flow rate; meanwhile, the coolant of steam generator (water) should have flow rate as large as possible. Finally, a four-step procedure with precooling process of steam generator was recommended; sensitive study for the main parameters was conducted. Xingtuan Yang, Yanfei Sun, Huaiming Ju, and Shengyao Jiang Copyright © 2014 Xingtuan Yang et al. All rights reserved. Nuclear Power Plants Safety and Maintenance Sun, 06 Jul 2014 06:55:29 +0000 Wael H. Ahmed, Atef Mohany, and Bing Li Copyright © 2014 Wael H. Ahmed et al. All rights reserved. Validation of NEPTUNE-CFD Two-Phase Flow Models Using Experimental Data Mon, 30 Jun 2014 09:33:17 +0000 This paper deals with the validation of the two-phase flow models of the CFD code NEPTUNEC-CFD using experimental data provided by the OECD BWR BFBT and PSBT Benchmark. Since the two-phase models of CFD codes are extensively being improved, the validation is a key step for the acceptability of such codes. The validation work is performed in the frame of the European NURISP Project and it was focused on the steady state and transient void fraction tests. The influence of different NEPTUNE-CFD model parameters on the void fraction prediction is investigated and discussed in detail. Due to the coupling of heat conduction solver SYRTHES with NEPTUNE-CFD, the description of the coupled fluid dynamics and heat transfer between the fuel rod and the fluid is improved significantly. The averaged void fraction predicted by NEPTUNE-CFD for selected PSBT and BFBT tests is in good agreement with the experimental data. Finally, areas for future improvements of the NEPTUNE-CFD code were identified, too. Jorge Pérez Mañes, Victor Hugo Sánchez Espinoza, Sergio Chiva Vicent, Michael Böttcher, and Robert Stieglitz Copyright © 2014 Jorge Pérez Mañes et al. All rights reserved. CFD Turbulence Study of PWR Spacer-Grids in a Rod Bundle Mon, 30 Jun 2014 07:01:01 +0000 Nuclear fuel bundles include spacers essentially for mechanical stability and to influence the flow dynamics and heat transfer phenomena along the fuel rods. This work presents the analysis of the turbulence effects of a split-type and swirl-type spacer-grid geometries on single phase in a PWR (pressurized water reactor) rod bundle. Various computational fluid dynamics (CFD) calculations have been performed and the results validated with the experiments of the OECD/NEA-KAERI rod bundle CFD blind benchmark exercise on turbulent mixing in a rod bundle with spacers at the MATiS-H facility. Simulation of turbulent phenomena downstream of the spacer-grid presents high complexity issues; a wide range of length scales are present in the domain increasing the difficulty of defining in detail the transient nature of turbulent flow with ordinary turbulence models. This paper contains a complete description of the procedure to obtain a validated CFD model for the simulation of the spacer-grids. Calculations were performed with the commercial code ANSYS CFX using large eddy simulation (LES) turbulence model and the CFD modeling procedure validated by comparison with measurements to determine their suitability in the prediction of the turbulence phenomena. C. Peña-Monferrer, J. L. Muñoz-Cobo, and S. Chiva Copyright © 2014 C. Peña-Monferrer et al. All rights reserved. Design of the VISTA-ITL Test Facility for an Integral Type Reactor of SMART and a Post-Test Simulation of a SBLOCA Test Thu, 26 Jun 2014 11:15:53 +0000 To validate the performance and safety of an integral type reactor of SMART, a thermal-hydraulic integral effect test facility, VISTA-ITL, is introduced with a discussion of its scientific design characteristics. The VISTA-ITL was used extensively to assess the safety and performance of the SMART design, especially for its passive safety system such as a passive residual heat removal system, and to validate various thermal-hydraulic analysis codes. The VISTA-ITL program includes several tests on the SBLOCA, CLOF, and PRHRS performances to support a verification of the SMART design and contribute to the SMART design licensing by providing proper test data for validating the system analysis codes. A typical scenario of SBLOCA was analyzed using the MARS-KS code to assess the thermal-hydraulic similarity between the SMART design and the VISTA-ITL facility, and a posttest simulation on a SBLOCA test for the shutdown cooling system line break has been performed with the MARS-KS code to assess its simulation capability for the SBLOCA scenario of the SMART design. The SBLOCA scenario in the SMART design was well reproduced using the VISTA-ITL facility, and the measured thermal-hydraulic data were properly simulated with the MARS-KS code. Hyun-Sik Park, Byung-Yeon Min, Youn-Gyu Jung, Yong-Cheol Shin, Yung-Joo Ko, and Sung-Jae Yi Copyright © 2014 Hyun-Sik Park et al. All rights reserved. Pore Scale Thermal Hydraulics Investigations of Molten Salt Cooled Pebble Bed High Temperature Reactor with BCC and FCC Configurations Mon, 23 Jun 2014 08:43:40 +0000 The present paper systematically investigated pore scale thermal hydraulics characteristics of molten salt cooled high temperature pebble bed reactor. By using computational fluid dynamics (CFD) methods and employing simplified body center cubic (BCC) and face center cubic (FCC) model, pressure drop and local mean Nusselt number are calculated. The simulation result shows that the high Prandtl number molten salt in packed bed has unique fluid-dynamics and thermodynamic properties. There are divergences between CFD results and empirical correlations’ predictions of pressure drop and local Nusselt numbers. Local pebble surface temperature distributions in several default conditions are investigated. Thermal removal capacities of molten salt are confirmed in the case of nominal condition; the pebble surface temperature under the condition of local power distortion shows the tolerance of pebble in extreme neutron dose exposure. The numerical experiments of local pebble insufficient cooling indicate that in the molten salt cooled pebble bed reactor, the pebble surface temperature is not very sensitive to loss of partial coolant. The methods and results of this paper would be useful for optimum designs and safety analysis of molten salt cooled pebble bed reactors. Shixiong Song, Xiangzhou Cai, Yafen Liu, Quan Wei, and Wei Guo Copyright © 2014 Shixiong Song et al. All rights reserved. Containment Depressurization Capabilities of Filtered Venting System in 1000 MWe PWR with Large Dry Containment Sun, 22 Jun 2014 12:50:59 +0000 After the Fukushima Daiichi nuclear power plant accident, the Korean government and nuclear industries performed comprehensive safety inspections on all domestic nuclear power plants against beyond design bases events. As a result, a total of 50 recommendations were defined as safety improvement action items. One of them is installation of a containment filtered venting system (CFVS) or portable backup containment spray system. In this paper, the applicability of CFVS is examined for OPR1000, a 1000 MWe PWR with large dry containment in Korea. Thermohydraulic analysis results show that a filtered discharge flow rate of 15 [kg/s] at 0.9 [MPa] is sufficient to depressurize the containment against representative containment overpressurization scenarios. Radiological release to the environment is reduced to considering the decontamination factor. Also, this cyclic venting strategy reduces noble gas release by 50% for 7 days. The probability of maintaining the containment integrity in level 2 probabilistic safety assessment (PSA) initiating events is improved twofold, from 43% to 87%. So, the CFVS can further improve the containment integrity in severe accident conditions. Sang-Won Lee, Tae-Hyub Hong, Yu-Jung Choi, Mi-Ro Seo, and Hyeong-Taek Kim Copyright © 2014 Sang-Won Lee et al. All rights reserved. Development of Instrument Transmitter Protecting Device against High-Temperature Condition during Severe Accidents Tue, 17 Jun 2014 06:23:48 +0000 Reliable information through instrumentation systems is essential in mitigating severe accidents such as the one that occurred at the Fukushima Daiichi nuclear power plant. There are five elements which might pose a potential threat to the reliability of parameter detection at nuclear power plants during a severe accident: high temperature, high pressure, high humidity, high radiation, and missiles generated during the evolution of a severe accident. Of these, high temperature apparently poses the most serious threat, since thin shielding can get rid of pressure, humidity, radiation (specifically, alpha and beta radiations), and missile effects. In view of this fact, our study focused on designing an instrument transmitter protecting device that can eliminate the high-temperature effect on transmitters to maintain their functional integrity. We present herein a novel concept for designing such a device in terms of heat transfer model that takes into account various heat transfer mechanisms associated with the device. Min Yoo, Sung Min Shin, and Hyun Gook Kang Copyright © 2014 Min Yoo et al. All rights reserved. Insights on Accident Information and System Operations during Fukushima Events Thu, 05 Jun 2014 11:37:17 +0000 As part of the development of an integrated perspective on lessons learned from the Fukushima Daiichi nuclear accident, this paper highlights lessons learned and implications relating to the accident information and system operational aspects during the events. Our analysis clearly indicates that the plant was neither designed nor prepared to withstand such an unexpected event, which included a complete loss of electrical power sources for a long period. The author focused on the accident information and system operational aspects of the Fukushima event, including lack of information, provision of wrong information, operator performance in life-threatening environments, and improvisation given lack of procedures and training. Suggestions for further improvement of the nuclear plant safety are then made with respect to preparation for beyond design basis events, provision of reliable essential information to operators, development of guidelines/procedures, training of operators, and development of operator support systems with consideration of severe accidents caused by unexpected events. It is hoped that the lessons learned from the accident will significantly contribute to the enhancement of nuclear plant safety. Man Cheol Kim Copyright © 2014 Man Cheol Kim. All rights reserved. Experimental Research on Passive Residual Heat Removal System of Chinese Advanced PWR Thu, 05 Jun 2014 07:54:51 +0000 Passive residual heat removal system (PRHRS) for the secondary loop is one of the important features for Chinese advance pressurized water reactor (CAPWR). To prove the safety characteristics of CAPWR, serials of experiments have been done on special designed PRHRS test facility in the former stage. The test facility was built up following the scaling laws to preserve the similarity to CAPWR. A total of more than 300 tests have been performed on the test facility, including 90% steady state cases and 10% transient cases. A semiempirical model was generated for passive heat removal functions based on the experimental results of steady state cases. The dynamic capability characteristics and reliability of passive safety system for CAPWR were evidently proved by transient cases. A new simulation code, MISAP2.0, has been developed and calibrated by experimental results. It will be applied in future design evaluation and optimization works. Zhuo Wenbin, Huang Yanping, Xiao Zejun, Peng Chuanxin, and Lu Sansan Copyright © 2014 Zhuo Wenbin et al. All rights reserved. Some Movement Mechanisms and Characteristics in Pebble Bed Reactor Sun, 01 Jun 2014 08:17:52 +0000 The pebblebed-type high temperature gas-cooled reactor is considered to be one of the promising solutions for generation IV advanced reactors, and the two-region arranged reactor core can enhance its advantages by flattening neutron flux. However, this application is held back by the existence of mixing zone between central and peripheral regions, which results from pebbles’ dispersion motions. In this study, experiments have been carried out to study the dispersion phenomenon, and the variation of dispersion region and radial distribution of pebbles in the specifically shaped flow field are shown. Most importantly, the standard deviation of pebbles’ radial positions in dispersion region, as a quantitative index to describe the size of dispersion region, is gotten through statistical analysis. Besides, discrete element method has been utilized to analyze the parameter influence on dispersion region, and this practice offers some strategies to eliminate or reduce mixing zone in practical reactors. Xingtuan Yang, Yu Li, Nan Gui, Xinlong Jia, Jiyuan Tu, and Shengyao Jiang Copyright © 2014 Xingtuan Yang et al. All rights reserved. Estimation of Intervention Distances for Urgent Protective Actions Using Comparative Approach of MACCS and InterRAS Wed, 28 May 2014 11:02:19 +0000 Distances for taking evacuation as a protective measure during early phase of a nuclear accident have been approximated using MELCOR Accident Consequence Code System (MACCS). As a reference data, the source term of Pakistan Research Reactor 1 (PARR-1) and meteorological data of Islamabad, Pakistan, have been considered. Based on comparison with published data and international radiological assessment (InterRAS) code results, it is concluded that MACCS is a rational tool for estimation of urgent protective actions during early phase of nuclear accident by taking into account the variations in meteorological and release concentrations parameters. Mazzammal Hussain, Salah Ud-Din Khan, Waqar A. Adil Syed, and Shahab Ud-Din Khan Copyright © 2014 Mazzammal Hussain et al. All rights reserved. Uniform versus Nonuniform Axial Power Distribution in Rod Bundle CHF Experiments Tue, 27 May 2014 06:20:03 +0000 Rod bundle experiments with axially uniform and nonuniform heat fluxes are examined to explore the potential limitations of using uniform rod bundle CHF data for CHF correlation development of light water reactors with nonuniform axial power distribution (APD). The case of upstream burnout is presented as an example of unique phenomena associated with nonuniform rod bundle CHF experiments. It is a result from combined effect of axial nonuniform power shape and different interchannel mixing mechanisms. In addition, several key parameters are investigated with respect to their potential impacts on the thermal-hydraulic behaviors between rod bundles with uniform and nonuniform APDs. This type of misrepresentation cannot be amended or compensated through the use of correction factors due to the lack of critical information in the uniform rod bundle CHF testing as well as the fundamental difference in the underlining driving mechanisms. Other potential issues involved with the use of uniform rod bundle CHF data for nonuniform APD system applications also present strong evidence concerning the limitations and inadequacy of using uniform rod bundle CHF data for the correlation, prediction, and design limit calculation for safety analysis. Baowen Yang, Jianqiang Shan, Junli Gou, Hui Zhang, Aiguo Liu, and Hu Mao Copyright © 2014 Baowen Yang et al. All rights reserved.