Science and Technology of Nuclear Installations The latest articles from Hindawi Publishing Corporation © 2014 , Hindawi Publishing Corporation . 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. Extended Station Blackout Coping Capabilities of APR1400 Sun, 25 May 2014 06:01:00 +0000 The Fukushima Dai-ichi nuclear power plant accident shows that an extreme natural disaster can prevent the proper restoration of electric power for several days, so-called extended SBO. In Korea, the government and industry performed comprehensive special safety inspections on all domestic nuclear power plants against beyond design bases external events. One of the safety improvement action items related to the extended SBO is installation of external water injection provision and equipment to RCS and SG. In this paper, the extended SBO coping capability of APR1400 is examined using MAAP4 to assess the effectiveness of the external water injection strategy. Results show that an external injection into SG is applicable to mitigate an extended SBO scenario. However, an external injection into RCS is only effective when RCS depressurization capacity is sufficiently provided in case of high pressure scenarios. Based on the above results, the technical basis of external injection strategy will be reflected on development of revised severe accident management guideline. Sang-Won Lee, Tae Hyub Hong, Mi-Ro Seo, Young-Seung Lee, and Hyeong-Taek Kim Copyright © 2014 Sang-Won Lee et al. All rights reserved. A New Coupled CFD/Neutron Kinetics System for High Fidelity Simulations of LWR Core Phenomena: Proof of Concept Wed, 21 May 2014 07:45:43 +0000 The Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is investigating the application of the meso- and microscale analysis for the prediction of local safety parameters for light water reactors (LWR). By applying codes like CFD (computational fluid dynamics) and SP3 (simplified transport) reactor dynamics it is possible to describe the underlying phenomena in a more accurate manner than by the nodal/coarse 1D thermal hydraulic coupled codes. By coupling the transport (SP3) based neutron kinetics (NK) code DYN3D with NEPTUNE-CFD, within a parallel MPI-environment, the NHESDYN platform is created. The newly developed system will allow high fidelity simulations of LWR fuel assemblies and cores. In NHESDYN, a heat conduction solver, SYRTHES, is coupled to NEPTUNE-CFD. The driver module of NHESDYN controls the sequence of execution of the solvers as well as the communication between the solvers based on MPI. In this paper, the main features of NHESDYN are discussed and the proof of the concept is done by solving a single pin problem. The prediction capability of NHESDYN is demonstrated by a code-to-code comparison with the DYNSUB code. Finally, the future developments and validation efforts are highlighted. Jorge Pérez Mañes, Victor Hugo Sánchez Espinoza, Sergio Chiva, and Robert Stieglitz Copyright © 2014 Jorge Pérez Mañes et al. All rights reserved. Interaction of Radiation with Matter and Related Topics Wed, 21 May 2014 05:44:19 +0000 Jakrapong Kaewkhao, Mitra Djamal, and Turgay Korkut Copyright © 2014 Jakrapong Kaewkhao et al. All rights reserved. Experimental and Numerical Study of Stagnant Zones in Pebble Bed Sun, 18 May 2014 10:26:39 +0000 The experimental method (side area method) and DEM simulation have been carried out to analyse the stagnant zone in the quasi-two-dimensional silos. The side area method is a phenomenological method by means of investigating the interface features of different areas composed of different coloured pebbles. Two methods have been discussed to define the stagnant zone. In particular, the area of the stagnant zone has been calculated with the mean-streamline method, and the tracking time of different marking pebbles has been investigated with the stagnant time method to explore the kinematics characteristics of the pebbles. The stagnant zone is crucial for the safety of the pebble-bed reactor, and the practical reactor core must avoid the existence of the stagnant zone. Furthermore, this paper also analyses the effects of bed configuration (the bed height, the base angle, and the friction coefficient) on stagnant zone with the two methods mentioned above. In detail, the bed height shows little impact on the stagnant zones when the bed height exceeds a certain limit, while the base angle has negative prominent correlation with the stagnant zone. The friction coefficient effect seems complicated and presents the great nonlinearity, which deserves to be deeply investigated. Xinlong Jia, Xingtuan Yang, Nan Gui, Yu Li, Jiyuan Tu, and Shengyao Jiang Copyright © 2014 Xinlong Jia et al. All rights reserved. A Critical Heat Generation for Safe Nuclear Fuels after a LOCA Sun, 18 May 2014 07:46:11 +0000 This study applies a thermo-elasto-plastic-creep finite element procedure to the analysis of an accidental behavior of nuclear fuel as well as normal behavior. The result will be used as basic data for the robust design of nuclear power plant and fuels. We extended the range of mechanical strain from small or medium to large adopting the Hencky logarithmic strain measure in addition to the Green-Lagrange strain and Almansi strain measures, for the possible large strain situation in accidental environments. We found that there is a critical heat generation after LOCA without ECCS (event category 5), under which the cladding of fuel sustains the internal pressure and temperature for the time being for the rescue of the power plant. With the heat generation above the critical value caused by malfunctioning of the control rods, the stiffness of cladding becomes zero due to the softening by high temperature. The weak position of cladding along the length continuously bulges radially to burst and to discharge radioactive substances. This kind of cases should be avoid by any means. Jae-Yong Kim, Dong-Bock Kim, Hee-Jeong Cho, Soon-Bum Kwon, and Young-Doo Kwon Copyright © 2014 Jae-Yong Kim et al. All rights reserved. Core Flow Distribution from Coupled Supercritical Water Reactor Analysis Thu, 15 May 2014 13:41:42 +0000 This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal. Po Hu and Paul P. H. Wilson Copyright © 2014 Po Hu and Paul P. H. Wilson. All rights reserved. Study on Nuclear Accident Precursors Using AHP and BBN Wed, 14 May 2014 13:36:05 +0000 Most of the nuclear accident reports used to indicate the implicit precursors which are not easily quantified as underlying factors. The current Probabilistic Safety Assessment (PSA) is capable of quantifying the importance of accident causes in limited scope. It was, therefore, difficult to achieve quantifiable decision-making for resource allocation. In this study, the methodology which facilitates quantifying these precursors and a case study were presented. First, four implicit precursors have been obtained by evaluating the causality and hierarchy structure of various accident factors. Eventually, it turned out that they represent the lack of knowledge. After four precursors are selected, subprecursors were investigated and their cause-consequence relationship was implemented by Bayesian Belief Network (BBN). To prioritize the precursors, the prior probability is initially estimated by expert judgment and updated upon observations. The pair-wise importance between precursors is calculated by Analytic Hierarchy Process (AHP) and the results are converted into node probability tables of the BBN model. Using this method, the sensitivity and the posterior probability of each precursor can be analyzed so that it enables making prioritization for the factors. We tried to prioritize the lessons learned from Fukushima accident to demonstrate the feasibility of the proposed methodology. Sujin Park, Huichang Yang, Gyunyoung Heo, Muhammad Zubair, and Rahman Khalil Ur Copyright © 2014 Sujin Park et al. All rights reserved. Code Development in Coupled PARCS/RELAP5 for Supercritical Water Reactor Thu, 08 May 2014 08:21:54 +0000 The new capability is added to the existing coupled code package PARCS/RELAP5, in order to analyze SCWR design under supercritical pressure with the separated water coolant and moderator channels. This expansion is carried out on both codes. In PARCS, modification is focused on extending the water property tables to supercritical pressure, modifying the variable mapping input file and related code module for processing thermal-hydraulic information from separated coolant/moderator channels, and modifying neutronics feedback module to deal with the separated coolant/moderator channels. In RELAP5, modification is focused on incorporating more accurate water properties near SCWR operation/transient pressure and temperature in the code. Confirming tests of the modifications is presented and the major analyzing results from the extended codes package are summarized. Po Hu and Paul Wilson Copyright © 2014 Po Hu and Paul Wilson. All rights reserved.