Science and Technology of Nuclear Installations The latest articles from Hindawi Publishing Corporation © 2016 , Hindawi Publishing Corporation . All rights reserved. Development and Validation of a Three-Dimensional Diffusion Code Based on a High Order Nodal Expansion Method for Hexagonal- Geometry Mon, 17 Oct 2016 09:13:41 +0000 A three-dimensional, multigroup, diffusion code based on a high order nodal expansion method for hexagonal- geometry (HNHEX) was developed to perform the neutronic analysis of hexagonal- geometry. In this method, one-dimensional radial and axial spatially flux of each node and energy group are defined as quadratic polynomial expansion and four-order polynomial expansion, respectively. The approximations for one-dimensional radial and axial spatially flux both have second-order accuracy. Moment weighting is used to obtain high order expansion coefficients of the polynomials of one-dimensional radial and axial spatially flux. The partially integrated radial and axial leakages are both approximated by the quadratic polynomial. The coarse-mesh rebalance method with the asymptotic source extrapolation is applied to accelerate the calculation. This code is used for calculation of effective multiplication factor, neutron flux distribution, and power distribution. The numerical calculation in this paper for three-dimensional SNR and VVER 440 benchmark problems demonstrates the accuracy of the code. In addition, the results show that the accuracy of the code is improved by applying quadratic approximation for partially integrated axial leakage and four-order approximation for one-dimensional axial spatially flux in comparison to flat approximation for partially integrated axial leakage and quadratic approximation for one-dimensional axial spatially flux. Daogang Lu and Chao Guo Copyright © 2016 Daogang Lu and Chao Guo. All rights reserved. Nonuniform Oxidation on the Surface of Fuel Element in HTR Tue, 11 Oct 2016 06:43:42 +0000 The graphite oxidation of fuel element has obtained high attention in air ingress accident analysis of high temperature gas-cooled reactor (HTR). The shape function, defined as the relationship between the maximum and the average of the oxidation, is an important factor to estimate the consequence of the accident. There are no detailed studies on the shape function currently except two experiments several decades ago. With the development of computer technology, CFD method is used in the numerical experiment about graphite oxidation in pebble bed of HTR in this paper. Structured packed beds are used in the calculation instead of random packed beds. The result shows the nonuniform distribution of oxidation on the sphere surface and the shape function in the condition of air ingress accident. Furthermore, the sensitive factors of shape function, such as temperature and Re number, are discussed in detail and the relationship between the shape function and sensitive factors is explained. According to the results in this paper, the shape function ranges from 1.05 to 4.7 under the condition of temperature varying from 600°C to 1200°C and Re varying from 16 to 1600. Peng Liu, Yanhua Zheng, Wei Xu, and Lei Shi Copyright © 2016 Peng Liu et al. All rights reserved. Nondestructive Evaluation of Functionally Graded Subsurface Damage on Cylinders in Nuclear Installations Based on Circumferential SH Waves Thu, 29 Sep 2016 13:01:14 +0000 Subsurface damage could affect the service life of structures. In nuclear engineering, nondestructive evaluation and detection of the evaluation of the subsurface damage region are of great importance to ensure the safety of nuclear installations. In this paper, we propose the use of circumferential horizontal shear (SH) waves to detect mechanical properties of subsurface regions of damage on cylindrical structures. The regions of surface damage are considered to be functionally graded material (FGM) and the cylinder is considered to be a layered structure. The Bessel functions and the power series technique are employed to solve the governing equations. By analyzing the SH waves in the 12Cr-ODS ferritic steel cylinder, which is frequently applied in the nuclear installations, we discuss the relationship between the phase velocities of SH waves in the cylinder with subsurface layers of damage and the mechanical properties of the subsurface damaged regions. The results show that the subsurface damage could lead to decrease of the SH waves’ phase velocity. The gradient parameters, which represent the degree of subsurface damage, can be evaluated by the variation of the SH waves’ phase velocity. Research results of this study can provide theoretical guidance in nondestructive evaluation for use in the analysis of the reliability and durability of nuclear installations. Zhen Qu, Xiaoqin Shen, and Xiaoshan Cao Copyright © 2016 Zhen Qu et al. All rights reserved. Fabrication of U-10 wt.%Zr Metallic Fuel Rodlets for Irradiation Test in BOR-60 Fast Reactor Tue, 27 Sep 2016 13:24:00 +0000 The fabrication technology for metallic fuel has been developed to produce the driver fuel in a PGSFR in Korea since 2007. In order to evaluate the irradiation integrity and validate the in-reactor of the starting metallic fuel with FMS cladding for the loading of the metallic fuel, U-10 wt.%Zr fuel rodlets were fabricated and evaluated for a verification of the starting driver fuel through an irradiation test in the BOR-60 fast reactor. The injection casting method was applied to U-10 wt.%Zr fuel slugs with a diameter of 5.5 mm. Consequently, fuel slugs per melting batch without casting defects were fabricated through the development of advanced casting technology and evaluation tests. The optimal GTAW welding conditions were also established through a number of experiments. In addition, a qualification test was carried out to prove the weld quality of the end plug welding of the metallic fuel rodlets. The wire wrapping of metallic fuel rodlets was successfully accomplished for the irradiation test. Thus, PGSFR fuel rodlets have been soundly fabricated for the irradiation test in a BOR-60 fast reactor. Ki-Hwan Kim, Jong-Hwan Kim, Seok-Jin Oh, Jung-Won Lee, Ho-Jin Lee, and Chan-Bock Lee Copyright © 2016 Ki-Hwan Kim et al. All rights reserved. Numerical Investigation on Bubble Growth and Sliding Process of Subcooled Flow Boiling in Narrow Rectangular Channel Tue, 27 Sep 2016 09:54:04 +0000 In order to investigate single bubble evolution, a boiling phase change model in subcooled flow boiling is proposed in this paper, and VOF model combined with phase change model is adopted to simulate the single bubble growth and movement. The effects of flow velocity, liquid subcooling, wall superheat, and vapor-liquid contact angle are considered in this model. The predicted bubble growth curve agrees well with the experimental result. Based on the analysis of bubble shape evolution and temperature field, it is found that the average bubble growth rate, flow velocity, and dynamic contact angle have significant effect on the bubble shape evolution during the bubble growth and movement while the temperature gradient in superheated liquid does not change with bubble growing. The character of dynamic contact angle during bubble growth and movement is also obtained in different working condition. De-wen Yuan, Zejun Xiao, Deqi Chen, Yunke Zhong, Xiao Yan, Jianjun Xu, and Yanping Huang Copyright © 2016 De-wen Yuan et al. All rights reserved. Time-Dependent Neutronic Analysis of a Power-Flattened Gas Cooled Accelerator Driven System Fuelled with Thorium, Uranium, Plutonium, and Curium Dioxides TRISO Particles Wed, 21 Sep 2016 10:06:19 +0000 This study presents the power flattening and time-dependent neutronic analysis of a conceptual helium gas cooled Accelerator Driven System (ADS) loaded with TRISO (tristructural-isotropic) fuel particles. Target material is lead-bismuth eutectic (LBE). ThO2, UO2, PuO2, and CmO2 TRISO particles are used as fuel. PuO2 and CmO2 fuels are extracted from PWR-MOX spent fuel. Subcritical core is radially divided into 10 equidistant subzones in order to flatten the power produced in the core. Tens of thousands of these TRISO fuel particles are embedded in the carbon matrix fuel pebbles as five different cases. The high-energy Monte Carlo code MCNPX 2.7 with the LA150 library is used for the neutronic calculations. Time-dependent burnup calculations are carried out for thermal fission power () of 1000 MW using the BURN card. The energy gain of the ADS is in the range of 99.98–148.64 at the beginning of a cycle. Furthermore, the peak-to-average fission power density ratio is obtained between 1.021 and 1.029 at the beginning of the cycle. These ratios show a good quasi-uniform power density for each case. Furthermore, up to 155.1 g 233U and 103.6 g 239Pu per day can be produced. The considered system has a high neutronic capability in terms of energy multiplication, fissile breeding, and spent fuel transmutation with thorium utilization. Gizem Bakır, Gamze Genç, and Hüseyin Yapıcı Copyright © 2016 Gizem Bakır et al. All rights reserved. Development and Implementation of an RFID-Based Tunnel Access Monitoring System Tue, 20 Sep 2016 13:37:05 +0000 Due to safety reasons, the exact number and location of people working in an underground tunnel need to be known all the time. This work introduces the development and implementation of an RFID-based access monitoring system for the ONKALO nuclear waste storage facility. This system was taken into use in 2010 and was systematically monitored for one year. The system principle and the used equipment are presented in this paper together with the reliability evaluation results of the implemented system. According to the field use evaluation of the ready system, the reading reliability at the end of the monitoring period was 100%. In addition, even after the successful monitoring period, the system has been updated and new features for safety improvement have been created based on fire department guidelines and achieved user experience. In the future, the RFID system has been planned to be used also in the final depositing of the used nuclear fuel and buffer materials. Kai Kordelin, Jaana Kordelin, Markku Johansson, Johanna Virkki, Leena Ukkonen, and Lauri Sydänheimo Copyright © 2016 Kai Kordelin et al. All rights reserved. Analysis for the Effect of Spatial Discretization Method on AP1000 Reactor Pressure Vessel Fluence Calculation Thu, 15 Sep 2016 09:48:52 +0000 Maintaining the structural integrity of the reactor pressure vessel (RPV) is a critical concern related to the safe operation of nuclear power plants. To estimate the structural integrity over the designed lifetime and to support analyses for a potential plant life extension, an accurate calculation of the fast neutron fluence ( MeV or  MeV) at the RPV is significant. The discrete ordinates method is one of the main methods to solve such problems. During the calculation process, many factors will affect the results. In this paper, the deviations introduced by different differencing schemes and mesh sizes on the AP1000 RPV fast neutron fluence have been studied, which are based on new discrete ordinates code ARES. The analysis shows that the differencing scheme (diamond difference with or without linear zero fix-up, theta weighted, directional theta weighted, and exponential directional weighted) introduces a deviation within 4%. The coarse mesh (4 × 4 cm meshes in plane) leads to approximately 23.7% calculation deviation compared to those of refined mesh (1 × 1 cm meshes in plane). Comprehensive study on the deviation introduced by differencing scheme and mesh size has great significance for reasoned evaluation of RPV fast neutron fluence calculation results. Junxiao Zheng, Bin Zhang, Shengchun Shi, and Yixue Chen Copyright © 2016 Junxiao Zheng et al. All rights reserved. Reliability Evaluation of NPP’s Power Supply System Based on Improved GO-FLOW Method Thu, 08 Sep 2016 08:10:29 +0000 NPP’s power supply system is repairable and there is common cause failure between the components. The repair rate is introduced and total signaling is considered in the improved GO-FLOW method, aimed at reliability analysis for NPP’s power supply system. Traditional GO-FLOW operators’ algorithms are improved. Comprehensively considering the effect of total signaling flow in the power supply system, the equivalent reliability parameter model and common cause failure probability model of multimodal repairable components are constructed. The improved GO-FLOW model of NPP’s power supply system is set up. Based on the proposed model, components’ reliability parameters are computed. The failure probability time-varying trend in thirty years, respectively, of NPP’s offsite power source and power supply system, is simulated and analyzed. Compared with calculation results of dynamic fault tree analysis method, the validity and the simplicity of the improved GO-FLOW method are verified. The effectiveness and applicability of the improved GO-FLOW model for NPP’s power supply system are proved by simulation examples. Jie Zhao, Tian Liu, Yu Zhao, Dichen Liu, Xiaodong Yang, Yi Lin, Zhangsui Lin, and Yong Lei Copyright © 2016 Jie Zhao et al. All rights reserved. Development of a Composite Technique for Preconditioning of 41Cr4 Steel Used as Gear Material: Examination of Its Microstructural Characteristics and Properties Mon, 05 Sep 2016 13:57:38 +0000 Commercial 41Cr4 (ISO standard) steel was treated by a composite technique. An intermediate layer was introduced firstly at the 41Cr4 steel surface by traditional carburizing and nitriding. Then a hard Cr coating was brush-plated on the intermediate layer. Finally, the coating layer was modified by high current pulsed electron beam (HCPEB), followed by quenching and subsequent tempering treatment. The microstructure, mechanical properties, and fracture behavior were characterized. The results show that a nanocrystalline Cr coating is formed at the 41Cr4 steel surface by the treatment of the new composite technique. Such nanocrystalline Cr coating has acceptable hardness and high corrosion resistance performance, which satisfies the demands of the gears working under high speed and corrosive environment. The composite process proposed in this study is considered as a new prospect method due to the multifunction layer design on the gear surface. Jianjun Hu, Chaoping Ma, Hongbin Xu, Ning Guo, and Tianfeng Hou Copyright © 2016 Jianjun Hu et al. All rights reserved. The Feasibility of Multidimensional CFD Applied to Calandria System in the Moderator of CANDU-6 PHWR Using Commercial and Open-Source Codes Mon, 05 Sep 2016 07:52:40 +0000 The moderator system of CANDU, a prototype of PHWR (pressurized heavy-water reactor), has been modeled in multidimension for the computation based on CFD (computational fluid dynamics) technique. Three CFD codes are tested in modeled hydrothermal systems of heavy-water reactors. Commercial codes, COMSOL Multiphysics and ANSYS-CFX with OpenFOAM, an open-source code, are introduced for the various simplified and practical problems. All the implemented computational codes are tested for a benchmark problem of STERN laboratory experiment with a precise modeling of tubes, compared with each other as well as the measured data and a porous model based on the experimental correlation of pressure drop. Also the effect of turbulence model is discussed for these low Reynolds number flows. As a result, they are shown to be successful for the analysis of three-dimensional numerical models related to the calandria system of CANDU reactors. Hyoung Tae Kim, Se-Myong Chang, Jong-Hyeon Shin, and Yong Gwon Kim Copyright © 2016 Hyoung Tae Kim et al. All rights reserved. Design, Construction, and Modeling of a 252Cf Neutron Irradiator Wed, 31 Aug 2016 07:48:07 +0000 Neutron production methods are an integral part of research and analysis for an array of applications. This paper examines methods of neutron production, and the advantages of constructing a radioisotopic neutron irradiator assembly using 252Cf. Characteristic neutron behavior and cost-benefit comparative analysis between alternative modes of neutron production are also examined. The irradiator is described from initial conception to the finished design. MCNP modeling shows a total neutron flux of 3 × 105 n/(cm2·s) in the irradiation chamber for a 25 μg source. Measurements of the gamma-ray and neutron dose rates near the external surface of the irradiator assembly are 120 μGy/h and 30 μSv/h, respectively, during irradiation. At completion of the project, total material, and labor costs remained below $50,000. Blake C. Anderson, Keith E. Holbert, and Herbert Bowler Copyright © 2016 Blake C. Anderson et al. All rights reserved. Wave Characteristics of Falling Film on Inclination Plate at Moderate Reynolds Number Mon, 29 Aug 2016 11:46:17 +0000 Falling water film on an inclined plane is studied by shadowgraphy. The ranges of inclination angle and the film Reynolds number are, respectively, up to 21° and 60. Water is used as working fluid. The scenario of wave regime evolution is identified as three distinctive regimes, namely, initial quiescent smooth film flow, two-dimensional regular solitary wave pattern riding on film flow, and three-dimensional irregular wave pattern. Three characteristic parameters of two-dimensional solitary wave pattern, namely, inception length, primary pulse spacing, and propagation velocity, are examined, which are significant in engineering applications for estimation of heat and mass transfer on film flow. The present experimental data are well in agreement with the Koizumi correlations, the deviation from which is limited to 20% and 15%, respectively, for primary pulse spacing and propagation velocity. Through the scrutiny of the present experimental observation, it is concluded that wave evolution on film flow at the moderate Reynolds number is controlled by gravity and drag and the Rayleigh-Taylor instability that occurred on the steep front of primary pulse triggers the disintegration of continuous two-dimensional regular solitary wave pattern into three-dimensional irregular wave pattern. Chuan Lu, Sheng-Yao Jiang, and Ri-Qiang Duan Copyright © 2016 Chuan Lu et al. All rights reserved. Analysis of Void Reactivity Coefficient for 3D BWR Assembly Model Wed, 24 Aug 2016 11:48:18 +0000 The effect of BWR fuel assembly 3D model on void reactivity coefficient (VRC) estimation is investigated. VRC values were calculated for different BWR assembly models applying deterministic T-NEWT and Monte Carlo KENO-VI functional modules of SCALE 6.1 code package. The difference between deterministic T-NEWT and Monte Carlo KENO-VI simulations is negligible (0.18 pcm/%). The influence of the assumed more detailed coolant density profile was estimated as well. VRC increases with the application of a larger number of coolant density values across fuel assembly height. It was shown that the coolant density profile described by 6 values per height could be considered sufficient from prospect of VRC estimation, as a more detailed density profile has impact below 1% on total assembly void effects. VRC values were decomposed to values for individual nodes and isotopes, since decomposition provides useful insights to describe the overall behaviour of VRC in detail. Andrius Slavickas, Raimondas Pabarčius, Aurimas Tonkūnas, and Eugenijus Ušpuras Copyright © 2016 Andrius Slavickas et al. All rights reserved. The Definition Method and Optimization of Atomic Strain Tensors for Nuclear Power Engineering Materials Thu, 18 Aug 2016 15:30:24 +0000 A common measure of deformation between atomic scale simulations and the continuum framework is provided and the strain tensors for multiscale simulations are defined in this paper. In order to compute the deformation gradient of any atom , the weight function is proposed to eliminate the different contributions within the neighbor atoms which have different distances to atom , and the weighted least squares error optimization model is established to seek the optimal coefficients of the weight function and the optimal local deformation gradient of each atom. The optimization model involves more than 9 parameters. To guarantee the reliability of subsequent parameters identification result and lighten the calculation workload of parameters identification, an overall analysis method of parameter sensitivity and an advanced genetic algorithm are also developed. Xiangguo Zeng, Ying Sheng, Huayan Chen, and Tixin Han Copyright © 2016 Xiangguo Zeng et al. All rights reserved. Calculation of Radioactivity and Dose Rate of Activated Corrosion Products in Water-Cooled Fusion Reactor Wed, 10 Aug 2016 11:19:37 +0000 In water-cooled reactor, the dominant radioactive source term under normal operation is activated corrosion products (ACPs), which have an important impact on reactor inspection and maintenance. A three-node transport model of ACPs was introduced into the new version of ACPs source term code CATE in this paper, which makes CATE capable of theoretically simulating the variation and the distribution of ACPs in a water-cooled reactor and suitable for more operating conditions. For code testing, MIT PWR coolant chemistry loop was simulated, and the calculation results from CATE are close to the experimental results from MIT, which means CATE is available and credible on ACPs analysis of water-cooled reactor. Then ACPs in the blanket cooling loop of water-cooled fusion reactor ITER under construction were analyzed using CATE and the results showed that the major contributors are the short-life nuclides, especially Mn-56. At last a point kernel integration code ARShield was coupled with CATE, and the dose rate around ITER blanket cooling loop was calculated. Results showed that after shutting down the reactor only for 8 days, the dose rate decreased nearly one order of magnitude, which was caused by the rapid decay of the short-life ACPs. Jingyu Zhang, Lu Li, Shuxiang He, and Yixue Chen Copyright © 2016 Jingyu Zhang et al. All rights reserved. Application of Nuclear Analytical Techniques in Elemental Characterization of Wadi El-Nakhil Alabaster, Central Eastern Desert, Egypt Thu, 04 Aug 2016 08:22:46 +0000 Instrumental neutron activation analysis (INAA) is a powerful technique for trace element determination in rocks. Nine alabaster samples were collected from Wadi El-Nakhil located at the intersection of lat. 26°10′50′′N and long. 34°03′40′′E, central Eastern Desert, Egypt, for investigation by INAA and Energy Depressive X-Ray Fluorescence (EDXRF). The samples were irradiated by thermal neutrons at the TRIGA Mainz research reactor at a neutron flux of 7 × 1011 n/cm2·s. Twenty-two elements were determined, namely, As, Ba, Ca, Co, Cr, Sc, Fe, Hf, K, Mg, Mn, Na, Rb, U, Zn, Zr, Lu, Ce, Sm, La, Yb, and Eu. The chemical analysis of alabaster indicated having high contents of CaO and MgO and LOI and low contents of SiO2, Al2O3, Na2O, K2O, MnO, and Fe2O3. Zain M. Alamoudi and A. El-Taher Copyright © 2016 Zain M. Alamoudi and A. El-Taher. All rights reserved. MCNP-X Monte Carlo Code Application for Mass Attenuation Coefficients of Concrete at Different Energies by Modeling 3 × 3 Inch NaI(Tl) Detector and Comparison with XCOM and Monte Carlo Data Sun, 31 Jul 2016 08:31:14 +0000 Gamma-ray measurements in various research fields require efficient detectors. One of these research fields is mass attenuation coefficients of different materials. Apart from experimental studies, the Monte Carlo (MC) method has become one of the most popular tools in detector studies. An NaI(Tl) detector has been modeled, and, for a validation study of the modeled NaI(Tl) detector, the absolute efficiency of 3 × 3 inch cylindrical NaI(Tl) detector has been calculated by using the general purpose Monte Carlo code MCNP-X (version 2.4.0) and compared with previous studies in literature in the range of 661–2620 keV. In the present work, the applicability of MCNP-X Monte Carlo code for mass attenuation of concrete sample material as building material at photon energies 59.5 keV, 80 keV, 356 keV, 661.6 keV, 1173.2 keV, and 1332.5 keV has been tested by using validated NaI(Tl) detector. The mass attenuation coefficients of concrete sample have been calculated. The calculated results agreed well with experimental and some other theoretical results. The results specify that this process can be followed to determine the data on the attenuation of gamma-rays with other required energies in other materials or in new complex materials. It can be concluded that data from Monte Carlo is a strong tool not only for efficiency studies but also for mass attenuation coefficients calculations. Huseyin Ozan Tekin Copyright © 2016 Huseyin Ozan Tekin. All rights reserved. Parametric Investigation and Thermoeconomic Optimization of a Combined Cycle for Recovering the Waste Heat from Nuclear Closed Brayton Cycle Tue, 19 Jul 2016 13:10:48 +0000 A combined cycle that combines AWM cycle with a nuclear closed Brayton cycle is proposed to recover the waste heat rejected from the precooler of a nuclear closed Brayton cycle in this paper. The detailed thermodynamic and economic analyses are carried out for the combined cycle. The effects of several important parameters, such as the absorber pressure, the turbine inlet pressure, the turbine inlet temperature, the ammonia mass fraction, and the ambient temperature, are investigated. The combined cycle performance is also optimized based on a multiobjective function. Compared with the closed Brayton cycle, the optimized power output and overall efficiency of the combined cycle are higher by 2.41% and 2.43%, respectively. The optimized LEC of the combined cycle is 0.73% lower than that of the closed Brayton cycle. Lihuang Luo, Hong Gao, Chao Liu, and Xiaoxiao Xu Copyright © 2016 Lihuang Luo et al. All rights reserved. Oxidation Analyses of Massive Air Ingress Accident of HTR-PM Mon, 18 Jul 2016 16:47:55 +0000 The double-ended guillotine break (DEGB) of the horizontal coaxial gas duct accident is a serious air ingress accident of the high temperature gas-cooled reactor pebble-bed module (HTR-PM). Because the graphite is widely used as the structure material and the fuel element matrix of HTR-PM, the oxidation analyses of this severe air ingress accident have got enough attention in the safety analyses of the HTR-PM. The DEGB of the horizontal coaxial gas duct accident is calculated by using the TINTE code in this paper. The results show that the maximum local oxidation of the matrix graphite of spherical fuel elements in the core will firstly reach  mol/m3 at about 120 h, which means that only the outer 5 mm fuel-free zone of matrix graphite will be oxidized out. Even at 150 h, the maximum local weight loss ratio of the nuclear grade graphite in the bottom reflectors is only 0.26. Besides, there is enough time to carry out some countermeasures to stop the air ingress during several days. Therefore, the nuclear grade graphite of the bottom reflectors will not be fractured in the DEGB of the horizontal coaxial gas duct accident and the integrity of the HTR-PM can be guaranteed. Wei Xu, Yanhua Zheng, Lei Shi, and Peng Liu Copyright © 2016 Wei Xu et al. All rights reserved. A Comparative Study for Modeling Displacement Instabilities due to TGO Formation in TBCs of High-Temperature Components in Nuclear Power Plant Thu, 14 Jul 2016 07:09:55 +0000 This paper reports two numerical simulation methods for modeling displacement instabilities around a surface groove in a metal substrate used in nuclear power plant. The amplitude change in the groove, the downward displacement at the base node, and the groove displacement at the periphery were simulated using ABAQUS to compare the results from two methods, as well as the tangential stress in the elements at the groove base and periphery. The comparison showed that for the tangential stress two methods were in close agreement for all thermal cycles. For the amplitude change, the downward displacement, the groove displacement, and the stress distribution, the two methods were in close agreement for the first 3 to 6 thermal cycles. After that, inconsistency increased with the number of thermal cycles. It is interesting that the thermal cycle at which the discrepancy between the two methods began to occur corresponded to a thermally grown oxide (TGO) thickness of 1 μm, which showed the accuracy of the present work over the classic method. It is concluded that the present work’s numerical simulation scheme worked better with a thinner TGO layer than the classic method and could overcome the limitation of TGO thickness by simulating any thickness. Xia Huang, Jian Wang, Kun Song, Feng Zhang, Tong Yi, and Jun Ding Copyright © 2016 Xia Huang et al. All rights reserved. NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR) Sun, 10 Jul 2016 09:39:04 +0000 This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR). The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers), heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s). The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use radial nodes per assembly, axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively. Surian Pinem, Tagor Malem Sembiring, and Peng Hong Liem Copyright © 2016 Surian Pinem et al. All rights reserved. The Sliding and Overturning Analysis of Spent Fuel Storage Rack Based on Dynamic Analysis Model Thu, 30 Jun 2016 12:32:56 +0000 Spent fuel rack is the key equipment for the storage of spent fuel after refueling. In order to investigate the performance of the spent fuel rack under the earthquake, the phenomena including sliding, collision, and overturning of the spent fuel rack were studied. An FEM model of spent fuel rack is built to simulate the transient response under seismic loading regarding fluid-structure interaction by ANSYS. Based on D’Alambert’s principle, the equilibriums of force and momentum were established to obtain the critical sliding and overturning accelerations. Then 5 characteristic transient loadings which were designed based on the critical sliding and overturning accelerations were applied to the rack FEM model. Finally, the transient displacement and impact force response of rack with different gap sizes and the supporting leg friction coefficients were analyzed. The result proves the FEM model is applicable for seismic response of spent fuel rack. This paper can guide the design of the future’s fluid-structure interaction experiment for spent fuel rack. Yu Liu, Daogang Lu, Yuanpeng Wang, and Hongda Liu Copyright © 2016 Yu Liu et al. All rights reserved. A Study on the Instantaneous Turbulent Flow Field in a 90-Degree Elbow Pipe with Circular Section Thu, 23 Jun 2016 09:01:39 +0000 Based on the special application of 90-degree elbow pipe in the HTR-PM, the large eddy simulation was selected to calculate the instantaneous flow field in the 90-degree elbow pipe combining with the experimental results. The characteristics of the instantaneous turbulent flow field under the influence of flow separation and secondary flow were studied by analyzing the instantaneous pressure information at specific monitoring points and the instantaneous velocity field on the cross section of the elbow. The pattern and the intensity of the Dean vortex and the small scale eddies change over time and induce the asymmetry of the flow field. The turbulent disturbance upstream and the flow separation near the intrados couple with the vortexes of various scales. Energy is transferred from large scale eddies to small scale eddies and dissipated by the viscous stress in the end. Shiming Wang, Cheng Ren, Yangfei Sun, Xingtuan Yang, and Jiyuan Tu Copyright © 2016 Shiming Wang et al. All rights reserved. An Approach for Integrated Analysis of Human Factors in Remote Handling Maintenance Wed, 22 Jun 2016 10:19:51 +0000 Considering dangerous environmental conditions, maintenance of radioactive equipment can be performed by remote handling maintenance (RHM) system. The RHM system is a sophisticated man-machine system. Therefore, human factors analysis is an inevitable aspect considered in guaranteeing successful and safe task performance. This study proposes an approach for integrated analysis of human factors in RHM so as to make the evaluating process more practical. In the approach, indicators of accessibility, health safety, and fatigue are analyzed using virtual human simulation technologies. The human error factors in the maintenance process are analyzed using the human error probability (HEP) based on the success likelihood index method- (SLIM-) analytic hierarchy process (AHP). The psychological factors level of maintenance personnel is determined with an expert scoring. The human factors for the entire RHM system are then evaluated using the interval method. An application example is present, and the application results show that the approach can support the evaluation of the human factors in RHM. Jianwen Guo, Zhenzhong Sun, Jiaxin He, Xuejun Jia, Hongjuan Li, Xiaohui Yan, Haibin Chen, Hong Tang, and GuoHong Wu Copyright © 2016 Jianwen Guo et al. All rights reserved. A Calculation Method for the Sloshing Impact Pressure Imposed on the Roof of a Passive Water Storage Tank of AP1000 Sun, 12 Jun 2016 08:14:30 +0000 There is a large water storage tank installed at the top of containment of AP1000, which can supply the passive cooling. In the extreme condition, sloshing of the free surface in the tank may impact on the roof under long-period earthquake. For the safety assessment of structure, it is necessary to calculate the impact pressure caused by water sloshing. Since the behavior of sloshing impacted on the roof is involved into a strong nonlinear phenomenon, it is a little difficult to calculate such pressure by theoretical or numerical method currently. But it is applicable to calculate the height of sloshing in a tank without roof. In the present paper, a simplified method was proposed to calculate the impact pressure using the sloshing wave height, in which we first marked the position of the height of roof, then produced sloshing in the tank without roof and recorded the maximum wave height, and finally regarded approximately the difference between maximum wave height and roof height as the impact pressure head. We also designed an experiment to verify this method. The experimental result showed that this method overpredicted the impact pressure with a certain error of no more than 35%. By the experiment, we conclude that this method is conservative and applicable for the engineering design. Daogang Lu, Xiaojia Zeng, Junjie Dang, and Yu Liu Copyright © 2016 Daogang Lu et al. All rights reserved. Effect of Chemical Corrosion on the Mechanical Characteristics of Parent Rocks for Nuclear Waste Storage Tue, 07 Jun 2016 08:47:39 +0000 Long-term immersion was adopted to explore the damage deterioration and mechanical properties of granite under different chemical solutions. Here, granite was selected as the candidate of parent rocks for nuclear waste storage. The physical and mechanical properties of variation regularity immersed in various chemical solutions were analyzed. Meanwhile, the damage variable based on the variation in porosity was used in the quantitative analysis of chemical damage deterioration degree. Experimental results show that granite has a significant weakening tendency after chemical corrosion. The fracture toughness , splitting tensile strength, and compressive strength all demonstrate the same deteriorating trend with chemical corrosion time. However, a difference exists in the deterioration degree of the mechanical parameters; that is, the deterioration degree of fracture toughness is the greatest followed by those of splitting tensile strength and compressive strength, which are relatively smaller. Strong acid solutions may aggravate chemical damage deterioration in granite. By contrast, strong alkaline solutions have a certain inhibiting effect on chemical damage deterioration. The chemical solutions that feature various compositions may have different effects on chemical damage degree; that is, ions have a greater effect on the chemical damage in granite than ions. Tielin Han, Junping Shi, Yunsheng Chen, and Zhihui Li Copyright © 2016 Tielin Han et al. All rights reserved. ROSA/LSTF Tests and RELAP5 Posttest Analyses for PWR Safety System Using Steam Generator Secondary-Side Depressurization against Effects of Release of Nitrogen Gas Dissolved in Accumulator Water Mon, 06 Jun 2016 11:42:21 +0000 Two tests related to a new safety system for a pressurized water reactor were performed with the ROSA/LSTF (rig of safety assessment/large scale test facility). The tests simulated cold leg small-break loss-of-coolant accidents with 2-inch diameter break using an early steam generator (SG) secondary-side depressurization with or without release of nitrogen gas dissolved in accumulator (ACC) water. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. The pressure difference between the primary and SG secondary sides after the actuation of ACC system was larger in the test with the dissolved gas release than that in the test without the dissolved gas release. No core uncovery and heatup took place because of the ACC coolant injection and two-phase natural circulation. Long-term core cooling was ensured by the actuation of low-pressure injection system. The RELAP5 code predicted most of the overall trends of the major thermal-hydraulic responses after adjusting a break discharge coefficient for two-phase discharge flow under the assumption of releasing all the dissolved gas at the vessel upper plenum. Takeshi Takeda, Akira Ohnuki, Daisuke Kanamori, and Iwao Ohtsu Copyright © 2016 Takeshi Takeda et al. All rights reserved. Assessment of Prediction Capabilities of COCOSYS and CFX Code for Simplified Containment Mon, 06 Jun 2016 07:28:26 +0000 The acceptable accuracy for simulation of severe accident scenarios in containments of nuclear power plants is required to investigate the consequences of severe accidents and effectiveness of potential counter measures. For this purpose, the actual capability of CFX tool and COCOSYS code is assessed in prototypical geometries for simplified physical process-plume (due to a heat source) under adiabatic and convection boundary condition, respectively. Results of the comparison under adiabatic boundary condition show that good agreement is obtained among the analytical solution, COCOSYS prediction, and CFX prediction for zone temperature. The general trend of the temperature distribution along the vertical direction predicted by COCOSYS agrees with the CFX prediction except in dome, and this phenomenon is predicted well by CFX and failed to be reproduced by COCOSYS. Both COCOSYS and CFX indicate that there is no temperature stratification inside dome. CFX prediction shows that temperature stratification area occurs beneath the dome and away from the heat source. Temperature stratification area under adiabatic boundary condition is bigger than that under convection boundary condition. The results indicate that the average temperature inside containment predicted with COCOSYS model is overestimated under adiabatic boundary condition, while it is underestimated under convection boundary condition compared to CFX prediction. Jia Zhu, Xiaohui Zhang, and Xu Cheng Copyright © 2016 Jia Zhu et al. All rights reserved. Particle Swarm Optimization-Based Direct Inverse Control for Controlling the Power Level of the Indonesian Multipurpose Reactor Tue, 31 May 2016 12:05:21 +0000 A neural network-direct inverse control (NN-DIC) has been simulated to automatically control the power level of nuclear reactors. This method has been tested on an Indonesian pool type multipurpose reactor, namely, Reaktor Serba Guna-GA Siwabessy (RSG-GAS). The result confirmed that this method still cannot minimize errors and shorten the learning process time. A new method is therefore needed which will improve the performance of the DIC. The objective of this study is to develop a particle swarm optimization-based direct inverse control (PSO-DIC) to overcome the weaknesses of the NN-DIC. In the proposed PSO-DIC, the PSO algorithm is integrated into the DIC technique to train the weights of the DIC controller. This integration is able to accelerate the learning process. To improve the performance of the system identification, a backpropagation (BP) algorithm is introduced into the PSO algorithm. To show the feasibility and effectiveness of this proposed PSO-DIC technique, a case study on power level control of RSG-GAS is performed. The simulation results confirm that the PSO-DIC has better performance than NN-DIC. The new developed PSO-DIC has smaller steady-state error and less overshoot and oscillation. Yoyok Dwi Setyo Pambudi, Wahidin Wahab, and Benyamin Kusumoputro Copyright © 2016 Yoyok Dwi Setyo Pambudi et al. All rights reserved.