Science and Technology of Nuclear Installations The latest articles from Hindawi Publishing Corporation © 2016 , Hindawi Publishing Corporation . 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. Advanced PHWR Safety Technology: PHWR Challenging Issues for Safe Operation and Long-Term Sustainability Tue, 17 May 2016 08:16:46 +0000 Jin Ho Song, Wei Shen, Malcolm Griffiths, Bo Wook Rhee, YongMann Song, and Masanori Naitoh Copyright © 2016 Jin Ho Song et al. All rights reserved. Scaled-Down Moderator Circulation Test Facility at Korea Atomic Energy Research Institute Tue, 10 May 2016 09:24:09 +0000 Korea Atomic Energy Research Institute (KAERI) started the experimental research on moderator circulation as one of a the national research and development programs from 2012. This research program includes the construction of the moderator circulation test (MCT) facility, production of the validation data for self-reliant computational fluid dynamics (CFD) tools, and development of optical measurement system using the particle image velocimetry (PIV). In the present paper we introduce the scaling analysis performed to extend the scaling criteria suitable for reproducing thermal-hydraulic phenomena in a scaled-down CANDU- (CANada Deuterium Uranium-) 6 moderator tank, a manufacturing status of the 1/4 scale moderator tank. Also, preliminary CFD analysis results for the full-size and scaled-down moderator tanks are carried out to check whether the moderator flow and temperature patterns of both the full-size reactor and scaled-down facility are identical. Hyoung Tae Kim and Bo Wook Rhee Copyright © 2016 Hyoung Tae Kim and Bo Wook Rhee. All rights reserved. Medical Radioisotope Production in a Power-Flattened ADS Fuelled with Uranium and Plutonium Dioxides Wed, 04 May 2016 16:40:31 +0000 This study presents the medical radioisotope production performance of a conceptual accelerator driven system (ADS). Lead-bismuth eutectic (LBE) is selected as target material. The subcritical fuel core is conceptually divided into ten equidistant subzones. The ceramic (natural U, Pu)O2 fuel mixture and the materials used for radioisotope production (copper, gold, cobalt, holmium, rhenium, thulium, mercury, palladium, thallium, molybdenum, and yttrium) are separately prepared as cylindrical rods cladded with carbon/carbon composite (C/C) and these rods are located in the subzones. In order to obtain the flattened power density, percentages of PuO2 in the mixture of UO2 and PuO2 in the subzones are adjusted in radial direction of the fuel zone. Time-dependent calculations are performed at 1000 MW thermal fission power () for one hour using the BURN card. The neutronic results show that the investigated ADS has a high neutronic capability, in terms of medical radioisotope productions, spent fuel transmutation and energy multiplication. Moreover, a good quasiuniform power density is achieved in each material case. The peak-to-average fission power density ratio is in the range of 1.02–1.28. Gizem Bakır, Saltuk Buğra Selçuklu, and Hüseyin Yapıcı Copyright © 2016 Gizem Bakır et al. All rights reserved. Prediction Study on PCI Failure of Reactor Fuel Based on a Radial Basis Function Neural Network Sun, 24 Apr 2016 12:50:04 +0000 Pellet-clad interaction (PCI) is one of the major issues in fuel rod design and reactor core operation in water cooled reactors. The prediction of fuel rod failure by PCI is studied in this paper by the method of radial basis function neural network (RBFNN). The neural network is built through the analysis of the existing experimental data. It is concluded that it is a suitable way to reduce the calculation complexity. A self-organized RBFNN is used in our study, which can vary its structure dynamically in order to maintain the prediction accuracy. For the purpose of the appropriate network complexity and overall computational efficiency, the hidden neurons in the RBFNN can be changed online based on the neuron activity and mutual information. The presented method is tested by the experimental data from the reference, and the results demonstrate its effectiveness. Xinyu Wei, Jiashuang Wan, and Fuyu Zhao Copyright © 2016 Xinyu Wei et al. All rights reserved. Elemental Analysis and Natural Radioactivity Levels of Clay by Gamma Ray Spectrometer and Instrumental Neutron Activation Analysis Mon, 18 Apr 2016 16:44:35 +0000 Due to increased global demand for clay, the present work involves the use of INAA for elemental analysis and pollutants concentration in clay. The samples were collected from Aswan in South Egypt. The samples were irradiated using the thermal neutrons “at the TRIGA Mainz research reactor” and at a neutron flux “of 7 × 10 n/cm s”. Twenty-six elements quantitatively and qualitatively were specified for the first time upon studying the samples. The elements determined are U, Th, Ta, Hf, Lu, Eu, Ce, Ba, Sn, Nb, Rb, Zn, Co, Fe, Cr, Sc, Sm, La, Yb, As, Ga, K, Mn, Na, Ti, and Mg. The concentrations of natural radionuclides 232Th, 226Ra, and 40K were also calculated. Based on these concentrations, to estimate the exposure risk for using clay as raw materials in building materials, the radiation hazard indices such as radium equivalent activities, effective doses rate, and the external hazard indices have been computed. The obtained results were compared with analogous studies carried out in other countries and with the UNSCEAR reports. W. R. Alharbi and A. El-Taher Copyright © 2016 W. R. Alharbi and A. El-Taher. All rights reserved. Statistical Analysis of Loss of Offsite Power Events Thu, 14 Apr 2016 06:12:17 +0000 This paper presents the results of the statistical analysis of the loss of offsite power events (LOOP) registered in four reviewed databases. The reviewed databases include the IRSN (Institut de Radioprotection et de Sûreté Nucléaire) SAPIDE database and the GRS (Gesellschaft für Anlagen- und Reaktorsicherheit mbH) VERA database reviewed over the period from 1992 to 2011. The US NRC (Nuclear Regulatory Commission) Licensee Event Reports (LERs) database and the IAEA International Reporting System (IRS) database were screened for relevant events registered over the period from 1990 to 2013. The number of LOOP events in each year in the analysed period and mode of operation are assessed during the screening. The LOOP frequencies obtained for the French and German nuclear power plants (NPPs) during critical operation are of the same order of magnitude with the plant related events as a dominant contributor. A frequency of one LOOP event per shutdown year is obtained for German NPPs in shutdown mode of operation. For the US NPPs, the obtained LOOP frequency for critical and shutdown mode is comparable to the one assessed in NUREG/CR-6890. Decreasing trend is obtained for the LOOP events registered in three databases (IRSN, GRS, and NRC). Andrija Volkanovski, Antonio Ballesteros Avila, and Miguel Peinador Veira Copyright © 2016 Andrija Volkanovski et al. All rights reserved. Nuclear Power Plant Construction Scheduling Problem with Time Restrictions: A Particle Swarm Optimization Approach Mon, 11 Apr 2016 06:05:17 +0000 In nuclear power plant construction scheduling, a project is generally defined by its dependent preparation time, the time required for construction, and its reactor installation time. The issues of multiple construction teams and multiple reactor installation teams are considered. In this paper, a hierarchical particle swarm optimization algorithm is proposed to solve the nuclear power plant construction scheduling problem and minimize the occurrence of projects failing to achieve deliverables within applicable due times and deadlines. Shang-Kuan Chen, Yen-Wu Ti, and Kuo-Yu Tsai Copyright © 2016 Shang-Kuan Chen et al. All rights reserved. Measurement of the Residual Stresses and Investigation of Their Effects on a Hardfaced Grid Plate due to Thermal Cycling in a Pool Type Sodium-Cooled Fast Reactor Sun, 03 Apr 2016 09:53:38 +0000 In sodium-cooled fast reactors (SFR), grid plate is a critical component which is made of 316 L(N) SS. It is supported on core support structure. The grid plate supports the core subassemblies and maintains their verticality. Most of the components of SFR are made of 316 L(N)/304 L(N) SS and they are in contact with the liquid-metal sodium which acts as a coolant. The peak operating temperature in SFR is 550°C. However, the self-welding starts at 500°C. To avoid self-welding and galling, hardfacing of the grid plate has become necessary. Nickel based cobalt-free colmonoy 5 has been identified as the hardfacing material due to its lower dose rate by Plasma Transferred Arc Welding (PTAW). This paper is concerned with the measurement and investigations of the effects of the residual stress generated due to thermal cycling on a scale-down physical model of the grid plate. Finite element analysis of the hardfaced grid plate model is performed for obtaining residual stresses using elastoplastic analysis and hence the results are validated. The effects of the residual stresses due to thermal cycling on the hardfaced grid plate model are studied. S. Balaguru, Vela Murali, and P. Chellapandi Copyright © 2016 S. Balaguru et al. All rights reserved. Preliminary Study on the Fabrication of Particulate Fuel through Pressureless Sintering Process Wed, 23 Mar 2016 12:20:30 +0000 U-10wt%Zr spherical particles for use as particulate fuel were prepared by centrifugal atomization and subjected to pressureless sintering, which is one of the simplest powder processing techniques. At sintering temperature of 1100°C for 30 or 60 min, all samples ranging from +50 to −325 mesh showed no apparent bonding between the particles. However, at 1150°C (80 min), all samples formed a bulk body and the microstructures showed apparent sintering stages. Particularly, sample B (50–70 mesh) and sample C (70–100 mesh) showed pore characteristics suitable for a particulate fuel. The results suggest that pressureless sinterability for U-10Zr particulate fuel can be improved by adding small-size (–325 mesh) particles. Jong-Hwan Kim, Jung-Won Lee, Ki-Hwan Kim, and Chan-Bock Lee Copyright © 2016 Jong-Hwan Kim et al. All rights reserved. A Denoising Based Autoassociative Model for Robust Sensor Monitoring in Nuclear Power Plants Tue, 22 Mar 2016 06:13:56 +0000 Sensors health monitoring is essentially important for reliable functioning of safety-critical chemical and nuclear power plants. Autoassociative neural network (AANN) based empirical sensor models have widely been reported for sensor calibration monitoring. However, such ill-posed data driven models may result in poor generalization and robustness. To address above-mentioned issues, several regularization heuristics such as training with jitter, weight decay, and cross-validation are suggested in literature. Apart from these regularization heuristics, traditional error gradient based supervised learning algorithms for multilayered AANN models are highly susceptible of being trapped in local optimum. In order to address poor regularization and robust learning issues, here, we propose a denoised autoassociative sensor model (DAASM) based on deep learning framework. Proposed DAASM model comprises multiple hidden layers which are pretrained greedily in an unsupervised fashion under denoising autoencoder architecture. In order to improve robustness, dropout heuristic and domain specific data corruption processes are exercised during unsupervised pretraining phase. The proposed sensor model is trained and tested on sensor data from a PWR type nuclear power plant. Accuracy, autosensitivity, spillover, and sequential probability ratio test (SPRT) based fault detectability metrics are used for performance assessment and comparison with extensively reported five-layer AANN model by Kramer. Ahmad Shaheryar, Xu-Cheng Yin, Hong-Wei Hao, Hazrat Ali, and Khalid Iqbal Copyright © 2016 Ahmad Shaheryar et al. All rights reserved. Analysis of the Relationship between Risk Perception and Willingness to Pay for Nuclear Power Plant Risk Reduction Sun, 20 Mar 2016 09:53:52 +0000 With the adoption of new technologies, more risk is introduced into modern society. Important decisions about new technologies tend to be made by specialists, which can lead to a mismatch of risk perception between citizens and specialists, resulting in high social cost. Using contingent valuation methods, this paper analyzes the relationship between willingness to pay (WTP) and the factors expressed through people’s image of nuclear power plants (NPP), their perception of NPP safety, and how these can be affected by their scientific background level. Results indicate that groups with a high scientific background level tend to have low risk perception level, represented through their image and safety levels. Further, the results show that mean WTP is dependent on scientific background and image levels. It is believed that these results could help decision makers address the mismatch of trust between the public and specialists in terms of new policy. Mirae Yun, Sang Hun Lee, and Hyun Gook Kang Copyright © 2016 Mirae Yun et al. All rights reserved.