Science and Technology of Nuclear Installations http://www.hindawi.com The latest articles from Hindawi Publishing Corporation © 2016 , Hindawi Publishing Corporation . 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 http://www.hindawi.com/journals/stni/2016/8368504/ 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 http://www.hindawi.com/journals/stni/2016/5265748/ 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 http://www.hindawi.com/journals/stni/2016/9108751/ 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 http://www.hindawi.com/journals/stni/2016/1613989/ 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 http://www.hindawi.com/journals/stni/2016/7853787/ 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 http://www.hindawi.com/journals/stni/2016/7481793/ 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 http://www.hindawi.com/journals/stni/2016/9542121/ 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 http://www.hindawi.com/journals/stni/2016/1065790/ 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 http://www.hindawi.com/journals/stni/2016/2689861/ 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 http://www.hindawi.com/journals/stni/2016/5903602/ 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 http://www.hindawi.com/journals/stni/2016/5302176/ 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 http://www.hindawi.com/journals/stni/2016/4720685/ 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 http://www.hindawi.com/journals/stni/2016/8726260/ 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 http://www.hindawi.com/journals/stni/2016/7692659/ 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 http://www.hindawi.com/journals/stni/2016/2150692/ 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 http://www.hindawi.com/journals/stni/2016/8353256/ 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 http://www.hindawi.com/journals/stni/2016/3717361/ 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 http://www.hindawi.com/journals/stni/2016/9746948/ 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 http://www.hindawi.com/journals/stni/2016/6293758/ 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. Emittance Measurement for Beamline Extension at the PET Cyclotron Wed, 16 Mar 2016 13:13:21 +0000 http://www.hindawi.com/journals/stni/2016/4697247/ Particle-induced X-ray emission is used for determining the elemental composition of materials. This method uses low-energy protons (of several MeV), which can be obtained from high-energy (of tens MeV) accelerators. Instead of manufacturing an accelerator for generating the MeV protons, the use of a PET cyclotron has been suggested for designing the beamline for multipurpose applications, especially for the PIXE experiment, which has a dedicated high-energy (of tens MeV) accelerator. The beam properties of the cyclotron were determined at this experimental facility by using an external beamline before transferring the ion beam to the experimental chamber. We measured the beam profile and calculated the emittance using the pepper-pot method. The beam profile was measured as the beam current using a wire scanner, and the emittance was measured as the beam distribution at the beam dump using a radiochromic film. We analyzed the measurement results and are planning to use the results obtained in the simulations of external beamline and aligned beamline components. We will consider energy degradation after computing the beamline simulation. The experimental study focused on measuring the emittance from the cyclotron, and the results of this study are presented in this paper. Sae-Hoon Park, Sang-Hoon Lee, and Yu-Seok Kim Copyright © 2016 Sae-Hoon Park et al. All rights reserved. New RELAP5-3D Lead and LBE Thermophysical Properties Implementation for Safety Analysis of Gen IV Reactors Wed, 09 Mar 2016 13:26:21 +0000 http://www.hindawi.com/journals/stni/2016/1687946/ The latest versions of RELAP5-3D© code allow the simulation of thermodynamic system, using different type of working fluids, that is, liquid metals, molten salt, diathermic oil, and so forth, thanks to the ATHENA code integration. The RELAP5-3D© water thermophysical properties are largely verified and validated; however there are not so many experiments to generate the liquid metals ones in particular for the Lead and the Lead Bismuth Eutectic. Recently, new and more accurate experimental data are available for liquid metals. The comparison between these state-of-the-art data and the RELAP5-3D© default thermophysical properties shows some discrepancy; therefore a tool for the generation of new properties binary files has been developed. All the available data came from experiments performed at atmospheric pressure. Therefore, to extend the pressure domain below and above this pressure, the tool fits a semiempirical model (soft sphere model with inverse-power-law potential), specific for the liquid metals. New binary files of thermophysical properties, with a detailed mesh grid of point to reduce the code mass error (especially for the Lead), were generated with this tool. Finally, calculations using a simple natural circulation loop were performed to understand the differences between the default and the new properties. P. Balestra, F. Giannetti, G. Caruso, and A. Alfonsi Copyright © 2016 P. Balestra et al. All rights reserved. Numerical Investigation of the Transient Behavior of a Hot Gas Duct under Rapid Depressurization Wed, 09 Mar 2016 12:02:07 +0000 http://www.hindawi.com/journals/stni/2016/6858425/ A hot gas duct is an indispensable component for the nuclear-process heat applications of the Very-High-Temperature Reactor (VHTR), which has to fulfill three requirements: to withstand high temperature, high pressure, and large pressure transient. In this paper, numerical investigation of pressure transient is performed for a hot gas duct under rapid depressurization. System depressurization imposes an imploding pressure differential on the internal structural elements of a hot gas duct, the structural integrity of which is susceptible to being damaged. Pressure differential and its imposed duration, which are two key factors to evaluate the damage severity of a hot gas duct under depressurization, are examined in regard to depressurization rate and insulation packing tightness. It is revealed that depressurization rate is a decisive parameter for controlling the pressure differential and its duration, whereas insulating-packing tightness has little effect on them. JingBao Liu, ShengYao Jiang, Tao Ma, and RiQiang Duan Copyright © 2016 JingBao Liu et al. All rights reserved. A Simple Formula for Local Burnup and Isotope Distributions Based on Approximately Constant Relative Reaction Rate Wed, 17 Feb 2016 11:37:59 +0000 http://www.hindawi.com/journals/stni/2016/6980547/ A simple and analytical formula is suggested to solve the problems of the local burnup and the isotope distributions. The present method considers two extreme conditions of neutrons penetrating the fuel rod. Based on these considerations, the formula is obtained to calculate the reaction rates of 235U, 238U, and 239Pu and straightforward the local burnup and the isotope distributions. Starting from an initial burnup level, the parameters of the formula are fitted to the reaction rates given by a Monte Carlo (MC) calculation. Then the present formula independently gives very similar results to the MC calculation from the starting to high burnup level but takes just a few minutes. The relative reaction rates are found to be almost independent of the radius (except of  238U) and the burnup, providing a solid background for the present formula. A more realistic examination is also performed when the fuel rods locate in an assembly. A combination of the present formula and the MC calculation is expected to have a nice balance between the numerical accuracy and time consumption. Cenxi Yuan, Xuming Wang, and Shengli Chen Copyright © 2016 Cenxi Yuan et al. All rights reserved. Study on the Impact of Thermal Agitation on Doppler Coefficient in Epithermal Range for Gd-Bearing Fuel Mon, 18 Jan 2016 07:37:36 +0000 http://www.hindawi.com/journals/stni/2016/5834614/ The impact of thermal agitation on Doppler coefficient for Gd-bearing fuel was analyzed. It was found through the analysis that the impact increases when a small amount of Gd2O3 is added to pure UO2 fuel although the impact decreases for a large amount of Gd2O3. This tendency was discussed with the usage of simplified expression for the difference of Doppler coefficient. The simplified expression was used to consider the tendency, and it was revealed that the tendency mainly comes from the rapid decrement of multiplication factor and the relatively slow decrement of the magnitude of sensitivity coefficient of U-238 capture cross section at low Gd2O3 concentration. Similar tendency which shows a maximum impact on Doppler coefficient at interior concentration is expected for other UO2 fuel with a slight content of strong absorber. This indicates that Doppler coefficient of UO2 fuel system with low content of strong absorber should be analyzed carefully by considering thermal agitation in epithermal range. Satoshi Takeda, Satoshi Ino, Kazuhiro Wada, Michitaka Ono, and Takanori Kitada Copyright © 2016 Satoshi Takeda et al. All rights reserved. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100 Sun, 17 Jan 2016 07:53:23 +0000 http://www.hindawi.com/journals/stni/2016/9612120/ Small modular reactor (SMR) has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR) is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100) is being developed by University of Science and Technology of China (USTC). In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS) transient simulation at beginning of the reactor cycle (BOC) has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance. Pengcheng Zhao, Kangli Shi, Shuzhou Li, Jingchao Feng, and Hongli Chen Copyright © 2016 Pengcheng Zhao et al. All rights reserved. Normal and Accidental Scenarios Analyses with MELCOR 1.8.2 and MELCOR 2.1 for the DEMO Helium-Cooled Pebble Bed Blanket Concept Wed, 30 Dec 2015 06:30:25 +0000 http://www.hindawi.com/journals/stni/2015/865829/ As for Light Water Reactors (LWRs), one of the most challenging accidents for the future DEMOnstration power plant is the Loss of Coolant Accident, which can trigger the pressurization of the confinement structures and components. Hence, careful analyses have to be executed to demonstrate that the confinement barriers are able to withstand the pressure peak within design limits and the residual cooling capabilities of the Primary Heat Transfer System are sufficient to remove the decay heat. To do so, severe accident codes, as MELCOR, can be employed. In detail, the MELCOR code has been developed to cope also with fusion reactors, but unfortunately, these fusion versions are based on the old 1.8.x source code. On the contrary, for LWRs, the newest 2.1.x versions are continuously updated. Thanks to the new features introduced in these latest 2.1.x versions, the main phenomena occurring in the helium-cooled blanket concepts of DEMO can be simulated in a basic manner. For this purpose, several analyses during normal and accidental DEMO conditions have been executed. The aim of these analyses is to compare the results obtained with MELCOR 1.8.2 and MELCOR 2.1 in order to highlight the differences among the results of the main thermal-hydraulic parameters. Bruno Gonfiotti and Sandro Paci Copyright © 2015 Bruno Gonfiotti and Sandro Paci. All rights reserved. Monitoring of 220Rn Concentrations in Buildings of Kufa Technical Institute, Iraq Thu, 17 Dec 2015 09:50:17 +0000 http://www.hindawi.com/journals/stni/2015/738019/ This paper presents the measurements of thoron and the progeny in fifteen buildings in Kufa Technical Institute, Iraq, from June 2015 to April 2015 using RAD-7 detectors. Also, annual effective dose rate was calculated in all buildings under study. The thoron concentration varies from  Bq/m3 to  Bq/m3 with an average  Bq/m3. The concentration of thoron daughters was found to vary from 0.14 mWL to 1.44 mWL with an average  mWL. The annual effective doses due to thoron mainly vary from 0.042 mSv/y to 0.81 mSv/y with an average  mSv/y. The preliminary results in this study indicate that they may be suitable for evaluating the indoor 220Rn and its progeny concentrations whenever the public exposure to 220Rn and its progeny is taken into account. During this survey, the continuous difficulty in measuring thoron was also pointed out, due to its short half-life and faults in the measuring system. Ali Abid Abojassim Al-Hamidawi Copyright © 2015 Ali Abid Abojassim Al-Hamidawi. All rights reserved. Method of Measuring the Efficiency of the Conversion of Nuclear Energy into Optical Energy Tue, 08 Dec 2015 09:03:54 +0000 http://www.hindawi.com/journals/stni/2015/161535/ A method of measuring the efficiency of converting nuclear energy into optical energy was developed based on correlations between intensities of the research line and the nitrogen second positive system in an Ar-N2 mixture. In addition, the values of the coefficient of the conversion of nuclear energy into radiation at the lines of a Hg triplet in mixtures of Хе-Hg and Kr-Hg were determined. The values measured correspond to a selectiveness of pumping of 73S1 that was close to 1 (). Erlan G. Batyrbekov, Yuriy N. Gordienko, Mendykhan U. Khasenov, and Yuriy V. Ponkratov Copyright © 2015 Erlan G. Batyrbekov et al. All rights reserved. Prediction of Flow Regimes and Thermal Hydraulic Parameters in Two-Phase Natural Circulation by RELAP5 and TRACE Codes Sun, 29 Nov 2015 12:03:55 +0000 http://www.hindawi.com/journals/stni/2015/296317/ In earlier study we have demonstrated that RELAP5 can predict flow instability parameters (flow rate, oscillation period, temperature, and pressure) in single channel tests in CIRCUS-IV facility. The main goals of this work are to (i) validate RELAP5 and TRACE capabilities in prediction of two-phase flow instability and flow regimes and (ii) assess the effect of improvement in flow regime identification on code predictions. Most of the results of RELAP5 and TRACE calculation are in reasonable agreement with experimental data from CIRCUS-IV. However, both codes misidentified instantaneous flow regimes which were observed in the test with high speed camera. One of the reasons for the incorrect identification of the flow regimes is the small tube flow regime transition model in RELAP5 and the combined bubbly-slug flow regime in TRACE. We found that calculation results are sensitive to flow regime boundaries of RELAP5 which were modified in order to match the experimental data on flow regimes. Although the flow regime became closer to the experimental one, other predicted thermal hydraulic parameters showed larger discrepancy with the experimental data than with the base case calculations where flow regimes were misidentified. Viet-Anh Phung and Pavel Kudinov Copyright © 2015 Viet-Anh Phung and Pavel Kudinov. All rights reserved. Study of the Reactor Control System of MSHIM in AP1000 Sun, 29 Nov 2015 12:03:35 +0000 http://www.hindawi.com/journals/stni/2015/620205/ According to the mechanism analysis and simulation of power control system of MSHIM in AP1000, a modified MSHIM (Mechanical Shim) control strategy is presented, which employs the error between the reactor coolant average temperature and its reference value as the unique control signal with a P-controller added. The modified MSHIM control strategy is verified by simulations of three typical working conditions. The results show that the modified power control system satisfies the needs of reactor core power control and power distribution control. The conclusions have reference value for the engineering practice. Xinyu Wei and Fuyu Zhao Copyright © 2015 Xinyu Wei and Fuyu Zhao. All rights reserved.