Science and Technology of Nuclear Installations https://www.hindawi.com The latest articles from Hindawi © 2017 , Hindawi Limited . All rights reserved. Dynamic Modeling and Control Characteristics of the Two-Modular HTR-PM Nuclear Plant Mon, 22 May 2017 08:32:03 +0000 http://www.hindawi.com/journals/stni/2017/6298037/ The modular high temperature gas-cooled reactor (MHTGR) is a typical small modular reactor (SMR) with inherent safety feature. Due to its high reactor outlet coolant temperature, the MHTGR can be applied not only for electricity production but also as a heat source for industrial complexes. Through multimodular scheme, that is, the superheated steam flows produced by multiple MHTGR-based nuclear supplying system (NSSS) modules combined together to drive a common thermal load, the inherent safety feature of MHTGR is applicable to large-scale nuclear plants at any desired power ratings. Since the plant power control technique of traditional single-modular nuclear plants cannot be directly applied to the multimodular plants, it is necessary to develop the power control method of multimodular plants, where dynamical modeling, control design, and performance verification are three main aspects of developing plant control method. In this paper, the study in the power control for two-modular HTR-PM plant is summarized, and the verification results based on numerical simulation are given. The simulation results in the cases of plant power step and ramp show that the plant control characteristics are satisfactory. Zhe Dong, Yifei Pan, Maoxuan Song, Xiaojing Huang, Yujie Dong, and Zuoyi Zhang Copyright © 2017 Zhe Dong et al. All rights reserved. The Electric Current Effect on Electrochemical Deconsolidation of Spherical Fuel Elements Thu, 18 May 2017 00:00:00 +0000 http://www.hindawi.com/journals/stni/2017/2126876/ For High-Temperature Gas-Cooled Reactor in China, fuel particles are bonded into spherical fuel elements by a carbonaceous matrix. For the study of fuel failure mechanism from individual fuel particles, an electrochemical deconsolidation apparatus was developed in this study to separate the particles from the carbonaceous matrix by disintegrating the matrix into fine graphite powder. The deconsolidated graphite powder and free particles were characterized by elemental analysis, X-ray photoelectron spectroscopy (XPS), X-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersive spectrometer (EDS), and ceramography. The results showed that the morphology, size distribution, and element content of deconsolidated graphite matrix and free particles were notably affected by electric current intensity. The electrochemical deconsolidation mechanism of spherical fuel element was also discussed. Xiaotong Chen, Zhenming Lu, Hongsheng Zhao, Bing Liu, Junguo Zhu, and Chunhe Tang Copyright © 2017 Xiaotong Chen et al. All rights reserved. Application of ASTEC, MELCOR, and MAAP Computer Codes for Thermal Hydraulic Analysis of a PWR Containment Equipped with the PCFV and PAR Systems Sun, 14 May 2017 07:34:02 +0000 http://www.hindawi.com/journals/stni/2017/8431934/ The integrity of the containment will be challenged during a severe accident due to pressurization caused by the accumulation of steam and other gases and possible ignition of hydrogen and carbon monoxide. Installation of a passive filtered venting system and passive autocatalytic recombiners allows control of the pressure, radioactive releases, and concentration of flammable gases. Thermal hydraulic analysis of the containment equipped with dedicated passive safety systems after a hypothetical station blackout event is performed for a two-loop pressurized water reactor NPP with three integral severe accident codes: ASTEC, MELCOR, and MAAP. MELCOR and MAAP are two major US codes for severe accident analyses, and the ASTEC code is the European code, joint property of Institut de Radioprotection et de Sûreté Nucléaire (IRSN, France) and Gesellschaft für Anlagen und Reaktorsicherheit (GRS, Germany). Codes’ overall characteristics, physics models, and the analysis results are compared herein. Despite considerable differences between the codes’ modelling features, the general trends of the NPP behaviour are found to be similar, although discrepancies related to simulation of the processes in the containment cavity are also observed and discussed in the paper. Siniša Šadek, Davor Grgić, and Zdenko Šimić Copyright © 2017 Siniša Šadek et al. All rights reserved. Effects of Different Operating Temperatures on the Tensile Properties of the Grid Plate Hardfaced with Colmonoy in a Pool Type Sodium Fast Reactor Sun, 30 Apr 2017 10:23:01 +0000 http://www.hindawi.com/journals/stni/2017/5926105/ In sodium-cooled fast reactors (SFRs), the grid plate is a critical component which is made of 316 L(N) SS. It is supported on a core support structure which is also made of 316 L(N) SS. This assembly is immersed in a pool of sodium which acts as a coolant. If there is a direct contact between the grid plate and the flange of core support structure, self-welding takes place between them at the high operating temperature of SFR by a thin sheet of liquid sodium which gets into the gap between them as this sodium acts as a metallic gum. To avoid self-welding, the bottom plate of the grid plate is hardfaced with Colmonoy 5 by PTAW so that the direct contact between those two components is avoided. Due to the difference in coefficients of thermal expansion between the base metal and the coating, the interface is subjected to tensile force which may weaken the bonding strength between them at higher temperatures. Therefore, the weldment should be able to withstand the tensile force at higher operating temperatures for which hot tensile properties of the base metal and the weldment have been determined to study the compatibility between them after hardfacing for the reliable operation of SFR. S. Balaguru, Vela Murali, and P. Chellapandi Copyright © 2017 S. Balaguru et al. All rights reserved. A New Method to Measure Crack Extension in Nuclear Graphite Based on Digital Image Correlation Wed, 26 Apr 2017 00:00:00 +0000 http://www.hindawi.com/journals/stni/2017/2416545/ Graphite components, used as moderators, reflectors, and core-support structures in a High-Temperature Gas-Cooled Reactor, play an important role in the safety of the reactor. Specifically, they provide channels for the fuel elements, control rods, and coolant flow. Fracture is the main failure mode for graphite, and breaching of the above channels by crack extension will seriously threaten the safety of a reactor. In this paper, a new method based on digital image correlation (DIC) is introduced for measuring crack extension in brittle materials. Cross-correlation of the displacements measured by DIC with a step function was employed to identify the advancing crack tip in a graphite beam specimen under three-point bending. The load-crack extension curve, which is required for analyzing the R-curve and tension softening behaviors, was obtained for this material. Furthermore, a sensitivity analysis of the threshold value employed for the cross-correlation parameter in the crack identification process was conducted. Finally, the results were verified using the finite element method. Shigang Lai, Li Shi, Alex Fok, Haiyan Li, Libin Sun, and Zhengming Zhang Copyright © 2017 Shigang Lai et al. All rights reserved. Research on the Computed Tomography Pebble Flow Detecting System for HTR-PM Mon, 24 Apr 2017 00:00:00 +0000 http://www.hindawi.com/journals/stni/2017/5403701/ Pebble dynamics is important for the safe operation of pebble-bed high temperature gas-cooled reactors and is a complicated problem of great concern. To investigate it more authentically, a computed tomography pebble flow detecting (CT-PFD) system has been constructed, in which a three-dimensional model is simulated according to the ratio of 1 : 5 with the core of HTR-PM. A multislice helical CT is utilized to acquire the reconstructed cross-sectional images of simulated pebbles, among which special tracer pebbles are designed to indicate pebble flow. Tracer pebbles can be recognized from many other background pebbles because of their heavy kernels that can be resolved in CT images. The detecting principle and design parameters of the system were demonstrated by a verification experiment on an existing CT system in this paper. Algorithms to automatically locate the three-dimensional coordinates of tracer pebbles and to rebuild the trajectory of each tracer pebble were presented and verified. The proposed pebble-detecting and tracking technique described in this paper will be implemented in the near future. Xin Wan, Ximing Liu, Jichen Miao, Peng Cong, Yuai Zhang, and Zhifang Wu Copyright © 2017 Xin Wan et al. All rights reserved. Physical Analysis of the Initial Core and Running-In Phase for Pebble-Bed Reactor HTR-PM Sun, 16 Apr 2017 09:32:47 +0000 http://www.hindawi.com/journals/stni/2017/8918424/ The pebble-bed reactor HTR-PM is being built in China and is planned to be critical in one or two years. At present, one emphasis of engineering design is to determine the fuel management scheme of the initial core and running-in phase. There are many possible schemes, and many factors need to be considered in the process of scheme evaluation and analysis. Based on the experience from the constructed or designed pebble-bed reactors, the fuel enrichment and the ratio of fuel spheres to graphite spheres are important. In this paper, some relevant physical considerations of the initial core and running-in phase of HTR-PM are given. Then a typical scheme of the initial core and running-in phase is proposed and simulated with VSOP code, and some key physical parameters, such as the maximum power per fuel sphere, the maximum fuel temperature, the refueling rate, and the discharge burnup, are calculated. Results of the physical parameters all satisfy the relevant design requirements, which means the proposed scheme is safe and reliable and can provide support for the fuel management of HTR-PM in the future. Jingyu Zhang, Fu Li, and Yuliang Sun Copyright © 2017 Jingyu Zhang et al. All rights reserved. Source Term Analysis of the Irradiated Graphite in the Core of HTR-10 Thu, 06 Apr 2017 00:00:00 +0000 http://www.hindawi.com/journals/stni/2017/2614890/ The high temperature gas-cooled reactor (HTGR) has potential utilization due to its featured characteristics such as inherent safety and wide diversity of utilization. One distinct difference between HTGR and traditional pressurized water reactor (PWR) is the large inventory of graphite in the core acting as reflector, moderator, or structure materials. Some radionuclides will be generated in graphite during the period of irradiation, which play significant roles in reactor safety, environmental release, waste disposal, and so forth. Based on the actual operation of the 10 MW pebble bed high temperature gas-cooled reactor (HTR-10) in Tsinghua University, China, an experimental study on source term analysis of the irradiated graphite has been done. An irradiated graphite sphere was randomly collected from the core of HTR-10 as sample in this study. This paper focuses on the analytical procedure and the establishment of the analytical methodology, including the sample collection, graphite sample preparation, and analytical parameters. The results reveal that the Co-60, Cs-137, Eu-152, and Eu-154 are the major contributors, while H-3 and C-14 are the dominating emitting nuclides in postirradiation graphite material of HTR-10. The distribution profiles of the above four nuclides are also presented. Xuegang Liu, Xin Huang, Feng Xie, Fuming Jia, Xiaogui Feng, and Hong Li Copyright © 2017 Xuegang Liu et al. All rights reserved. Specific Features of Structural-Phase State and Properties of Reactor Pressure Vessel Steel at Elevated Irradiation Temperature Mon, 03 Apr 2017 00:00:00 +0000 http://www.hindawi.com/journals/stni/2017/1064182/ This paper considers influence of elevated irradiation temperature on structure and properties of 15Kh2NMFAA reactor pressure vessel (RPV) steel. The steel is investigated after accelerated irradiation at 300°C (operating temperature of VVER-1000-type RPV) and 400°C supposed to be the operating temperature of advanced RPVs. Irradiation at 300°C leads to formation of radiation-induced precipitates and radiation defects-dislocation loops, while no carbide phase transformation is observed. Irradiation at a higher temperature (400°C) neither causes formation of radiation-induced precipitates nor provides formation of dislocation loops, but it does increase the number density of the main initial hardening phase—of the carbonitrides. Increase of phosphorus concentration in grain boundaries is more pronounced for irradiation at 400°C as compared to irradiation at 300°C due to influence of thermally enhanced diffusion at a higher temperature. The structural-phase changes determine the changes of mechanical properties: at both irradiation temperatures irradiation embrittlement is mainly due to the hardening mechanism with some contribution of the nonhardening one for irradiation at 400°C. Lack of formation of radiation-induced precipitates at = 400°C provides a small shift (17°C). The obtained results demonstrate that the investigated 15Kh2NMFAA steel may be a promising material for advanced reactors with an elevated operating temperature. E. A. Kuleshova, B. A. Gurovich, E. V. Krikun, A. S. Frolov, D. A. Maltsev, Z. V. Bukina, M. A. Saltykov, and A. G. Balikoev Copyright © 2017 E. A. Kuleshova et al. All rights reserved. Neutron Flux Monitoring Based on Blind Source Separation Algorithms in Moroccan TRIGA MARK II Reactor Thu, 30 Mar 2017 14:05:31 +0000 http://www.hindawi.com/journals/stni/2017/5369614/ We present an overview of fission chamber’s functioning modes, theoretical aspects of the nonnegative matrix factorization methods, and the opportunities that offer neutron data processing in order to achieve neutron flux monitoring tasks. Indeed, it is a part of research project that aimed at applying Blind Source Separation methods for in-core and ex-core neutron flux monitoring while analyzing the outputs of fission chamber. The latter could be used as a key issue for control, fuel management, safety concerns, and material irradiation experiments. The Blind Source Separation methods had been used in many scientific fields such as biomedical engineering and telecommunications. Recently, they were used for gamma spectrometry data processing. The originality of this research work is to apply these powerful methods to process the fission chamber output signals. We illustrated the effectiveness of this tool using simulated fission chamber signals. Hanane Arahmane, El-Mehdi Hamzaoui, and Rajaa Cherkaoui El Moursli Copyright © 2017 Hanane Arahmane et al. All rights reserved. Experimental Study on Interfacial Area Transport of Two-Phase Flow under Vibration Conditions Wed, 22 Mar 2017 08:05:07 +0000 http://www.hindawi.com/journals/stni/2017/5809541/ An experimental study on air-water two-phase flow under vibration condition has been conducted using double-sensor conductivity probe. The test section is an annular geometry with hydraulic diameter of 19.1 mm. The vibration frequency ranges from 0.47 Hz to 2.47 Hz. Local measurements of void fraction, interfacial area concentration (IAC), and Sauter mean diameter have been performed along one radius in the vibration direction. The result shows that local parameters fluctuate continuously around the base values in the vibration cycle. Additional bubble force due to inertia is used to explain lateral bubble motions. The fluctuation amplitudes of local void fraction and IAC increase significantly with vibration frequency. The radial distribution of local parameters at the maximum vibration displacement is specifically analyzed. In the void fraction and IAC profiles, the peak near the inner wall is weakened or even disappearing and a strong peak skewed to outer wall is gradually observed with the increase of vibration frequency. The nondimensional peak void fraction can reach a maximum of 49% and the mean relative variation of local void fraction can increase to more than 29% as the vibration frequency increases to 2.47 Hz. But the increase of vibration frequency does not bring significant change to bubble diameter. Xiu Xiao, Qingzi Zhu, Shao-Wen Chen, Mamoru Ishii, Yajun Zhang, and Haijun Jia Copyright © 2017 Xiu Xiao et al. All rights reserved. Adsorption Behaviors of Cobalt on the Graphite and SiC Surface: A First-Principles Study Mon, 20 Mar 2017 00:00:00 +0000 http://www.hindawi.com/journals/stni/2017/8296387/ Graphite and silicon carbide (SiC) are important materials of fuel elements in High Temperature Reactor-Pebble-bed Modules (HTR-PM) and it is essential to analyze the source term about the radioactive products adsorbed on graphite and SiC surface in HTR-PM. In this article, the adsorption behaviors of activation product Cobalt (Co) on graphite and SiC surface have been studied with the first-principle calculation, including the adsorption energy, charge density difference, density of states, and adsorption ratios. It shows that the adsorption behaviors of Co on graphite and SiC both belong to chemisorption, with an adsorption energy 2.971 eV located at the Hollow site and 6.677 eV located at the hcp-Hollow site, respectively. Combining the charge density difference and density of states, it indicates that the interaction of Co-SiC system is stronger than Co-graphite system. Furthermore, the variation of adsorption ratios of Co on different substrate is obtained by a model of grand canonical ensemble, and it is found that when the temperature is close to 650 K and 1700 K for graphite surface and SiC surface, respectively, the Co adatom on the substrate will desorb dramatically. These results show that SiC layer in fuel element could obstruct the diffusion of Co effectively in normal and accidental operation conditions, but the graphite may become a carrier of Co radioactivity nuclide in the primary circuit of HTR-PM. Wenyi Wang, Chuan Li, Jianzhu Cao, and Chao Fang Copyright © 2017 Wenyi Wang et al. All rights reserved. 3D Nondestructive Visualization and Evaluation of TRISO Particles Distribution in HTGR Fuel Pebbles Using Cone-Beam Computed Tomography Tue, 07 Mar 2017 09:18:13 +0000 http://www.hindawi.com/journals/stni/2017/3857075/ A nonuniform distribution of tristructural isotropic (TRISO) particles within a high-temperature gas-cooled reactor (HTGR) pebble may lead to excessive thermal gradients and nonuniform thermal expansion during operation. If the particles are closely clustered, local hotspots may form, leading to excessive stresses on particle layers and an increased probability of particle failure. Although X-ray digital radiography (DR) is currently used to evaluate the TRISO distributions in pebbles, X-ray DR projection images are two-dimensional in nature, which would potentially miss some details for 3D evaluation. This paper proposes a method of 3D visualization and evaluation of the TRISO distribution in HTGR pebbles using cone-beam computed tomography (CBCT): first, a pebble is scanned on our high-resolution CBCT, and 2D cross-sectional images are reconstructed; secondly, all cross-sectional images are restructured to form the 3D model of the pebble; then, volume rendering is applied to segment and display the TRISO particles in 3D for visualization and distribution evaluation. For method validation, several pebbles were scanned and the 3D distributions of the TRISO particles within the pebbles were produced. Experiment results show that the proposed method provides more 3D than DR, which will facilitate pebble fabrication research and production quality control. Gongyi Yu, Yi Du, Xincheng Xiang, Yuan Liu, Ziqiang Li, and Xiangang Wang Copyright © 2017 Gongyi Yu et al. All rights reserved. Heat Transfer Analysis of Passive Residual Heat Removal Heat Exchanger under Tube outside Boiling Condition Mon, 06 Mar 2017 08:06:11 +0000 http://www.hindawi.com/journals/stni/2017/3497103/ The Passive Residual Heat Removal Heat Exchanger (PRHR HX) is an important part of the Passive Residual Heat Removal System (PRHRS). The C-shaped bundle is being used in the PRHR HX. A test facility of C-shaped tube immerged in a water tank was built to research the heat transfer of the PRHR HX. Through the experiments, three regions were found within a particular period of time during the heating process in the tank: natural convection region, transition region, and saturation boiling region. For the tube outside saturation boiling, comparisons of three different correlations in literatures with the experimental data were carried out. Results show that the Rohsenow correlation provides a best-estimate fit with the experimental results. For the tube outside transition region, a formulation is put forward to reduce error based on the Rohsenow subcooled boiling correlation. Yanbin Liu, Xuesheng Wang, Qiming Men, Xiangyu Meng, and Qing Zhang Copyright © 2017 Yanbin Liu et al. All rights reserved. ARES: A Parallel Discrete Ordinates Transport Code for Radiation Shielding Applications and Reactor Physics Analysis Mon, 06 Mar 2017 07:24:03 +0000 http://www.hindawi.com/journals/stni/2017/2596727/ ARES is a multidimensional parallel discrete ordinates particle transport code with arbitrary order anisotropic scattering. It can be applied to a wide variety of radiation shielding calculations and reactor physics analysis. ARES uses state-of-the-art solution methods to obtain accurate solutions to the linear Boltzmann transport equation. A multigroup discretization is applied in energy. The code allows multiple spatial discretization schemes and solution methodologies. ARES currently provides diamond difference with or without linear-zero flux fixup, theta weighted, directional theta weighted, exponential directional weighted, and linear discontinuous finite element spatial differencing schemes. Discrete ordinates differencing in angle and spherical harmonics expansion of the scattering source are adopted. First collision source method is used to eliminate or mitigate the ray effects. Traditional source iteration and Krylov iterative method preconditioned with diffusion synthetic acceleration are applied to solve the linear system of equations. ARES uses the Koch-Baker-Alcouffe parallel sweep algorithm to obtain high parallel efficiency. Verification and validation for the ARES transport code system have been done by lots of benchmarks. In this paper, ARES solutions to the HBR-2 benchmark and C5G7 benchmarks are in excellent agreement with published results. Numerical results are presented which demonstrate the accuracy and efficiency of these methods. Yixue Chen, Bin Zhang, Liang Zhang, Junxiao Zheng, Ying Zheng, and Cong Liu Copyright © 2017 Yixue Chen et al. All rights reserved. Experimental Studies on Breakup and Fragmentation Behavior of Molten Tin and Coolant Interaction Tue, 28 Feb 2017 13:56:22 +0000 http://www.hindawi.com/journals/stni/2017/4576328/ Jet breakup and fragmentation behavior significantly affect the likelihood (and ultimate strength) of steam explosion, but it is very challenging to assess the potential damage to reactor cavity due to general lack of knowledge regarding jet breakup phenomena. In this study, the METRIC (mechanism study test apparatus for melt-coolant interaction) was launched at Shanghai Jiao Tong University to investigate FCI physics. The first five tests on molten tin and water interactions are analyzed in this paper. Significant breakup and fragmentation were observed without considerable pressure pulse, and intense expansion of droplets in local areas was observed at melt temperature higher than 600°C. The chain interactions of expansion all ceased, however, and there was no energetic steam explosion observed. Quantitative analysis on jet breakup length and debris was studied to investigate the effect of the melt temperature, initial diameter of the jet, and so on. Furthermore, the results of tests were compared with current theories. It is found that melt temperature has strong impact on fragmentation that need to be embodied in advanced fragmentation models. Yankai Li, Zefeng Wang, Meng Lin, Mingjun Zhong, Yueshan Zhou, and Yanhua Yang Copyright © 2017 Yankai Li et al. All rights reserved. HTR-10GT Dual Bypass Valve Control Features and Decoupling Strategy for Power Regulation Mon, 13 Feb 2017 00:00:00 +0000 http://www.hindawi.com/journals/stni/2017/9404636/ HTR-10GT is the development of HTR-10 reactor, which PCU will be a closed Brayton cycle with two-stage compression and heat recuperation. Bypass control method is adopted for rapid power regulation and safety protection. But quick opening of single bypass valve would inevitably lead to temperature shocks in multiple components especially at the reactor inlet and the recuperator core. Based on the regulating characteristics of each possible bypass valve, a dual bypass valve control scheme was proposed along with MIMO decoupling controller designed with diagonal matrix method. The system was modeled with Modelica; the DASSL code was used to solve the Differential and Algebraic Equations during simulations. System’s control characteristic was analyzed with classical linear control theory and theory applied on linearized system model. Further numerical simulations showed that cooperative functioning of two bypass valves could effectively limit the temperature variation during power regulation, while the decoupler could improve the control effect and the stability of the system. The results will be helpful for the future design of the control system of HTR-10GT or other closed Brayton cycle of the same kind. Xiao Li, Xiaoyong Yang, Youjie Zhang, and Jie Wang Copyright © 2017 Xiao Li et al. All rights reserved. RELAP5 Simulation of PKL Facility Experiments under Midloop Conditions Thu, 09 Feb 2017 00:00:00 +0000 http://www.hindawi.com/journals/stni/2017/6140323/ Nuclear power plant risk has to be quantified in full power and in other modes of operation. This latter situation corresponds to low power and shutdown modes of operation in which the residual heat removal (RHR) system is required to extract the heat generated in the core. These accidental sequences are great contributors to the total plant risk. Thus, it is important to analyze the plant behavior to establish the accident mitigation measures required. In this way, PKL facility experimental series were undertaken to analyze the plant behavior in other modes of operation when the RHR is lost. In these experiments, the plant configurations were changed to analyze the influence of steam generators secondary side configurations, the temperature inside the pressurizer, and the inventory level on the plant behavior. Moreover, different accident management measures were proposed in each experiment to reach the conditions to restart the RHR. To understand the physical phenomena that takes place inside the reactor, the experiments are simulated with thermal-hydraulic codes, and this makes it possible to analyze the code capabilities to predict the plant behavior. This work presents the simulation results of four experiments included in PKL experimental series obtained using RELAP5/Mod3.3. J. F. Villanueva, S. Carlos, F. Sanchez-Saez, I. Martón, and S. Martorell Copyright © 2017 J. F. Villanueva et al. All rights reserved. Manufacturing Data Uncertainties Propagation Method in Burn-Up Problems Thu, 26 Jan 2017 00:00:00 +0000 http://www.hindawi.com/journals/stni/2017/7275346/ A nuclear data-based uncertainty propagation methodology is extended to enable propagation of manufacturing/technological data (TD) uncertainties in a burn-up calculation problem, taking into account correlation terms between Boltzmann and Bateman terms. The methodology is applied to reactivity and power distributions in a Material Testing Reactor benchmark. Due to the inherent statistical behavior of manufacturing tolerances, Monte Carlo sampling method is used for determining output perturbations on integral quantities. A global sensitivity analysis (GSA) is performed for each manufacturing parameter and allows identifying and ranking the influential parameters whose tolerances need to be better controlled. We show that the overall impact of some TD uncertainties, such as uranium enrichment, or fuel plate thickness, on the reactivity is negligible because the different core areas induce compensating effects on the global quantity. However, local quantities, such as power distributions, are strongly impacted by TD uncertainty propagations. For isotopic concentrations, no clear trends appear on the results. Thomas Frosio, Thomas Bonaccorsi, and Patrick Blaise Copyright © 2017 Thomas Frosio et al. All rights reserved. Unbalance Compensation of a Full Scale Test Rig Designed for HTR-10GT: A Frequency-Domain Approach Based on Iterative Learning Control Thu, 26 Jan 2017 00:00:00 +0000 http://www.hindawi.com/journals/stni/2017/3126738/ Unbalance vibrations are crucial problems in heavy rotational machinery, especially for the systems with high operation speed, like turbine machinery. For the program of 10 MW High Temperature gas-cooled Reactor with direct Gas-Turbine cycle (HTR-10GT), the rated operation speed of the turbine system is 15000 RPM which is beyond the second bending frequency. In that case, even a small residual mass will lead to large unbalance vibrations. Thus, it is of great significance to study balancing methods for the system. As the turbine rotor is designed to be suspended by active magnetic bearings (AMBs), unbalance compensation could be achieved by adequate control strategies. In the paper, unbalance compensation for the Multi-Input and Multi-Output (MIMO) active magnetic bearing (AMB) system using frequency-domain iterative learning control (ILC) is analyzed. Based on the analysis, an ILC controller for unbalance compensation of the full scale test rig, which is designed for the rotor and AMBs in HTR-10GT, is designed. Simulation results are reported which show the efficiency of the ILC controller for attenuating the unbalance vibration of the full scale test rig. This research can offer valuable design criterion for unbalance compensation of the turbine machinery in HTR-10GT. Ying He, Lei Shi, Zhengang Shi, and Zhe Sun Copyright © 2017 Ying He et al. All rights reserved. Challenge Analysis and Schemes Design for the CFD Simulation of PWR Tue, 24 Jan 2017 00:00:00 +0000 http://www.hindawi.com/journals/stni/2017/5695809/ CFD simulation for a PWR is an important part for the development of Numerical Virtual Reactor (NVR) in Harbin Engineering University of China. CFD simulation can provide the detailed information of the flow and heat transfer process in a PWR. However, a large number of narrow flow channels with numerous complex structures (mixing vanes, dimples, springs, etc.) are located in a typical PWR. To obtain a better CFD simulation, the challenges created by these structural features were analyzed and some quantitative regularity and estimation were given in this paper. It was found that both computing resources and time are in great need for the CFD simulation of a whole reactor. These challenges have to be resolved, so two schemes were designed to assist/realize the reduction of the simulation burden on resources and time. One scheme is used to predict the combined efficiency of the simulation conditions (configuration of computing resources and application of simulation schemes), so it can assist the better choice/decision of the combination of the simulation conditions. The other scheme is based on the suitable simplification and modification, and it can directly reduce great computing burden. Guangliang Chen, Zhijian Zhang, Zhaofei Tian, Lei Li, and Xiaomeng Dong Copyright © 2017 Guangliang Chen et al. All rights reserved. Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U3Si2-FeCrAl Wed, 18 Jan 2017 14:29:13 +0000 http://www.hindawi.com/journals/stni/2017/3146985/ Neutronic performance is investigated for a potential accident tolerant fuel (ATF), which consists of U3Si2 fuel and FeCrAl cladding. In comparison with current UO2-Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U3Si2 has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U3Si2-FeCrAl fuel-cladding system is taken into consideration. Fundamental properties of the suggested fuel-cladding combination are investigated in a fuel assembly. These properties include moderator and fuel temperature coefficients, control rods worth, radial power distribution (in a fuel rod), and different void reactivity coefficients. The present work proves that the new combination has less reactivity variation during its service lifetime. Although, compared with the current system, it has a little larger deviation on power distribution and a little less negative temperature coefficient and void reactivity coefficient and its control rods worth is less important, variations of these parameters are less important during the service lifetime of fuel. Hence, U3Si2-FeCrAl system is a potential ATF candidate from a neutronic view. Shengli Chen and Cenxi Yuan Copyright © 2017 Shengli Chen and Cenxi Yuan. All rights reserved. Evaluation of ACPs in China Fusion Engineering Test Reactor Using CATE 2.1 Code Mon, 09 Jan 2017 12:57:57 +0000 http://www.hindawi.com/journals/stni/2017/2936069/ Activated corrosion products (ACPs) are the dominant radiation hazard in water-cooled fusion reactor under normal operation conditions and directly determine the occupational radiation exposure during operation and maintenance. Recently, the preliminary design of China Fusion Engineering Test Reactor (CFETR) has been just completed. Evaluation of ACPs is an important work for the safety of CFETR. In this paper, the ACPs analysis code CATE 2.1 was used to simulate the spatial distribution of ACPs along the blanket cooling loop of CFETR, in which the influence of adopting different pulse handling methods was researched. At last, the dose rate caused by ACPs around the blanket cooling loop was calculated using the point kernel code ARShield. The results showed that the dose rate under normal operation for 1.2 years at contact is 1.02 mSv/h and at 1 m away from pipe is 0.45 mSv/h. And after shutting down the reactor, there will be a rapid decrease of dose rate, because of the rapid decay of short-lived ACPs. Lu Li, Jingyu Zhang, Qingyang Guo, Xiaokang Zhang, Songlin Liu, and Yixue Chen Copyright © 2017 Lu Li et al. All rights reserved. Separation of Transformers for Class 1E Systems in Nuclear Power Plants Mon, 02 Jan 2017 13:00:39 +0000 http://www.hindawi.com/journals/stni/2017/3976049/ In order to supply electric power to the safety related loads, safety and reliability of onsite power have to be ensured for the safety function performance in nuclear power plants. Even though the existing electric power system of APR1400 meets the requirements of codes regarding Class 1E system, there is a room for improvement in the design margin against the voltage drop and short circuit current. This paper discusses the amount that the voltage drop and short circuit current occur in the existing electric power system of APR1400. Additionally, this paper studies with regard to the improved model that has the extra margin against the high voltage drop and short circuit current by separation of unit auxiliary transformer (UAT) and standby auxiliary transformer (SAT) for the Class 1E loads. The improved model of the electric power system by separation of UAT and SAT has been suggested through this paper. Additionally, effects of reliability and cost caused by the electric power system modification are considered. Sang-Hyun Lee and Choong-Koo Chang Copyright © 2017 Sang-Hyun Lee and Choong-Koo Chang. All rights reserved. Mechanical Properties in Nuclear Installation and the Relevant Measurement Methods Sun, 25 Dec 2016 12:20:04 +0000 http://www.hindawi.com/journals/stni/2016/1948507/ Yan Yang, Alejandro Clausse, Leon Cizelj, Xing Chen, and Parashuram Sahoo Copyright © 2016 Yan Yang et al. All rights reserved. Using CFD as Preventative Maintenance Tool for the Cold Neutron Source Thermosiphon System Mon, 19 Dec 2016 13:55:52 +0000 http://www.hindawi.com/journals/stni/2016/5452085/ The cold neutron source (CNS) system of the Open Pool Australian Light-Water (OPAL) reactor is a 20 L cryogenically cooled liquid deuterium thermosiphon system. The CNS is cooled by forced convective helium which is held at room temperature during stand-by (SO) mode and at ~20 K during normal operation (NO) mode. When helium cooling stops, the reactor is shut down to prevent the moderator cell from overheating. This computational fluid dynamics (CFD) study aims to determine whether the combined effects of conduction and natural convection would provide sufficient cooling for the moderator cell under the influence of reactor decay heat after reactor shutdown. To achieve this, two commercial CFD software packages using an identical CFD mesh were first assessed against an experimental heat transfer test of the CNS. It was found that both numerical models were valid within the bounds of experimental uncertainty. After this, one CFD model was used to simulate the thermosiphon transient condition after the reactor is shut down. Two independent shutdown conditions of different decay-heat power profiles were simulated. It was found that the natural convection and conduction cooling in the thermosiphon were sufficient for dissipating both decay-heat profiles, with the moderator cell remaining below the safe temperature of 200°C. Mark Ho, Yeongshin Jeong, Haneol Park, Guan Heng Yeoh, and Weijian Lu Copyright © 2016 Mark Ho et al. All rights reserved. Sizing of the Vacuum Vessel Pressure Suppression System of a Fusion Reactor Based on a Water-Cooled Blanket, for the Purpose of the Preconceptual Design Mon, 19 Dec 2016 11:11:26 +0000 http://www.hindawi.com/journals/stni/2016/8719695/ A methodology to preliminarily evaluate the size of the suppression tank and the relief pipes for a Vacuum Vessel Pressure Suppression System, to be adopted in a fusion reactor based on a water cooled blanket, is presented. The volume of the ST depends on the total energy of the water cooling system and it can be sized based on a required final pressure at equilibrium, by a simple energy balance. The pressure peak in the VV depends mainly on break area and the flow area of the relief pipes and some suggestions about the method for a preliminarily evaluation of their size are discussed. The computer code CONSEN has been used to perform a parametric study and to verify the methodology. Gianfranco Caruso and Fabio Giannetti Copyright © 2016 Gianfranco Caruso and Fabio Giannetti. All rights reserved. Analysis of Density Wave Oscillations in Helically Coiled Tube Once-Through Steam Generator Thu, 15 Dec 2016 14:26:31 +0000 http://www.hindawi.com/journals/stni/2016/3071686/ Helically coiled tube Once-Through Steam Generator (H-OTSG) is one of the key equipment types for small modular reactors. The flow instability of the secondary side of the H-OTSG is particularly serious, because the working condition is in the range of low and medium pressure. This paper presents research on density wave oscillations (DWO) in a typical countercurrent H-OTSG. Based on the steady-state calculation, the mathematical model of single-channel system was established, and the transfer function was derived. Using Nyquist stability criterion of the single variable, the stability cases were studied with an in-house computer program. According to the analyses, the impact law of the geometrical parameters to the system stability was obtained. RELAP5/MOD3.2 code was also used to simulate DWO in H-OTSG. The theoretical analyses of the in-house program were compared to the simulation results of RELAP5. A correction factor was introduced to reduce the error of RELAP5 when modeling helical geometry. The comparison results agreed well which showed that the correction is effective. Junwei Hao, Yaoli Zhang, Jianxiang Zheng, Zhiwei Zhou, Xuanyu Sheng, Gang Hong, Kai Ye, and Ning Li Copyright © 2016 Junwei Hao et al. All rights reserved. Cost Estimation and Efficiency Analysis of Korean CANDU Spent Fuel Disposal Alternatives in Consideration of Future Price Volatility Thu, 15 Dec 2016 13:21:29 +0000 http://www.hindawi.com/journals/stni/2016/3967572/ In Korea, spent fuel is temporarily stored in spent fuel pools at nuclear reactor sites and it is predicted to become saturated between 2020 and 2024. For this reason, four disposal alternatives (KRS-1, A-KRS-1, A-KRS-21, and A-KRS-22) have been developed in order to carry out the direct disposal of the CANDU spent fuel. The objective of this study is to conduct cost efficiency analysis of the disposal alternatives in consideration of price volatility for the radioactive waste repository. To derive future price volatility, this study used the ARIMA model. As a result, A-KRS-1 is the most efficient in terms of price per bundle using 2015 price. As for the results using ARIMA model, except in the case of KRS-1, the cost per bundle of A-KRS-1, A-KRS-21, and A-KRS-22 is decreased. Cost estimation using ARIMA model shows little change or decreases in cost while cost estimation using inflation rates for 2020 resulted in approximately 7.2% increases compared to 2015 for all options. As for the results of scenario analysis, A-KRS-1 earned 8,160 points, while A-KRS-22 followed closely behind with 7,980 points among the total 24,300 points. The results of this study provide invaluable policy data for any nation considering the construction of spent nuclear fuel repository. Sungsig Bang, Yanghon Chung, Dongphil Chun, Chulhong Kwon, and Sungjun Hong Copyright © 2016 Sungsig Bang et al. All rights reserved. Development of End Plug Welding Technique for SFR Fuel Rod Fabrication Thu, 15 Dec 2016 10:10:55 +0000 http://www.hindawi.com/journals/stni/2016/9549805/ In Korea, R&D on a sodium-cooled fast reactor (SFR) was begun in 1997, as one of the national long-term nuclear R&D programs. As one fuel option for a prototype SFR, a metallic fuel, U-Zr alloy fuel, was selected and is currently being developed. For the fabrication of SFR metallic fuel rods, the end plug welding is a crucial process. The sealing of the end plug to the cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions, and parameters were developed for the end plug welding of SFR metallic fuel rods. A gas tungsten arc welding (GTAW) technique was adopted and the welding joint design was developed. In addition, the optimal welding conditions and parameters were established. Based on the establishment of the welding conditions, the GTAW technique was qualified for the end plug welding of SFR metallic fuel rods. Jung Won Lee, Jong Hwan Kim, Ki Hwan Kim, Jeong Yong Park, and Sung Ho Kim Copyright © 2016 Jung Won Lee et al. All rights reserved.