Science and Technology of Nuclear Installations The latest articles from Hindawi © 2017 , Hindawi Limited . All rights reserved. Development and Application of a New High-Efficiency Sparse Linear System Solver in the Thermal-Hydraulic System Analysis Code Tue, 19 Sep 2017 00:00:00 +0000 This paper presents a faster solver named NRLU (Node Reordering Lower Upper) factorization solver to improve the solution speed for the pressure equations, which are formed by RELAP5/MOD3.3. The NRLU solver uses the oriented graph method and minimal fill-ins rule to reorder the structure of the nonsymmetry sparse pressure matrix. It solves the pressure matrix by LU factorization. Then the solver is embedded into the large scale advanced thermal-hydraulic system analysis program RELAP5/MOD3.3. The comparisons of the original solver and the NRLU solver show that the NRLU solver is faster than the original solver in RELAP5/MOD3.3, and the rate enhancement can be 44.44%. The results also show that the NRLU solver can reduce the number of fill-ins effectively. This can improve the calculation speed. Li Ge, Wei Liu, and Jianqiang Shan Copyright © 2017 Li Ge et al. All rights reserved. Comprehensive Uncertainty Quantification in Nuclear Safeguards Tue, 12 Sep 2017 00:00:00 +0000 Nuclear safeguards aim to confirm that nuclear materials and activities are used for peaceful purposes. To ensure that States are honoring their safeguards obligations, quantitative conclusions regarding nuclear material inventories and transfers are needed. Statistical analyses used to support these conclusions require uncertainty quantification (UQ), usually by estimating the relative standard deviation (RSD) in random and systematic errors associated with each measurement method. This paper has two main components. First, it reviews why UQ is needed in nuclear safeguards and examines recent efforts to improve both top-down (empirical) UQ and bottom-up (first-principles) UQ for calibration data. Second, simulation is used to evaluate the impact of uncertainty in measurement error RSDs on estimated nuclear material loss detection probabilities in sequences of measured material balances. E. Bonner, T. Burr, T. Krieger, K. Martin, and C. Norman Copyright © 2017 E. Bonner et al. All rights reserved. First-Principles Study on Various Point Defects Formed by Hydrogen and Helium Atoms in Tungsten Thu, 07 Sep 2017 08:24:49 +0000 The different point defects formed by two hydrogen atoms or two helium atoms in tungsten were investigated through first-principles calculation. The energetically favorable site for a hydrogen atom is tetrahedral interstitial site while substitutional site is the most preferred site for a helium atom. The formation energies of two hydrogen or helium atoms are determined by their positions, and they are not simply 2 times the formation energy of a single hydrogen or helium atom’s defect. After relaxation, two adjacent hydrogen atoms are away from each other while helium atoms are close to each other. The reasons for the interaction between two hydrogen or helium atoms are also discussed. Qiang Zhao, Zheng Zhang, Yang Li, and Xiaoping Ouyang Copyright © 2017 Qiang Zhao et al. All rights reserved. Quantitative Analysis of Oxygen Gas Exhausted from Anode through In Situ Measurement during Electrolytic Reduction Mon, 21 Aug 2017 00:00:00 +0000 Quantitative analysis by in situ measurement of oxygen gas evolved from an anode was employed to monitor the progress of electrolytic reduction of simulated oxide fuel in a molten Li2O–LiCl salt. The electrolytic reduction of 0.6 kg of simulated oxide fuel was performed in 5 kg of 1.5 wt.% Li2O–LiCl molten salt at 650°C. Porous cylindrical pellets of simulated oxide fuel were used as the cathode by loading a stainless steel wire mesh cathode basket. A platinum plate was employed as the anode. The oxygen gas evolved from the anode was exhausted to the instrumentation for in situ measurement during electrolytic reduction. The instrumentation consisted of a mass flow controller, pump, wet gas meter, and oxygen gas sensor. The oxygen gas was successfully measured using the instrumentation in real time. The measured volume of the oxygen gas was comparable to the theoretically calculated volume generated by the charge applied to the simulated oxide fuel. Eun-Young Choi, Jeong Lee, Dong Hyun Heo, and Jin-Mok Hur Copyright © 2017 Eun-Young Choi et al. All rights reserved. CFD Analysis of a Decay Tank and a Siphon Breaker for an Innovative Integrated Passive Safety System for a Research Reactor Thu, 10 Aug 2017 00:00:00 +0000 An innovative integrated passive safety system for a research reactor is proposed in this study to improve the safety of the research reactor. This integrated system has three functions in the facility as a decay tank, siphon breaker, and long-term cooling tank. This paper also deals with the process of designing and optimizing the decay tank and the siphon breaker of the integrated passive safety system. At first, the decay tank was designed and improved step by step, while considering the computational fluid dynamics analysis results. Consequently, we could satisfy the design requirements of the decay tank. In addition, the performance of a new type of siphon breaker that was installed in the final decay tank model was tested. We designed an 18-inch diameter siphon breaker at the top of the decay tank’s third section, and we could observe the breaking of the siphon that prevented the occurrence of a severe accident in the research reactor. By locating the siphon breaker at the third section of the decay tank, we could also use the coolant of the front three sections for long-term cooling of the research reactor. Kwon-Yeong Lee, Hyun-Gi Yoon, and Dong Kyou Park Copyright © 2017 Kwon-Yeong Lee et al. All rights reserved. Probabilistic Dose Assessment from SB-LOCA Accident in Ujung Lemahabang Using TMI-2 Source Term Wed, 09 Aug 2017 00:00:00 +0000 Probabilistic dose assessment and mapping for nuclear accident condition are performed for Ujung Lemahabang site in Muria Peninsula region in Indonesia. Source term is obtained from Three-Mile Island unit 2 (TMI-2) PWR-type SB-LOCA reactor accident inverse modeling. Effluent consisted of Xe-133, Kr-88, I-131, and Cs-137 released from a 50 m stack. Lagrangian Particle Dispersion Method (LPDM) and 3-dimensional mass-consistent wind field are employed to obtain surface-level time-integrated air concentration and spatial distribution of ground-level total dose in dry condition. Site-specific meteorological data is obtained from hourly records obtained during the Site Feasibility Study period in Ujung Lemahabang. Effluent is released from a height of 50 meters in uniform rate during a 6-hour period and the dose is integrated during this period in a neutrally stable atmospheric condition. Maximum dose noted is below regulatory limit of 1 mSv and radioactive plume is spread mostly to the W-SW inland and to N-NE from the proposed plant to Java Sea. This paper has demonstrated for the first time a probabilistic analysis method for assessing possible spatial dose distribution, a hypothetical release, and a set of meteorological data for Ujung Lemahabang region. Sunarko, Zaki Su’ud, Idam Arif, and Syeilendra Pramuditya Copyright © 2017 Sunarko et al. All rights reserved. Oxidation Behavior of Matrix Graphite and Its Effect on Compressive Strength Tue, 01 Aug 2017 09:34:34 +0000 Matrix graphite (MG) with incompletely graphitized binder used in high-temperature gas-cooled reactors (HTGRs) is commonly suspected to exhibit lower oxidation resistance in air. In order to reveal the oxidation performance, the oxidation behavior of newly developed A3-3 MG at the temperature range from 500 to 950°C in air was studied and the effect of oxidation on the compressive strength of oxidized MG specimens was characterized. Results show that temperature has a significant influence on the oxidation behavior of MG. The transition temperature between Regimes I and II is ~700°C and the activation energy () in Regime I is around 185 kJ/mol, a little lower than that of nuclear graphite, which indicates MG is more vulnerable to oxidation. Oxidation at 550°C causes more damage to compressive strength of MG than oxidation at 900°C. Comparing with the strength of pristine MG specimens, the rate of compressive strength loss is 77.3% after oxidation at 550°C and only 12.5% for oxidation at 900°C. Microstructure images of SEM and porosity measurement by Mercury Porosimetry indicate that the significant compressive strength loss of MG oxidized at 550°C may be attributed to both the uniform pore formation throughout the bulk and the preferential oxidation of the binder. Xiangwen Zhou, Cristian I. Contescu, Xi Zhao, Zhenming Lu, Jie Zhang, Yutai Katoh, Yanli Wang, Bing Liu, Yaping Tang, and Chunhe Tang Copyright © 2017 Xiangwen Zhou et al. All rights reserved. The Optimization of Radiation Protection in the Design of the High Temperature Reactor-Pebble-Bed Module Mon, 31 Jul 2017 09:21:50 +0000 The optimization of radiation protection is an important task in both the design and operation of a nuclear power plant. Although this topic has been considerably investigated for pressurized water reactors, there are very few public reports on it for pebble-bed reactors. This paper proposes a routine that jointly optimizes the system design and radiation protection of High Temperature Reactor-Pebble-Bed Module (HTR-PM) towards the As Low As Reasonably Achievable (ALARA) principle. A systematic framework is also established for the optimization of radiation protection for pebble-bed reactors. Typical calculations for the radiation protection of radioactivity-related systems are presented to quantitatively evaluate the efficiency of the optimization routine, which achieve 23.3%~90.6% reduction of either dose rate or shielding or both of them. The annual collective doses of different systems are reduced through iterative optimization of the dose rates, designs, maintenance procedures, and work durations and compared against the previous estimates. The comparison demonstrates that the annual collective dose of HTR-PM is reduced from 0.490 man-Sv/a before optimization to 0.445 man-Sv/a after optimization, which complies with the requirements of the Chinese regulatory guide and proves the effectiveness of the proposed routine and framework. Sida Sun, Hong Li, and Sheng Fang Copyright © 2017 Sida Sun et al. All rights reserved. Source Term Study on Tritium in HTR-PM: Theoretical Calculations and Experimental Design Mon, 31 Jul 2017 00:00:00 +0000 The high temperature gas-cooled reactor pebble-bed module (HTR-PM) in China received much attention for its inherent safety performance and high thermal efficiency. The generation mechanism, distribution, reduction route, and release type of tritium (H-3) in HTR-PM are presented with a complete theoretical model. The calculation results indicate the majority of H-3 in the core is generated by the activation reaction of B-10. The activity concentration of H-3 in the primary loop and the specific activity of H-3 in the secondary loop at the operating equilibrium are computed as 3.69 × 106  of helium and 4.22 × 104 Bq/kg of water, respectively. The H-3 sampling measurement in HTR-PM has been designed to collect data from the primary coolant, from the liquid waste storage tank, from the secondary coolant, and from the liquid and gaseous effluents, separately. In this paper, the design of H-3 sampling positions in the helium purification system is discussed. The H-3 sampling measurement from the primary helium in HTR-PM has been improved, which can provide reliable activity concentration data of H-3 in the primary loop and supply accurate evaluation for the efficiency of the helium purification system. Jianzhu Cao, Liguo Zhang, Feng Xie, Bing Xia, and Stephen Tsz Tang Lam Copyright © 2017 Jianzhu Cao et al. All rights reserved. Analysis of Precooling Injection Transient of Steam Generator for High Temperature Gas Cooled Reactor Thu, 27 Jul 2017 06:38:14 +0000 After a postulated design basis accident leads high temperature gas cooled reactor to emergency shutdown, steam generator still remains with high temperature level and needs to be cooled down by a precooling before reactor restarts with clearing of fault. For the large difference of coolant temperature between inlet and outlet of steam generator in normal operation, the temperature distribution on the components of steam generator is very complicated. Therefore, the temperature descending rate of the components in steam generator needs to be limited to avoid the potential damage during the precooling stage. In this paper, a pebble-bed high temperature gas cooled reactor is modeled by thermal-hydraulic system analysis code and several postulated precooling injection transients are simulated and compared to evaluate their effects, which will provide support for the precooling design. The analysis results show that enough precooling injection is necessary to satisfy the precooling requirements, and larger mass flow rate of precooling water injection will accelerate the precooling process. The temperature decrease of steam generator is related to the precooling injection scenarios, and the maximal mass flow rate of the precooling injection should be limited to avoid the excessively quick temperature change of the structures in steam generator. Yan Wang, Lei Shi, and Yanhua Zheng Copyright © 2017 Yan Wang et al. All rights reserved. Nuclear Waste Management Decision-Making Support with MCDA Tue, 25 Jul 2017 00:00:00 +0000 The paper proposes a multicriteria decision analysis (MCDA) framework for a comparative evaluation of nuclear waste management strategies taking into account different local perspectives (expert and stakeholder opinions). Of note, a novel approach is taken using a multiple-criteria formulation that is methodologically adapted to tackle various conflicting criteria and a large number of expert/stakeholder groups involved in the decision-making process. The purpose is to develop a framework and to show its application to qualitative comparison and ranking of options in a hypothetical case of three waste management alternatives: interim storage at and/or away from the reactor site for the next 100 years, interim decay storage followed in midterm by disposal in a national repository, and disposal in a multinational repository. Additionally, major aspects of a decision-making aid are identified and discussed in separate paper sections dedicated to application context, decision supporting process, in particular problem structuring, objective hierarchy, performance evaluation modeling, sensitivity/robustness analyses, and interpretation of results (practical impact). The aim of the paper is to demonstrate the application of the MCDA framework developed to a generic hypothetical case and indicate how MCDA could support a decision on nuclear waste management policies in a “small” newcomer country embarking on nuclear technology in the future. A. Schwenk-Ferrero and A. Andrianov Copyright © 2017 A. Schwenk-Ferrero and A. Andrianov. All rights reserved. Analyses of the TIARA 43 MeV Proton Benchmark Shielding Experiments Using the ARES Transport Code Mon, 24 Jul 2017 00:00:00 +0000 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. To validate the applicability of the code to accelerator shielding problems, ARES is adopted to simulate a series of accelerator shielding experiments for 43 MeV proton-7Li quasi-monoenergetic neutrons, which is performed at Takasaki Ion Accelerator for Advanced Radiation Application. These experiments on iron and concrete were analyzed using the ARES code with FENDL/MG-3.0 multigroup libraries and compared to direct measurements from the BC501A detector. The simulations show good agreement with the experimental data. The ratios of calculated values to experimental data for integrated neutron flux at peak and continuum energy regions are within 64% and 25% discrepancy for the concrete and iron experiments, respectively. The results demonstrate the accuracy and efficiency of ARES code for accelerator shielding calculation. Bin Zhang, Liang Zhang, and Yixue Chen Copyright © 2017 Bin Zhang et al. All rights reserved. Assessment of the MARS Code Using the Two-Phase Natural Circulation Experiments at a Core Catcher Test Facility Mon, 17 Jul 2017 00:00:00 +0000 A core catcher has been developed to maintain the integrity of nuclear reactor containment from molten corium during a severe accident. It uses a two-phase natural circulation for cooling molten corium. Flow in a typical core catcher is unique because (i) it has an inclined cooling channel with downwards-facing heating surface, of which flow processes are not fully exploited, (ii) it is usually exposed to a low-pressure condition, where phase change causes dramatic changes in the flow, and (iii) the effects of a multidimensional flow are very large in the upper part of the core catcher. These features make computational analysis more difficult. In this study, the MARS code is assessed using the two-phase natural circulation experiments that had been conducted at the CE-PECS facility to verify the cooling performance of a core catcher. The code is a system-scale thermal-hydraulic (TH) code and has a multidimensional TH component. The facility was modeled by using both one- and three-dimensional components. Six experiments at the facility were selected to investigate the parametric effects of heat flux, pressure, and form loss. The results show that MARS can predict the two-phase flow at the facility reasonably well. However, some limitations are obviously revealed. Dong Hun Lee, Su Ryong Choi, Kwang Soon Ha, Han Young Yoon, and Jae Jun Jeong Copyright © 2017 Dong Hun Lee et al. All rights reserved. Sensitivity Evaluation of AP1000 Nuclear Power Plant Best Estimation Model Wed, 28 Jun 2017 00:00:00 +0000 The best estimation process of AP1000 Nuclear Power Plant (NPP) requires proper selections of parameters and models so as to obtain the most accurate results compared with the actual design parameters. Therefore, it is necessary to identify and evaluate the influences of these parameters and modeling approaches quantitatively and qualitatively. Based on the best estimate thermal-hydraulic system code RELAP5/MOD3.2, sensitivity analysis has been performed on core partition methods, parameters, and model selections in AP1000 Nuclear Power Plant, like the core channel number, pressurizer node number, feedwater temperature, and so forth. The results show that core channel number, core channel node number, and the pressurizer node number have apparent influences on the coolant temperature variation and pressure drop through the reactor. The feedwater temperature is a sensitive factor to the Steam Generator (SG) outlet temperature and the Steam Generator outlet pressure. In addition, the cross-flow model nearly has no effects on the coolant temperature variation and pressure drop in the reactor, in both the steady state and the loss of power transient. Furthermore, some fittest parameters with which the most accurate results could be obtained have been put forward for the nuclear system simulation. Hao Shi, Qi Cai, and Yuqing Chen Copyright © 2017 Hao Shi et al. All rights reserved. Design and Development Framework of Safety-Critical Software in HTR-PM Wed, 28 Jun 2017 00:00:00 +0000 With the development of information technology, the instrumentation and control system of nuclear power plant nowadays rely heavily on the massive and complex software to ensure the safe and efficient operation of the power plant. The improvement of the software design and development for the safety systems has been a research focus for its decisive impact on the nuclear safety. The framework of the software design and development for reactor protection system in High Temperature Gas-Cooled Reactor-Pebble bed Module was introduced in this paper. Firstly, during the design period, in addition to multichannel redundancy, grouping of protection variables and diverse 2-out-of-4 logics were adopted by different subsystems of each channel in case of common cause failure. Then a series of development characteristics together with strict software verification and validation were performed. Thirdly, during the software test period, an improved software reliability growth model based on the Goel-Okumoto model according to the analysis of fault severity was proposed to help in estimating the reliability of the software product and identifying the software release time. Chao Guo, Huasheng Xiong, Xiaojin Huang, and Duo Li Copyright © 2017 Chao Guo et al. All rights reserved. R&D on a Nonlinear Dynamics Analysis Code for the Drop Time of the Control Rod Mon, 19 Jun 2017 08:12:26 +0000 Whether the control rod can drop down in time is one of the important guarantees for the safe operation of the nuclear power plant. The drop-down process of the control rod is very complicated. For a long time, the researchers have done a lot of work on that, but it is hard to consider all the nonlinear factors. This paper considers the main factors together. Based on the theoretical analysis, we developed the nonlinear dynamics response analysis software for the nuclear power plant, which can be used to calculate the rod’s drop-down time. Compared with the results of the experiments, the software we developed proves to be applicable and reliable. Daogang Lu, Yuanpeng Wang, Qingyu Xie, Huimin Zhang, and Muhammed Ali Copyright © 2017 Daogang Lu et al. All rights reserved. Simulation and Analysis of Small Break LOCA for AP1000 Using RELAP5-MV and Its Comparison with NOTRUMP Code Mon, 19 Jun 2017 07:30:39 +0000 Many reactor safety simulation codes for nuclear power plants (NPPs) have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions. In reactor analysis, small break loss of coolant accident (SBLOCA) is an important safety issue. RELAP5-MV Visualized Modularization software is recognized as one of the best estimate transient simulation programs of light water reactors (LWR). RELAP5-MV has new options for improved modeling methods and interactive graphics display. Though the same models incorporated in RELAP5/MOD 4.0 are in RELAP5-MV, the significant difference of the latter is the interface for preparing the input deck. In this paper, RELAP5-MV is applied for the transient analysis of the primary system variation of thermal hydraulics parameters in primary loop under SBLOCA in AP1000 NPP. The upper limit of SBLOCA (10 inches) is simulated in the cold leg of the reactor and the calculations performed up to a transient time of 450,000.0 s. The results obtained from RELAP5-MV are in good agreement with those of NOTRUMP code obtained by Westinghouse when compared under the same conditions. It can be easily inferred that RELAP5-MV, in a similar manner to RELAP5/MOD4.0, is suitable for simulating a SBLOCA scenario. Eltayeb Yousif, Zhijian Zhang, Zhaofei Tian, and Hao-ran Ju Copyright © 2017 Eltayeb Yousif et al. All rights reserved. Analysis of Seismic Soil-Structure Interaction for a Nuclear Power Plant (HTR-10) Tue, 13 Jun 2017 00:00:00 +0000 The response of nuclear power plants (NPPs) to seismic events is affected by soil-structure interactions (SSI). In the present paper, a finite element (FE) model with transmitting boundaries is used to analyse the SSI effect on the response of NPP buildings subjected to vertically incident seismic excitation. Analysis parameters that affect the accuracy of the calculations, including the dimension of the domain and artificial boundary types, are investigated through a set of models. A numerical SSI analysis for the 10 MW High Temperature Gas Cooled Test Reactor (HTR-10) under seismic excitation was carried out using the developed model. The floor response spectra (FRS) produced by the SSI analysis are compared with a fixed-base model to investigate the SSI effect on the dynamic response of the reactor building. The results show that the FRS at foundation level are reduced and those at higher floor levels are altered significantly when taking SSI into account. The peak frequencies of the FRS are reduced due to the SSI, whereas the acceleration at high floor levels is increased at a certain frequency range. The seismic response of the primary system components, however, is reduced by the analysed SSI for the HTR-10 on the current soil site. Xiaoxin Wang, Qin Zhou, Kaixin Zhu, Li Shi, Xiaotian Li, and Haitao Wang Copyright © 2017 Xiaoxin Wang et al. All rights reserved. Temperature Dependence of Thermophysical Properties of Full-Scale Corium of Fast Energy Reactor Mon, 12 Jun 2017 10:06:34 +0000 For the first time the paper determines thermophysical properties (specific heat capacity, thermal diffusivity, and heat conductivity) of the full-scale corium of the fast energy nuclear reactor within the temperature range from ~30°С to ~400°С. Obtained data are to be used in temperature fields calculations during modeling the processes of corium melt retention inside of the fast reactor vessel. Mazhyn K. Skakov, Nurzhan Ye. Mukhamedov, Alexander D. Vurim, and Ilya I. Deryavko Copyright © 2017 Mazhyn K. Skakov et al. All rights reserved. FM-DBEM Simulation of 3D Microvoid and Microcrack Graphite Models Tue, 06 Jun 2017 08:20:35 +0000 The graphite is porous medium, and the geometry and size distribution of its structural deficiencies such as microcracks and microvoids at different oxidation degrees have a great influence on the overall performance. In this paper, we adopt the FM-DBEM to study 3D models which contain spheroidal microvoids and circular microcracks. The accuracy of this method is tested by a comparison to the theoretical solution to the problem of 2D microcrack and microvoid interaction problem. Two simulations are conducted: the simulation of graphite model containing a large number of randomly distributed microcracks and microvoids and the simulation of graphite model containing microcracks and growing microvoids. The simulations investigate the effective moduli versus the two microstructures’ density and the effect of microvoid’s growth on the SIF of microcrack. Houdi Lu, Hongtao Wang, Haitao Wang, Lie Jin, Xinxin Wu, and Yu Zhou Copyright © 2017 Houdi Lu et al. All rights reserved. Three Design Basis Accidents’ Analysis on the HTR-10GT Mon, 05 Jun 2017 10:45:53 +0000 The study simulated the design basis accidents (DBAs) sequences of the HTR-10GT core with THERMIX. When a DBA happens, the protection system will receive a scram signal which shall lead to active measures to shut down the reactor following it. In the paper, three typical DBA cases were studied. They include an accident induced by station blackout, a case caused by the withdrawal of one control rod out of the core by a mistake, and a case resulting from an earthquake, respectively. The simulation results illustrate that the fuel peak temperatures in the core during these accidents are 1066°C, 1201°C and 1067°C, respectively. It is shown that the HTR-10GT has a good safety characteristic. Minggang Lang, Heng Xie, and Yujie Dong Copyright © 2017 Minggang Lang et al. All rights reserved. Preliminary CFD Assessment of an Experimental Test Facility Operating with Heavy Liquid Metals Thu, 01 Jun 2017 06:36:38 +0000 The CFD analysis of a Venturi nozzle operating in LBE (key component of the CIRCE facility, owned by ENEA) is presented in this paper. CIRCE is a facility developed to investigate in detail the fluid-dynamic behavior of ADS and/or LFR reactor plants. The initial CFD simulations have been developed hand in hand with the comparison with experimental data: the test results were used to confirm the reliability of the CFD model, which, in turn, was used to improve the interpretation of the experimental data. The Venturi nozzle is modeled with a 3D CFD code (STAR-CCM+). Later on, the CFD model has been used to assess the performance of the component in conditions different from the ones tested in CIRCE: the performance of the Venturi is presented, in terms of pressure drops, for various operating conditions. Finally, the CFD analysis has been focused on the evaluation of the effects of the injection of an inert gas in the flow of the liquid coolant on the performance of the Venturi nozzle. Matteo Lizzoli, Walter Borreani, Francesco Devia, Guglielmo Lomonaco, and Mariano Tarantino Copyright © 2017 Matteo Lizzoli et al. All rights reserved. Using Wireless Sensor Networks to Achieve Intelligent Monitoring for High-Temperature Gas-Cooled Reactor Tue, 30 May 2017 00:00:00 +0000 High-temperature gas-cooled reactors (HTGR) can incorporate wireless sensor network (WSN) technology to improve safety and economic competitiveness. WSN has great potential in monitoring the equipment and processes within nuclear power plants (NPPs). This technology not only reduces the cost of regular monitoring but also enables intelligent monitoring. In intelligent monitoring, large sets of heterogeneous data collected by the WSN can be used to optimize the operation and maintenance of the HTGR. In this paper, WSN-based intelligent monitoring schemes that are specific for applications of HTGR are proposed. Three major concerns regarding wireless technology in HTGR are addressed: wireless devices interference, cybersecurity of wireless networks, and wireless standards selected for wireless platform. To process nonlinear and non-Gaussian data obtained by WSN for fault diagnosis, novel algorithms combining Kernel Entropy Component Analysis (KECA) and support vector machine (SVM) are developed. Jianghai Li, Jia Meng, Xiaojing Kang, Zhenhai Long, and Xiaojin Huang Copyright © 2017 Jianghai Li et al. All rights reserved. Dynamic Modeling and Control Characteristics of the Two-Modular HTR-PM Nuclear Plant Mon, 22 May 2017 08:32:03 +0000 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 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 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 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 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 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 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.