Science and Technology of Nuclear Installations The latest articles from Hindawi Publishing Corporation © 2015 , Hindawi Publishing Corporation . All rights reserved. Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems Sun, 26 Jul 2015 11:57:34 +0000 A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2. Wonkyeong Kim, Jinsu Park, Tomasz Kozlowski, Hyun Chul Lee, and Deokjung Lee Copyright © 2015 Wonkyeong Kim et al. All rights reserved. Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL Wed, 01 Jul 2015 10:35:47 +0000 In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor. A. Rais, D. Siefman, G. Girardin, M. Hursin, and A. Pautz Copyright © 2015 A. Rais et al. All rights reserved. Properties of Neutron Noise Induced by Localized Perturbations in an SFR Wed, 24 Jun 2015 09:00:34 +0000 Investigation of the properties of neutron noise induced by localized perturbations in a sodium-cooled fast reactor has been performed using a multigroup neutron noise simulator. Three representations of the noise source associated with the perturbations of absorption, fission, and scattering cross sections, respectively, were assumed to be located at the first fuel ring around the central assembly. The energy- and space-dependent noise, that is, the amplitude and the phase, was calculated in a wide range of frequencies, for example, 0.1–100 Hz. The results show that in the important energy range (>1.0 keV) where the noise amplitude is significant the phase is almost constant with energy at the calculated frequencies despite the source types. At low frequencies, the variation of the phase is negligibly small at a large distance from the source. The perturbation in several fast groups has a significant contribution and dominates the amplitude and the phase of the induced noise. Hoai-Nam Tran Copyright © 2015 Hoai-Nam Tran. All rights reserved. Full Scope Modeling and Analysis on the Secondary Circuit of Chinese Large-Capacity Advanced PWR Based on RELAP5 Code Wed, 17 Jun 2015 11:51:12 +0000 Chinese large-capacity advanced PWR under construction in China is a new and indispensable reactor type in the developing process of NPP fields. At the same time of NPP construction, accident sequences prediction and operators training are in progress. Since there are some possible events such as feedwater pumps trip in secondary circuit may lead to severe accident in NPP, training simulators and engineering simulators of CI are necessary. And, with an increasing proportion of nuclear power in China, NPP will participate in regulating peak load in power network, which requires accuracy calculation and control of secondary circuit. In order to achieve real-time and full scope simulation in the power change transient and accident scenarios, RELAP5/MOD 3.4 code has been adopted to model the secondary circuit for its advantage of high calculation accuracy. This paper describes the model of steady state and turbine load transient from 100% to 40% of secondary circuit using RELAP5 and provides a reasonable equivalent method to solve the calculation divergence problem caused by dramatic two-phase condition change while guaranteeing the heat transfer efficiency. The validation of the parameters shows that all the errors between the calculation values and design values are reasonable and acceptable. Dao-gang Lu, Fan Zhang, Dan-ting Sui, Xue-zhang Xi, and Lei-bo Yu Copyright © 2015 Dao-gang Lu et al. All rights reserved. Analysis of Wetting Characteristics on Microstructured Hydrophobic Surfaces for the Passive Containment Cooling System Thu, 11 Jun 2015 07:52:12 +0000 As the heat transfer surface in the passive containment cooling system, the anticorrosion coating (AC) of steel containment vessel (CV) must meet the requirements on heat transfer performance. One of the wall surface ACs with simple structure, high mechanical strength, and well hydrophobic characteristics, which is conductive to form dropwise condensation, is significant for the heat removal of the CV. In this paper, the grooved structures on silicon wafers by lithographic methods are systematically prepared to investigate the effects of microstructures on the hydrophobic property of the surfaces. The results show that the hydrophobicity is dramatically improved in comparison with the conventional Wenzel and Cassie-Baxter model. In addition, the experimental results are successfully explained by the interface state effect. As a consequence, it is indicated that favorable hydrophobicity can be obtained even if the surface is with lower roughness and without any chemical modifications, which provides feasible solutions for improving the heat transfer performance of CV. Wei Zhao, Xiang Zhang, Chunlai Tian, and Zhan Gao Copyright © 2015 Wei Zhao et al. All rights reserved. Multilayered Pipe Cutting Test for Remote Handling Maintenance Sun, 07 Jun 2015 08:24:12 +0000 Based on the requirements for remote handling maintenance (RHM) of China Spallation Neutron Source (CSNS) multilayered pipes, pipes cutting tests were performed under remote handling maintenance conditions. In this study, the results were obtained from different cutting directions and supporting intensities of pipe baseplates comparisons: When enough power was provided and the blade gripper did not slip, the cutting direction had little impact on the cutting capacity but more on the fault surface; the clearance between the blades caused the rotating torque; for remote handling maintenance, good horizontal support of the long-handled lever of the hydraulic cutter was required. Significant conclusions were made for multilayered pipe cutting that are crucial for auxiliary tools development for remote handling maintenance. Haibin Chen, Jianwen Guo, Zhenzhong Sun, Xuejun Jia, and Hong Tang Copyright © 2015 Haibin Chen et al. All rights reserved. Heat Transfer Calculation on Plate-Type Fuel Assembly of High Flux Research Reactor Mon, 25 May 2015 08:53:27 +0000 Heat transfer characteristics of fuel assemblies for a high flux research reactor with a neutron trap are numerically investigated in this study. Single-phase turbulence flow is calculated by a commercial code, FLUENT, where the computational objective covers standard and control fuel assemblies. The simulation is carried out with an inlet coolant velocity varying from 4.5 m/s to 7.5 m/s in hot assemblies. The results indicate that the cladding temperature is always lower than the saturation temperature in the calculated ranges. The temperature rise in the control fuel assembly is smaller than that of the standard fuel assembly. Additionally, the assembly with a hot spot is specially studied, and the safety of the research reactor is also approved. Daxin Gong, Shanfang Huang, Guanbo Wang, and Kan Wang Copyright © 2015 Daxin Gong et al. All rights reserved. Transuranics Transmutation Using Neutrons Spectrum from Spallation Reactions Wed, 20 May 2015 11:39:22 +0000 The aim is to analyse the neutron spectrum influence in a hybrid system ADS-fission inducing transuranics (TRUs) transmutation. A simple model consisting of an Accelerator-Driven Subcritical (ADS) system containing spallation target, moderator or coolant, and spheres of actinides, “fuel,” at different locations in the system was modelled. The simulation was performed using the MCNPX 2.6.0 particles transport code evaluating capture and fission reactions, as well as the burnup of actinides. The goal is to examine the behaviour and influences of the hard neutron spectrum from spallation reactions in the transmutation, without the contribution or interference of multiplier subcritical medium, and compare the results with those obtained from the neutron fission spectrum. The results show that the transmutation efficiency is independent of the spallation target material used, and the neutrons spectrum from spallation does not contribute to increased rates of actinides transmutation even in the vicinity of the target. Maurício Gilberti, Claubia Pereira, Maria Auxiliadora F. Veloso, and Antonella Lombardi Costa Copyright © 2015 Maurício Gilberti et al. All rights reserved. CANDU Safety R&D Status, Challenges, and Prospects in Canada Mon, 04 May 2015 13:57:35 +0000 In Canada, safe operation of CANDU (CANada Deuterium Uranium; it is a registered trademark of Atomic Energy of Canada Limited) reactors is supported by a full-scope program of nuclear safety research and development (R&D) in key technical areas. Key nuclear R&D programs, facilities, and expertise are maintained in order to address the unique features of the CANDU as well as generic technology areas common to CANDU and LWR (light water reactor). This paper presents an overview of the CANDU safety R&D which includes background, drivers, current status, challenges, and future directions. This overview of the Canadian nuclear safety R&D programs includes those currently conducted by the COG (CANDU Owners Group), AECL (Atomic Energy of Canada Limited), Candu Energy Inc., and the CNSC (Canadian Nuclear Safety Commission) and by universities via UNENE (University Network of Excellence in Nuclear Engineering) sponsorship. In particular, the nuclear safety R&D program related to the emerging CANDU ageing issues is discussed. The paper concludes by identifying directions for the future nuclear safety R&D. W. Shen and F. Doyle Copyright © 2015 W. Shen and F. Doyle. All rights reserved. Void Reactivity Coefficient Analysis during Void Fraction Changes in Innovative BWR Assemblies Thu, 30 Apr 2015 12:31:55 +0000 The study of the void reactivity variation in innovative BWR fuel assemblies is presented in this paper. The innovative assemblies are loaded with high enrichment fresh UO2 and MOX fuels. UO2 fuel enrichment is increased above existing design limitations for LWR fuels (>5%). MOX fuel enrichment with fissile Pu content is established to achieve the same burnup level as that of high enrichment UO2 fuel. For the numerical analysis, the TRITON functional module of SCALE 6.1 code with the 238-group ENDF/B-VI cross section data library was applied. The investigation of the void reactivity feedback is performed in the entire 0–100% void fraction range. Higher values of void reactivity coefficient for assembly loaded with MOX fuel are found in comparison with values for assembly loaded with UO2 fuel. Moreover, coefficient values for MOX fuel are positive over 75% void fraction. The variation of the void reactivity coefficient is explained by the results of the decomposition analysis based on four-factor formula and neutron absorption reactions for main isotopes. Additionally, the impact of the moderation enhancement on the void reactivity coefficient was investigated for the innovative assembly with MOX fuel. Andrius Slavickas, Raimondas Pabarčius, Aurimas Tonkūnas, Gediminas Stankūnas, and Eugenijus Ušpuras Copyright © 2015 Andrius Slavickas et al. All rights reserved. Measurement of Velocity and Temperature Profiles in the 1/40 Scaled-Down CANDU-6 Moderator Tank Mon, 27 Apr 2015 13:15:08 +0000 In order to simulate the CANDU-6 moderator circulation phenomena during steady state operating and accident conditions, a scaled-down moderator test facility has been constructed at Korea Atomic Energy Institute (KAERI). In the present work an experiment using a 1/40 scaled-down moderator tank has been performed to identify the potential problems of the flow visualization and measurement in the scaled-down moderator test facility. With a transparent moderator tank model, a flow field is visualized with a particle image velocimetry (PIV) technique under an isothermal state, and the temperature field is measured using a laser induced fluorescence (LIF) technique. A preliminary CFD analysis is also performed to find out the flow, thermal, and heating boundary conditions with which the various flow patterns expected in the prototype CANDU-6 moderator tank can be reproduced in the experiment. Hyoung Tae Kim, Jae Eun Cha, Han Seo, and In Cheol Bang Copyright © 2015 Hyoung Tae Kim et al. All rights reserved. Safety Enhancements for PHWRs Based on Macroscopic Losses of the Fukushima Accident Mon, 27 Apr 2015 12:57:55 +0000 The role of nuclear energy is to supply electric power on a stable basis to meet increasing demands, reduce carbon dioxide emissions, and maintain stable electric power costs while ensuring safety. The Fukushima accident taught us many lessons for creating safer nuclear power plants. Considering the design of systems, the areas of weakness at the Fukushima nuclear power plants can be divided into three categories: plant protection, electricity supply, and cooling of the nuclear fuel. In this paper, focusing on these three areas, the lessons learned are proposed and applied for pressurized heavy water reactors. Firstly, hard protection against external risks ensures the integrity of components and systems such that they can perform their original functions. Secondly, additional emergency power supply systems for electrical redundancy and diversity can improve the response capabilities for an accident by increasing the availability of active components. Thirdly, cooling for removing decay heat can be augmented by adopting diverse safety systems derived from other types of reactors. This study is expected to contribute to the safety enhancement of pressurized heavy water reactors by applying design changes based on the lessons learned from the Fukushima accident. Sang Ho Kim, Tsuneo Futami, Soon Heung Chang, and Yong Hoon Jeong Copyright © 2015 Sang Ho Kim et al. All rights reserved. CHF Enhancement of Advanced 37-Element Fuel Bundles Mon, 27 Apr 2015 11:33:20 +0000 A standard 37-element fuel bundle (37S fuel bundle) has been used in commercial CANDU reactors for over 40 years as a reference fuel bundle. Most CHF of a 37S fuel bundle have occurred at the elements arranged in the inner pitch circle for high flows and at the elements arranged in the outer pitch circle for low flows. It should be noted that a 37S fuel bundle has a relatively small flow area and high flow resistance at the peripheral subchannels of its center element compared to the other subchannels. The configuration of a fuel bundle is one of the important factors affecting the local CHF occurrence. Considering the CHF characteristics of a 37S fuel bundle in terms of CHF enhancement, there can be two approaches to enlarge the flow areas of the peripheral subchannels of a center element in order to enhance CHF of a 37S fuel bundle. To increase the center subchannel areas, one approach is the reduction of the diameter of a center element, and the other is an increase of the inner pitch circle. The former can increase the total flow area of a fuel bundle and redistributes the power density of all fuel elements as well as the CHF. On the other hand, the latter can reduce the gap between the elements located in the middle and inner pitch circles owing to the increasing inner pitch circle. This can also affect the enthalpy redistribution of the fuel bundle and finally enhance CHF or dry-out power. In this study, the above two approaches, which are proposed to enlarge the flow areas of the center subchannels, were considered to investigate the impact of the flow area changes of the center subchannels on the CHF enhancement as well as the thermal characteristics by applying a subchannel analysis method. Joo Hwan Park, Jong Yoeb Jung, and Eun Hyun Ryu Copyright © 2015 Joo Hwan Park et al. All rights reserved. The Effects of the Treatment of the Periodic Boundary Condition in TRIAINA Codes with a Pressure Tube Creep Problem Mon, 27 Apr 2015 10:40:36 +0000 To verify the periodic boundary condition (PBC) treatment which was implemented in a TRI-angle elements induced numerical analyzer (TRIAINA), the pressure tube creep problem is chosen and examined with three cases of normal, 2.5% creep, and 5.0% creep on the aspects of the multiplication factor and relative pin power. The McCARD code is used for the homogenized group constants generation. It is shown that the differences are nearly negligible for the pressure tube creep problem. E. H. Ryu, S. Y. Yoo, B. Y. Chung, and J. Y. Jung Copyright © 2015 E. H. Ryu et al. All rights reserved. Monte Carlo Few-Group Constant Generation for CANDU 6 Core Analysis Mon, 27 Apr 2015 10:39:08 +0000 The current neutronics design methodology of CANDU-PHWRs based on the two-step calculations requires determining not only homogenized two-group constants for ordinary fuel bundle lattice cells by the WIMS-AECL lattice cell code but also incremental two-group constants arising from the penetration of control devices into the fuel bundle cells by a supercell analysis code like MULTICELL or DRAGON. As an alternative way to generate the two-group constants necessary for the CANDU-PHWR core analysis, this paper proposes utilizing a B1 theory augmented Monte Carlo (MC) few-group constant generation method (B1 MC method) which has been devised for the PWR fuel assembly analysis method. To examine the applicability of the B1 MC method for the CANDU 6 core analysis, the fuel bundle cell and supercell calculations are performed using it to obtain the two-group constants. By showing that the two-group constants from the B1 MC method agree well with those from WIMS-AECL and that core neutronics calculations for hypothetical CANDU 6 cores by a deterministic diffusion theory code SCAN with B1 MC method generated two-group constants also agree well with whole core MC analyses, it is concluded that the B1 MC method is well qualified for both fuel bundle cell and supercell analyses. Seung Yeol Yoo, Hyung Jin Shim, and Chang Hyo Kim Copyright © 2015 Seung Yeol Yoo et al. All rights reserved. Uncertainty Analysis of the Potential Hazard of MCCI during Severe Accidents for the CANDU6 Plant Mon, 27 Apr 2015 10:32:45 +0000 This paper illustrates the application of a severe accident analysis computer program to the uncertainty analysis of molten corium-concrete interaction (MCCI) phenomena in cases of severe accidents in CANDU6 type plant. The potential hazard of MCCI is a failure of the reactor building owing to the possibility of a calandria vault floor melt-through even though the containment filtered vent system is operated. Meanwhile, the MCCI still has large uncertainties in several phenomena such as a melt spreading area and the extent of water ingression into a continuous debris layer. The purpose of this study is to evaluate the MCCI in the calandria vault floor via an uncertainty analysis using the ISAAC program for the CANDU6. Sooyong Park, Kwang-Il Ahn, and YongMann Song Copyright © 2015 Sooyong Park et al. All rights reserved. Alpha Stable Distribution Based Morphological Filter for Bearing and Gear Fault Diagnosis in Nuclear Power Plant Thu, 23 Apr 2015 13:36:50 +0000 Gear and bearing play an important role as key components of rotating machinery power transmission systems in nuclear power plants. Their state conditions are very important for safety and normal operation of entire nuclear power plant. Vibration based condition monitoring is more complicated for the gear and bearing of planetary gearbox than those of fixed-axis gearbox. Many theoretical and engineering challenges in planetary gearbox fault diagnosis have not yet been resolved which are of great importance for nuclear power plants. A detailed vibration condition monitoring review of planetary gearbox used in nuclear power plants is conducted in this paper. A new fault diagnosis method of planetary gearbox gears is proposed. Bearing fault data, bearing simulation data, and gear fault data are used to test the new method. Signals preprocessed using dilation-erosion gradient filter and fast Fourier transform for fault information extraction. The length of structuring element (SE) of dilation-erosion gradient filter is optimized by alpha stable distribution. Method experimental verification confirmed that parameter alpha is superior compared to kurtosis since it can reflect the form of entire signal and it cannot be influenced by noise similar to impulse. Xinghui Zhang, Jianshe Kang, Lei Xiao, and Jianmin Zhao Copyright © 2015 Xinghui Zhang et al. All rights reserved. An Improved Shuffled Frog Leaping Algorithm for Assembly Sequence Planning of Remote Handling Maintenance in Radioactive Environment Sun, 19 Apr 2015 08:33:40 +0000 Assembly sequence planning (ASP) of remote handling maintenance in radioactive environment is a combinatorial optimization problem. This study proposes an improved shuffled frog leaping algorithm (SFLA) for the combinatorial optimization problem of ASP. An ASP experiment is conducted to verify the feasibility and stability of the improved SFLA. Simultaneously, the improved SFLA is compared with SFLA, genetic algorithm, particle swarm optimization, and adaptive mutation particle swarm optimization in terms of efficiency and capability of locating the best global assembly sequence. Experiment results show that the proposed algorithm exhibits outstanding performance in solving the ASP problem. The application of the proposed algorithm should increase the level of ASP in a radioactive environment. Jianwen Guo, Hong Tang, Zhenzhong Sun, Song Wang, Xuejun Jia, Haibin Chen, and Zhicong Zhang Copyright © 2015 Jianwen Guo et al. All rights reserved. On Some Fundamental Peculiarities of the Traveling Wave Reactor Sun, 05 Apr 2015 07:32:00 +0000 On the basis of the condition for nuclear burning wave existence in the neutron-multiplying media (U-Pu and Th-U cycles) we show the possibility of surmounting the so-called dpa-parameter problem and suggest an algorithm of the optimal nuclear burning wave mode adjustment, which is supposed to yield the wave parameters (fluence/neutron flux, width and speed of nuclear burning wave) that satisfy the dpa-condition associated with the tolerable level of the reactor materials radioactive stability, in particular that of the cladding materials. It is shown for the first time that the capture and fission cross sections of 238U and 239Pu increase with temperature within 1000–3000 K range, which under certain conditions may lead to a global loss of the nuclear burning wave stability. Some variants of the possible stability loss due to the so-called blow-up modes (anomalous nuclear fuel temperature and neutron flow evolution) are discussed and are found to possibly become a reason for a trivial violation of the traveling wave reactor internal safety. V. D. Rusov, V. A. Tarasov, I. V. Sharph, V. N. Vashchenko, E. P. Linnik, T. N. Zelentsova, M. E. Beglaryan, S. A. Chernegenko, S. I. Kosenko, and V. P. Smolyar Copyright © 2015 V. D. Rusov et al. All rights reserved. Influence of the Saturation Ratio on Concrete Behavior under Triaxial Compressive Loading Mon, 30 Mar 2015 13:29:18 +0000 When a concrete structure is subjected to an impact, the material is subjected to high triaxial compressive stresses. Furthermore, the water saturation ratio in massive concrete structures may reach nearly 100% at the core, whereas the material dries quickly on the skin. The impact response of a massive concrete wall may thus depend on the state of water saturation in the material. This paper presents some triaxial tests performed at a maximum confining pressure of 600 MPa on concrete representative of a nuclear power plant containment building. Experimental results show the concrete constitutive behavior and its dependence on the water saturation ratio. It is observed that as the degree of saturation increases, a decrease in the volumetric strains as well as in the shear strength is observed. The coupled PRM constitutive model does not accurately reproduce the response of concrete specimens observed during the test. The differences between experimental and numerical results can be explained by both the influence of the saturation state of concrete and the effect of deviatoric stresses, which are not accurately taken into account. The PRM model was modified in order to improve the numerical prediction of concrete behavior under high stresses at various saturation states. Xuan-Dung Vu, Matthieu Briffaut, Yann Malecot, Laurent Daudeville, and Bertrand Ciree Copyright © 2015 Xuan-Dung Vu et al. All rights reserved. Integrated Cost and Schedule Control Systems for Nuclear Power Plant Construction: Leveraging Strategic Advantages to Owners and EPC Firms Wed, 04 Mar 2015 08:32:26 +0000 As the owners expect that the cost and time for nuclear power plant construction would decrease with new entrants into the market, there will be severer competition in the nuclear industry. In order to achieve performance improvement and to attain competitive advantages under the globalized competition, practitioners and researchers in the nuclear industry have recently exerted efforts to develop an advanced and efficient management methodology for the nuclear mega-projects. Among several candidates, integrated cost and schedule control system is of great concern because it can effectively manage the three most important project performances including cost, time, and quality. In this context, the purpose of this paper is to develop a project numbering system (PNS) of integrated cost and schedule control system for nuclear power plant construction. Distinct attributes of nuclear power plant construction were investigated first in order to identify influencing variables that characterize real-world implementation of advanced cost and schedule controls. A scenario was then developed and analysed to simulate a case-project. By using this case-project, proposed management requirements, management methods, measurement techniques, data structure, and data collection methods for integrated cost and schedule PNS were illustrated. Finally, findings and implications are outlined, and recommendations for further research are presented. Youngsoo Jung, Byeong-Suk Moon, Yun-Myung Kim, and Woojoong Kim Copyright © 2015 Youngsoo Jung et al. All rights reserved. HTR-Based Power Plants’ Performance Analysis Applied on Conventional Combined Cycles Tue, 03 Mar 2015 11:45:17 +0000 In high temperature reactors including gas cooled fast reactors and gas turbine modular helium reactors (GT-MHR) specifically designed to operate as power plant heat sources, efficiency enhancement at effective cost under safe conditions can be achieved. Mentioned improvements concern the implementation of two cycle structures: (a), a stand alone Brayton operating with helium and a stand alone Rankine cycle (RC) with regeneration, operating with carbon dioxide at ultrasupercritical pressure as working fluid (WF), where condensation is carried out at quasicritical conditions, and (b), a combined cycle (CC), in which the topping closed Brayton cycle (CBC) operates with helium as WF, while the bottoming RC is operated with one of the following WFs: carbon dioxide, xenon, ethane, ammonia, or water. In both cases, an intermediate heat exchanger (IHE) is proposed to provide thermal energy to the closed Brayton or to the Rankine cycles. The results of the case study show that the thermal efficiency, through the use of a CC, is slightly improved (from 45.79% for BC and from 50.17% for RC to 53.63 for the proposed CC with He-H2O operating under safety standards). José Carbia Carril, Álvaro Baaliña Insua, Javier Romero Gómez, and Manuel Romero Gómez Copyright © 2015 José Carbia Carril et al. All rights reserved. Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code Wed, 25 Feb 2015 06:41:19 +0000 Cases of core coolant flow blockage transient have been simulated and analysed for the TRIGA IPR-R1 research reactor using the RELAP5-MOD3.3 code. The transients are related to partial and to total obstruction of the core coolant channels. The reactor behaviour after the loss of flow was analysed as well as the changes in the coolant and fuel temperatures. The behaviour of the thermal hydraulic parameters from the transient simulations was analysed. For a partial blockage, it was observed that the reactor reaches a new steady state operation with new values for the thermal hydraulic parameters. The total core blockage brings the reactor to an abnormal operation causing increase in core temperature. Patrícia A. L. Reis, Antonella L. Costa, Claubia Pereira, Maria Auxiliadora F. Veloso, and Amir Z. Mesquita Copyright © 2015 Patrícia A. L. Reis et al. All rights reserved. Safety Assessment of Low-Contaminated Equipment Dismantling at Nuclear Power Plants Mon, 23 Feb 2015 08:49:56 +0000 The decommissioning of nuclear facilities requires adequate planning and demonstration that dismantling and decontamination activities can be conducted safely. Existing safety standards require that an appropriate safety assessment be performed to support the decommissioning plan for each facility (International Atomic Energy Agency, 2006). This paper presents safety assessment approach used in Lithuania during the development of the first dismantling and decontamination project for Ignalina NPP. The paper will mainly focus on the identification and assessment of the hazards raised due to dismantling and decontamination activities at Ignalina Nuclear Power Plant and on the assessment of the nonradiological and radiological consequences of the indicated most dangerous initiating event. The drop of heavy item was indicated as one of most dangerous initiating events for the discussed Ignalina Nuclear Power Plant dismantling and decontamination project. For the analysis of the nonradiological impact the finite element model for the load drop force calculation was developed. The radiological impact was evaluated in those accident cases which would lead to the worst radiological consequences. The assessments results show that structural integrity of the building and supporting columns of building structures will be maintained and radiological consequences are lower than the annual regulatory operator dose limit. Egidijus Babilas, Eugenijus Ušpuras, Sigitas Rimkevičius, Gintautas Dundulis, and Mindaugas Vaišnoras Copyright © 2015 Egidijus Babilas et al. All rights reserved. AP1000 Shield Building Dynamic Response for Different Water Levels of PCCWST Subjected to Seismic Loading considering FSI Wed, 18 Feb 2015 08:29:59 +0000 Huge water storage tank on the top of many buildings may affect the safety of the structure caused by fluid-structure interaction (FSI) under the earthquake. AP1000 passive containment cooling system water storage tank (PCCWST) placed at the top of shield building is a key component to ensure the safety of nuclear facilities. Under seismic loading, water will impact the wall of PCCWST, which may pose a threat to the integrity of the shield building. In the present study, an FE model of AP1000 shield building is built for the modal and transient seismic analysis considering the FSI. Six different water levels in PCCWST were discussed by comparing the modal frequency, seismic acceleration response, and von Mises stress distribution. The results show the maximum von Mises stress emerges at the joint of shield building roof and water around the air inlet. However, the maximum von Mises stress is below the yield strength of reinforced concrete. The results may provide a reference for design of the AP1000 and CAP1400 in the future. Daogang Lu, Yu Liu, and Xiaojia Zeng Copyright © 2015 Daogang Lu et al. All rights reserved. Development of the Noncontact Temperature Sensor Using the Infrared Optical Fiber Coated with Antifog Solution Wed, 11 Feb 2015 11:24:15 +0000 This study developed a noncontact fiber-optic temperature sensor that can be installed in a spent nuclear fuel pool. This fiber-optic temperature sensor was fabricated using an infrared optical fiber to transmit the infrared light emitted from water at a certain temperature. To minimize the decrease in the detection efficiency of the fiber-optic temperature sensor due to vapor generation, its surface was coated by spraying an antifog solution and drying several times. The measurement data of the fiber-optic temperature sensor was almost linear in the range of 30~70°C. This sensor could be used as an auxiliary temperature monitoring system in a spent nuclear fuel pool. Rinah Kim, Chan Hee Park, Arim Lee, and Joo Hyun Moon Copyright © 2015 Rinah Kim et al. All rights reserved. Scientific and Engineering Literature Mini Review of Molten Salt Oxidation for Radioactive Waste Treatment and Organic Compound Gasification as well as Spent Salt Treatment Tue, 10 Feb 2015 08:35:18 +0000 Literature review was performed for the molten salt oxidation technology in order to collect all available scientific and engineering information for further use of this technology in nuclear applications. This report provides a summary of a review of scientific and engineering literature on MSO treatment of a wide variety of radioactive wastes, organic compound gasification, and related studies such as radioactive spent salt processing that was found important for further development of the MSO technology in the nuclear field for radioactive waste treatment. Miscellaneous nonnuclear uses of molten salts, such as converting carbon monoxide to carbon dioxide, are not discussed. Petr Kovařík, James D. Navratil, and Jan John Copyright © 2015 Petr Kovařík et al. All rights reserved. Study on the Use of Hydride Fuel in High-Performance Light Water Reactor Concept Sun, 01 Feb 2015 07:22:33 +0000 Hydride fuels have features which could make their use attractive in future advanced power reactors. The potential benefit of use of hydride fuel in HPLWR without introducing significant modification in the current core design concept of the high-performance light water reactor (HPLWR) has been evaluated. Neutronics and thermal hydraulic analyses were performed for a single assembly model of HPLWR with oxide and hydride fuels. The hydride assembly shows higher moderation with softer neutron spectrum and slightly more uniform axial power distribution. It achieves a cycle length of 18 months with sufficient excess reactivity. At Beginning of Cycle the fuel temperature coefficient of the hydride assembly is higher whereas the moderator and void coefficients are lower. The thermal hydraulic results show that the achievable fuel temperature in the hydride assembly is well below the design limits. The potential benefits of the use of hydride fuel in the current design of the HPLWR with the achieved improvements in the core neutronics characteristics are not sufficient to justify the replacement of the oxide fuel. Therefore for a final evaluation of the use of hydride fuels in HPLWR concepts additional studies which include modification of subassembly and core layout designs are required. Haileyesus Tsige-Tamirat and Luca Ammirabile Copyright © 2015 Haileyesus Tsige-Tamirat and Luca Ammirabile. All rights reserved. Research and Evaluation for Passive Safety System in Low Pressure Reactor Mon, 26 Jan 2015 06:42:18 +0000 Low pressure reactor is a small size advanced reactor with power of 180 MWt, which is under development at Nuclear Power Institute of China. In order to assess the ability and feasibility of passive safety system, several tests have been implemented on the passive safety system (PSS) test facility. During the LOCA and SBO accident, the adequate core cooling is provided by the performance of passive safety system. In addition the best-estimate thermal hydraulic code, CATHARE V2.1, has been assessed against cold leg LOCA test. The calculation results show that CATHARE is in a satisfactory agreement with the test for the steady state and transient test. Peng Chuanxin, Zhuo Wenbin, Chen Bingde, Nie Changhua, and Huang Yanping Copyright © 2015 Peng Chuanxin et al. All rights reserved. Single Phase Natural Circulation Behaviors of the Integral Type Marine Reactor Simulator under Rolling Motion Condition Mon, 26 Jan 2015 06:37:38 +0000 During operation in the sea the reactor natural circulation behaviors are affected by ship rolling motion. The development of an analysis code and the natural circulation behaviors of a reactor simulator under rolling motion are described in this paper. In the case of rolling motion, the primary coolant flow rates in the hot legs and heating channels oscillated periodically, and the amplitude of flow rate oscillation was in direct proportion to rolling amplitude, but in inverse proportion to rolling period. The total mass flow rate also oscillated with half the rolling period, and the average total mass flow rate was less than that in steady state. In the natural circulation under a rolling motion, the flow rate oscillations in the hot legs were controlled by the tangential force; however, the mass flow rate oscillations in the total natural circulation and the heating channels were a result of the combined action of the change of inclination angle, flow resistance, and the extra force arising from the rolling motion. The extra tangential force brought about intense flow rate oscillations in the hot legs, which resulted in increasing total flow resistance; however the extra centrifugal force played a role in increasing thermal driving head. Hou-jun Gong, Xing-tuan Yang, Yan-ping Huang, and Sheng-yao Jiang Copyright © 2015 Hou-jun Gong et al. All rights reserved.