Science and Technology of Nuclear Installations http://www.hindawi.com The latest articles from Hindawi Publishing Corporation © 2015 , Hindawi Publishing Corporation . All rights reserved. Input Calibration and Validation of RELAP5 Against CIRCUS-IV Single Channel Tests on Natural Circulation Two-Phase Flow Instability Wed, 26 Aug 2015 09:53:22 +0000 http://www.hindawi.com/journals/stni/2015/130741/ RELAP5 is a system thermal-hydraulic code that is used to perform safety analysis on nuclear reactors. Since the code is based on steady state, two-phase flow regime maps, there is a concern that RELAP5 may provide significant errors for rapid transient conditions. In this work, the capability of RELAP5 code to predict the oscillatory behavior of a natural circulation driven, two-phase flow at low pressure is investigated. The simulations are compared with a series of experiments that were performed in the CIRCUS-IV facility at the Delft University of Technology. For this purpose, we developed a procedure for calibration of the input and code validation. The procedure employs (i) multiple parameters measured in different regimes, (ii) independent consideration of the subsections of the loop, and (iii) assessment of importance of the uncertain input parameters. We found that predicted system parameters are less sensitive to variations of the uncertain input and boundary conditions in high frequency oscillations regime. It is shown that calculation results overlap experimental values, except for the high frequency oscillations regime where the maximum inlet flow rate was overestimated. This finding agrees with the idea that steady state, two-phase flow regime maps might be one of the possible reasons for the discrepancy in case of rapid transients in two-phase systems. Viet-Anh Phung, Pavel Kudinov, Dmitry Grishchenko, and Martin Rohde Copyright © 2015 Viet-Anh Phung et al. All rights reserved. Study on the Effects of Liquid Thermal Media on the Irradiation Capsule of High-Temperature Materials Mon, 24 Aug 2015 12:14:47 +0000 http://www.hindawi.com/journals/stni/2015/723081/ Irradiation tests of materials at HANARO have usually been conducted using a standard capsule at temperatures of about 300°C for irradiation of materials used at PWR. Thus, the standard capsule uses aluminum as the specimen holder, which acts to dissipate the thermal energy. Future nuclear systems such as a VHTR and SFR require the irradiation tests at a relatively high temperature. As an alternative to aluminum which has been used as the thermal media in a standard material capsule, the characteristics of liquid metals such as NaK and LBE are reviewed. The temperatures of the capsule are affected by the variation of parameters such as the gap and wall thickness of the container. In particular, the external gap is most important in determining the temperature of the specimen. LBE raises the temperature of the specimen higher than NaK at the same configuration of the capsule. Thus, LBE can lessen the gap of the parts and reduce the vibration for a stable long-term test in reactor. Man Soon Cho, Sung Ryul Kim, Seong Woo Yang, and Kee Nam Choo Copyright © 2015 Man Soon Cho et al. All rights reserved. Integrated Deterministic and Probabilistic Safety Analysis for Safety Assessment of Nuclear Power Plants Wed, 05 Aug 2015 12:47:07 +0000 http://www.hindawi.com/journals/stni/2015/136940/ Francesco Di Maio, Enrico Zio, Curtis Smith, and Valentin Rychkov Copyright © 2015 Francesco Di Maio et al. All rights reserved. Improved Modelling and Assessment of the Performance of Firefighting Means in the Frame of a Fire PSA Tue, 04 Aug 2015 13:12:55 +0000 http://www.hindawi.com/journals/stni/2015/238723/ An integrated deterministic and probabilistic safety analysis (IDPSA) was carried out to assess the performances of the firefighting means to be applied in a nuclear power plant. The tools used in the analysis are the code FDS (Fire Dynamics Simulator) for fire simulation and the tool MCDET (Monte Carlo Dynamic Event Tree) for handling epistemic and aleatory uncertainties. The combination of both tools allowed for an improved modelling of a fire interacting with firefighting means while epistemic uncertainties because lack of knowledge and aleatory uncertainties due to the stochastic aspects of the performances of the firefighting means are simultaneously taken into account. The MCDET-FDS simulations provided a huge spectrum of fire sequences each associated with a conditional occurrence probability at each point in time. These results were used to derive probabilities of damage states based on failure criteria considering high temperatures of safety related targets and critical exposure times. The influence of epistemic uncertainties on the resulting probabilities was quantified. The paper describes the steps of the IDPSA and presents a selection of results. Focus is laid on the consideration of epistemic and aleatory uncertainties. Insights and lessons learned from the analysis are discussed. Martina Kloos and Joerg Peschke Copyright © 2015 Martina Kloos and Joerg Peschke. All rights reserved. A Flooding Induced Station Blackout Analysis for a Pressurized Water Reactor Using the RISMC Toolkit Tue, 04 Aug 2015 11:31:25 +0000 http://www.hindawi.com/journals/stni/2015/308163/ In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. In addition, the impact of power uprate is determined in terms of both core damage probability and safety margins. Diego Mandelli, Steven Prescott, Curtis Smith, Andrea Alfonsi, Cristian Rabiti, Joshua Cogliati, and Robert Kinoshita Copyright © 2015 Diego Mandelli et al. All rights reserved. Scenario Grouping and Classification Methodology for Postprocessing of Data Generated by Integrated Deterministic-Probabilistic Safety Analysis Tue, 04 Aug 2015 11:29:46 +0000 http://www.hindawi.com/journals/stni/2015/278638/ Integrated Deterministic-Probabilistic Safety Assessment (IDPSA) combines deterministic model of a nuclear power plant with a method for exploration of the uncertainty space. Huge amount of data is generated in the process of such exploration. It is very difficult to “manually” process and extract from such data information that can be used by a decision maker for risk-informed characterization, understanding, and eventually decision making on improvement of the system safety and performance. Such understanding requires an approach for interpretation, grouping of similar scenario evolutions, and classification of the principal characteristics of the events that contribute to the risk. In this work, we develop an approach for classification and characterization of failure domains. The method is based on scenario grouping, clustering, and application of decision trees for characterization of the influence of timing and order of events. We demonstrate how the proposed approach is used to classify scenarios that are amenable to treatment with Boolean logic in classical Probabilistic Safety Assessment (PSA) from those where timing and order of events determine process evolution and eventually violation of safety criteria. The efficiency of the approach has been verified with application to the SARNET benchmark exercise on the effectiveness of hydrogen management in the containment. Sergey Galushin and Pavel Kudinov Copyright © 2015 Sergey Galushin and Pavel Kudinov. All rights reserved. An Enhanced Preventive Maintenance Optimization Model Based on a Three-Stage Failure Process Tue, 04 Aug 2015 11:28:52 +0000 http://www.hindawi.com/journals/stni/2015/193075/ Nuclear power plants are highly complex systems and the issues related to their safety are of primary importance. Probabilistic safety assessment is regarded as the most widespread methodology for studying the safety of nuclear power plants. As maintenance is one of the most important factors for affecting the reliability and safety, an enhanced preventive maintenance optimization model based on a three-stage failure process is proposed. Preventive maintenance is still a dominant maintenance policy due to its easy implementation. In order to correspond to the three-color scheme commonly used in practice, the lifetime of system before failure is divided into three stages, namely, normal, minor defective, and severe defective stages. When the minor defective stage is identified, two measures are considered for comparison: one is that halving the inspection interval only when the minor defective stage is identified at the first time; the other one is that if only identifying the minor defective stage, the subsequent inspection interval is halved. Maintenance is implemented immediately once the severe defective stage is identified. Minimizing the expected cost per unit time is our objective function to optimize the inspection interval. Finally, a numerical example is presented to illustrate the effectiveness of the proposed models. Ruifeng Yang, Jianshe Kang, and Zhenya Quan Copyright © 2015 Ruifeng Yang et al. All rights reserved. Safety Assessment of Nuclear Power Plants for Liquefaction Consequences Tue, 04 Aug 2015 11:25:31 +0000 http://www.hindawi.com/journals/stni/2015/727291/ In case of some nuclear power plants constructed at the soft soil sites, liquefaction should be analysed as beyond design basis hazard. The aim of the analysis is to define the postevent condition of the plant, definition of plant vulnerabilities, and identification of the necessary measures for accident management. In the paper, the methodology of the analysis of liquefaction effects for nuclear power plants is outlined. The procedure includes identification of the scope of the safety analysis and the acceptable limit cases for plant structures having different role from accident management point of view. Considerations are made for identification of dominating effects of liquefaction. The possibility of the decoupling of the analysis of liquefaction effects from the analysis of vibratory ground motion is discussed. It is shown in the paper that the practicable empirical methods for definition of liquefaction susceptibility provide rather controversial results. Selection of method for assessment of soil behaviour that affects the integrity of structures requires specific considerations. The case of nuclear power plant at Paks, Hungary, is used as an example for demonstration of practical importance of the presented results and considerations. Tamás János Katona, Zoltán Bán, Erzsébet Győri, László Tóth, and András Mahler Copyright © 2015 Tamás János Katona et al. All rights reserved. Probabilistic Dynamics for Integrated Analysis of Accident Sequences considering Uncertain Events Tue, 04 Aug 2015 11:15:43 +0000 http://www.hindawi.com/journals/stni/2015/892502/ The analytical/deterministic modelling and simulation/probabilistic methods are used separately as a rule in order to analyse the physical processes and random or uncertain events. However, in the currently used probabilistic safety assessment this is an issue. The lack of treatment of dynamic interactions between the physical processes on one hand and random events on the other hand causes the limited assessment. In general, there are a lot of mathematical modelling theories, which can be used separately or integrated in order to extend possibilities of modelling and analysis. The Theory of Probabilistic Dynamics (TPD) and its augmented version based on the concept of stimulus and delay are introduced for the dynamic reliability modelling and the simulation of accidents in hybrid (continuous-discrete) systems considering uncertain events. An approach of non-Markovian simulation and uncertainty analysis is discussed in order to adapt the Stimulus-Driven TPD for practical applications. The developed approach and related methods are used as a basis for a test case simulation in view of various methods applications for severe accident scenario simulation and uncertainty analysis. For this and for wider analysis of accident sequences the initial test case specification is then extended and discussed. Finally, it is concluded that enhancing the modelling of stimulated dynamics with uncertainty and sensitivity analysis allows the detailed simulation of complex system characteristics and representation of their uncertainty. The developed approach of accident modelling and analysis can be efficiently used to estimate the reliability of hybrid systems and at the same time to analyze and possibly decrease the uncertainty of this estimate. Robertas Alzbutas Copyright © 2015 Robertas Alzbutas. All rights reserved. Demonstration of Emulator-Based Bayesian Calibration of Safety Analysis Codes: Theory and Formulation Tue, 04 Aug 2015 11:13:11 +0000 http://www.hindawi.com/journals/stni/2015/839249/ System codes for simulation of safety performance of nuclear plants may contain parameters whose values are not known very accurately. New information from tests or operating experience is incorporated into safety codes by a process known as calibration, which reduces uncertainty in the output of the code and thereby improves its support for decision-making. The work reported here implements several improvements on classic calibration techniques afforded by modern analysis techniques. The key innovation has come from development of code surrogate model (or code emulator) construction and prediction algorithms. Use of a fast emulator makes the calibration processes used here with Markov Chain Monte Carlo (MCMC) sampling feasible. This work uses Gaussian Process (GP) based emulators, which have been used previously to emulate computer codes in the nuclear field. The present work describes the formulation of an emulator that incorporates GPs into a factor analysis-type or pattern recognition-type model. This “function factorization” Gaussian Process (FFGP) model allows overcoming limitations present in standard GP emulators, thereby improving both accuracy and speed of the emulator-based calibration process. Calibration of a friction-factor example using a Method of Manufactured Solution is performed to illustrate key properties of the FFGP based process. Joseph P. Yurko, Jacopo Buongiorno, and Robert Youngblood Copyright © 2015 Joseph P. Yurko et al. All rights reserved. Delayed Station Blackout Event and Nuclear Safety Tue, 04 Aug 2015 11:11:43 +0000 http://www.hindawi.com/journals/stni/2015/192601/ The loss of off-site power (LOOP) event occurs when all electrical power to the nuclear power plant from the power grid is lost. Complete failure of both off-site and on-site alternating current (AC) power sources is referred to as a station blackout (SBO). Combined LOOP and SBO events are analyzed in this paper. The analysis is done for different time delays between the LOOP and SBO events. Deterministic safety analysis is utilized for the assessment of the plant parameters for different time delays of the SBO event. Obtained plant parameters are used for the assessment of the probabilities of the functional events in the SBO event tree. The results show that the time delay of the SBO after the LOOP leads to a decrease of the core damage frequency (CDF) from the SBO event tree. The reduction of the CDF depends on the time delay of the SBO after the LOOP event. The results show the importance of the safety systems to operate after the plant shutdown when the decay heat is large. Small changes of the basic events importance measures are identified with the introduction of the delay of the SBO event. Andrija Volkanovski and Andrej Prošek Copyright © 2015 Andrija Volkanovski and Andrej Prošek. All rights reserved. Risk-Based Clustering for Near Misses Identification in Integrated Deterministic and Probabilistic Safety Analysis Tue, 04 Aug 2015 11:11:20 +0000 http://www.hindawi.com/journals/stni/2015/693891/ In integrated deterministic and probabilistic safety analysis (IDPSA), safe scenarios and prime implicants (PIs) are generated by simulation. In this paper, we propose a novel postprocessing method, which resorts to a risk-based clustering method for identifying Near Misses among the safe scenarios. This is important because the possibility of recovering these combinations of failures within a tolerable grace time allows avoiding deviations to accident and, thus, reducing the downtime (and the risk) of the system. The postprocessing risk-significant features for the clustering are extracted from the following: (i) the probability of a scenario to develop into an accidental scenario, (ii) the severity of the consequences that the developing scenario would cause to the system, and (iii) the combination of (i) and (ii) into the overall risk of the developing scenario. The optimal selection of the extracted features is done by a wrapper approach, whereby a modified binary differential evolution (MBDE) embeds a -means clustering algorithm. The characteristics of the Near Misses scenarios are identified solving a multiobjective optimization problem, using the Hamming distance as a measure of similarity. The feasibility of the analysis is shown with respect to fault scenarios in a dynamic steam generator (SG) of a nuclear power plant (NPP). Francesco Di Maio, Matteo Vagnoli, and Enrico Zio Copyright © 2015 Francesco Di Maio et al. 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 http://www.hindawi.com/journals/stni/2015/180979/ 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 http://www.hindawi.com/journals/stni/2015/237646/ 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 http://www.hindawi.com/journals/stni/2015/140979/ 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 http://www.hindawi.com/journals/stni/2015/913274/ 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 http://www.hindawi.com/journals/stni/2015/652731/ 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 http://www.hindawi.com/journals/stni/2015/920183/ 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 http://www.hindawi.com/journals/stni/2015/198654/ 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 http://www.hindawi.com/journals/stni/2015/104739/ 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 http://www.hindawi.com/journals/stni/2015/143767/ 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 http://www.hindawi.com/journals/stni/2015/757201/ 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 http://www.hindawi.com/journals/stni/2015/439863/ 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 http://www.hindawi.com/journals/stni/2015/520756/ 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 http://www.hindawi.com/journals/stni/2015/243867/ 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 http://www.hindawi.com/journals/stni/2015/571547/ 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 http://www.hindawi.com/journals/stni/2015/284642/ 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 http://www.hindawi.com/journals/stni/2015/462941/ 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 http://www.hindawi.com/journals/stni/2015/460131/ 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 http://www.hindawi.com/journals/stni/2015/516470/ 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.