Science and Technology of Nuclear Installations The latest articles from Hindawi Publishing Corporation © 2016 , Hindawi Publishing Corporation . All rights reserved. Study on the Impact of Thermal Agitation on Doppler Coefficient in Epithermal Range for Gd-Bearing Fuel Mon, 18 Jan 2016 07:37:36 +0000 The impact of thermal agitation on Doppler coefficient for Gd-bearing fuel was analyzed. It was found through the analysis that the impact increases when a small amount of Gd2O3 is added to pure UO2 fuel although the impact decreases for a large amount of Gd2O3. This tendency was discussed with the usage of simplified expression for the difference of Doppler coefficient. The simplified expression was used to consider the tendency, and it was revealed that the tendency mainly comes from the rapid decrement of multiplication factor and the relatively slow decrement of the magnitude of sensitivity coefficient of U-238 capture cross section at low Gd2O3 concentration. Similar tendency which shows a maximum impact on Doppler coefficient at interior concentration is expected for other UO2 fuel with a slight content of strong absorber. This indicates that Doppler coefficient of UO2 fuel system with low content of strong absorber should be analyzed carefully by considering thermal agitation in epithermal range. Satoshi Takeda, Satoshi Ino, Kazuhiro Wada, Michitaka Ono, and Takanori Kitada Copyright © 2016 Satoshi Takeda et al. All rights reserved. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100 Sun, 17 Jan 2016 07:53:23 +0000 Small modular reactor (SMR) has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR) is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100) is being developed by University of Science and Technology of China (USTC). In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS) transient simulation at beginning of the reactor cycle (BOC) has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance. Pengcheng Zhao, Kangli Shi, Shuzhou Li, Jingchao Feng, and Hongli Chen Copyright © 2016 Pengcheng Zhao et al. All rights reserved. Normal and Accidental Scenarios Analyses with MELCOR 1.8.2 and MELCOR 2.1 for the DEMO Helium-Cooled Pebble Bed Blanket Concept Wed, 30 Dec 2015 06:30:25 +0000 As for Light Water Reactors (LWRs), one of the most challenging accidents for the future DEMOnstration power plant is the Loss of Coolant Accident, which can trigger the pressurization of the confinement structures and components. Hence, careful analyses have to be executed to demonstrate that the confinement barriers are able to withstand the pressure peak within design limits and the residual cooling capabilities of the Primary Heat Transfer System are sufficient to remove the decay heat. To do so, severe accident codes, as MELCOR, can be employed. In detail, the MELCOR code has been developed to cope also with fusion reactors, but unfortunately, these fusion versions are based on the old 1.8.x source code. On the contrary, for LWRs, the newest 2.1.x versions are continuously updated. Thanks to the new features introduced in these latest 2.1.x versions, the main phenomena occurring in the helium-cooled blanket concepts of DEMO can be simulated in a basic manner. For this purpose, several analyses during normal and accidental DEMO conditions have been executed. The aim of these analyses is to compare the results obtained with MELCOR 1.8.2 and MELCOR 2.1 in order to highlight the differences among the results of the main thermal-hydraulic parameters. Bruno Gonfiotti and Sandro Paci Copyright © 2015 Bruno Gonfiotti and Sandro Paci. All rights reserved. Monitoring of 220Rn Concentrations in Buildings of Kufa Technical Institute, Iraq Thu, 17 Dec 2015 09:50:17 +0000 This paper presents the measurements of thoron and the progeny in fifteen buildings in Kufa Technical Institute, Iraq, from June 2015 to April 2015 using RAD-7 detectors. Also, annual effective dose rate was calculated in all buildings under study. The thoron concentration varies from  Bq/m3 to  Bq/m3 with an average  Bq/m3. The concentration of thoron daughters was found to vary from 0.14 mWL to 1.44 mWL with an average  mWL. The annual effective doses due to thoron mainly vary from 0.042 mSv/y to 0.81 mSv/y with an average  mSv/y. The preliminary results in this study indicate that they may be suitable for evaluating the indoor 220Rn and its progeny concentrations whenever the public exposure to 220Rn and its progeny is taken into account. During this survey, the continuous difficulty in measuring thoron was also pointed out, due to its short half-life and faults in the measuring system. Ali Abid Abojassim Al-Hamidawi Copyright © 2015 Ali Abid Abojassim Al-Hamidawi. All rights reserved. Method of Measuring the Efficiency of the Conversion of Nuclear Energy into Optical Energy Tue, 08 Dec 2015 09:03:54 +0000 A method of measuring the efficiency of converting nuclear energy into optical energy was developed based on correlations between intensities of the research line and the nitrogen second positive system in an Ar-N2 mixture. In addition, the values of the coefficient of the conversion of nuclear energy into radiation at the lines of a Hg triplet in mixtures of Хе-Hg and Kr-Hg were determined. The values measured correspond to a selectiveness of pumping of 73S1 that was close to 1 (). Erlan G. Batyrbekov, Yuriy N. Gordienko, Mendykhan U. Khasenov, and Yuriy V. Ponkratov Copyright © 2015 Erlan G. Batyrbekov et al. All rights reserved. Prediction of Flow Regimes and Thermal Hydraulic Parameters in Two-Phase Natural Circulation by RELAP5 and TRACE Codes Sun, 29 Nov 2015 12:03:55 +0000 In earlier study we have demonstrated that RELAP5 can predict flow instability parameters (flow rate, oscillation period, temperature, and pressure) in single channel tests in CIRCUS-IV facility. The main goals of this work are to (i) validate RELAP5 and TRACE capabilities in prediction of two-phase flow instability and flow regimes and (ii) assess the effect of improvement in flow regime identification on code predictions. Most of the results of RELAP5 and TRACE calculation are in reasonable agreement with experimental data from CIRCUS-IV. However, both codes misidentified instantaneous flow regimes which were observed in the test with high speed camera. One of the reasons for the incorrect identification of the flow regimes is the small tube flow regime transition model in RELAP5 and the combined bubbly-slug flow regime in TRACE. We found that calculation results are sensitive to flow regime boundaries of RELAP5 which were modified in order to match the experimental data on flow regimes. Although the flow regime became closer to the experimental one, other predicted thermal hydraulic parameters showed larger discrepancy with the experimental data than with the base case calculations where flow regimes were misidentified. Viet-Anh Phung and Pavel Kudinov Copyright © 2015 Viet-Anh Phung and Pavel Kudinov. All rights reserved. Study of the Reactor Control System of MSHIM in AP1000 Sun, 29 Nov 2015 12:03:35 +0000 According to the mechanism analysis and simulation of power control system of MSHIM in AP1000, a modified MSHIM (Mechanical Shim) control strategy is presented, which employs the error between the reactor coolant average temperature and its reference value as the unique control signal with a P-controller added. The modified MSHIM control strategy is verified by simulations of three typical working conditions. The results show that the modified power control system satisfies the needs of reactor core power control and power distribution control. The conclusions have reference value for the engineering practice. Xinyu Wei and Fuyu Zhao Copyright © 2015 Xinyu Wei and Fuyu Zhao. All rights reserved. NaI(Tl) Detector Efficiency Computation Using Radioactive Parallelepiped Sources Based on Efficiency Transfer Principle Thu, 12 Nov 2015 08:00:03 +0000 The efficiency transfer (ET) principle is considered as a simple numerical simulation method, which can be used to calculate the full-energy peak efficiency (FEPE) of NaI(Tl) scintillation detector over a wide energy range. In this work, the calculations of FEPE are based on computing the effective solid angle ratio between a radioactive point and parallelepiped sources located at various distances from the detector surface. Besides, the attenuation of the photon by the source-to-detector system (detector material, detector end cap, and holder material) was considered and determined. This method is straightforwardly useful in setting up the efficiency calibration curve for NaI(Tl) scintillation detector, when no calibration sources exist in volume shape. The values of the efficiency calculations using theoretical method are compared with the measured ones and the results show that the discrepancies in general for all the measurements are found to be less than 6%. Mohamed S. Badawi, Mona M. Gouda, Ahmed M. El-Khatib, Abouzeid A. Thabet, Ahmed A. Salim, and Mahmoud I. Abbas Copyright © 2015 Mohamed S. Badawi et al. All rights reserved. Thermal-Hydraulic Assessment of W7-X Plasma Vessel Venting System in Case of 40 mm In-Vessel LOCA Tue, 10 Nov 2015 07:07:49 +0000 This paper presents assessment of the capacity of W7-X venting system in response to in-vessel LOCA, rupture of 40 mm diameter pipe during operation mode “baking.” The integral analysis of the coolant release from the cooling system, pressurisation of PV, and response of the venting system is performed using RELAP5 code. The same coolant release rate was introduced to the COCOSYS code, which is a lumped-parameter code developed for analysis of processes in containment of the light water reactors and the detailed analysis of the plasma vessel and the venting system is performed. Different options of coolant release modeling available in COCOSYS are compared to define the base case model, which is further used for assessment of the other parameters, that is, the failure of one burst disk, the temperature in the environment, and the pressure losses in the piping of venting system. The performed analysis identified the best option for coolant release modeling and showed that the capacity of the W7-X venting system is enough to prevent overpressure of the plasma vessel in the case of in-vessel LOCA. E. Urbonavičius and T. Kaliatka Copyright © 2015 E. Urbonavičius and T. Kaliatka. All rights reserved. An Innovative Passive Residual Heat Removal System of an Open-Pool Type Research Reactor with Pump Flywheel and Gravity Core Cooling Tank Mon, 05 Oct 2015 14:14:24 +0000 In an open-pool type research reactor, the primary cooling system can be designed to have a downward flow inside the core during normal operation because of the plate type fuel geometry. There is a flow inversion inside the core from the downward flow by the inertia force of the primary coolant to the upward flow by the natural circulation when the pump is turned off. To delay the flow inversion time, an innovative passive system with pump flywheel and GCCT is developed to remove the residual heat. Before the primary cooling pump starts up, the water level of the GCCT is the same as that of the reactor pool. During the primary cooling pump operation, the water in the GCCT is moved into the reactor pool because of the pump suction head. After the pump stops, the potential head generates a downward flow inside the core by moving the water from the reactor pool to the GCCT and removes the residual heat. When the water levels of the two pools are the same again, the core flow has an inversion of the flow direction, and natural circulation is developed through the flap valves. Kwon-Yeong Lee and Hyun-Gi Yoon Copyright © 2015 Kwon-Yeong Lee and Hyun-Gi Yoon. All rights reserved. Analysis of the EBR-II SHRT-45R Unprotected Loss of Flow Experiment with ERANOS and RELAP Wed, 16 Sep 2015 08:01:43 +0000 This paper presents the results of the analysis of the Unprotected Loss of Flow (ULOF) experiment SHRT-45R performed in the EBR-II fast reactor. These experiments are being analyzed in the scope of a benchmark exercise coordinated by the IAEA. The SHRT-45R benchmark contains a neutronic and a thermal-hydraulic part and results are presented for both. Neutronic calculations are performed with the ERANOS2.0 code in combination with various sets of nuclear data. The thermal-hydraulic evaluation is done with RELAP5-3D. The results are that the major neutronic parameters are well predicted with error margins on the order of 1%. The thermal-hydraulic results are less favourable: a consistent overestimation of the outlet temperature occurs in combination with erroneous flow distribution. Observed differences with measured data cannot be explained easily. The work presented in this paper was undertaken to investigate and validate the effectiveness of the calculational tools and data that are commonly used in our lab for the design and analysis of liquid metal cooled fast reactors. W. F. G. van Rooijen and H. Mochizuki Copyright © 2015 W. F. G. van Rooijen and H. Mochizuki. All rights reserved. Temperature Response of the HTR-10 during the Power Ascension Test Sun, 13 Sep 2015 12:58:08 +0000 The 10 MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10) is the first High Temperature Gas-Cooled Reactor in China. With the objective of raising the reactor power from 30% to 100% rated power, the power ascension test was planned and performed in January 2003. The test results verified the practicability and validity of the HTR-10 power regulation methods. In this study, the power ascension process is preliminarily simulated using the THERMIX code. The code satisfactorily reproduces the reactor transient parameters, including the reactor power, the primary helium pressure, and the primary helium outlet temperature. Reactor internals temperatures are also calculated and compared with the test values recorded by a number of thermocouples. THERMIX correctly simulates the temperature variation tendency for different measuring points, with good to fair agreement between the calculated temperatures and the measured ones. Based on the comparison results, the THERMIX simulation capability for the HTR-10 dynamic characteristics during the power ascension process can be demonstrated. With respect to the reactor safety features, it is of utmost importance that the maximum fuel center temperature during the test process is always much lower than the fuel temperature limit of 1620°C. Fubing Chen, Yujie Dong, and Zuoyi Zhang Copyright © 2015 Fubing Chen et al. All rights reserved. Monte Carlo Alpha Iteration Algorithm for a Subcritical System Analysis Thu, 03 Sep 2015 06:10:03 +0000 The α-k iteration method which searches the fundamental mode alpha-eigenvalue via iterative updates of the fission source distribution has been successfully used for the Monte Carlo (MC) alpha-static calculations of supercritical systems. However, the α-k iteration method for the deep subcritical system analysis suffers from a gigantic number of neutron generations or a huge neutron weight, which leads to an abnormal termination of the MC calculations. In order to stably estimate the prompt neutron decay constant (α) of prompt subcritical systems regardless of subcriticality, we propose a new MC alpha-static calculation method named as the α iteration algorithm. The new method is derived by directly applying the power method for the α-mode eigenvalue equation and its calculation stability is achieved by controlling the number of time source neutrons which are generated in proportion to α divided by neutron speed in MC neutron transport simulations. The effectiveness of the α iteration algorithm is demonstrated for two-group homogeneous problems with varying the subcriticality by comparisons with analytic solutions. The applicability of the proposed method is evaluated for an experimental benchmark of the thorium-loaded accelerator-driven system. Hyung Jin Shim, Sang Hoon Jang, and Soo Min Kang Copyright © 2015 Hyung Jin Shim et al. All rights reserved. Advanced Integral Type Reactors: Passive Safety Design and Experiment Wed, 02 Sep 2015 14:04:11 +0000 Li Shengqiang, Yanping Huang, Luciano Burgazzi, and Annalisa Manera Copyright © 2015 Li Shengqiang et al. 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 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 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 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 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 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 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 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 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 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 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 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 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 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.