Science and Technology of Nuclear Installations The latest articles from Hindawi Publishing Corporation © 2016 , Hindawi Publishing Corporation . All rights reserved. Advanced PHWR Safety Technology: PHWR Challenging Issues for Safe Operation and Long-Term Sustainability Tue, 17 May 2016 08:16:46 +0000 Jin Ho Song, Wei Shen, Malcolm Griffiths, Bo Wook Rhee, YongMann Song, and Masanori Naitoh Copyright © 2016 Jin Ho Song et al. All rights reserved. Scaled-Down Moderator Circulation Test Facility at Korea Atomic Energy Research Institute Tue, 10 May 2016 09:24:09 +0000 Korea Atomic Energy Research Institute (KAERI) started the experimental research on moderator circulation as one of a the national research and development programs from 2012. This research program includes the construction of the moderator circulation test (MCT) facility, production of the validation data for self-reliant computational fluid dynamics (CFD) tools, and development of optical measurement system using the particle image velocimetry (PIV). In the present paper we introduce the scaling analysis performed to extend the scaling criteria suitable for reproducing thermal-hydraulic phenomena in a scaled-down CANDU- (CANada Deuterium Uranium-) 6 moderator tank, a manufacturing status of the 1/4 scale moderator tank. Also, preliminary CFD analysis results for the full-size and scaled-down moderator tanks are carried out to check whether the moderator flow and temperature patterns of both the full-size reactor and scaled-down facility are identical. Hyoung Tae Kim and Bo Wook Rhee Copyright © 2016 Hyoung Tae Kim and Bo Wook Rhee. All rights reserved. Medical Radioisotope Production in a Power-Flattened ADS Fuelled with Uranium and Plutonium Dioxides Wed, 04 May 2016 16:40:31 +0000 This study presents the medical radioisotope production performance of a conceptual accelerator driven system (ADS). Lead-bismuth eutectic (LBE) is selected as target material. The subcritical fuel core is conceptually divided into ten equidistant subzones. The ceramic (natural U, Pu)O2 fuel mixture and the materials used for radioisotope production (copper, gold, cobalt, holmium, rhenium, thulium, mercury, palladium, thallium, molybdenum, and yttrium) are separately prepared as cylindrical rods cladded with carbon/carbon composite (C/C) and these rods are located in the subzones. In order to obtain the flattened power density, percentages of PuO2 in the mixture of UO2 and PuO2 in the subzones are adjusted in radial direction of the fuel zone. Time-dependent calculations are performed at 1000 MW thermal fission power () for one hour using the BURN card. The neutronic results show that the investigated ADS has a high neutronic capability, in terms of medical radioisotope productions, spent fuel transmutation and energy multiplication. Moreover, a good quasiuniform power density is achieved in each material case. The peak-to-average fission power density ratio is in the range of 1.02–1.28. Gizem Bakır, Saltuk Buğra Selçuklu, and Hüseyin Yapıcı Copyright © 2016 Gizem Bakır et al. All rights reserved. Prediction Study on PCI Failure of Reactor Fuel Based on a Radial Basis Function Neural Network Sun, 24 Apr 2016 12:50:04 +0000 Pellet-clad interaction (PCI) is one of the major issues in fuel rod design and reactor core operation in water cooled reactors. The prediction of fuel rod failure by PCI is studied in this paper by the method of radial basis function neural network (RBFNN). The neural network is built through the analysis of the existing experimental data. It is concluded that it is a suitable way to reduce the calculation complexity. A self-organized RBFNN is used in our study, which can vary its structure dynamically in order to maintain the prediction accuracy. For the purpose of the appropriate network complexity and overall computational efficiency, the hidden neurons in the RBFNN can be changed online based on the neuron activity and mutual information. The presented method is tested by the experimental data from the reference, and the results demonstrate its effectiveness. Xinyu Wei, Jiashuang Wan, and Fuyu Zhao Copyright © 2016 Xinyu Wei et al. All rights reserved. Elemental Analysis and Natural Radioactivity Levels of Clay by Gamma Ray Spectrometer and Instrumental Neutron Activation Analysis Mon, 18 Apr 2016 16:44:35 +0000 Due to increased global demand for clay, the present work involves the use of INAA for elemental analysis and pollutants concentration in clay. The samples were collected from Aswan in South Egypt. The samples were irradiated using the thermal neutrons “at the TRIGA Mainz research reactor” and at a neutron flux “of 7 × 10 n/cm s”. Twenty-six elements quantitatively and qualitatively were specified for the first time upon studying the samples. The elements determined are U, Th, Ta, Hf, Lu, Eu, Ce, Ba, Sn, Nb, Rb, Zn, Co, Fe, Cr, Sc, Sm, La, Yb, As, Ga, K, Mn, Na, Ti, and Mg. The concentrations of natural radionuclides 232Th, 226Ra, and 40K were also calculated. Based on these concentrations, to estimate the exposure risk for using clay as raw materials in building materials, the radiation hazard indices such as radium equivalent activities, effective doses rate, and the external hazard indices have been computed. The obtained results were compared with analogous studies carried out in other countries and with the UNSCEAR reports. W. R. Alharbi and A. El-Taher Copyright © 2016 W. R. Alharbi and A. El-Taher. All rights reserved. Statistical Analysis of Loss of Offsite Power Events Thu, 14 Apr 2016 06:12:17 +0000 This paper presents the results of the statistical analysis of the loss of offsite power events (LOOP) registered in four reviewed databases. The reviewed databases include the IRSN (Institut de Radioprotection et de Sûreté Nucléaire) SAPIDE database and the GRS (Gesellschaft für Anlagen- und Reaktorsicherheit mbH) VERA database reviewed over the period from 1992 to 2011. The US NRC (Nuclear Regulatory Commission) Licensee Event Reports (LERs) database and the IAEA International Reporting System (IRS) database were screened for relevant events registered over the period from 1990 to 2013. The number of LOOP events in each year in the analysed period and mode of operation are assessed during the screening. The LOOP frequencies obtained for the French and German nuclear power plants (NPPs) during critical operation are of the same order of magnitude with the plant related events as a dominant contributor. A frequency of one LOOP event per shutdown year is obtained for German NPPs in shutdown mode of operation. For the US NPPs, the obtained LOOP frequency for critical and shutdown mode is comparable to the one assessed in NUREG/CR-6890. Decreasing trend is obtained for the LOOP events registered in three databases (IRSN, GRS, and NRC). Andrija Volkanovski, Antonio Ballesteros Avila, and Miguel Peinador Veira Copyright © 2016 Andrija Volkanovski et al. All rights reserved. Nuclear Power Plant Construction Scheduling Problem with Time Restrictions: A Particle Swarm Optimization Approach Mon, 11 Apr 2016 06:05:17 +0000 In nuclear power plant construction scheduling, a project is generally defined by its dependent preparation time, the time required for construction, and its reactor installation time. The issues of multiple construction teams and multiple reactor installation teams are considered. In this paper, a hierarchical particle swarm optimization algorithm is proposed to solve the nuclear power plant construction scheduling problem and minimize the occurrence of projects failing to achieve deliverables within applicable due times and deadlines. Shang-Kuan Chen, Yen-Wu Ti, and Kuo-Yu Tsai Copyright © 2016 Shang-Kuan Chen et al. All rights reserved. Measurement of the Residual Stresses and Investigation of Their Effects on a Hardfaced Grid Plate due to Thermal Cycling in a Pool Type Sodium-Cooled Fast Reactor Sun, 03 Apr 2016 09:53:38 +0000 In sodium-cooled fast reactors (SFR), grid plate is a critical component which is made of 316 L(N) SS. It is supported on core support structure. The grid plate supports the core subassemblies and maintains their verticality. Most of the components of SFR are made of 316 L(N)/304 L(N) SS and they are in contact with the liquid-metal sodium which acts as a coolant. The peak operating temperature in SFR is 550°C. However, the self-welding starts at 500°C. To avoid self-welding and galling, hardfacing of the grid plate has become necessary. Nickel based cobalt-free colmonoy 5 has been identified as the hardfacing material due to its lower dose rate by Plasma Transferred Arc Welding (PTAW). This paper is concerned with the measurement and investigations of the effects of the residual stress generated due to thermal cycling on a scale-down physical model of the grid plate. Finite element analysis of the hardfaced grid plate model is performed for obtaining residual stresses using elastoplastic analysis and hence the results are validated. The effects of the residual stresses due to thermal cycling on the hardfaced grid plate model are studied. S. Balaguru, Vela Murali, and P. Chellapandi Copyright © 2016 S. Balaguru et al. All rights reserved. Preliminary Study on the Fabrication of Particulate Fuel through Pressureless Sintering Process Wed, 23 Mar 2016 12:20:30 +0000 U-10wt%Zr spherical particles for use as particulate fuel were prepared by centrifugal atomization and subjected to pressureless sintering, which is one of the simplest powder processing techniques. At sintering temperature of 1100°C for 30 or 60 min, all samples ranging from +50 to −325 mesh showed no apparent bonding between the particles. However, at 1150°C (80 min), all samples formed a bulk body and the microstructures showed apparent sintering stages. Particularly, sample B (50–70 mesh) and sample C (70–100 mesh) showed pore characteristics suitable for a particulate fuel. The results suggest that pressureless sinterability for U-10Zr particulate fuel can be improved by adding small-size (–325 mesh) particles. Jong-Hwan Kim, Jung-Won Lee, Ki-Hwan Kim, and Chan-Bock Lee Copyright © 2016 Jong-Hwan Kim et al. All rights reserved. A Denoising Based Autoassociative Model for Robust Sensor Monitoring in Nuclear Power Plants Tue, 22 Mar 2016 06:13:56 +0000 Sensors health monitoring is essentially important for reliable functioning of safety-critical chemical and nuclear power plants. Autoassociative neural network (AANN) based empirical sensor models have widely been reported for sensor calibration monitoring. However, such ill-posed data driven models may result in poor generalization and robustness. To address above-mentioned issues, several regularization heuristics such as training with jitter, weight decay, and cross-validation are suggested in literature. Apart from these regularization heuristics, traditional error gradient based supervised learning algorithms for multilayered AANN models are highly susceptible of being trapped in local optimum. In order to address poor regularization and robust learning issues, here, we propose a denoised autoassociative sensor model (DAASM) based on deep learning framework. Proposed DAASM model comprises multiple hidden layers which are pretrained greedily in an unsupervised fashion under denoising autoencoder architecture. In order to improve robustness, dropout heuristic and domain specific data corruption processes are exercised during unsupervised pretraining phase. The proposed sensor model is trained and tested on sensor data from a PWR type nuclear power plant. Accuracy, autosensitivity, spillover, and sequential probability ratio test (SPRT) based fault detectability metrics are used for performance assessment and comparison with extensively reported five-layer AANN model by Kramer. Ahmad Shaheryar, Xu-Cheng Yin, Hong-Wei Hao, Hazrat Ali, and Khalid Iqbal Copyright © 2016 Ahmad Shaheryar et al. All rights reserved. Analysis of the Relationship between Risk Perception and Willingness to Pay for Nuclear Power Plant Risk Reduction Sun, 20 Mar 2016 09:53:52 +0000 With the adoption of new technologies, more risk is introduced into modern society. Important decisions about new technologies tend to be made by specialists, which can lead to a mismatch of risk perception between citizens and specialists, resulting in high social cost. Using contingent valuation methods, this paper analyzes the relationship between willingness to pay (WTP) and the factors expressed through people’s image of nuclear power plants (NPP), their perception of NPP safety, and how these can be affected by their scientific background level. Results indicate that groups with a high scientific background level tend to have low risk perception level, represented through their image and safety levels. Further, the results show that mean WTP is dependent on scientific background and image levels. It is believed that these results could help decision makers address the mismatch of trust between the public and specialists in terms of new policy. Mirae Yun, Sang Hun Lee, and Hyun Gook Kang Copyright © 2016 Mirae Yun et al. All rights reserved. Emittance Measurement for Beamline Extension at the PET Cyclotron Wed, 16 Mar 2016 13:13:21 +0000 Particle-induced X-ray emission is used for determining the elemental composition of materials. This method uses low-energy protons (of several MeV), which can be obtained from high-energy (of tens MeV) accelerators. Instead of manufacturing an accelerator for generating the MeV protons, the use of a PET cyclotron has been suggested for designing the beamline for multipurpose applications, especially for the PIXE experiment, which has a dedicated high-energy (of tens MeV) accelerator. The beam properties of the cyclotron were determined at this experimental facility by using an external beamline before transferring the ion beam to the experimental chamber. We measured the beam profile and calculated the emittance using the pepper-pot method. The beam profile was measured as the beam current using a wire scanner, and the emittance was measured as the beam distribution at the beam dump using a radiochromic film. We analyzed the measurement results and are planning to use the results obtained in the simulations of external beamline and aligned beamline components. We will consider energy degradation after computing the beamline simulation. The experimental study focused on measuring the emittance from the cyclotron, and the results of this study are presented in this paper. Sae-Hoon Park, Sang-Hoon Lee, and Yu-Seok Kim Copyright © 2016 Sae-Hoon Park et al. All rights reserved. New RELAP5-3D Lead and LBE Thermophysical Properties Implementation for Safety Analysis of Gen IV Reactors Wed, 09 Mar 2016 13:26:21 +0000 The latest versions of RELAP5-3D© code allow the simulation of thermodynamic system, using different type of working fluids, that is, liquid metals, molten salt, diathermic oil, and so forth, thanks to the ATHENA code integration. The RELAP5-3D© water thermophysical properties are largely verified and validated; however there are not so many experiments to generate the liquid metals ones in particular for the Lead and the Lead Bismuth Eutectic. Recently, new and more accurate experimental data are available for liquid metals. The comparison between these state-of-the-art data and the RELAP5-3D© default thermophysical properties shows some discrepancy; therefore a tool for the generation of new properties binary files has been developed. All the available data came from experiments performed at atmospheric pressure. Therefore, to extend the pressure domain below and above this pressure, the tool fits a semiempirical model (soft sphere model with inverse-power-law potential), specific for the liquid metals. New binary files of thermophysical properties, with a detailed mesh grid of point to reduce the code mass error (especially for the Lead), were generated with this tool. Finally, calculations using a simple natural circulation loop were performed to understand the differences between the default and the new properties. P. Balestra, F. Giannetti, G. Caruso, and A. Alfonsi Copyright © 2016 P. Balestra et al. All rights reserved. Numerical Investigation of the Transient Behavior of a Hot Gas Duct under Rapid Depressurization Wed, 09 Mar 2016 12:02:07 +0000 A hot gas duct is an indispensable component for the nuclear-process heat applications of the Very-High-Temperature Reactor (VHTR), which has to fulfill three requirements: to withstand high temperature, high pressure, and large pressure transient. In this paper, numerical investigation of pressure transient is performed for a hot gas duct under rapid depressurization. System depressurization imposes an imploding pressure differential on the internal structural elements of a hot gas duct, the structural integrity of which is susceptible to being damaged. Pressure differential and its imposed duration, which are two key factors to evaluate the damage severity of a hot gas duct under depressurization, are examined in regard to depressurization rate and insulation packing tightness. It is revealed that depressurization rate is a decisive parameter for controlling the pressure differential and its duration, whereas insulating-packing tightness has little effect on them. JingBao Liu, ShengYao Jiang, Tao Ma, and RiQiang Duan Copyright © 2016 JingBao Liu et al. All rights reserved. A Simple Formula for Local Burnup and Isotope Distributions Based on Approximately Constant Relative Reaction Rate Wed, 17 Feb 2016 11:37:59 +0000 A simple and analytical formula is suggested to solve the problems of the local burnup and the isotope distributions. The present method considers two extreme conditions of neutrons penetrating the fuel rod. Based on these considerations, the formula is obtained to calculate the reaction rates of 235U, 238U, and 239Pu and straightforward the local burnup and the isotope distributions. Starting from an initial burnup level, the parameters of the formula are fitted to the reaction rates given by a Monte Carlo (MC) calculation. Then the present formula independently gives very similar results to the MC calculation from the starting to high burnup level but takes just a few minutes. The relative reaction rates are found to be almost independent of the radius (except of  238U) and the burnup, providing a solid background for the present formula. A more realistic examination is also performed when the fuel rods locate in an assembly. A combination of the present formula and the MC calculation is expected to have a nice balance between the numerical accuracy and time consumption. Cenxi Yuan, Xuming Wang, and Shengli Chen Copyright © 2016 Cenxi Yuan et al. 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.