Science and Technology of Nuclear Installations The latest articles from Hindawi Publishing Corporation © 2014 , Hindawi Publishing Corporation . All rights reserved. Simplification of State Transition Diagrams in Average Unavailability Analysis by Using Generalized Perturbation Theory Thu, 17 Apr 2014 14:57:19 +0000 Safety analysis studies in nuclear engineering, more specifically system reliability, usually handle a great number of components, so that computational difficulties may arise. To face the problem of many component systems a method for simplifying the state transition diagram in Markovian reliability analyses has been proposed, using the edges which can be cut, since these latter have a smaller influence on system failure probability. In order to extend the application of GPT (Generalized Perturbation Theory), this work uses GPT formalism to reduce the number of states in a transition diagram, not considering the state probability as the integral quantity of interest, but the mean system unavailability instead. Therefore, after simplifying the original diagram, the mean unavailability for the new system was calculated and the results were very close to those of the original diagram integral quantity (giving a relative error of about 2%), showing that the proposed simplification is quite reasonable and simple to apply. R. C. Vicente, F. C. Silva, P. F. Frutuoso e Melo, and A. C. M. Alvim Copyright © 2014 R. C. Vicente et al. All rights reserved. The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark Mon, 14 Apr 2014 16:53:07 +0000 A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised). Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement. Surian Pinem, Tagor Malem Sembiring, and Peng Hong Liem Copyright © 2014 Surian Pinem et al. All rights reserved. Neutron Radiation Tests about FeCr Slag and Natural Zeolite Loaded Brick Samples Thu, 10 Apr 2014 08:56:39 +0000 Neutron shielding performances of new brick samples are investigated. Brick samples including 10, 20, and 30 percentages of ferrochromium slag (FeCr waste) and natural zeolite are prepared and mechanical properties are obtained. Total macroscopic cross sections are calculated by using results of 4.5 MeV neutron transmission experiments. As a result, neutron shielding capacity of brick samples increases with increasing FeCr slag and natural zeolite percentages. This information could be useful in the area of neutron shielding. Vedat Veli Cay, Mucahit Sutcu, Osman Gencel, and Turgay Korkut Copyright © 2014 Vedat Veli Cay et al. All rights reserved. Study on Thermal-Hydraulic Behavior of an Integral Type Reactor under Heaving Condition Tue, 08 Apr 2014 08:41:05 +0000 A self-developed program was used to study the thermal-hydraulic behavior of an integral type reactor under heaving condition. Comparison of calculated results with the data of experiments performed on a natural circulation loop designed with reference to an integral type reactor of Tsinghua University in inclination, heaving, and rolling motions was carried out. Characteristics of natural circulation in heaving motion and effect of motion parameters on natural circulation were investigated. Results indicated that: (1) long-period heaving motion would lead to more significant influence than inclination and rolling motion; (2) it was an alternating force field which consisted of gravity and an additional force that decided the flow temperature and density difference of natural circulation; (3) effect of strength and cycle of heaving motion on flow fluctuation of natural circulation and condensate depression of heating section outlet was performed. Beibei Feng, Xingtuan Yang, Shengqiang Li, Yanfei Sun, Jiyuan Tu, and Shengyao Jiang Copyright © 2014 Beibei Feng et al. All rights reserved. Review on Recent Developments in Laser Driven Inertial Fusion Mon, 07 Apr 2014 14:01:30 +0000 Discovery of the laser in 1960 hopes were based on using its very high energy concentration within very short pulses of time and very small volumes for energy generation from nuclear fusion as “Inertial Fusion Energy” (IFE), parallel to the efforts to produce energy from “Magnetic Confinement Fusion” (MCF), by burning deuterium-tritium (DT) in high temperature plasmas to helium. Over the years the fusion gain was increased by a number of magnitudes and has reached nearly break-even after numerous difficulties in physics and technology had been solved. After briefly summarizing laser driven IFE, we report how the recently developed lasers with pulses of petawatt power and picosecond duration may open new alternatives for IFE with the goal to possibly ignite solid or low compressed DT fuel thereby creating a simplified reactor scheme. Ultrahigh acceleration of plasma blocks after irradiation of picosecond (PS) laser pulses of around terawatt (TW) power in the range of 1020 cm/s2 was discovered by Sauerbrey (1996) as measured by Doppler effect where the laser intensity was up to about 1018 W/cm2. This is several orders of magnitude higher than acceleration by irradiation based on thermal interaction of lasers has produced. M. Ghoranneviss and A. Salar Elahi Copyright © 2014 M. Ghoranneviss and A. Salar Elahi. All rights reserved. Evaluation of the Detection Efficiency of LYSO Scintillator in the Fiber-Optic Radiation Sensor Sun, 06 Apr 2014 12:23:45 +0000 The aim of this study was to develop and evaluate fiber-optic sensors for the remote detection of gamma rays in areas that are difficult to access, such as a spent fuel pool. The fiber-optic sensor consists of a light-generating probe, such as scintillators for radiation detection, plastic optical fibers, and light-measuring devices, such as PMT. The (Lu,Y)2SiO5:Ce(LYSO:Ce) scintillator was chosen as the light-generating probe. The (Lu,Y)2SiO5:Ce(LYSO:Ce) scintillator has higher scintillation efficiency than the others and transmits light well through an optical fiber because its refraction index is similar to the refractive index of the optical fiber. The fiber-optic radiation sensor using the (Lu,Y)2SiO5:Ce(LYSO:Ce) scintillator was evaluated in terms of the detection efficiency and reproducibility for examining its applicability as a radiation sensor. Chan Hee Park, Arim Lee, Rinah Kim, and Joo Hyun Moon Copyright © 2014 Chan Hee Park et al. All rights reserved. Identification of Industrial Furnace Temperature for Sintering Process in Nuclear Fuel Fabrication Using NARX Neural Networks Thu, 03 Apr 2014 13:39:26 +0000 Nonlinear system identification is becoming an important tool which can be used to improve control performance and achieve robust fault-tolerant behavior. Among the different nonlinear identification techniques, methods based on neural network model are gradually becoming established not only in the academia, but also in industrial application. An identification scheme of nonlinear systems for sintering furnace temperature in nuclear fuel fabrication using neural network autoregressive with exogenous inputs (NNARX) model investigated in this paper. The main contribution of this paper is to identify the appropriate model and structure to be applied in control temperature in the sintering process in nuclear fuel fabrication, that is, a nonlinear dynamical system. Satisfactory agreement between identified and experimental data is found with normalized sum square error 1 for heating step and for soaking step. That result shows the model successfully predict the evolution of the temperature in the furnace. Dede Sutarya and Benyamin Kusumoputro Copyright © 2014 Dede Sutarya and Benyamin Kusumoputro. All rights reserved. Natural Radioactivity Measurements and Radiation Dose Estimation in Some Sedimentary Rock Samples in Turkey Wed, 02 Apr 2014 14:29:27 +0000 The natural radioactivity existed since creation of the universe due to the long life time of some radionuclides. This natural radioactivity is caused by γ-radiation originating from the uranium and thorium series and 40K. In this study, the gamma radiation has been measured to determine natural radioactivity of 238U, 232Th, and 40K in collected sedimentary rock samples in different places of Turkey. The measurements have been performed using γ-ray spectrometer containing NaI(Tl) detector and multichannel analyser (MCA). Absorbed dose rate (D), annual effective dose (AED), radium equivalent activities (), external hazard index (), and internal hazard index () associated with the natural radionuclide were calculated to assess the radiation hazard of the natural radioactivity in the sedimentary rock samples. The average values of absorbed dose rate in air (D), annual effective dose (AED), radium equivalent activity (), external hazard index (), and internal hazard index () were calculated and these were 45.425 nGy/h, 0.056 mSv/y, 99.014 Bq/kg, 0.267, and 0.361, respectively. I. Akkurt and K. Günoğlu Copyright © 2014 I. Akkurt and K. Günoğlu. All rights reserved. Effect of Flow Blockage on the Coolability during Reflood in a 2 2 Rod Bundle Tue, 01 Apr 2014 16:10:22 +0000 During the reflood phase of a large-break loss-of-coolant accident (LBLOCA) in a pressurized-water reactor (PWR), the fuel rods can be ballooned or rearranged owing to an increase in the temperature and internal pressure of the fuel rods. In this study, an experimental study was performed to understand the thermal behavior and effect of the ballooned region on the coolability using a 2 × 2 rod bundle test facility. The electrically heated rod bundle was used and the ballooning shape of the rods was simulated by superimposing hollow sleeves, which have a 90% blockage ratio. Forced reflood tests were performed to examine the transient two-phase heat transfer behavior for different reflood rates and rod powers. The droplet behaviors were also investigated by measuring the velocity and size of droplets near the blockage region. The results showed that the heat transfer was enhanced in the downstream of the blockage region, owing to the reduced flow area of the subchannel, intensification of turbulence, and deposition of the droplet. Kihwan Kim, Byung-Jae Kim, Young-Jung Youn, Hae-Seob Choi, Sang-Ki Moon, and Chul-Hwa Song Copyright © 2014 Kihwan Kim et al. All rights reserved. Study on the Behaviors of a Conceptual Passive Containment Cooling System Wed, 26 Mar 2014 10:36:35 +0000 The containment is an ultimate and important barrier to mitigate the consequences after the release of mass and energy during such scenarios as loss of coolant accident (LOCA) or main steam line break (MSLB). In this investigation, a passive containment cooling system (PCCS) concept is proposed for a large dry concrete containment. The system is composed of series of heat exchangers, long connecting pipes with relatively large diameter, valves, and a water tank, which is located at the top of the system and serves as the final heat sink. The performance of the system is numerically studied in detail under different conditions. In addition, the influences of condensation heat transfer conditions and containment environment temperature conditions are also studied on the behaviors of the system. The results reveal that four distinct operating stages could be experienced as follows: startup stage, single phase quasisteady stage, flashing speed-up transient stage, and flashing dominated quasisteady operating stage. Furthermore, the mechanisms of system behaviors are thus analyzed. Moreover, the feasibility of the system is also discussed to meet the design purpose for the containment integrity requirement. Considering the passive feature and the compactness of the system, the proposed PCCS is promising for the advanced integral type reactor. Jianjun Wang, Xueqing Guo, Shengzhi Yu, Baowei Cai, Zhongning Sun, and Changqi Yan Copyright © 2014 Jianjun Wang et al. All rights reserved. Preliminary Development of Thermal Power Calculation Code H-Power for a Supercritical Water Reactor Wed, 26 Mar 2014 10:36:03 +0000 SCWR (Supercritical Water Reactor) is one of the promising Generation IV nuclear systems, which has higher thermal power efficiency than current pressurized water reactor. It is necessary to perform the thermal equilibrium and thermal power calculation for the conceptual design and further monitoring and calibration of the SCWR. One visual software named H-Power was developed to calculate thermal power and its uncertainty of SCWR, in which the advanced IAPWS-IF97 industrial formulation was used to calculate the thermodynamic properties of water and steam. The ISO-5167-4: 2003 standard was incorporated in the code as the basis of orifice plate to compute the flow rate. New heat balance model and uncertainty estimate have also been included in the code. In order to validate H-Power, an assessment was carried out by using data published by US and Qinshan Phase II. The results showed that H-Power was able to estimate the thermal power of SCWR. Fan Zhang, Dao-gang Lu, Dan-ting Sui, Bo Yuan, and Chao Guo Copyright © 2014 Fan Zhang et al. All rights reserved. Research of Dosimetry Parameters in Small Electron Beams Tue, 25 Mar 2014 11:14:19 +0000 In this study, dose distributions and outputs of circular fields with dimensions of 5 cm and smaller, for 6 and 9 MeV nominal energies from the Siemens ONCOR Linac, were measured and compared with data from a treatment planning system using the pencil beam algorithm in electron beam calculations. All dose distribution measurements were performed using the GafChromic EBT film; these measurements were compared with data that were obtained from the Computerized Medical Systems (CMS) XiO treatment planning system (TPS). Output measurements were performed using GafChromic EBT film, an Advanced Markus ion chamber, and thermoluminescent dosimetry (TLD). Although it is used in many clinics, there is not a substantial amount of detailed information in the literature about use of the pencil beam algorithm to model electron beams. Output factors were consistent; differences from the values obtained from the TPS were at maximum. When the dose distributions from the TPS were compared with the measurements from the ion chamber and GafChromic EBT films, it was observed that the results were consistent with 2 cm diameter fields and larger, but the outputs for 1 cm diameter fields and smaller were not consistent. Nazmiye Donmez Kesen, Aydin Cakir, Murat Okutan, and Hatice Bilge Copyright © 2014 Nazmiye Donmez Kesen et al. All rights reserved. Operation of Shared Systems via a Common Control System in a Multi-Modular Plant Thu, 20 Mar 2014 09:42:56 +0000 Integral type reactors may need to be grouped to produce as much energy as a utility demands due to the small electrical output of an individual reactor. Sharing of systems among modules at a nuclear plant site is economically beneficial. Operation of systems shared between modules in a multi-modular plant is an issue never met in current NPPs, which may impact human performance. A design of operation of the shared systems via a common control system is presented as a technical approach to solve the problem. Modules and shared systems are controlled in independent network domains, respectively. Different from current NPPs, a limitation of operation authorities corresponding to certain modules and shared systems is defined to minimize the operation confusion between modules by one operator and to minimize the operation confusion of shared systems by different operators. Different characteristics of the shared system are analyzed, and different operation and control strategies are presented. An example is given as an application of the operation strategies. The operation design of the multi-modular system is in the preliminary stage, and, as an concept design, more verification and validation is needed in further works. Jia Qianqian, Huang Xiaojin, and Zhang Liangju Copyright © 2014 Jia Qianqian et al. All rights reserved. The Effect of Beam Intensity on Temperature Distribution in ADS Windowless Lead-Bismuth Eutectic Spallation Target Wed, 19 Mar 2014 07:31:17 +0000 The spallation target is the component coupling the accelerator and the reactor and is regarded as the “heart” of the accelerator driven system (ADS). Heavy liquid metal lead-bismuth eutectic (LBE) is served as core coolant and spallation material to carry away heat deposition of spallation reaction and produce high flux neutron. So it is very important to study the heat transfer process in the target. In this paper, the steady-state flow pattern has been numerically obtained and taken as the input for the nuclear physics calculation, and then the distribution of the extreme large power density of the heat load is imported back to the computational fluid dynamics as the source term in the energy equation. Through the coupling, the transient and steady-state temperature distribution in the windowless spallation target is obtained and analyzed based on the flow process and heat transfer. Comparison of the temperature distribution with the different beam intensity shows that its shape is the same as broken wing of the butterfly. Nevertheless, the maximum temperature as well as the temperature gradient is different. The results play an important role and can be applied to the further design and optimization of the ADS windowless spallation target. Jie Liu, Lei Gao, and Wen-qiang Lu Copyright © 2014 Jie Liu et al. All rights reserved. Decomposition Analysis of Void Reactivity Coefficient for Innovative and Modified BWR Assemblies Tue, 18 Mar 2014 12:16:37 +0000 The decomposition analysis of void reactivity coefficient for innovative BWR assemblies is presented in this paper. The innovative assemblies were loaded with high enrichment UO2 and MOX fuels. Additionally the impact of the moderation enhancement on the void reactivity coefficient through a full fuel burnup discharge interval was investigated for the innovative assembly with MOX fuel. For the numerical analysis the TRITON functional module of SCALE code with ENDF/B-VI cross section library was applied. The obtained results indicate the influence of the most important isotopes to the void reactivity behaviour over a fuel burnup interval of 70 GWd/t for both UO2 and MOX fuels. From the neutronic safety concern positive void reactivity coefficient values are observed for MOX fuel at the beginning of the fuel irradiation cycle. For extra-moderated assembly designs, implementing 8 and 12 water holes, the neutron spectrum softening is achieved and consequently the lower void reactivity values. Variations in void reactivity coefficient values are explained by fulfilled decomposition analysis based on neutrons absorption reactions for separate isotopes. Andrius Slavickas, Raimondas Pabarčius, Aurimas Tonkūnas, and Gediminas Stankūnas Copyright © 2014 Andrius Slavickas et al. All rights reserved. The Stress and Reliability Analysis of HTR’s Graphite Component Tue, 18 Mar 2014 07:07:42 +0000 The high temperature gas cooled reactor (HTR) is developing rapidly toward a modular, compact, and integral direction. As the main structure material, graphite plays a very important role in HTR engineering, and the reliability of graphite component has a close relationship with the integrity of reactor core. The graphite components are subjected to high temperature and fast neutron irradiation simultaneously during normal operation of the reactor. With the stress accumulation induced by high temperature and irradiation, the failure risk of graphite components increases constantly. Therefore it is necessary to study and simulate the mechanical behavior of graphite component under in-core working conditions and forecast the internal stress accumulation history and the variation of reliability. The work of this paper focuses on the mechanical analysis of pebble-bed type HTR's graphite brick. The analysis process is comprised of two procedures, stress analysis and reliability analysis. Three different creep models and two different reliability models are reviewed and taken into account in simulation. The stress and failure probability calculation results are obtained and discussed. The results gained with various models are highly consistent, and the discrepancies are acceptable. Xiang Fang, Haitao Wang, and Suyuan Yu Copyright © 2014 Xiang Fang et al. All rights reserved. Feasibility Study of 1/3 Thorium-Plutonium Mixed Oxide Core Mon, 17 Mar 2014 11:12:47 +0000 Thorium-plutonium mixed oxide (Th-MOX) fuel has become one of the most promising solutions to reduce a large and increasing plutonium stockpile. Compared with traditional uranium-plutonium mixed oxide (U-MOX) fuels, Th-MOX fuel has higher consumption rate of plutonium in LWRs. Besides, thorium based fuels have improved thermomechanical material properties compared with traditional U-MOX fuels. Previous studies on a full Th-MOX core have shown reduced efficiency in reactivity control mechanisms, stronger reactivity feedback, and a significantly lower fraction of delayed neutrons compared with a traditional uranium oxide (UOX) core. These problems complicate the implementation of a full Th-MOX core in a similar way as for a traditional U-MOX core. In order to reduce and avoid some of these issues, the introduction of a lower fraction of Th-MOX fuel in the core is proposed. In this study, one-third of the assemblies are Th-MOX fuel, and the rest are traditional UOX fuel. The feasibility study is based on the Swedish Ringhals-3 PWR. The results show that the core characteristics are more similar to a traditional UOX core, and the fraction of delayed neutrons is within acceptable limits. Moreover, the damping of axial xenon oscillations induced by control rod insertions is almost 5 times more effective for the 1/3 Th-MOX core compared with the standard core. Cheuk Wah Lau, Henrik Nylén, Klara Insulander Björk, and Urban Sandberg Copyright © 2014 Cheuk Wah Lau et al. All rights reserved. Detection Efficiency of NaI(Tl) Detector in 511–1332 keV Energy Range Wed, 12 Mar 2014 07:03:17 +0000 As it is important to obtain accurate analytical result in an experimental research, this required quality control of the experimental system. Gamma spectrometry system can be used in a variety of different fields such as radiation and medical physics. In this paper the absolute efficiency, peak to valley ratio, and energy resolution of a NaI(Tl) detector were determined experimentally for 511, 662, 835, 1173, 1275, and 1332 keV photon energies obtained from 22Na, 54Mn, 60Co, and 137Cs radioactive sources. I. Akkurt, K. Gunoglu, and S. S. Arda Copyright © 2014 I. Akkurt et al. All rights reserved. A Simplified Supercritical Fast Reactor with Thorium Fuel Mon, 10 Mar 2014 16:25:09 +0000 Super-Critical water-cooled Fast Reactor (SCFR) is a feasible option for the Gen-IV SCWR designs, in which much less moderator and thus coolant are needed for transferring the fission heat from the core compared with the traditional LWRs. The fast spectrum of SCFR is useful for fuel breeding and thorium utilization, which is then beneficial for enhancing the sustainability of the nuclear fuel cycle. A SCFR core is constructed in this work, with the aim of simplifying the mechanical structure and keeping negative coolant void reactivity during the whole core life. A core burnup simulation scheme based on Monte Carlo lattice homogenization is adopted in this study, and the reactor physics analysis has been performed with DU-MOX and Th-MOX fuel. The main issues discussed include the fuel conversion ratio and the coolant void reactivity. The analysis shows that thorium-based fuel can provide inherent safety for SCFR without use of blanket, which is favorable for the mechanical design of SCFR. Peng Zhang, Kan Wang, and Ganglin Yu Copyright © 2014 Peng Zhang et al. All rights reserved. Applying UPC Scaling-Up Methodology to the LSTF-PKL Counterpart Test Sun, 02 Mar 2014 12:55:15 +0000 In the framework of the nodalization qualification process and quality guarantee procedures and following the guidelines of Kv-scaled analysis and UMAE methodology, further development has been performed by UPC team resulting in a scaling-up methodology. Such methodology has been applied in this paper for analyzing discrepancies that appear between the simulations of two counterpart tests. It allows the analysis of scaling-down criterion used for the design of an ITF and also the investigation of the differences of configuration between an ITF and a particular NPP. For analyzing both, it applies two concepts “scaled-up nodalizations” and “hybrid nodalizations.” The result of this activity is the explanation of appeared distortions and its final goal is to qualify nodalizations for their use in the analysis of equivalent scenarios at an NPP scale. In this sense, the experimental data obtained in the OECD/NEA PKL-2 and ROSA-2 projects as counterpart test are of a great value for the testing of the present methodology. The results of the posttest calculations of LSTF-PKL counterpart tests have allowed the analyst to define which phenomena could be well reproduced by their nodalizations and which not, in this way establishing the basis for a future extrapolation to an NPP scaled calculation. The application of the UPC scaling up methodology has demonstrated that selected phenomena can be scaled-up and explained between counterpart simulations by carefully considering the differences in scale and design. V. Martinez-Quiroga, F. Reventos, and J. Freixa Copyright © 2014 V. Martinez-Quiroga et al. All rights reserved. Selected Papers from OECD-NEA PSBT Benchmark Wed, 26 Feb 2014 11:17:43 +0000 Maria Avramova, Annalisa Manera, David Novog, Diana Cuervo, and Alessandro Petruzzi Copyright © 2014 Maria Avramova et al. All rights reserved. Experimental Investigation on Flow-Induced Vibration of Fuel Rods in Supercritical Water Loop Mon, 24 Feb 2014 11:07:24 +0000 The supercritical water-cooled reactor (SCWR) is one of the most promising Generation IV reactors. In order to make the fuel qualification test for SCWR, a research plan is proposed to test a small scale fuel assembly in a supercritical water loop. To ensure the structure safety of fuel assembly in the loop, a flow-induced vibration experiment was carried out to investigate the vibration behavior of fuel rods, especially the vibration caused by leakage flow. From the experiment result, it can be found that: the vibration of rods is mainly caused by turbulence when flow rate is low. However, the effects of leakage flow become obvious as flow rate increases, which could changes the distribution of vibrational energy in spectrum, increasing the vibrational energy in high-frequency band. That is detrimental to the structure safety of fuel rods. Therefore, it is more reasonable to improve the design by using the spacers with blind hole, which can eliminate the leakage flow, to assemble the fuel rods in supercritical water loop. On the other hand, the experimental result could provide a benchmark for the theoretical studies to validate the applicability of boundary condition set for the leakage-flow-induced vibration. Licun Wu, Daogang Lu, and Yu Liu Copyright © 2014 Licun Wu et al. All rights reserved. Radiation Shielding Properties Comparison of Pb-Based Silicate, Borate, and Phosphate Glass Matrices Sun, 23 Feb 2014 12:47:31 +0000 Theoretical calculations of mass attenuation coefficients, partial interactions, atomic cross-section, and effective atomic numbers of PbO-based silicate, borate, and phosphate glass systems have been investigated at 662 keV. PbO-based silicate glass has been found with the highest total mass attenuation coefficient and then phosphate and borate glasses, respectively. Compton scattering has been the dominate interaction contributed to the different total attenuation coefficients in each of the glass matrices. The silicate and phosphate glass systems are more appropriate choices as lead-based radiation shielding glass than the borate glass system. Moreover, comparison of results has shown that the glasses possess better shielding properties than standard shielding concretes, suggesting a smaller size requirement in addition to transparency in the visible region. Suwimon Ruengsri Copyright © 2014 Suwimon Ruengsri. All rights reserved. Subchannel Analysis of Wire Wrapped SCWR Assembly Sun, 23 Feb 2014 10:42:22 +0000 Application of wire wrap spacers in SCWR can reduce pressure drop and obtain better mixing capability. As a consequence, the required coolant pumping power is decreased and the coolant temperature profile inside the fuel bundle is flattened which will obviously decrease the peak cladding temperature. The distributed resistance model for wire wrap was developed and implemented in ATHAS subchannel analysis code. The HPLWR wire wrapped assembly was analyzed. The results show that: (1) the assembly with wire wrap can obtain a more uniform coolant temperature profile than the grid spaced assembly, which will result in a lower peak cladding temperature; (2) the pressure drop in a wire wrapped assembly is less than that in a grid spaced assembly, which can reduce the operating power of pump effectively; (3) the wire wrap pitch has significant effect on the flow in the assembly. Smaller will result in stronger cross flow a more uniform coolant temperature profile, and also a higher pressure drop. Jianqiang Shan, Henan Wang, Wei Liu, Linxing Song, Xuanxiang Chen, and Yang Jiang Copyright © 2014 Jianqiang Shan et al. All rights reserved. Direct Numerical Simulation and Visualization of Subcooled Pool Boiling Wed, 19 Feb 2014 08:06:08 +0000 A direct numerical simulation of the boiling phenomena is one of the promising approaches in order to clarify their heat transfer characteristics and discuss the mechanism. During these decades, many DNS procedures have been developed according to the recent high performance computers and computational technologies. In this paper, the state of the art of direct numerical simulation of the pool boiling phenomena during mostly two decades is briefly summarized at first, and then the nonempirical boiling and condensation model proposed by the authors is introduced into the MARS (MultiInterface Advection and Reconstruction Solver developed by the authors). On the other hand, in order to clarify the boiling bubble behaviors under the subcooled conditions, the subcooled pool boiling experiments are also performed by using a high speed and high spatial resolution camera with a highly magnified telescope. Resulting from the numerical simulations of the subcooled pool boiling phenomena, the numerical results obtained by the MARS are validated by being compared to the experimental ones and the existing analytical solutions. The numerical results regarding the time evolution of the boiling bubble departure process under the subcooled conditions show a very good agreement with the experimental results. In conclusion, it can be said that the proposed nonempirical boiling and condensation model combined with the MARS has been validated. Tomoaki Kunugi and Yasuo Ose Copyright © 2014 Tomoaki Kunugi and Yasuo Ose. All rights reserved. A Performance Evaluation of a Notebook PC under a High Dose-Rate Gamma Ray Irradiation Test Wed, 12 Feb 2014 08:43:49 +0000 We describe the performance of a notebook PC under a high dose-rate gamma ray irradiation test. A notebook PC, which is small and light weight, is generally used as the control unit of a robot system and loaded onto the robot body. Using TEPCO’s CAMS (containment atmospheric monitoring system) data, the gamma ray dose rate before and after a hydrogen explosion in reactor units 1–3 of the Fukushima nuclear power plant was more than 150 Gy/h. To use a notebook PC as the control unit of a robot system entering a reactor building to mitigate the severe accident situation of a nuclear power plant, the performance of the notebook PC under such intense gamma-irradiation fields should be evaluated. Under a similar dose-rate (150 Gy/h) gamma ray environment, the performances of different notebook PCs were evaluated. In addition, a simple method for a performance evaluation of a notebook PC under a high dose-rate gamma ray irradiation test is proposed. Three notebook PCs were tested to verify the method proposed in this paper. Jai Wan Cho and Kyung Min Jeong Copyright © 2014 Jai Wan Cho and Kyung Min Jeong. All rights reserved. Study and Evaluation of Innovative Fuel Handling Systems for Sodium-Cooled Fast Reactors: Fuel Handling Route Optimization Tue, 11 Feb 2014 14:14:36 +0000 The research for technological improvement and innovation in sodium-cooled fast reactor is a matter of concern in fuel handling systems in a view to perform a better load factor of the reactor thanks to a quicker fuelling/defueling process. An optimized fuel handling route will also limit its investment cost. In that field, CEA has engaged some innovation study either of complete FHR or on the optimization of some specific components. This paper presents the study of three SFR fuel handling route fully described and compared to a reference FHR option. In those three FHR, two use a gas corridor to transfer spent and fresh fuel assembly and the third uses two casks with a sodium pot to evacuate and load an assembly in parallel. All of them are designed for the ASTRID reactor (1500 MWth) but can be extrapolated to power reactors and are compatible with the mutualisation of one FHS coupled with two reactors. These three concepts are then intercompared and evaluated with the reference FHR according to four criteria: performances, risk assessment, investment cost, and qualification time. This analysis reveals that the “mixed way” FHR presents interesting solutions mainly in terms of design simplicity and time reduction. Therefore its study will be pursued for ASTRID as an alternative option. Franck Dechelette, Franck Morin, Guy Laffont, Gilles Rodriguez, Emmanuel Sanseigne, Sébastien Christin, Xavier Mognot, and Aurélien Morcillo Copyright © 2014 Franck Dechelette et al. All rights reserved. Influence of Spacer Grid Outer Strap on Fuel Assembly Thermal Hydraulic Performance Tue, 11 Feb 2014 06:18:41 +0000 The outer strap as a typical structure of a spacer grid enhances the mechanical strength, decreases hang-up susceptibility, and also influences thermal hydraulic performance, for example, pressure loss, mixing performance, and flow distribution. In the present study, a typical grid spacer with different outer strap designs is adopted to investigate the influence of outer strap design on fuel assembly thermal hydraulic performance by using a commercial computational fluid dynamics (CFD) code, ANSYS CFX, and a subchannel analysis code, FLICA. To simulate the outer straps’ influence between fuel assemblies downstream, four quarter-bundles from neighboring fuel assemblies are constructed to form the computational domain. The results show that the outer strap design has a major impact on cross-flow between fuel assemblies and temperature distribution within the fuel assembly. Jingwen Yan, Yuxiang Zhang, Baowen Yang, Weicai Li, and Yuemin Zhou Copyright © 2014 Jingwen Yan et al. All rights reserved. The Use of System Codes in Scaling Studies: Relevant Techniques for Qualifying NPP Nodalizations for Particular Scenarios Mon, 10 Feb 2014 13:19:46 +0000 System codes along with necessary nodalizations are valuable tools for thermal hydraulic safety analysis. Qualifying both codes and nodalizations is an essential step prior to their use in any significant study involving code calculations. Since most existing experimental data come from tests performed on the small scale, any qualification process must therefore address scale considerations. This paper describes the methodology developed at the Technical University of Catalonia in order to contribute to the qualification of Nuclear Power Plant nodalizations by means of scale disquisitions. The techniques that are presented include the so-called -scaled calculation approach as well as the use of “hybrid nodalizations” and “scaled-up nodalizations.” These methods have revealed themselves to be very helpful in producing the required qualification and in promoting further improvements in nodalization. The paper explains both the concepts and the general guidelines of the method, while an accompanying paper will complete the presentation of the methodology as well as showing the results of the analysis of scaling discrepancies that appeared during the posttest simulations of PKL-LSTF counterpart tests performed on the PKL-III and ROSA-2 OECD/NEA Projects. Both articles together produce the complete description of the methodology that has been developed in the framework of the use of NPP nodalizations in the support to plant operation and control. V. Martinez-Quiroga and F. Reventos Copyright © 2014 V. Martinez-Quiroga and F. Reventos. All rights reserved. Nuclear Power Plant Maintenance Optimization with Heuristic Algorithm Sun, 09 Feb 2014 07:29:59 +0000 The test and maintenance activities are conducted in the nuclear power plants in order to prevent or limit failures resulting from the ageing or deterioration. The components and systems are partially or fully unavailable during the maintenance activities. This is especially important for the safety systems and corresponding equipment because they are important contributors to the overall nuclear power plant safety. A novel method for optimization of the maintenance activities in the nuclear power plant considering the plant safety is developed and presented. The objective function of the optimization is the mean value of the selected risk measure. The risk measure is assessed from the minimal cut sets identified in the Probabilistic Safety Assessment. The optimal solution of the objective function is estimated with genetic algorithm. The proposed method is applied on probabilistic safety analysis model of the selected safety system of the reference nuclear power plant. Obtained results show that optimization of maintenance decreases the risk and thus improves the plant safety. The implications of the consideration of different constraints on the obtained results are investigated and presented. The future prospects for the optimization of the maintenance activities in the nuclear power plants with the presented method are discussed. Andrija Volkanovski and Leon Cizelj Copyright © 2014 Andrija Volkanovski and Leon Cizelj. All rights reserved.