bluuebunny/crossref_metadata_2025_split huggingface dataset:
Best estimate analysis of PHEBUS FPT1 experiment bundle phase using ASTEC code ICARE module
Authors: Vileiniskis V. , Kaliatka A. 
Published: December 2013

Abstract
The main objective of the PHEBUS FPT1 experiment was to study the release of fission products and their subsequent transport and deposition in the primary circuit and the containment under the conditions representative of a severe accident of a Pressurised Water Reactor. The FPT1 test was divided into the bundle degradation, aerosol, washing and chemistry phases. The objective of this paper is related to the best estimate analysis of the bundle degradation phase using ASTEC code ICARE module. GRS (Germany) best estimate method was used for the simulation of FPT1 test by employing SUSA 3.5 package. The performed best estimate analysis showed that ICARE module is capable to model the main severe accidents phenomena in the fuel bundle, heating and melting, taking into account the used physical and code parameters uncertainties.
An analytical benchmark of MYRRHA ADS in cylindrical geometry
Authors: Atak H. , Yilmazer A. 
Published: December 2013

Abstract
In this study, the steady and transient neutronic behaviour of MYRRHA ADS is investigated. For this purpose, a recently proposed analytical benchmark of the diffusion kinetics as 1D slab model of the MYRRHA ADS concept developed in Belgium has been extended to the cylindrical geometry which represents the system more realistically. Analytical calculations are performed using the Customized Solution Method and numerical Laplace inversion techniques such as Fixed-Talbot and Gaver-Wynn-Rho algorithms. Results are compared with the finite element program FLEXPDE® and they are found to be in complete agreement. The necessity of modeling the MYRHHA reactor in cylindrical geometry rather than slab geometry to obtain more realistic benchmark results is demonstrated.
Emergency exercises conducted by the BMU
Authors: Bühling A. , Kuhlen J. 
Published: December 2013

Abstract
According to the Precautionary Radiation Protection Act, responsibility for crisis management in the case of events which do not pose an imminent danger lies with the Federal Ministry for the Environment (BMU). To fulfil this obligation, technical as well as organisational arrangements are required, and the BMU staff has to be made familiar with the relevant tasks. To accomplish this, suitable exercises are essential. In recent years exercise series have been initiated by international organisations and by the EU, which deal with practical experience in managing urgent radiological information exchange and crisis management according to international agreements. They cover both the early phase of nuclear accidents and the specific requirements of the late phase.
Maintaining competence in nuclear safety and waste management research by BMBF
Authors: Ehrlich A. 
Published: December 2013

Abstract
Germany is to undertake a structured phasing-out of power generation from nuclear energy. Until the last nuclear power plant is shut down, safety must be guaranteed in line with the very latest developments in science and technology. The R&D work performed is in accord with the resolution for the structured phasing-out of the use of nuclear power.
The Federal Ministry of Education and Research (BMBF) with its “Basic Energy Research 2020+” funding concept supplements institutionally funded work of Helmholtz Institutes in a few core areas to further extend co-operation with universities. Close coordination between institutional and project funding will be ensured via the Alliance for Competence in Nuclear Technology in Germany (“Kompetenzverbund Kerntechnik”).
In the area of nuclear safety and disposal research, R&D is carried out on the scientific and technological aspects of safety in existing nuclear reactors, the safety of nuclear disposal, the minimisation of highly radioactive substances ultimately requiring disposal and radiation research. Special attention is to be paid within this concept to the funding of young scientists. In addition to doctorate posts in research projects, special funding instruments are to be offered to promote the next generation of scientists.
Condensation induced water hammer and steam assisted gravity drainage in the Athabasca oil sands
Authors: Carlson M. R. 
Published: December 2013

Abstract
Most people will have been exposed to some aspect of the debate about the Athabasca Oil Sands in North-Eastern Alberta and the significant role that the oil sands are expected to play in supplying conventional fossil fuels. Part of the bitumen is recovered from mines and part is recovered from in situ projects utilizing the Steam Assisted Gravity Drainage Process (SAGD). SAGD utilizes a considerable amount of steam, that is injected into geological formations. Hot water, bitumen and some vapour are recovered from the production wells. With significant steam generation, transmission and injection, there is the very real possibility of condensation induced water hammers. There have been a number of catastrophic failures to date. The intent of the paper is to provide interesting background information on the in situ oil sands industry. More importantly, to show some interesting and broader applications of thermalhydraulics developed in the nuclear industry. The expertise developed may have potential markets, with some adaptation, to the oil sands industry. Finally, there has been some discussion about using nuclear power for steam generation in the oil sands.
An analytical solution for the consideration of the effect of adjacent fuel assemblies; extension to VVER-440 type fuel assemblies
Authors: Merk B. , Rohde U. 
Published: December 2013

Abstract
A new analytical method is described to deal with the Leakage Environmental Effect for the cross section production. The method is based on the one dimensional analytical solution of the two-group diffusion equation for two adjacent, homogenized fuel assemblies. The high quality of the results for this highly efficient method is demonstrated for square fuel assemblies. In additional tests the limiting case for the transferability of the concept to the small hexagonal VVER-440 type fuel assemblies is shown.
Effect of spacer grid mixing vanes on coolant outlet temperature distribution
Authors: Rämä T. , Lahtinen T. , Brandt T. , Toppila T. 
Published: December 2013

Abstract
In Loviisa VVER-440-type NPP the coolant outlet temperature of the hot subchannel is constantly monitored during the operation. According to the authority requirement the maximum subchannel outlet temperature must not exceed the saturation temperature. Coolant temperature distribution inside the fuel assembly is affected by the efficiency of the coolant mixing. In order to enhance the coolant mixing the fuel manufacturer is introducing the additional mixing vanes on the fuel bundle spacer grids. In the paper the effect of the different mixing vane modifications is studied with computational fluid dynamics (CFD) simulation. Goal of the modelling is to find vane modifications with which sufficient mixing is reached with acceptable increase in the spacer grid pressure loss. The results of the studies are discussed in the paper.
Analytical investigation of the properties of the neutron noise induced by vibrating absorber and fuel rods
Authors: Jonsson A. , Tran H. N. , Dykin V. , Pázsit I. 
Published: December 2013

Abstract
Analytical solution methods for the neutron noise in a one-dimensional multi-region system in two-group theory, which have so far been based on the adjoint function technique, are extended here to using the forward Green's function technique. The forward Green's functions were calculated analytically for a noise source in a core surrounded by reflector regions at both sides. It is shown that with symbolic computation methods, the forward Green's function can be used for the calculation of the space- and frequency-dependent noise in the first order approximation for arbitrary noise sources which have an analytical representation. The properties of the induced neutron noise were investigated for vibrations of both absorbers and fuel assemblies, with two representations of the noise sources: a point-like source which corresponds to the vibrations of a fuel rod, and a finite width source which corresponds to vibrations of a fuel assembly. The contributions of the components induced by the fluctuations of the various types of macroscopic cross sections in the total noise are also discussed and the information content of the noise in the fast group is explored for the identification of fuel assembly vibrations.
Dynamic assessment for life extension of nuclear power plants (NPPs) using system dynamics (SD) method
Authors: Woo T. H. 
Published: December 2013

Abstract
It has been proposed to extend the life of nuclear power plants (NPPs) for the economic purpose. Especially, the primary systems in reactor are considered in the thermohydraulic and neutronic aspect, which is related to the safety system. The electric power and the lifespan of components are expressed as economic situation. In addition, political considerations are given by the presidential change and the nuclear non-proliferation characteristics. The dynamical investigation using system dynamics (SD) shows the effective time for the life extension of the NPPs by Monte-Carlo simulations. This non-linear algorithm is incorporated with the feedback loop of the event sequences. The expected event is related to the past event, which affects to the dynamical simulations of lifetime in the NPPs. In the conclusions, the safety guarantee as well as the economic profit in the re-use of long term operated power plants is presented, which is mentioned as the transient time between 2019 and 2021 in this paper.
Effects on criticality of selected scattering phase functions in neutron transport equation using the Chebyshev approximation
Authors: Kara A. , Anlı F. 
Published: December 2013

Abstract
Criticality and diffusion problems have been widely studied in neutron transport theory using polynomial expansion techniques in different geometries. In this study, instead of conventional scattering functions, two different phase function will be used in slab geometry transport equation by using the TN approximation and Marshak boundary conditions. Critical half thicknesses of the slab will be calculated. The numerical results obtained will be presented in tables.
Effect of Cu2+/Al3+ mole ratio on structure of Cu – Al bimetallic nanoparticles prepared by radiation induced method
Authors: Abedini A. , Larki F. , Saion E. , Noroozi M. 
Published: December 2013

Abstract
Cu–Al bimetallic nanoparticles were synthesized by gamma irradiation technique in aqueous solutions containing metal chlorides as precursors, polyvinyl alcohol (PVA) as a capping agent, isopropanol as a radical scavenger, and distilled water as a solvent. The Cu–Al bimetallic nanoparticles were characterized by transmission electron microscopy (TEM), UV-visible absorption spectrometry, powder X-ray diffractometer (XRD), and Energy-dispersive X-ray spectroscopy (EDX). The TEM, XRD, EDX, and absorption analyses confirmed the formation of core-shell structure of Cu–Al bimetallic nanoparticles at lower Cu2+/Al3+ mole ratio, and the formation of Cu–Al alloy nanoparticles at higher Cu2+/Al3+ mole ratio. The TEM analysis for particle size and size distribution revealed that the average particle size of Cu–Al bimetallic nanoparticles decreased with the increase of absorbed dose. It may be explained due to the competition between nucleation and aggregation processes in the formation of metallic nanoparticles under irradiation.
Experiments of condensation-induced water hammers at the UniBw Munich
Authors: Dirndorfer S. , Kulisch H. , Malcherek A. 
Published: December 2013

Abstract
Condensation-induced water hammers belong to the most serious and complex pressure surges. Experimental data from literature are scanty, incomplete, not public or simply missing. A new test facility at the UniBw (University of the German Armed Forces) was constructed to perform own experiments within a research alliance. This new test facility uses a complete new approach. In contrast to other test stations, the UniBw water hammer test facility is a closed system. This leads on the one hand to a more complex handling with respect of conducting experiments but on the other hand also to a more realistic characterization of thermo-hydraulic phenomena which can occur in a power plant. The first results of a measured condensation-induced water hammer are presented.
On-site power system reliability of a nuclear power plant after the earthquake
Authors: Volkanovski A. 
Published: December 2013

Abstract
Continuous and reliable sources of electrical energy are required during all modes of operation of a nuclear power plant in order to assure the safety of the plant. The electrical energy to the safety systems of the nuclear power plant is provided from the on-site power system. The objective of this paper is to analyze the nuclear power plant on-site power system reliability during normal operation and after the earthquakes with given intensity. The obtained results show the importance of the on-site power system reliability in general and battery supported section of the Class 1E power system in particular for the safety of the nuclear power plant. High level of reliability is obtained for the Class 1E and Non-1E power system during normal operation and safe shutdown earthquake. Obtained results show high level of reliability of a Class 1E power system after the earthquake with peak ground acceleration equal to the one measured at the Fukushima Daiichi nuclear power plant. Recommendation for revision of the current guideline for assessment of the acceptable station blackout duration capability is recommended considering the results of the analysis and identified deficiencies in the current guideline.
How to construct a seismic risk model?
Authors: Sperbeck S. T. , Türschmann M. 
Published: December 2013

Abstract
The German PSA Guideline and its technical documents on PSA methods and data require probabilistic safety analyses (PSA) to be carried out in the frame of safety reviews for nuclear power plants. Since 2005 this also includes a seismic PSA (SPSA) for sites with design earthquake intensities exceeding the value VII (MSK-64/EMS-98). It is shown how the plant model of PSA Level 1 for internal events can be extended on the level of fault tree basic events to get a quantifiable seismic plant model. A two-step screening procedure can be applied to derive the seismic equipment list (SEL) and a list of all possibly seismic-induced dependent equipment failures. The screening procedure is supported by a database. The database keeps at hand all the data and information to extend the plant model of PSA Level 1 in a proper manner.
Albedo and constant source problems for extremely anisotropic scattering
Authors: Koçmen M. A. , Teğmen A. , Türeci D. , Güleçyüz M. Ç. , Türeci R. G. 
Published: December 2013

Abstract
The half-space albedo problem and the constant source problem have been solved for a combination of the linearly anisotropic scattering and İnönü’s scattering functions. The linear transport equation for extremely anisotropic scattering kernel can be converted into an equivalent equation with a linearly anisotropic scattering kernel and the modified FN method can be used for albedo calculations.
Burnup calculations using serpent code in accelerator driven thorium reactors
Authors: Korkmaz M. E. , Yiğit M. , Ağar O. 
Published: December 2013

Abstract
In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed 232Th and mixed 233U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period.
The estimation of the control rods absorber burn-up during the VVER-1000 operation
Authors: Bolshagin S. , Gorodkov S. , Sukhino-Khomenko E. 
Published: December 2013

Abstract
The isotopic composition of the control rods absorber changes under the neutron flux influence, so the control rods efficiency can decrease. In the VVER-1000 control rods boron carbide and dysprosium titanate are used as absorbing materials. In boric part the efficiency decreases due to the 10B isotope burn-up. Dysprosium isotopes turn into other absorbing isotopes, so the absorbing properties of dysprosium part decrease to a lesser degree. Also the control rod's shells may be deformed as a consequence of boron carbide radiation swelling. This fact should be considered in substantiation of control rods durability. For the estimation of the control rods absorber burn-up two models are developed: VVER-1000 3-D fuel assembly with control rods partially immersed (imitation of the control rods operation in the working group) and VVER-1000 3-D fuel assembly with control rods, located at the upper limit switch (imitation of the control rods operation in groups of the emergency shutdown system).
The main characteristic of the evolution project SuperVVER with spectrum shift regulation
Authors: Alekseev P. N. , Chibinyaev A. V. , Gorohov V. F. , Teplov P. S. 
Published: December 2013

Abstract
The researches on the increase of achievable fuel burn-up in light water reactors (LWR) with so-called “spectrum shift” have been carried out in the world for a long time. The main idea of the “spectrum shift” is based on neutron spectrum shifting from the resonance energy region at the beginning of the cycle to the thermal region at the end of the cycle. There are many different ways of such regulation in the core – starting from coolant density variation during reactor cycle, to changing water-uranium ratio with some mechanical equipment. Spectrum shift regulation with depleted uranium or zirconium displacers is presented in this paper for uranium and mixed uranium and plutonium oxide (MOX) fuel loadings. New fuel assemblies with special guide tubes for displacers were developed for these purposes.
Application of statistical uncertainty and sensitivity evaluations to a PWR LBLOCA analysis calculated with the code ATHLET. Part 1: uncertainty analysis
Authors: Kozmenkov Y. , Rohde U. 
Published: December 2013

Abstract
Usually, reactor safety analyses are based on the traditional conservative deterministic approach. An alternative approach to reactor safety analyses uses best estimate computer codes in combination with quantification of uncertainties in model and plant parameters. The German Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) has applied the methodology of uncertainty analysis based on the Wilks’ approach. According to this approach, the number of calculations required to perform a probabilistic estimation of an output parameter with a given coverage/confidence level (e.g., 95%/95%) does not depend on the number of input uncertainty parameters. In this paper, the results of an uncertainty analysis for a Large Break LOCA scenario for a generic German PWR of the Konvoi design are reported. All calculations were performed using the system code ATHLET Mod 2.2 Cycle A. The EXCEL-integrated software SUSA developed by GRS was used for generation of the uncertainty parameter vectors, their incorporation into ATHLET input files and for statistical evaluation of the results.
Tracer transport modeling with the Alliances platform in the presence of evapotranspiration
Authors: Constantin A. , Genty A. , Diaconu D. , Bucur C. 
Published: December 2013

AbstractThe knowledge and understanding of water flow and solute transport in the unsaturated zone is becoming increasingly important especially in mitigation of groundwater pollution. Fate of radionuclide in the geological environment is a topic to address in performance and safety assessment studies for nuclear waste disposal and may be modeled considering flow and transport in porous media. However, often, due to the heterogeneity and anisotropy of the real systems, the computer simulations may be difficult to render the real behavior. This paper addresses the simulation of a tracer transport in the unsaturated zone of the Saligny site, the potential location for the Romanian low and intermediate level waste (LILW) disposal. Computation was based on experimental data and was performed with the Alliances platform, a numerical tool developed by French organizations CEA, ANDRA and EDF. In order to obtain information regarding the solute migration in depth and the solute lateral dispersion, the dispersivity coefficients of iodine were investigated in order to match the experimental concentration determined on samples from different locations of the site. A close fit of the simulation over experimental data for the water saturation profile at a depth of 0.5 m in transient state was targeted by taking into account evapotranspiration in order to obtain a realistic estimation of the water infiltration in the porous media. Dispersivity coefficients obtained from the simulation of the tracer transport are in good order of magnitude for the unsaturated area and allow to have a good preview of the tracer plume. However, further investigations are recommended on new samples in order to validate the migration of the tracer plume as expected.
Thermodynamic assessment of the Mg–Nd system
Authors: Guo Cuiping , Du Zhenmin 
Published: December 2013

Abstract
The Mg–Nd system was critically assessed by means of the CALPHAD technique. The solution phases (liquid, body-centered cubic, double hexagonal close-packed and hexagonal close-packed) were modeled with the Redlich–Kister equation. The intermetallic compounds Mg41Nd5, Mg3Nd and MgNd, which have a homogeneity range, were treated as the formulae (Mg, Nd)41(Mg, Nd)5, Mg3(Mg, Nd) and Mg(Mg, Nd) by a two-sublattice model with Mg and Nd or Mg on the first sublattice, and Mg and Nd on the second one, respectively. The compound Mg2Nd was treated as a stoichiometric compound. A set of self-consistent thermodynamic parameters of the Mg–Nd system was obtained.
Phase diagram of the Co–Cu–Ti system at 850°C
Authors: Xu Honghui , Du Yong , Huang Baiyun , Jin Zhanpeng , Li Shitong , Zhang Xiang 
Published: December 2013

Abstract
The phase equilibria of the Co–Cu–Ti system at 850°C are investigated using a high-efficiency diffusion couple and six key ternary alloys. The samples are examined by means of optical microscopy, scanning electron microscopy, and electron probe microanalysis. At 850°C, no ternary compound is found. The Ti3Cu phase is a metastable phase. The solubilities of Cu in Ti2Co, TiCo, TiCo2(c) and TiCo3 were determined to be 8.7, 35.8, 2.1, and 5.4 at.%, respectively. And the solubilities of Co in Ti2Cu, TiCu, Ti3Cu4, and TiCu2 were measured to be 6.6, 0.7, 2.4, and 10.2 at.%, respectively. The following 9 three-phase equilibria are observed: (Ti) + Ti2Co + Ti2Cu, Ti2Co + Ti2Cu + TiCo, Ti2Cu + TiCu + TiCo, TiCu + Ti3Cu4 + TiCo, Ti3Cu4 + TiCu2 + TiCo, TiCu2 + TiCo + (Cu), TiCo + TiCo2(c) + (Cu), TiCo2(h) + TiCo3 + (Cu), and TiCo3 + (Co) + (Cu).