bluuebunny/crossref_metadata_2025_split huggingface dataset:
Meteorological influences on the results of area dose rate measurements
Authors: Dombrowski H. , Wissmann F. 
Published: December 2013

Abstract
The area dose rate (ADR) of the natural radiation is subject to strong fluctuations caused by meteorological influences. Rain in particular can strongly increase the local dose rate values on the earth's surface for a short time because radon progenies can be washed out of the atmosphere and accumulated on the Earth's surface. The Physikalisch-Technische Bundesanstalt (PTB) has reference measuring facilities at its disposal which allow permanent measurement of the individual components of natural ambient radiation. Results of long-term measurements and their correlation to weather data, seasonal fluctuations, but also to solar activity, are presented. In addition to the quantification of the radiation level caused by photons, the contributions of the charged components of cosmic radiation (myons and electrons) as well as neutron radiation were measured. The terrestrial component of natural ambient radiation results from the radionuclides in the soil and in the air. In order to determine which nuclides cause the corresponding dose rate, the in-situ-gamma spectrometry method is especially suited. Measurements of the variation of the nuclide-specific photon flux as a function of the precipitation prove that the activity concentrations of radon progenies have an important influence on the local dose rate.
On the determination of the ingestion dose in drinking and mineral water
Authors: Flesch K. , Gellermann R. , Gerler J. , Knappik R. , Köhler M. , Schulz H. 
Published: December 2013

Abstract
In Europe and Germany administrative regulations exist for the surveillance of the Total Indicative Dose of water supplied for human consumption. This parameter, which can not be analysed directly, has to be calculated using nuclide specific activity concentrations, age-specific dose conversion factors and consumption rates. Available calculation methods differ regarding the used radionuclides, consumption rates and whether they use age specific dose conversion factors or not. In Germany, administrative guidelines for the determination of the Total Ingestion Dose are still not available. From a large number of water sample analyses, a graded practical concept for the determination of the ingestion dose is derived, which considers radiological, analytical and hydrochemical aspects as well. Finally, it is suggested to consider mineral water in the same manner as drinking water.
Chebyshev polynomial (TN
) approximation to neutron transport theory and application to the critical slab problem
Authors: Bülbül A. , Anlı F. 
Published: December 2013

Abstract
The critical slab problem has been studied in the one-speed neutron transport equation with isotropic scattering by using the first kind of Chebyshev Polynomials. The moment criticality solutions were obtained for the uniform finite slab using Mark and Marshak type vacuum boundary conditions. The results obtained by this approximation are presented in tables which also include the results obtained by the PN method for comparison.
Deterministic calculation of grey Dancoff factors in cluster cells with cylindrical outer boundaries
Authors: Rodrigues L. Jenisch , Leite S. Q. Bogado , de Vilhena M. T. 
Published: December 2013

Abstract
In the present work, the WIMSD code routine PIJM is modified to compute deterministic Dancoff factors by the collision probability definition in general arrangements of partially absorbing fuel rods. Collision probabilities are calculated by an efficient integration scheme of the third-order Bickley functions, which considers each cell region separately. The effectiveness of the method is assessed by comparing grey Dancoff factors as calculated by PIJM, with those available in the literature by the Monte Carlo method, for the irregular geometry of the Canadian CANDU and CANFLEX assemblies. Dancoff factors at several different fuel pin positions are found in very good agreement with the literature results.
New calculations of excitation functions of some positron emitting and single photon emitting radioisotopes
Authors: Aydin E. G. , Tel E. , Kaplan A. , Aydin A. 
Published: December 2013

Abstract
In this study, new calculations of the excitation functions of 56Fe(p,2n)55Co,58Fe(p,2n)57Co,60Ni(p,n)60Cu,62Ni(p,2n)61Cu, 68Zn(p,2n)67Ga, 72Ge(p,n)72As,82Kr(p,2n)81Rb, 124Xe(p,2n)123Cs, 112Cd(p,2n)111In and 203Tl(p,3n)201Pb reactions have been carried for incident proton energies up to 50 MeV. In these calculations, the pre-equilibrium and equilibrium effects have been investigated. The pre-equilibrium calculations involve the newly evaluated geometry-dependent hybrid model, hybrid model and the cascade exciton model. Equilibrium effects are calculated according to the Weisskopf–Ewing model. The calculated results are compared with the experimental data taken from the literature.
Potential advantages and disadvantages of sequentially building small nuclear units instead of a large nuclear plant
Authors: Feretić D. , Čavlina N. , Grgić D. 
Published: December 2013

Abstract
Renewal of nuclear power programs in countries with modest electricity consumptions and weak electrical grid interconnections has raised the question of optimal nuclear power plants sizes for such countries. The same question would be also valid for isolated or weakly connected regions within a large country. Building large size nuclear power plant could be prevented by technical or financial limits. Research programs have been initiated in the International Atomic Energy Agency and in the USA (within the framework of the Global Nuclear Energy Partnership (GNEP) program) with the aim to inspect under which circumstances small and medium reactors could be the preferred option compared to large nuclear plants. The economy of scale is a clear advantage of large plants. This paper compares, by using probabilistic methods, the net cash flow of large and medium size plants, taking as example a large nuclear plant (around 1200 MW) and four sequentially built smaller plants (300 MW). Potential advantages and disadvantageous of both options have been considered. Main advantages of the sequential construction of several identical small units could be the reduced investor risk and reduced investment costs due to the learning effect. This analysis is a part of studies for the Croatian power generating system development.
Solution of half space and slab albedo problems for linearly anisotropic scattering with the modified FN method
Authors: Türeci D. , Güleçyüz M. Ç. 
Published: December 2013

Abstract
In this work, the FN method with a different numerical approximation is used to solve both, half-space and slab albedo problems for a linearly anisotropic scattering kernel. The numerical results are compared with data available in the literature and the accuracy of the method is verified.
Plate heat exchanger – inertia flywheel performance in loss of flow transient
Authors: Abou-El-Maaty T. , Abd-El-Hady A. 
Published: December 2013

Abstract
One of the most versatile types of heat exchangers used is the plate heat exchanger. It has principal advantages over other heat exchangers in that plates can be added and/or removed easily in order to change the area available for heat transfer and therefore its overall performance. The cooling systems of Egypt's second research reactor (ETRR 2) use this type of heat exchanger for cooling purposes in its primary core cooling and pool cooling systems. In addition to the change in the number of heat exchanger cooling channels, the effect of changing the amount of mass flow rate on the heat exchanger performance is an important issues in this study. The inertia flywheel mounted on the primary core cooling system pump with the plate heat exchanger plays an important role in the case of loss of flow transients. The PARET code is used to simulate the effect of loss of flow transients on the reactor core. Hence, the core outlet temperature with the pump-flywheel flow coast down is fed into the plate heat exchanger model developed to estimate the total energy transferred to the cooling tower, the primary side heat exchanger temperature variation, the transmitted heat exchanger power, and the heat exchanger effectiveness. In addition, the pressure drop in both, the primary side and secondary side of the plate heat exchanger is calculated in all simulated transients because their values have limits beyond which the heat exchanger is useless.
Assessment for inside building dispersion of aerosol from an outdoor radiological release
Authors: Wang W. H. 
Published: December 2013

Abstract
Estimation of indoor aerosol concentrations during outdoor release of radioactive materials is important since most people spend a lot of their time indoors. Whereas most of the mass transfer models are too complex to assess the indoor air quality by the first responders or radiation safety officers, this report uses the IAQX code to simulate the airborne concentration and inhalation exposure in a multi-zone building in the event of an outdoor radiological release. The way to establish a multi-zone well-mixed model is introduced first. The contamination in each zone is demonstrated by scenario and case study. Finally, the inhalation exposure will be derived from the air concentration versus time results. The analysis results show the major factors that influence the indoor air quality and health effects from an ambient release and the effectiveness of using IAQX to simulate complex multi-zone mass transfer model.
Modified Chebyshev polynomial (U

N
) approximation in the neutron transport theory and computation of critical half thicknesses
Authors: Bülbül A. , Anlı F. 
Published: December 2013

Abstract
A general theoretical scheme for the UN method and the evolution of a general eigenvalue spectrum are described. The applicability of UN method to one dimensional slab geometry neutron transport problems is discussed. The eigenvalue spectrum is calculated for isotropic scattering with different values of the parameter c, the mean number of secondary neutrons per collision, known as the fundamental eigenvalue. Then the critical slab problem has been studied. The critical half thicknesses are computed for different values of c. For the solution, Mark and Marshak boundary conditions are used. Results are obtained for both, UN and PN approximations for comparison.
Experience with quantitative safety targets on international level
Authors: Berg H.-P. 
Published: December 2013

Abstract
On international level, probabilistic safety assessment represents the state of the art in the licensing process for new industrial facilities, but increasingly also for evaluating the safety level of older industrial plants, e. g. as part of periodic safety reviews in case of nuclear installations. Quantitative safety goals or criteria have not yet reached the same level of acceptance. However, the role of such goals/criteria depends on the type of industry. Most of the countries consider such criteria as safety targets rather than as sharply defined boundary values. The Netherlands and the United Kingdom are exceptions because they require demonstration of compliance with legally binding safety goals in the licensing procedure for all types of industrial activities.
Erfahrungen aus der Übung „MERLIN“ zur Detektion radioaktiver Stoffe unter Beteiligung verschiedenen Einsatzfahrzeuge
Authors: Griesbach M. 
Published: December 2013

Kurzfassung
Es werden Erfahrungen aus einer Gefahrstoff-Übung, insbesondere aus einer Lage mit radioaktiven Stoffen (Messungen und Probennahmen) geschildert. Hierbei kamen verschiedene Spezialfahrzeuge mit Beladung und der ihrer Besatzung zum Einsatz. Ebenso waren die nach Landeskonzept notwendigen Strahlenschutzfachkräfte eingebunden. Es war die bislang größte Übung in Hessen, die zum Komplex „Gefahrgut“, insbesondere auch im Hinblick auf „Zwischenfall mit radioaktiven Stoffen“ in dieser Art stattfand.
The Swiss disaster management plan for coping with the aftermath of radiological dispersal devices – “dirty bomb” operational concept
Authors: Stoffel F. , Blättler M. , Leonardi A. 
Published: December 2013

Abstract
In 2007 the Swiss Federal Commission for NRBC Protection released a disaster management plan for coping with the aftermath of radiological dispersal devices. This paper summarises the basic concept and outlines the relevant bodies and agencies as well as their responsibilities. It also sets out the strategy to monitor radioactive contamination and the measures to prevent public radiation exposure.
Non-destructive measuring techniques for on-site detection and identification of illicit nuclear material
Authors: Rosenstock W. , Köble T. , Risse M. 
Published: December 2013

Abstract
In case of a terrorist's threat to release nuclear or radioactive material methods and procedures are needed for fast search and, after detection, identification of the substance. For searching the German Fraunhofer-INT in Euskirchen has a car at its disposal, which is equipped with very sensitive detectors suitable for a first identification of the material. Advanced measurements are performed with portable detectors of high energy resolution or by scanning of buildings or objects with a gamma camera. Initially such material may be stored under water in preparation for an attack. In this case we have a special sensor available for detection and subsequent identification. In case of no suspicious material being detected with all these passive methods, even though a well-founded suspicion exists, active methods will be employed. Therefore neutron interrogation is briefly presented. With all these techniques it is possible to determine the presence of radioactive or fissionable materials, as well as, within certain limits, type and quantity of material. Hence information will be gained on the possible risk potential and, by actions taking radiological aspects into account, recommendations can be given on further actions to be taken.
A model for calculation of forward isotropic scattering with application to transport equation in slab geometry
Authors: Yaşa F. , Anli F. 
Published: December 2013

Abstract
In order to develop a general procedure, the scattering kernel is described in a consistent manner. By using this scattering kernel, eigenvalues and eigenfunctions and critical slab thicknesses are computed with a conventional spherical harmonics method (PN). The basis of this method relates to the generating function of Legendre polynomials. This article has improved analytical methods and numerical values. Numerical results indicate that the critical thickness varies with forward isotropic scattering. Results are discussed and compared with isotropic and anisotropic results using various methods available in the literature. Eigenvalues and critical thicknesses are calculated using the new phase function instead of conventional expansions.
Heuristic geometric “eigenvalue universality” in a one-dimensional neutron transport problem with anisotropic scattering
Authors: Gonçalves G. A. , de Vilhena M. T. , Bodmann B. E. J. 
Published: December 2013

Abstract
In the present work we propose a heuristic construction of a transport equation for neutrons with anisotropic scattering considering only the radial cylinder dimension. The eigenvalues of the solutions of the equation correspond to the positive values for the one dimensional case. The central idea of the procedure is the application of the SN method for the discretisation of the angular variable followed by the application of the zero order Hankel transformation. The basis the construction of the scattering terms in form of an integro-differential equation for stationary transport resides in the hypothesis that the eigenvalues that compose the elementary solutions are independent of geometry for a homogeneous medium. We compare the solutions for the cartesian one dimensional problem for an infinite cylinder with azimuthal symmetry and linear anisotropic scattering for two cases.
European clearinghouse on nuclear power plants operational experience feedback
Authors: Ranguelova V. , Bruynooghe C. , Noël M. 
Published: December 2013

Abstract
Learning from operational experience and applying this knowledge promptly and intelligently is one of the ways to improve the safety of Nuclear Power Plant (NPP). Recent reviews of the effectiveness of Operational Experience Feedback (OEF) systems have pointed to the need for further improvement, with importance being placed on tailoring the information to the needs of the regulators. In 2007, at the request of a number of nuclear safety regulatory authorities in Europe, the Institute for Energy of the European Commission's Joint Research Centre (EC JRC) initiated a project on Nuclear Power Plant operational experience feedback, which adopts an integrated approach to the research needed to strengthen the European capabilities for assessment of NPP operational events and to promote the development of tools and mechanisms for the improved application of the lessons learned. Consequently, a so-called “European Clearinghouse” on NPP OEF was established, which includes scientific officers from the EC JRC, a number of European nuclear safety regulatory authorities and some of their Technical Support Organizations (TSOs). The paper discusses the activities implemented in 2008 within the framework of the European Clearinghouse on NPP OEF (hereinafter called the European NPP Clearinghouse) and provides an overview of the main conclusions drawn from the safety studies performed. Outlook of the activities carried out in 2009 are given.
Steady state thermal hydraulic calculations for MTR plate-type research reactors
Authors: Abdelrazek I. D. , Zidan W. I. , Shokr A. M. , Gaheen M. A. 
Published: December 2013

Abstract
A computation code THERMOHYD has been developed to perform the steady state thermal-hydraulic calculations of the MTR type research reactors fueled with plate-type-fuel, in the ranges of pressures and temperatures typical for MTR reactors. The code calculates the coolant and clad surface temperatures along the cooling channel, as well as the critical heat flux and provides values for the critical phenomena parameters (RDR, ONBR and DNBR ratios) for reactor cores. The code takes into account the variation of the physical properties of the coolant with temperature. It also accounts for the uncertainties due to fuel fabrication tolerances, possible deviations in the operational measurements, the errors in the experimental correlations used in the analysis, and the errors due to simplifications made in the calculations which have been considered in the analysis. The THERMOHYD code was tested against the well known PARET code and proved to be accurate and more conservative in some aspects.
Adaptation of the FEMAXI-6 code and RBMK fuel rods model testing employing the best estimate approach
Authors: Marao A. , Kaliatka T. , Kaliatka A. , Ušpuras E. 
Published: December 2013

Abstract
This paper provides information about the possibility to apply the FEMAXI-6 code for RBMK-1500 reactors. New thermal properties and correlations responsible for thermal analysis (thermal conductivity and heat capacity) according to RBMK-1500 specification were included into the FEMAXI-6 code. The RBMK-1500 fuel rod model using the adapted FEMAXI-6 code was developed and tested by employing uncertainty and sensitivity analysis. The processes in fuel rods during normal plant operation were modelled. The results received were compared with calculations performed by specialists from the Kurchatov Institute (designers of RBMK). The reasonable agreement of both calculation results shows that the adapted FEMAXI-6 code and the developed model are suitable for future analysis of processes in fuel rods of RBMK-1500 reactors. Also this work demonstrates what sensitivity and uncertainty analysis can be used for the model development.
Interactions of radionuclides in solution with ions and with trace concentrations of solid phase
Authors: Kepák F. 
Published: December 2013

Abstract
Interactions of radionuclides in extremely low concentrations were determined by measurement of their apparent self-diffusion coefficients. Formation of trace amounts of hydrolysis products, hydroxides of RuIV, MnIV, FeIII, EuIII and of ionic precipitate of AgI were determined from apparent self-diffusion coefficients. Transition of 85Sr2+ by ion exchange or by coprecipitation into trace amounts of FeIII-hydroxide and MnIV-hydrated oxide was revealed by this method. 85Sr2+ sorbed on trace amounts of FeIII-hydroxide and MnIV-hydrated oxides lost its individual chemical properties. This can negatively affect its separation from solution. Formation of unstable embryos of Ag131I trace precipitate in unsaturated solution at concentration of Ag+ 1 · 10−9 mol · l−1 and 131I−1 · 10−8 mol · l−1 was detected by self-diffusion. Formation of trace amounts of Ag131I led to the decrease of sorption of 131I− on hydrated ferric oxide.
Investigation on primary side oriented accident management measures in a hypothetical station blackout scenario for a VVER-1000 pressurized water reactor
Authors: Tusheva P. , Schäfer F. , Reinke N. , Altstadt E. , Rohde U. , Weiss F. P. , Hurtado A. 
Published: December 2013

Abstract
As a consequence of a total loss of AC power supply (station blackout) at a VVER-1000 leading to unavailability of major active safety systems, the safety criteria ensuring the safe operation of the nuclear power plant would be violated and core heat-up with possible core degradation could occur. A dedicated accident management measure (primary side depressurization) can be applied to reduce the primary pressure and to activate the injection from the passive emergency core cooling systems (accumulators). The analyses presented in this paper are aiming at both a detailed investigation of the accident sequence, taking into account the depressurization of the primary circuit, and the possibilities to prevent or at least to mitigate a damage of the reactor core so as to gain additional time for taking necessary countermeasures. The analyses are performed using the codes ASTEC and ATHLET developed by IRSN (Institut de Radioprotection et de Sûreté Nucléaire) and GRS (Gesellschaft für Anlagen- und Reaktorsicherheit mbH).
Numerical investigation of heat transfer in the vertical annulus between pressure tube and calandria tube of the advanced water cooled reactor
Authors: Vaidya A. M. , Borgohain A. , Maheshwari N. K. , Govindan D. , Vijayan P. K. 
Published: December 2013

Abstract
In advanced water cooled reactors, an annular gap exists between pressure tube and calandria tube. The gap is closed from top but is open from bottom. Due to differential temperature between pressure tube and calandria tube, air flow is induced by natural convection. This leads to heat transfer from pressure tube to calandria tube. The quantification of the heat transfer between pressure tube and calandria tube is numerically carried out with the help of the CFD code PHOENICS. Validation of the CFD code with experimental results and some established computational work from the literature has been done in order to verify the accuracy of the code. The natural convection phenomenon in the annular gap is then simulated. The velocity and temperature fields obtained from the CFD simulation are used to compute local and average heat transfer coefficients. Heat transfer coefficients for various pressure tube temperatures are computed. The effect of water on the heat transfer in the annular gas is also studied.