bluuebunny/crossref_metadata_2025_split huggingface dataset:
Lumped parameters analysis of the IAEA research reactor benchmark problem
Authors: Gaheen M. A. , Elaraby S. , Aly M. N. , Nagy M. S. 
Published: December 2013

Abstract
A simple model has been developed to predict the transient behaviour of research reactors with simple means. The developed model uses a lumped parameters approach for the kinetics and heat transfer modeling with continuous feedback reactivities. The model is used for the analysis of the International Atomic Energy Agency (IAEA) 10MW MTR research reactor benchmark problem. Transient responses to reactivity insertion and loss of coolant flow are presented and analyzed. The model predictions are accurate enough compared with the calculations conducted in various institutions using different codes. It is shown that the model can provide accurate predictions as long as the clad temperature does not exceed the ONB (Onset of Nucleate Boiling) temperature. However, the results are very encouraging and the model is useful for practical purposes.
Minor actinide burning in a CANDU thorium reactor
Authors: Şahin S. , Yıldız K. , Şahin H. M. , Acır A. , Şahin N. , Altınok T. 
Published: December 2013

Abstract
Nuclear waste actinides can be used as a booster fissile fuel material in form of mixed fuel with thorium in a CANDU reactor in order to assure the initial criticality at startup. Two different fuel compositions have been found useful to provide sufficient reactor criticality over a long operation period: 1) 95% thoria (ThO2)+5% minor actinides MAO2 and 2) 90% ThO2+5% MAO2+5% UO2. The latter allows a higher degree of nuclear safeguarding through denaturing the new 233U fuel with 238U. The temporal variation of the criticality k∞ and the burn-up values of the reactor have been calculated by full power operation for a period of 10 years. The criticality starts by k∞>1.3 for both fuel compositions. A sharp decrease of the criticality has been observed in the first year as a consequence of rapid plutonium burnout in the actinide fuel. The criticality becomes quasi constant after the 2nd year and remains close to k∞ = ∼1.06 for ∼10 years. After the 2nd year, the CANDU reactor begins to operate practically as a thorium burner. Very high burn up could be achieved with the same fuel material (up to 200000 MW.D/MT), provided that the fuel rod claddings would be replaced periodically (after every 50000 or 100000 MW.D/MT). The reactor criticality can be maintained until a great fraction of the thorium fuel is burnt up. This would reduce fuel fabrication costs and nuclear waste mass for final disposal per unit energy drastically.
Thermal-hydraulic modeling of the onset of flow instability in MTR reactors
Authors: Khater H. A. , El-Morshdy S. El-Din , Ibrahim M. A. 
Published: December 2013

Abstract
Prediction of the onset of the flow instability (OFI) in steady and transient sub-cooled flow boiling is an important consideration in the design and operation of nuclear reactors especially material test reactors (MTR). In this study, a predictive model of OFI in the MTR has been developed. The model is based on both heat balance of bubble generation and condensation and force balance for the detached bubbles at the onset of significant void (OSV). Only one coefficient involved in the proposed model is identified by experimental data that covers the wide range of MTR operating conditions. The model results are compared with results of some previous models where the present model gives the lower deviation from the experimental data. A correlation for the heat flux at OFI is also developed based on the present model. The developed correlation gives lower deviation from the experimental data than the well-known correlation of Whittle and Forgan. The model is also used to predict the OFI locus during transient where it shows good agreement with the short transient of Lee and Bankoff as well.
Radiation transfer in inhomogeneous exponential media
Authors: Tezcan C. , Akçay H. 
Published: December 2013

Abstract
The angular distribution of the radiation intensity and the related constants C and D∗ are calculated for a new choice of c(x) having the form of the Morse potential in quantum mechanics c(x) = 1 – a – be−2αx
+ de−αx
. We use the modified Eddington method. The radiation intensity in this method is given in terms of the unknown even and odd functions of the space variable × and the direction cosine. The coefficients of these functions depend only on the space variables and satisfy second order differential equations. We solve the resulting second order differential equation using the Nikiforov-Uvarov method. The method provides exact analytical expressions and is not previously used to solve radiation problems. The numerical results are listed in a table for both the constants C and D∗ and the albedo and in the limiting cases are compared with the homogeneous values.
Neutron transport problems for extremely anisotropic scattering
Authors: Kaşkaş A. , Karahasanoğlu M. , Ç. Güleçyüz M. 
Published: December 2013

Abstract
Half-space albedo, slab albedo and critical slab problems will be solved for extremely anisotropic scattering using the analytical expressions of the FN method with a different numerical approximation. The infinite medium Green's function for the anisotropic scattering case obtained in terms of the infinite medium Green's function for the isotropic scattering is used. The numerical results are compared with data available in the literature.
The effects of different expansions of the exit distribution on the extrapolation length for linearly anisotropic scattering
Authors: Bulut S. , Güleçyüz M. Ç. , Kaşkaş A. , Tezcan C. 
Published: December 2013

Abstract
HN and singular eigenfunction methods are used to determine the neutron distribution everywhere in a source-free half space with zero incident flux for a linearly anisotropic scattering kernel. The singular eigenfunction expansion of the method of elementary solutions is used. The orthogonality relations of the discrete and continuous eigenfunctions for linearly anisotropic scattering provides the determination of the expansion coefficients. Different expansions of the exit distribution are used: the expansion in powers of μ, the expansion in terms of Legendre polynomials and the expansion in powers of 1/(1 + μ). The results are compared to each other. In the second part of our work, the transport equation and the infinite medium Green function are used. The numerical results of the extrapolation length obtained for the different expansions is discussed.
The RES/EPRI concensus – fire probabilistic risk assessment method∗
Authors: Nowlen S. P. 
Published: December 2013

Abstract
The U.S. Nuclear Regulatory Commission (USNRC) Office of Nuclear Regulatory Research (RES) and the Electric Power Research Institute (EPRI), in cooperation with nuclear utilities in the U.S., embarked on a joint effort to document and demonstrate state of the art fire probabilistic risk assessment (PRA) methods. The principal objective of this project, known as the Fire Risk Requantification Study, was to develop a technical basis and guidance to clarify issues affecting application of fire risk methods. The project culminated in the publication of a joint RES/EPRI consensus FPRA method. The methodology document was finalized in September 2005 and is now being used to support a range of risk informed/performance based (RI/PB) applications in nuclear power plant fire protection.
Fire PSA of the French PWR 900 MWe series
Authors: Barrachin G. , Corenwinder F. , Vial E. 
Published: December 2013

Abstract
The PSA approach described in part 1 is also used to assess the core damage frequency due to a fire in the control room of a 900 MWe Nuclear Power plant. The specifics of the control room include the cable failure modes due to the fire (spurious order in particular), two different kinds of fire detection systems in the control room and the use of controls and instrumentation of the remote shutdown panel to reach the safe shutdown state if the control room has to be evacuated. This paper will focus on the detailed study of critical compartments and on the results and insights of this study. The purpose of this paper is also to present the methodology used to assess the core damage frequency due to a fire in the control room of a French Nuclear Power Plant, some qualitative results and the main insights further to the discussions with the utility.
Methods for a Fire PSA exemplarily applied to a German BWR-69 type nuclear power plant
Authors: von Linden J. , Röwekamp M. , Türschmann M. , Berg H. P. 
Published: December 2013

Abstract
In the following, an advanced German approach for state-of-the-art probabilistic fire risk assessment (Fire PSA) to be performed within comprehensive safety reviews of nuclear power plants is presented on the background of the current German PSA Guide. This approach has been successfully applied in the frame of a Fire PSA recently performed for a nuclear power plant with boiling water reactor (BWR-69 type). The set of all compartments in a nuclear power plant is the starting point for determining the annual frequency of fire induced plant hazard and core damage states. Fire PSA has to be carried out in several steps. First step of the analysis is a selection process (‘screening’) providing critical fire zones, where a fully developed fire has the potential to both cause an initiating event and impair the function of at least one nuclear safety related component or system. Detailed analyses providing compartment specific fire induced hazard state frequencies are performed only for those compartments, where in case of fire a relevant contribution to the overall hazard state frequency can be expected. Both analytical steps, selection of relevant compartments as well as detailed analyses, are based on a comprehensive compartment related compilation of fire specific data and information.
Improvement, extension and integration of operational Decision Support Systems for nuclear emergency management (DSSNET)
Authors: Raskob W. 
Published: December 2013

Abstract
The DSSNET network was established in October 2000 with the overall objective to create an effective and accepted framework for better communication and understanding between the community of institutions involved in operational off-site emergency management and the many and diverse Research and Technological Development (RTD) institutes further developing methods and tools in this area, in particular decision support systems (DSS), for making well informed and consistent judgements with respect to practical improvements of emergency response in Europe. 37 institutions from 21 countries of East and West Europe have been members of the network with about half of them responsible for operational emergency management. To stimulate the communication and feedback between the operational and the RTD community, problem-oriented emergency exercises were performed, which covered the various time phases of an accident and which extended from the near range to farther distances with frontier crossing transport of radionuclides. This paper concentrates on the five emergency exercises conducted in the frame of the project.
Potentials and limits of electronic situation displays
Authors: Hable K. , Höbler Ch. , Bieringer P. 
Published: December 2013

Abstract
In the field of precautionary radiation protection and disaster response, the amount of information and the number of its sources is steadily increasing. Central compilation and instant availability of information for all involved authorities becomes more and more important. In this context the structured presentation of documents on a web server in terms of an “Electronic Situation Display” (ESD) has been established. In Germany, there are several Electronic Situation Display prototypes run by federal and state authorities for various purposes and using different technical approaches. The systems of the federal states focus on workflow management. They provide emergency log books and information on the implementation of countermeasures for disaster response authorities. The Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) tasked the Federal Office for Radiation Protection (BfS) to develop the Federal Electronic Situation Display (ELAN) that is mainly used as an information platform for decision making in case of nuclear emergencies. Documents containing radiological data or results from Decision Support Systems are almost automatically integrated into ELAN. Read and write permissions for users are managed by a concept of specific roles. The federal ESD and the systems of the federal states will be coordinated in order to achieve a better harmonisation. Interfaces for data exchange and data processing must be defined. Authorities of the federal states can use protected areas within the federal ESD where information is kept confidential for internal use before it is distributed to other states and federal authorities. In addition, key information of Electronic Situation Displays from the federal states can automatically be integrated into ELAN. Several task groups of the IAEA are working on the development and realization of a prototype for an information platform harmonising ESDs on an international level.
Requirements of emergency control managements on data and information for assessment of the radiological situation in case of a severe accident in a nuclear power plant
Authors: Urbach M. , Koch D. , Weiß H. , Pauly J. 
Published: December 2013

Abstract
In case of a severe accident in one of the nuclear power plants in northern Germany consequences may reach across the borders of counties and Länder and several local and regional authorities will have responsibilities for disaster response measures. Consistent assessment of the situation and coordinated action are essential for protection of the population and task personnel as effective as possible. Disaster response managements have to come to a comprehensive assessment of the developing radiological situation at an early stage to be able to initiate appropriate protective measures. The results of a working group from several local and regional authorities from Schleswig-Holstein, Hamburg, Lower Saxony and Mecklenburg-West Pomerania on this topic will be presented here. The working group has compiled sets of data and information which are relevant in decision making for protective measures and has developed a concept for the efficient distribution of data and information to the disaster response managements and other recipients with responsibilities in emergency control.
deNIS IIplus – computer-assisted crisis management system
Authors: Corr B. 
Published: December 2013

Abstract
The management of catastrophes, as natural disasters or man-made disasters, will only be effective and successful if all relevant information is available in time for decision-makers. During previous large-scale disasters it became evident that information needed for the disaster management was only partially available and that there were fundamental deficits in regard to the flow of information between federal states (“Laender”) as well as communication problems between disaster response teams in the effected regions. On this account in summer 2001 the Federal Ministry of the Interior has decided to develop the “German Emergency Preparedness Information System (deNIS)”. The aim of the enhanced version deNIS IIplus is to built up a network for the civil and disaster response and to assist as an information and communication system for decision-makers of the Federal Government and the Laender Governments to better co-ordination relief and rescue teams in the event of a natural disaster or technical accident. Correspondingly the primary task of deNIS is to support the decision-making of disaster management authorities.
Effect of fuel particles' size and position variations on multiplication factor in pebble-bed nuclear reactors
Authors: Snoj L. , Ravnik M. 
Published: December 2013

Abstract
The effect of TRISO particles' size variations and inter-particle distance on infinite multiplication factor (kinf) of a pebble bed reactor (PBR) is studied using MCNP code. It is observed that kinf decreases with increasing TRISO particles' size, mainly due to increased total particle surface leading to stronger resonance absorption. Negative temperature reactivity effect is observed at all particle sizes. The kinf increases strongly when particles form in clusters as the clusters behave as a single bigger particle. It is also observed that the white boundary condition is not an appropriate approximation of the unit-cell flux boundary condition in the calculations as it yields significantly higher values of kinf than periodic and reflective boundary conditions.
Investigation of nuclear model predictions for proton induced reaction cross-sections up to 150 MeV
Authors: Broeders C. H. M. , Konobeyev A. Y. , Mercatali L. 
Published: December 2013

Abstract
An extensive analysis of the accuracy in the theoretical reproduction of experimental proton induced reaction cross-sections up to 150 MeV is presented. The predictive capabilities of different models for the description of the nuclear level densities as implemented in the TALYS code and in the modified ALICE code have been assessed by means of a systematic comparison of the theoretical results with experimental EXFOR data relative to target nuclei from 24Mg to 209Bi. The results obtained give the possibility to find the best approaches for the cross-section calculation for nuclei from different mass ranges and for different channels.
The albedo problem for pure-triplet scattering
Authors: Türeci R. G. 
Published: December 2013

Abstract
The singular eigenfunction approach using CN equations can be extended for high order anisotropic scattering (triplet scattering) to the Albedo problem. Case's eigenfunctions and the orthogonality relations of this scattering can be derived mathematically. Thereby, the infinite medium Green function can be written for an homogeneous medium in terms of singular eigenfunctions. Also using the HN method, the constant source albedo problem can be investigated for this type of scattering.
The preignition problem in nuclear explosive devices (NEDs) for a sigmoidal Rossi-α and a high neutron background
Authors: Seifritz W. 
Published: December 2013

Abstract
The preignition probability for a sigmoidal reactivity input in Nuclear Explosive Devices (NEDs) with a strong inherent neutron source strength (i. e. using reactor grade plutonium) is treated comprehensively.
The cumulative preignition probability, its probability density as well as the most probable time for preignition during the compression phase are presented in a closed analytical form. This general formalism can be used to calculate the preignition probability for any metallic core composition (either Pu or U or hybrid) possessing an inherently strong or weak background of neutrons.
Comparison of economy and emissions of nuclear power plants and combined gas and wind electricity generators
Authors: Feretic D. , Grgic D. 
Published: December 2013

Abstract
As part of the analysis of options to generate additional 6 – 8 TWh of electrical energy for Croatian consumers in the next decade, a comparison between natural gas combined cycle plants jointly with wind electricity generators and nuclear power plants has been performed. The criteria for the optimal option are minimal electricity cost and minimal carbon dioxide emissions. Since the comparison has to include the analysis of electricity cost generated during plant life time and since potential long term fuel cost variations (gas, uranium concentrate and uranium enrichment) containing substantial uncertainties, the best solution is to calculate life time levelized electricity costs with probabilistic methods. The results, obtained in form of probabilistic distributions, show that the option with nuclear plants leads to remarkably lower electricity cost and CO2 emissions than the option with gas and wind plants. The difference is more pronounced when wind generators have low capacity factors.
Uncertainty in activation cross-section calculations at intermediate proton energies
Authors: Konobeyev A. Yu. , Broeders C. H. M. , Fischer U. , Mercatali L. 
Published: December 2013

Abstract
The uncertainty in calculations of proton induced reaction cross-sections using modern nuclear models and codes has been investigated. Cross-sections calculated with TALYS-0.72, ALICE/ASH, DISCA, and MCNPX codes have been compared with experimental data available in EXFOR for (p, x) reactions at proton incident energies up to 150 MeV for target nuclei from 24Mg to 209Bi. In addition, the comparison of evaluated data from ENDF/B-VII, JENDL-HE, and PADF calculations with measured cross-sections has been performed. Various statistical criteria were applied for the quantification of the difference between results of calculations, evaluated and experimental data.
Application of the HN method to radiative heat transfer for a non-conservative slab
Authors: Bulut S. , Güleçyüz M. Ç. 
Published: December 2013

Abstract
The HN method is used to calculate the partial fluxes relevant to radiative transfer in a plane parallel medium for linearly anisotropic scattering with specularly and diffusely reflecting boundaries. The solutions are obtained for the non-conservative case. The numerical values of the heat flux functions at the boundaries of the medium are calculated for the two source functions. The numerical results and the convergence of the HN method are compared with earlier results.
Prediction of indoor radon in Germany on a regional scale
Authors: Kemski J. , Klingel R. 
Published: December 2013

Abstract
Measurements of indoor radon activity concentration have been carried out in more than 10 000 dwellings in different regions of Germany. The statistical correlations between geology, building ground, indoor radon and the house characteristics were investigated. The geographical location – and therefore the geological setting – can be used as a proxy for both region-specific geological properties and building characteristics. Together with different variables of house type and living habits it significantly contribute to the explained variance for both ground floor and basement radon activity concentrations. Soil gas and indoor radon activity concentrations are connected by a transfer factor. The study has shown, that this factor can vary depending on regional differences in building characteristics, especially between the western and eastern part of Germany. Radon risk prediction maps for radon in houses based on the German soil gas map and the regional distribution of radon transfer, indicating the probability of exceeding indoor threshold values, can be designed. They can be used as a tool for indoor radon prediction, especially in regions without a sufficient number of indoor radon data.