bluuebunny/crossref_metadata_2025 huggingface dataset:
Updated requirements on PSA methods and data for comprehensive safety reviews in Germany
Authors: Berg H. P. , Fröhmel T. , Görtz R. , Schimetschka E.
Published: December 2013

The German regulatory guide on probabilistic safety assessment (PSA) describes the fundamental requirements concerning the performance of PSA in the frame of comprehensive safety reviews. The technical details regarding the performance of PSA are set out in two technical documents (PSA Methods and PSA Data) that have been developed by a working group of PSA experts (FAK PSA). Based on the experiences from the first series of periodic safety reviews in Germany, international experiences and the fact that comprehensive safety reviews are now mandatory since April 2002 all guides have to be revised. The first guide updated was the PSA guide together with the corresponding technical documents, issued in October and November 2005. In the following the key points of the revised requirements on PSA methods and data are described.
OECD/NEA International Common Cause Failure Data Exchange (ICDE) Project – insights and lessons learnt
Authors: Johanson G. , Kreuser A. , Pyy P. , Rasmuson D. , Werner W.
Published: December 2013

Events initiated by common-cause-failure (CCF) can significantly affect the availability and reliability of nuclear power plant safety systems. In recognition of this, CCF data are systematically collected and analysed in the International Common-Cause Data Exchange (ICDE) Project, which was initiated in August 1994. Since April 1998, the NEA has formally operated the project. Currently eleven countries participate in the project. The ICDE collects all events where two or more identical, redundant components of a group, fulfilling the same function, have failed or were impaired due to a shared cause (ICDE events). Complete CCFs, i.e. failure of all identical, redundant components in the group due to a shared cause are an important subset of the collected data. Currently, data exchange and analysis covers the following components: centrifugal pumps, diesel generators, motor-operated valves, safety and relief valves, check valves, reactor protection system components (level measurement, control rod drives, etc), circuit breakers, and batteries. The main findings of the ICDE reports issued by 2005 show averaged over all components that about two thirds of all complete CCF events involve faulty actions by plant personnel and contractors. The single largest contribution is from faulty testing and maintenance work due to deficient and/or incomplete procedures. Other important causes are insufficient testing and requalification of components or systems after maintenance, repair, modifications or backfitting work, as well as operator errors of commission. The probability that a reported ICDE event is a complete CCF decreases strongly with increasing number of redundant components, demonstrating the effectiveness of redundancy as a powerful defence against CCFs. However, complete CCFs cannot be completely prevented by high redundancy only.
Extension of the German database for common cause failure events
Authors: Holtschmidt H. , Kreuser A. , Verstegen C.
Published: December 2013

The first German data base for Common Cause Failures (CCF) had been set up for the German Risk study, Phase B, published in 1989. It was based on three sources: international operating experience from the OECD Incident Reporting System, the national operating experience reflected in the collection of reportable events and plant specific operating experience from the Nuclear Power Plant Biblis unit B. This data base was subsequently substantially extended in terms of additional events and extended observation periods where additional information was collected in the frame of further PSA studies und CCF projects. The reportable events were the key source for this extension. Using this improved data base an updated calculation of CCF probabilities has been performed which was published in the document on PSA data associated to the regulatory PSA guideline.
International network on incorporation of ageing effects into PSA
Authors: Kirchsteiger C. , Patrik M.
Published: December 2013

This paper describes the background and status of a new International Network on “Incorporating Ageing Effects into Probabilistic Safety Assessment”. The Joint Research Centre (JRC) of the European Commission organized in September 2004 the kickoff meeting of this Network at JRC's Institute for Energy in Petten, Netherlands, with the aims to open the APSA Network, to start discussion of ageing issues in relation to incorporating ageing effects into PSA tools and to come to consensus on objectives and work packages of the Network, taking into account the specific expectations of potential Network partners. The presentations and discussions at the meeting confirmed the main conclusion from the previously organized PSAM7 pre-conference workshop on “Incorporating PSA into Ageing Management”, Budapest, June 2004, namely that incorporating ageing effects into PSA seems to be more and more a hot topic particularly for risk assessment and ageing management of nuclear power plants operating at advanced age (more than 25–30 years) and for the purpose of plant life extension. However, it also appeared that, especially regarding the situation in Europe, at present there are several on-going feasibility or full studies in this area, but not yet a completed Ageing PSA leading to applications. The project's working method is a NETWORK of operators, industry, research, academia and consultants with an active interest in the area (physical networking via a series of workshops and virtual networking via the Internet). The resulting knowledge should help PSA developers and users to incorporate the effects of equipment ageing into current PSA tools and models, to identify and/or develop most effective corresponding methods, to focus on dominant ageing contributors and components and to promote the use of PSA for ageing management of Nuclear Power Plants.
Current use of probabilistic safety evaluations for advanced reactor concepts
Authors: Kirchsteiger C. , Bolado-Lavin R.
Published: December 2013

There is international consensus that for the various nuclear Advanced Reactor Concepts (ARCs), which are currently under development, design and operational safety goals as well as modern review frameworks for risk assessment required for licensing related decision-making shall be developed. The objective of this paper is to screen information available from different countries and international organisations on such efforts as far as they consider the use of Probabilistic Safety Assessment (PSA) or specific probabilistic evaluations. The screening of available information is done for many different types of ARCs (so-called Generation III, III+ and IV), including either active, passive or both types of systems. Main conclusion is that very little information is available for full probabilistic assessments, only recently some activities started on partial probabilistic evaluations, such as on the reliability of certain types of passive systems. Therefore, published statements on significantly lower risk levels achieved by certain ARCs are at least not verifiable.
Experimental investigation and CFD simulation of horizontal air/water slug flow
Authors: Vallée Christophe , Höhne Thomas , Prasser Horst-Michael , Sühnel Tobias
Published: December 2013

For the investigation of air/water slug flow, a horizontal channel with rectangular cross-section was built at Forschungszentrum Rossendorf. The channel allows the investigation of air/water co- and counter-current flows at atmospheric pressure, especially the slug behaviour. Optical measurements were performed with a high-speed video camera, and were complemented by simultaneous dynamic pressure measurements. Moreover velocity-fields were measured using Particle Image Velocimetry (PIV). A CFD simulation of the stratified co-current flow was performed using the code CFX-5, applying the Euler-Euler two fluid model with the free surface option. The grid contains 4105 control volumes. The turbulence was modelled separately for each phase using the k- based shear stress transport (SST) turbulence model. To achieve wave generation in such a short channel, the inlet water level had to be varied in time. For this purpose, the water level history was taken from a recorded image sequence and set as time-dependent boundary condition at the model inlet. The results show a wave formation up to slug development with closure of the whole channel cross-section and consequently an increase of the pressure level behind the slug. Despite unsteady conditions at the inlet of the test channel and simplified initial conditions in the model, the slug simulation with CFX is in good qualitative agreement with the experiment, while the slug length increases during its progression, witch was not observed in reality.
Performance analyses of the pool-top radiation level reduction systems at the ETRR-2 research reactor
Authors: Shokr A. M.
Published: December 2013

An analysis of the performance of the Hot Water Layer System, HWLS, and the Interconnection System, IS, at the ETRR-2 research reactor is presented. The behavior of the HWLS during the formation period of the hot water layer has been studied and is presented together with the pool-top measured radiation level associated with that behavior. Two different designs of the HWLS have been experimentally evaluated. For these two different designs, the radiation level at the ETRR-2 pool-top were characterized and discussed. The effect of adding a water purification system to the HWLS is demonstrated as well as the effect of the IS on the pool-top radiation level considering different operational conditions of the HWLS. The study shows the importance of these systems from the radiological protection point of view.
Uncertainty in cross-section calculations for reactions induced by neutrons with energy above 0.1 MeV
Authors: Broeders C. H. M. , Konobeyev A. Yu. , Mercatali L.
Published: December 2013

The uncertainty in the calculation of neutron induced reaction cross-sections using modern nuclear models and codes has been investigated. The cross-sections have been calculated with the help of the TALYS code and the modified ALICE code using different models for the calculation of nuclear level density. The experimental data from EXFOR for neutron induced reactions for nuclei from 27Al to 209Bi and incident neutron energies above 0.1 MeV have been used for the comparison with calculations. The results obtained give the possibility to find the best approaches for the cross-section calculation for nuclei from different mass ranges.
Determination of the exposure build-up factor in a slab using the LTSN method
Authors: Rodriguez B. A. , de Vilhena M. T. , Borges V.
Published: December 2013

In this paper an exposure build-up factor formulation is reported by solving the photon transport equation in a heterogeneous slab by the LTSN method, assuming the Klein-Nishina scattering kernel as the scattering differential cross section as well as the multigroup model in the wavelength variable. Numerical simulations and comparisons with available results in the literature are presented for different compositions containing water, iron and lead.
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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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.