bluuebunny/crossref_metadata_2025_split huggingface dataset:
Regulatory requirements on level 2 PSA in Germany and their associated potential to improve emergency management
Authors: Weil L. 
Published: December 2013

Abstract
The German regulatory PSA guide describes the fundamental requirements with respect to probabilistic safety assessments as a central element of comprehensive safety reviews. The technical details regarding the performance of PSA are set out in two technical documents (PSA Methods and PSA Data), which have been developed by a working group of PSA experts (FAK PSA). Based on the experiences gained from the first series of periodic safety reviews in Germany, on international experience and on the fact that comprehensive safety reviews are now mandatory since April 2002, all guides hat to be revised. The first guide updated was the PSA guide together with the corresponding technical documents, issued October and November 2005. This article describes the key points of the revised requirements on PSA methods and data as well as their associated potential to improve emergency planning on nuclear sites.
Determination of the RBMK-1500 reactor passport characteristics
Authors: Pabarcius R. , Tonkunas A. , Listopadskis N. 
Published: December 2013

Abstract
During the whole life of reactor equipment it has to be shown that safety relevant limitations of general reactivity characteristics are not exceeded. Usually, safety margins of these characteristics are listed in a technical certificate (reactor passport). This paper presents the calculation results of the Ignalina nuclear power plant unit 2 reactor passport characteristics obtained by the best estimate 3D computer code QUABOX/CUBBOX, adjusted for the RBMK-1500 reactor core. The characteristics determined are supported by an uncertainty and sensitivity analysis which was performed using the software package SUSA. The uncertainty of the reactivity coefficients and associated effects were quantified in terms of statistical tolerance limits. The sensitivity analysis provided a ranking of the uncertain input parameters with respect to their contribution to the uncertainty of the reactivity characteristics.
Application of the spectral Green's function method for numerically solving discrete ordinates problems in cylindrical geometry
Authors: Öztürk H. , Anlı F. , Güngör S. 
Published: December 2013

Abstract
The spectral Green's function (SGF) method is used to solve numerically the transport of neutrons in an infinite homogeneous cylindrical domain. In the method, the scattering of neutrons is assumed to be isotropic and the transport equation in cylindrical geometry is reduced to plane-like transport equation by using our reasonable approach. Then, the plane-like transport equation is solved by using the SGF method and the results are compared with those already obtained by the Diamond Difference (DD) scheme to check the accuracy of the results. The agreement of the results is good and it can be concluded that the SGF method is very effective for the numerical solution of the transport equation.
Phenomenological model of TENORM formation in industrial processes
Authors: Gellermann R. 
Published: December 2013

Abstract
This paper discusses a simple phenomenological model of TENORM formation. The model is based on a mass and activity balance in an industrial process. It characterises the radionuclide enrichment of a residue in relation to the input materials according to three parameters (mass transfer factor MTF, activity transfer factor ATF and enrichment factor EF). The model shows that high enrichment is only possible if mass transfer into the residue is low. By analysing data from several real industrial processes, the co-precipitation of dissolved radium with barium salts (EF about 104), the volatilisation of lead and polonium in thermal processes (EF about 300) and the combustion of fuels (EF 10–100) are identified as very effective enrichment processes. However, in real industrial processes the specific activity of residues may be influenced by admixtures of masses after the decisive step of radionuclide redistribution. Therefore, focus should be placed on the identification and characterisation of elementary processes of TENORM formation. Several such processes are discussed in this paper.
Development of an object-specific remediation solution for the waste rock dumps at the formerly abandoned Wismut sites
Authors: Regner J. , Schulz H. 
Published: December 2013

Abstract
The investigation of the radiation situation is an essential part of preparing the remediation of the formerly abandoned Wismut sites. If the exposure is relevant, then the exposure pathway “inhalation of radon/radon daughters” plays the dominant role. Therefore, the development of a straightforward approach suitable for the evaluation of the involved radon situation is one of the most important tasks. It has to include the whole chain beginning with the characterization of the relevant exposure situation and ending up with the identification of remedial options. For that purpose a new technique of visualizing the radon flux from the waste rock dumps has been developed. The specific conditions at formerly abandoned Wismut sites will be demonstrated and the remediation solutions envisaged to reduce the radon flux will be explained.
Radon in Austria: metrology and practice
Authors: Maringer F. J. 
Published: December 2013

Abstract
The implementation of radon mitigation and precaution standards needs continuously scientific attendance and research networking on international level. Otherwise the radon issue could degrade easily to a simplified techno-economical exercise without sustainable results in public health. In this paper the radon investigations in Austria which have been carried out in the last 20 years and the applied methods and derived standards for mitigation and precaution at home and workplaces are discussed. Future strategies, scientific and social necessities to solve the radon problem are outlined comprehensively. The agreed and proposed future research objectives and cooperations in Austria that seem absolutely necessary are discussed in consideration of the medium-term perspective of the European radiation protection.
Workplaces with natural radionuclides – investigation results from Bavaria
Authors: Reifenhäuser C. , Körner S. , Mallick R. 
Published: December 2013

Abstract
Since the amendment of the German Radiation Protection Ordinance in the year 2001 special workplaces with increased exposure due to natural terrestrial radiation have been subject to legal regulations. After the Radiation Protection Ordinance came into force the Bavarian Environment Agency informed all water supply facilities, breweries, dairies, mineral water manufactures, operators of underground workplaces and radon spas that the annual radon exposure of their staff has to be estimated. In Bavaria approximately 3330 companies with workplaces with exposure to radon were investigated. Herein the water supply facilities represented the largest group with 2550 companies. In about 96 % of all investigated companies the radon exposure of the staff is below the action level of 6 mSv per year. In 75 water supply facilities and two mines the radon exposure of the staff exceeded the action level (6 mSv/a) or the annual limit (20 mSv/a). Workplaces with increased exposure due to uranium and thorium are of secondary importance in Bavaria. Within the scope of a research project 13 companies were found, that still use thoriated welding electrodes. Only a few facilities use thoriated mantles, thorium alloys or natural thorium and uranium. The effective dose of all examined workers was below the action level of 6 mSv per year.
Study on the thermal-hydraulics characteristics of a boiling two-phase natural circulation loop with nanofluids
Authors: Nayak A. K. , Gartia M. R. 
Published: December 2013

Abstract
In this paper, we present the steady state and stability behaviour of a boiling two-phase natural circulation loop with a small concentration of Al2O3 nanofluids. For this, experiments were conducted in a natural circulation loop at different operating powers, pressures, and water levels in the steam drum. The test results revealed that while the steady state natural circulation behaviour is similar, however, the buoyancy induced flow rates are found to be relatively higher with nanofluids than with water alone. In addition, with nanofluids, the amplitudes of the boiling induced Type I instabilities are found to be significantly reduced, However, the Type II instabilities are found to appear at relatively lower power with addition of nanofluids as compared to that with water alone. The most interesting result is that with addition of nanofluids, the amplitudes of Type II instabilities are found to get suppressed with increase in power similar to that of low quality Type I instability.
The effect of combination of different materials on neutron absorption in a nuclear research reactor spent fuel pool
Authors: Logar M. 
Published: December 2013

Abstract
So far, only individual absorbing materials have been studied for neutron absorption in the pool-type reactor spent fuel storages. In this paper, the effect of combining various materials, which are already known as good neutron absorbers by themselves, on lowering the multiplication factor of spent fuel pool, is described. The results of Monte Carlo computer code MCNP calculations for the spent fuel pool of the TRIGA research reactor in Ljubljana, Slovenia, are specified. Due to property completion of the chosen absorber materials, some of their combinations partially decrease the multiplication factor and thus contribute to the enhancement of the safety conditions of the pool.
Re-evaluation of the criticality experiments of the “Otto Hahn Nuclear Ship” reactor
Authors: Lengar I. , Snoj L. , Rogan P. , Ravnik M. 
Published: December 2013

Abstract
Several series of experiments with a “FDR” reactor (advanced pressurized light water reactor) were performed in 1972 in the Geesthacht critical facility ANEX. The experiments were performed to test the core prior to its usage for the propulsion of the first German nuclear merchant ship “Otto-Hahn”. In the present paper a calculational re-evaluation of the experiments is described with the use of the up-to date computer codes (Monte-Carlo code MCNP5) and nuclear data (ENDF/B-VI release 6). It is focused on the determination of uncertainties in the benchmark model of the experimental set-up, originating mainly from the limited set of information still available about the experiments. Effects of the identified uncertainties on the multiplication factor were studied. The sensitivity studies include parametric variation of material composition and geometry. The combined total uncertainty being found 0.0050 in keff, the experiments are qualified as criticality safety benchmark experiments.
Behavior of CANDU fuel under power pulse conditions at the TRIGA reactor of INR Pitesti
Authors: Horhoianu G. , Dobrea D. , Parvan M. , Stefan V. 
Published: December 2013

Abstract
Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensors for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280 cal g–1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti.
Effect of using various grades of plutonium in the protective liquid wall of an IFE type fusion reactor
Authors: Übeyli M. 
Published: December 2013

Abstract
One of the main issues in a fusion reactor is the need for the frequent replacement of first wall structures during its lifetime. In order to eliminate this problem completely, use of a protective liquid wall between the fusion plasma and solid first wall has been proposed by various researchers. One of the major fusion energy reactor designs is called HYLIFE-II, an inertial fusion energy (IFE) type reactor that also utilizes such a flowing liquid wall to extend the lifetime of the first wall to its lifetime. This study presents the neutronic analysis of the HYLIFE-II reactor using Li2BeF4 (Flibe) with various grades of plutonium as a flowing liquid wall. For this purpose, four different coolants containing Flibe with 0.5 % or 1 % of weapon grade (WG) PuF4 and Flibe with 0.5 % or 1 % reactor grade (RG) PuF4 were considered. Main neutronic parameters: tritium breeding ratio (TBR), energy multiplication factor (M) and radiation damage at the first wall were computed with respect to the liquid wall thickness and coolant type. The highest TBR and M values were reached when the Flibe with 1 % WG PuF4 was used in the liquid wall.
A discrete ordinates scheme for void fraction evaluation with nonstandard reflective conditions and weakly divergent beams
Authors: de Abreu M. P. 
Published: December 2013

Abstract
In this article, a number of due changes in both a discrete ordinates neutron transport model and a companion void fraction evaluation scheme are introduced. These changes are due to the explicit consideration of a weakly divergent neutron beam, i. e. a neutron beam consisting of a monodirectional component with normal incidence, which is modeled by a Dirac delta distribution, and an angularly continuous component, which is modeled by a smooth function of the angular variable. Computational tests are performed in order to illustrate the numerical consistency of the evaluation scheme for weakly divergent beams with respect to the order of angular quadrature, as well as its low sensitivity to experimental inaccuracies in the detector responses.
Methodology for Fire PSA during design process
Authors: Blombach J. , Kollasko H. 
Published: December 2013

Abstract
Fire PSA is an essential part of a full scope level 1 PSA. Cable fires play an important role in fire PSA. Usually, cable routing is therefore modeled in detail. During the design of new nuclear power plants the information on cable routing is not yet available. However, for the use of probabilistic safety insights during the design and for licensing purposes a fire PSA may be requested. Therefore a methodology has been developed which makes use of the strictly divisional separation of redundancies in the design of modern nuclear power plants: cable routing is not needed within one division but replaced by the conservative assumption that all equipment fails due to a fire in the concerned division; critical fire areas are defined where components belonging to different divisions may be affected by a fire. For the determination of fire frequencies a component based approach is proposed. The resulting core damage frequencies due to fire are conservative.
The Open PSA Initiative for next generation probabilistic safety assessment
Authors: Epstein S. , Rauzy A. , Reinhart F. M. 
Published: December 2013

Abstract
The Open PSA Initiative round table discussion session is to bring the PSA community up-to-date on our meetings, workshops, and working groups during the last year. We will have members of the initiative speak on their point of view of the initiative, what they have gained, and where we as a group want to go during the next year. The main idea is to use appropriate modules of different computer codes for a specific purpose which requires a common platform and a common PSA format. In this short paper we introduce our main project for this year: the Open PSA Model Exchange Format.
Overview on the different applications of probabilistic safety assessment for nuclear power plants
Authors: Berg H.-P. 
Published: December 2013

Abstract
Worldwide it can be recognised that the use of probabilistic safety assessment (PSA) in regulatory as well as operational decision-making is state of the art and seen as a successful development. Therefore, in most cases the regulator encourages the performance of PSAs to provide information to complement and support the defence in depth philosophy as well as operational configuration decisions. The main application of the PSA is still as part of integrated safety reviews, in particular in the frame of comprehensive (periodic) safety reviews. Other more specific applications areas of PSA are, among others, design evaluation, event analysis with aid of PSA, evaluation of technical specifications; risk-informed in-service inspection, risk monitoring and accident management. The extent of these applications vary from country to country but has been increasing during the last years.
Practicle example of the defence of a contamination incident
Authors: Höfer H. 
Published: December 2013

Abstract
By routine examinations of a radiation worker, a contamination incident with transuranium nuclides was discovered. Immediate investigations showed a serious degree of the incident. First countermeasures were done to narrow the effects down and to avoid a spread of contamination. A complex decontamination followed for which the radiation protection infrastructure had to be built up.
Background information about polonium-210 – considerations on its biokinetics and internal dosimetry
Authors: Oeh U. , Li W. B. , Gerstmann U. , Giussani A. , Paretzke H. G. 
Published: December 2013

Abstract
The mysterious death of the former secret service agent Alexander Litvinenko on 23 November 2006 suddenly attracted the attention of the public to the rather unknown radionuclide 210Po. In this respect, this paper presents some general background information about this radionuclide as well as some details about its biokinetic behaviour and internal dosimetry. The results are discussed in the context of the Litvinenko case with regard to possible health effects on people from Germany being potentially contaminated. However, no evidence for contamination with 210Po could be found for any people tested at the GSF-National Research Center for Environment and Health (since 01.01.2008 Helmholtz Zentrum München).
Hazard and precautionary measures in connection with negligent or wilful contamination
Authors: Ahlfänger R. , Scheidegger R. 
Published: December 2013

Abstract
Handling open radioactive substances requires extensive precautionary measures in order to avoid any danger. During the past 100 years many persons were injured or killed due to unintended external irradiation, contamination of the skin or uptake of radioactive substances. With the help of organisational measures with users of sources and with operators of radionuclide laboratories or nuclear installations numerous accidents can be avoided or their consequences can be mitigated. Also the government itself is challenged in order to register and secure radioactive sources and maintaining the national safeguard. Education of personnel handling radioactive substances plays an important role when it comes to precautionary measures. In combination with suitable technical equipment it is possible to obtain a high degree of protection against the hazard from negligent or wilful contamination.
Prediction of time dependent standby failure rates for periodically tested components taking into account the operational history
Authors: Šimić Z. , Vrbanić I. , Vuković I. 
Published: December 2013

Abstract
The prediction of time dependent standby failure rates was studied, taking into account the operational history of a component. These studies are important for applications such as system modeling in probabilistic safety analysis to evaluate the impact of equipment aging and maintenance strategies on the risk measures (e. g. reactor core damage frequency) considered. The time dependent model for the standby failure rate is defined based on the Weibull distribution and the principles of proportional age reduction by equipment overhauls. The parameters which determine the standby failure rate are estimated, including the definition of the operational history model and likelihood function for Bayesian analysis of parameters for periodically tested standby repairable components. The operational history is provided as time axis with defined times of overhauls, surveillance tests and failures. Assessment of time dependent unavailability due to the failure of periodically tested standby components is described. As an example, the prediction of the future behavior of components for seven different operational histories is described.
Numerical simulation of passive catalytic recombiner
Authors: Gera B. , Sharma Pavan K. , Singh R. K. , Ghosh A. K. 
Published: December 2013

Abstract
Resolving hydrogen related safety issues, pertaining to nuclear reactor safety have been an important area of research world over for the past decade. Studies on hydrogen transport behavior and development of hydrogen mitigation systems are still being pursued actively in various research labs, including Bhabha Atomic Research Centre (BARC), in India. The Passive Catalytic Recombiner is one of such hydrogen mitigating devices consisting of catalyst surfaces arranged in an open-ended enclosure. In the presence of hydrogen with available oxygen, a catalytic reaction occurs spontaneously at the catalyst surfaces and the heat of the reaction produces a natural convection flow through the enclosure. The present study aims for performing a CFD simulation to obtain the temperature distribution inside the recombiner. A 3D CFD model has been developed to study the mechanism of catalytic recombination and has been tested for literature quoted experiments. A parametric study has been performed for a particular recombiner geometry for various inlet conditions. Salient features of the simplified CFD model developed at BARC and results of the present model calculations are presented in this paper.
An analytical solution for the one-dimensional time-dependent S

N
transport equation for bounded and unbounded domains in cartesian geometry
Authors: Segatto C. F. , Vilhena M. T. , Gonçalez T. T. 
Published: December 2013

Abstract
In this work a general analytical solution is reported, expressed in integral form for the time-dependent, one-dimensional SN transport equation in cartesian geometry valid for bounded and unbounded domains (0 < x <), using the double Laplace transform technique. The main idea consists in the application of the Laplace transform technique in time variable and the solution of the resulting equation by the LTSN method using appropriate boundary conditions for bounded and unbounded domain problems. We also report about the numerical simulations carried out.