bluuebunny/crossref_metadata_2025_split huggingface dataset:
Regulation of nuclear decommissioning
Authors: Wendling R. D. , Paul M. 
Published: December 2013

Abstract
Being an integral part of the life of all nuclear facilities, the decommissioning of these facilities is an important task to the owners as well as to the regulators and authorities. This applies not only to large nuclear facilities such as power plants and fuel cycle facilities but is also referring to the multitude of small nuclear applications in medicine, industry and research. The article provides an overview of the statutory and legal framework, the regulatory body and the licensing procedure for the decommissioning and dismantling of nuclear facilities in Germany. Furthermore, the respective international regulatory framework as it is laid down in the Joint Convention and the IAEA Safety Standards is outlined and a brief description of the current activities of WENRA is given.
The role of clearance in Germany – release of materials, buildings and sites
Authors: Thierfeldt S. 
Published: December 2013

Abstract
Clearance in Germany is far advanced. A complex regulatory framework exists in the Radiation Protection Ordinance (RPO/Strahlenschutzverordnung, StrlSchV) since 2001 which has replaced previous recommendations on clearance issued by the German Radiation Protection Commission (Strahlenschutzkommission, SSK). A number of options exist both for unconditional clearance (e. g. of all kinds of materials with no restrictions on the destination or future use) and for clearance for a specific purpose (e. g. recycling of metal scrap). For each clearance option there is a set of nuclide specific clearance levels which have been derived on the basis of the 10 Sv/a individual dose criterion using complex radiological models which are tailored to the respective material cycles (metal scrap, buildings, building rubble, waste for disposal, sites). The clearance levels have been thoroughly reviewed by the SSK. An extensive comparison to sets of clearance levels used in other countries or issued as guidance by international bodies revealed that the correspondence between values is between good and acceptable taking into account country-specific approaches and special assumptions which would not necessarily pertain to German situations. Clearance is a major factor in the material management, especially of nuclear installations undergoing decommissioning. The variety of clearance options leaves the operator of a nuclear installation sufficient flexibility for optimisation of the material management. Clearance is of particular importance in a country like Germany where the estimated costs for a future repository are very high and the interim storage facilities for radioactive material are limited and costly. While the licensee is responsible for his material management, the issuance of a clearance permit is done by the regulatory body upon application. The regulators may choose to use the clearance levels and regulations as stipulated by the RPO, or to impose site-specific regulations on the licensee. The actual clearance measurement, i. e. checking the compliance of the residual radionuclide content in or on a particular material, building or site with the clearance levels, is performed under the responsibility of the licensee's appointed radiation protection officer, with additional checks usually made by the regulator or an independent expert acting on his behalf. After compliance verification, the material will be reused, recycled or disposed of according to the clearance option, being then subject to the conventional waste legislation.
Decommissioning of back-end nuclear facilities
Authors: Dagbjartsson S. 
Published: December 2013

Abstract
Concerning the decommissioning of facilities related to reprocessing of spent nuclear fuel and processing and disposal of radioactive waste, i. e., facilities for waste conditioning or storage, specific features prevail. Such facilities are generally optimised for the handling of radioactive substances and, therefore, are frequently highly contaminated, but do not contain any activated material. One main concern of their decommissioning will usually be the decontamination of extended surfaces of process equipment and building structures. Depending on the extent and success of the decontamination, dismantling tasks might have to be remotely performed. Based on a survey of the number of such facilities, typical issues associated with their decommissioning were found to be well presented by focussing on the examples of the decommissioning of the two prototype spent fuel reprocessing plants at Dessel/Belgium and Karlsruhe/Germany. Such issues are mainly associated with the presence of alpha-contamination of the process equipment and large building structures. Experience gained from solving associated problems is discussed.
Decommissioning and rehabilitation of uranium and thorium production facilities
Authors: Hagen M. , Kunze C. , Schmidt P. 
Published: December 2013

Abstract
The use of nuclear energy for military as well as for peaceful purposes was and remains closely connected with the mining and processing of uranium ore and, to a lesser extent, of thorium ore. Mining and processing of radioactive ores are characterised by the generation of huge amounts of radioactive residues, massive impacts upon ecosystems, landscape reshaping (or devastation in some places), and the monostructural socio-economic orientation of human settlement areas. However, a great number of the mines and mills commissioned during the cold war have been already closed, either for deposits being depleted of economically recoverable resources or on political grounds. The specifics of uranium/thorium mining and milling make high demands on the decommissioning and rehabilitation of the facilities which in addition to radiological aspects would have to address issues such as water pollution control and soil conservation, future site re-use, landscaping, and infrastructure development. The present paper gives an overview of the state of decommissioning and rehabilitation. Radiological specifics and their integration into the decommissioning and rehabilitation management are demonstrated for the rehabilitation of uranium mining legacies in Saxony and Thuringia.
Measurement of the neutron fluence on the spallation source at Dubna
Authors: Adam J. , Barashenkov V. S. , Ganesan S. , Golovatiouk S. , Gupta S. K. , Kailas S. , Krivopustov M. I. , Kumar V. , Kumawat H. , Palsania H. S. , Pronskikh V. , Sharma Manish , Tsoupko-Sitnikov V. M. , Vladimirova N. , Westmeier H. , Westmeier W. 
Published: December 2013

AbstractA232Th activation sample was exposed directly to the spallation neutrons produced by 1GeV proton colliding with a lead target and other samples of197Au,209Bi,59Co,115In and181Ta were irradiated by moderated neutrons after passing through 6 cm of paraffin. Different (n, γ) and (n, xn) reaction products in232Th,197Au and209Bi samples are analyzed by gamma-ray spectrometry to measure the neutron fluence at the positions of the samples. For this purpose a histogram of ‘flat’ spectrum averaged neutron cross-sections in over thirty given energy groups was determined from the PREPRO-2000 software. Results of measurements are compared with Monte-Carlo simulations by the Dubna CASCADE code. Experimental values for both spallation and moderated neutron fluences are in reasonable agreement with the CASCADE code. Thus, the experimentally measured fluence of moderated neutrons was used to infer the ‘one group’ reaction cross-sections of209Bi (n, xn) reactions for x = 6 to 9.
Neutron reflection and leakage models for one-dimensional reactor computations
Authors: de Abreu M. P. 
Published: December 2013

Abstract
Exact discrete ordinates interface conditions were derived for neutron reflection and leakage for one-dimensional nuclear reactor computations with anisotropic scattering. The key to our derivation is a suitable reformulation of exact discretized equations – spectral nodal equations defined in the nonactive surrounding regions of nuclear reactor cores. As a result, interface conditions are obtained that can advantageously replace the nonactive surrounding regions of a nuclear reactor core in reactor computations. Numerical experiments with a representative one-dimensional code and an axial core model with anisotropic scattering were performed in order to illustrate the improvements that can be experienced by most one-dimensional discrete ordinates nuclear reactor codes with our interface conditions.
The singular eigenfunction method: the critical slab problem for linearly anisotropic scattering
Authors: Türeci R. G. , Güleçyüz M. Ç. , Kaşkaş A. , Tezcan C. 
Published: December 2013

Abstract
The critical slab problem for linearly anisotropic scattering is investigated using the singular eigenfunction method. The third form of the transport equation is considered. The singular eigenfunctions for linearly anisotropic scattering are inserted into the Green's function. This Green's function with the full range orthogonality relations of the singular eigenfunctions together with the appropriate boundary conditions provide the criticality equation. This equation is exact and leads as shown in tables to fast converging numerical results.
Integrated management systems in the nuclear field
Authors: Beckmerhagen I. A. , Berg H. P. , Karapetrovic S. V. , Willborn W. O. 
Published: December 2013

Abstract
In the last years several internationally accepted standards such as the ISO 9000 and ISO 14000 series and other function-specific management systems standards have been developed. At the same time, it has become imperative for organisations to continuously improve their overall quality, environmental and safety performance. Therefore, the need to create integrated management systems is of growing importance to enable an easier handling of the different management systems. This paper has two main objectives. The first one is to address the key issues in the underlying theory of integrated management systems including benefits and limits, the second one is to illustrate the importance of an integrated (in particular safety) management system and the experience feedback providing examples from different areas and different organisations in the nuclear field.
Phenomenological model for non-equilibrium deuteron emission in nucleon induced reactions
Authors: Broeders C. H. M. , Konobeyev A. Yu. 
Published: December 2013

Abstract
A new approach is proposed for the calculation of non-equilibrium deuteron energy distributions in nuclear reactions induced by nucleons of intermediate energies. It combines the model of the nucleon pick-up, the coalescence and the deuteron knock-out. Emission and absorption rates for excited particles are described by the preequilibrium hybrid model. The model of Sato, Iwamoto, Harada is used to describe the nucleon pick-up and the coalescence of nucleons from the exciton configurations starting from (2p, 1h). The model of deuteron knock-out is formulated taking into account the Pauli principle for the nucleon-deuteron interaction inside a nucleus. The contribution of the direct nucleon pick-up is described phenomenologically. The multiple pre-equilibrium emission of particles is taken into account. The calculated deuteron energy distributions are compared with experimental data from 12C to 209Bi
Thermal-hydraulic assessment of topaz irradiation processes at the ETRR-2 research reactor
Authors: El-Morshdy S. El-Din 
Published: December 2013

Abstract
The present work is carried out in order to evaluate the cooling process of irradiated topaz at the Egyptian Second Research Reactor (ETRR-2). A mathematical model is developed to simulate the cooling process of a part of the irradiation grid containing one topaz irradiation box. The coolant flow rate distribution through the model channels is estimated taken into consideration the buoyancy effect in each channel. The coolant temperature distributions along the model channels are predicted as well. The COSMOS/M finite element code is used to estimate the temperature distribution through one irradiated topaz stone. The predicted coolant and surface temperature values show the efficient cooling of topaz stones irradiated at the ETRR-2 research reactor.
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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.